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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

SU-E-I-43: Photoelectric Cross Section Revisited

Purpose: The importance of the precision in photoelectric cross-section value increases for recent developed technology such as dual energy computed tomography, in which some reconstruction algorithms require the energy dependence of the photo-absorption in each material composition of human being. In this study, we revisited the photoelectric cross-section calculation by self-consistent relativistic Hartree-Fock (HF) atomic model and compared with that widely distributed as “XCOM database” in National Institute of Standards and Technology, which was evaluated with localdensity approximation for electron-exchange (Fock)z potential. Methods: The photoelectric cross section can be calculated with the electron wave functions in initial atomic state (bound electron) and final continuum state (photoelectron). These electron states were constructed based on the selfconsistent HF calculation, where the repulsive Coulomb potential from the electron charge distribution (Hartree term) and the electron exchange potential with full electromagnetic interaction (Fock term) were included for the electron-electron interaction. The photoelectric cross sections were evaluated for He (Z=2), Be (Z=4), C (Z=6), O (Z=8), and Ne (Z=10) in energy range of 10keV to 1MeV. The Result was compared with XCOM database. Results: The difference of the photoelectric cross section between the present calculation and XCOM database was 8% at a maximum (in 10keV for Be). The agreement tends to be better as the atomic number increases. The contribution from each atomic shell has a considerable discrepancy with XCOM database except for K-shell. However, because the photoelectric cross section arising from K-shell is dominant, the net photoelectric cross section was almost insensitive to the different handling in Fock potential. Conclusion: The photoelectric cross-section program has been developed based on the fully self-consistent relativistic HF atomic model. Due to small effect on the Fock potential for K-shell electrons, the difference from XCOM database was limited: 1% to 8% for low-Z elements in 10keV-1MeV energy ranges. This work was partly supported by the JSPS Core-to-Core Program (No. 23003)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

MANTA. An Integral Reactor Physics Experiment to Infer the Neutron Capture Cross Sections of Actinides and Fission Products in Fast and Epithermal Spectra

Neutron cross-sections characterize the way neutrons interact with matter. They are essential to most nuclear engineering projects and, even though theoretical progress has been made as far as the predictability of neutron cross-section models, measurements are still indispensable to meet tight design requirements for reduced uncertainties. Within the field of fission reactor technology, one can identify the following specializations that rely on the availability of accurate neutron cross-sections: (1) fission reactor design, (2) nuclear fuel cycles, (3) nuclear safety, (4) nuclear safeguards, (5) reactor monitoring and neutron fluence determination and (6) waste disposal and transmutation. In particular, the assessment of advanced fuel cycles requires an extensive knowledge of transuranics cross sections. Plutonium isotopes, but also americium, curium and up to californium isotope data are required with a small uncertainty in order to optimize significant features of the fuel cycle that have an impact on feasibility studies (e.g. neutron doses at fuel fabrication, decay heat in a repository, etc.). Different techniques are available to determine neutron cross sections experimentally, with the common denominator that a source of neutrons is necessary. It can either come from an accelerator that produces neutrons as a result of interactions between charged particles and a target, or it can come from a nuclear reactor. When the measurements are performed with an accelerator, they are referred to as differential since the analysis of the data provides the cross-sections for different discrete energies, i.e. σ(E i ), and for the diffusion cross sections for different discrete angles. Another approach is to irradiate a very pure sample in a test reactor such as the Advanced Test Reactor (ATR) at INL and, after a given time, determine the amount of the different transmutation products. The precise characterization of the nuclide densities before and after neutron irradiation allows to infer energy-integrated neutron cross sections, i.e. ∫₀ ∞ σ(E)φ(E)dE, where φ(E) is the neutron flux “seen” by the sample. This approach, which is usually defined and led by reactor physicists, is referred to as integral and is the object of this report. These two sources of information, i.e. differential and integral, are complementary and are used by the nuclear physicists in charge of producing the evaluated nuclear data files used by the nuclear community (ENDF, JEFF…). The generation of accurate nuclear data files requires an iterative process involving reactor physicists and nuclear data evaluators. This experimental program has been funded by the ATR National Scientific User Facility (ATR-NSUF) and by the DOE Office of Science in the framework of the Recovery Act. It has been given the name MANTRA for Measurement of Actinides Neutron TRAnsmutation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Improved Cross Section Generation Capability of Griffin in FY22

The Griffin code is a Multiphysics Object-Oriented Simulation Environment (MOOSE) based reactor multiphysics analysis application jointly developed by Idaho National Laboratory and Argonne National Laboratory. The code includes a variety of deterministic steady-state transport solvers for fixed source, k-eigenvalue, adjoint, and subcritical multiplication as well as transient solvers for spatial dynamics with the improved quasi-static method. Griffin uses cross section data in the ISOXML format generated from external deterministic or Monte Carlo cross section generation codes. In recent years, the MC 2 -3 modules have been added to Griffin for fast reactor cross section generation, and the self-shielding application programming interface (SSAPI) was implemented in the ISOXML module for thermal reactor cross section generation. The on-the-fly slowing down method and double-heterogeneity treatment have been implemented to SSAPI and verified against particulate fuel-bearing graphite-moderated thermal reactor problems with high accuracy. This year, work has been focused on improving the cross section generation capability of ISOXML and streamlining the cross section generation procedures. In addition, the form function data were added to ISOXML in order to support the pin power reconstruction capability that was newly implemented in Griffin in this fiscal year. To facilitate the cross section generation using MC 2 -3 and SSAPI in Griffin, the cross section generation workflows have been set up for both fast and thermal spectrum reactors. The MOOSE action system tool was devised for fast spectrum problems, and the MOOSE stochastic tool was adopted to the branch calculation procedure for thermal spectrum problems. Meanwhile, to ensure the accuracy of group-constants, the thermal up-scattering kernel calculator accounting for resonance scattering was implemented in ISOXML, demonstrating the accurate computation of a Doppler-broadened scattering kernel of any Legendre order within a reasonable timescale. Other aspects of ISOXML, such as deletion solver and data, documentation, ISOXML file management, and the interface for Mixture, were improved as well.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Electron Scattering from 1-Methyl-5-Nitroimidazole: Cross-Sections for Modeling Electron Transport through Potential Radiosensitizers

In this study, we present a complete set of electron scattering cross-sections from 1-Methyl-5-Nitroimidazole (1M5NI) molecules for impact energies ranging from 0.1 to 1000 eV. This information is relevant to evaluate the potential role of 1M5NI as a molecular radiosensitizers. The total electron scattering cross-sections (TCS) that we previously measured with a magnetically confined electron transmission apparatus were considered as the reference values for the present analysis. Elastic scattering cross-sections were calculated by means of two different schemes: The Schwinger multichannel (SMC) method for the lower energies (below 15 eV) and the independent atom model-based screening-corrected additivity rule with interferences (IAM-SCARI) for higher energies (above 15 eV). The latter was also applied to calculate the total ionization cross-sections, which were complemented with experimental values of the induced cationic fragmentation by electron impact. Double differential ionization cross-sections were measured with a reaction microscope multi-particle coincidence spectrometer. Using a momentum imaging spectrometer, direct measurements of the anion fragment yields and kinetic energies by the dissociative electron attachment are also presented. Cross-sections for the other inelastic channels were derived with a self-consistent procedure by sampling their values at a given energy to ensure that the sum of the cross-sections of all the scattering processes available at that energy coincides with the corresponding TCS. This cross-section data set is ready to be used for modelling electron-induced radiation damage at the molecular level to biologically relevant media containing 1M5NI as a potential radiosensitizer. Nonetheless, a proper evaluation of its radiosensitizing effects would require further radiobiological experiments.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Performance Improvements to the Cross Section Calculation in MPACT

The MPACT code is the primary deterministic neutron transport solver available in the Virtual Environment for Reactor Applications (VERA) as part of the Consortium for Advanced Simulation of Light Water Reactors (CASL). In contrast to the conventional two-step method using homogenization techniques, MPACT employs a direct transport methodology, where the transport calculation is performed with the explicit reactor core geometry, material compositions, and temperature distribution. As the direct whole-core transport calculation requires significant computational resources, performance improvements have been recently conducted in various components of MPACT. The initial MPACT profiling was performed on the EOS cluster at the Oak Ridge Leadership Computing Facility. Table I shows that the cross section processing accounted for 74% and 33% of the total runtime for VERA Progression Problems 5 and 7. Starting with a problem-independent multigroup (MG) library, two primary functionalities are performed in MPACT to obtain the problem-dependent MG cross sections: (1) the resonance self-shielding calculation, which currently MPACT is approaching with the subgroup method; (2) use of equivalence cross section from resonance calculation to compute the microscopic cross sections, and thus, the macroscopic cross sections of material regions. As shown in Table I, the time distribution in cross section processing varies from case to case. For Problem 5, a 2-D quarter core depletion case, the macroscopic XS dominates due to the increased number of isotopes in the depleted fuel. For Problem 7, a 3-D quarter core PWR at HFP and BOL, more subgroup calculations are performed to obtain updated self-shielded cross sections during the iteration. In this summary, the approaches to reduce the computing time in cross section processing are presented. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

EXSAN: An Interactive Version of the Nuclear Cross Section Data Handbook - 25687

Neutron cross section measurements and data files have advanced significantly since 1989 when Los Alamos published 'A Nuclear Cross Section Data Handbook', a detailed compendium of point-wise cross section plots and pie charts of relative contributions of individual reactions to the total cross section for 129 isotopes. Other compendiums with similar data and analysis tools have also been published. The intent of this work is to develop a modern, open-source application for interactive exploration and analysis of Evaluated Nuclear Data File (ENDF) libraries that also allows users to create their own 'Nuclear Cross Section Data Handbooks'. EXSAN is an easy-to-use graphical user interface application for exploring ENDF photon and neutron cross sections developed in Python. In addition to interactive plotting, EXSAN includes methods for computing rank-order lists of cross section for any energy and calculating composite cross sections for isotopic and elemental compositions and additional batch automation features. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

Generating An Advanced Cross-section Library For HTGR Pebble Bed Depletion Calculations Using Reduced-Order Model Generation Techniques

For code development, Advanced Reactor Technologies - Gas Cooled Reactors Program (ART-GCR) rely on a collaboration with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, but the cross sections generation and the methodology definition is part of this program area goals. Based on previous studies in FY23, the size of microscopic cross section libraries increases rapidly with the number of tabulations, requiring significant amount of memory and drastically slowing down the Griffin calculations when evaluating cross sections via the multivariate linear interpolation approach. Rising to these challenges, this work investigates constructing Reduced-order Models (ROMs) for the multi-group microscopic cross sections to accelerate the cross section evaluation in Griffin. A database of multigroup cross sections is first collected considering all possible parameters that a designer could change for optimization. Down-selection of the ROM techniques afterward shows Deep Neural Network (DNN) as the best candidate when jointly consider memory efficiency, predictive accuracy, computational cost, scalability, flexibility and ease of implementation of the algorithms in comparison to the multidimensional interpolation. This work develops a specific interface that enables the cross section predictions using pre-trained DNN models into Griffin leveraging the existing ROM capabilities. DNNs have been trained for all isotopes for use in Griffin. Preliminary Griffin testing shows that DNNs exhibit exceptional predictive accuracy and the use of DNNs provides orders of magnitude improvement in memory efficiency compared to conventional interpolation techniques. With such ROM techniques, it holds great promise to further increase the fidelity of the Pebble Bed Reactor (PBR) simulation by increasing the number of tabulations/state variables during cross section evaluation, while maintaining the computational cost affordable in Griffin.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Reduce-Order Modeling of Multigroup Neutron Cross Sections for High-Temperature Gas-cooled Reactors

Deterministic neutronics calculations rely on multigroup neutron cross section libraries, which usually consists of a database of tabulated values, used to calculate the cross sections through multivariate linear interpolation. However, interpolation of the multidimensional cross section data becomes memory inefficient and time consuming as the number of tabulations increases, significantly slowing down the neutronics calculation, especially in the case of micro cross section libraries where every isotope (on the order of hundreds) has its own set of specific reactions and cross sections. To address this challenge, this work constructs efficient and robust reduced-order models (ROMs) of the multi-group cross sections to support the Griffin simulation of high-temperature gas-cooled reactors (HTGRs). The first part of the study investigates the linearity of the multi-group cross section data across isotopes, reaction types and energy groups on pre-generated datasets for the purpose of dimensionality reduction. Secondly, a down-selection of ROM techniques is presented on representative classical machine learning (ML) techniques, including variants of linear regression, kernel-based methods, tree-based algorithms, and artificial neural networks. The selection criteria jointly consider the memory efficiency, predictive accuracy, prediction speed, and scalability in comparison to the multidimensional interpolation. Among all the ML techniques, deep neural networks (DNNs) have proven to be the best selection with sufficient accuracy, high robustness, good memory efficiency, great scalability, and superior flexibility. DNNs for have been trained for all isotopes in this work and systematic Griffin testing is ongoing at this moment to ensure the feasibility of this ROM technique for cross section predictions.

42 - ENGINEERING↗
DOE OSTI2024

Advanced Cross Section Library Generation using Reduced Order Models

Deterministic neutronics calculations rely on multigroup neutron cross section libraries, which consist of databases of tabulated values, used to calculate the neutron cross sections through multivariate linear interpolation. However, interpolation of the multidimensional cross section data becomes memory inefficient and time consuming as the number of tabulations increases, significantly slowing down the neutronics calculation, especially in the case of microscopic cross section libraries where every isotope (on the order of hundreds) has its own set of specific reactions and cross sections. In order to address this challenge, this work constructs efficient and robust reduced-order models (ROMs) of the multi-group cross sections to support the Griffin simulation of high-temperature gas-cooled reactors (HTGRs). The first part of the study investigates the linearity of the multigroup cross section data across isotopes, reaction types, and energy groups on pre-generated datasets for the purpose of dimensionality reduction. Secondly, a down-selection of ROM techniques is presented on representative classical machine learning (ML) techniques, including variants of linear regression, kernel-based methods, tree-based algorithms, and artificial neural networks. The selection criteria jointly consider the memory efficiency, predictive accuracy, prediction speed, and scalability in comparison to the multidimensional interpolation. Among all the ML techniques, deep neural networks (DNNs) have proven to be the best selection with sufficient accuracy, high robustness, good memory efficiency, great scalability, and superior flexibility. DNNs have been trained for all isotopes in this work and systematic Griffin testing is ongoing to ensure the feasibility of this ROM technique for predicting cross section and reducing memory requirements without a significant sacrifice in computational performance.

42 - ENGINEERING↗
DOE OSTI2025

Characterization of uncertainties in electron-argon collision cross sections

Abstract The predictive capability of a plasma discharge model depends on accurate representations of electron-impact collision cross sections, which determine the corresponding reaction rates and electron transport properties. The values of cross sections can be known only approximately either through experiments or simulations and are thus subject to uncertainties. Quantifying the uncertainties in plasma simulations allows us to assess the reliability of simulations and to provide a basis for interpreting discrepancies between simulations and experiments. For such uncertainty quantification of plasma simulations, it is essential to quantify the uncertainties of the underlying cross sections. Although much effort has been committed to calibrate the cross section values, their uncertainties are not well investigated. We characterize uncertainties in electron-argon atom collision cross sections using a Bayesian framework. Six collision processes—elastic momentum transfer, ionization, and four excitations—are characterized with semi-empirical models, which effectively capture the features important to the macroscopic properties of the plasma. A probability model for the uncertain parameters of these semi-empirical models is developed. Specifically, a Gaussian-process likelihood model is proposed to capture discrepancies among data sets, as well as the model-form inadequacies of the semi-empirical models. Two other likelihood models are compared with the proposed Gaussian-process model, to illustrate the importance of the choice of the likelihood model. The cross section models are calibrated using the electron-beam experiments and ab-inito quantum simulations. The resulting calibrated uncertainties capture well the scattering among the data sets. The calibrated cross section models are further validated against swarm-parameter experiments and zero-dimensional Boltzmann equation simulations of widely used cross section datasets.

Chung, Seung Whan (ORCID:0000000302501549)↗
DOE OSTI2024

Determining the 9 $\mathrm{Be}$($n, γ$) 10 $\mathrm{Be}$ integral cross section at fission neutron energies

The 9 B neutron capture cross section has significant implications for Be materials in the nuclear industry as well as the α process in stellar nucleosynthesis. While the cross section is well constrained at thermal neutron energies, there is a lack of experimental data at higher neutron energies, and the evaluated nuclear data libraries can differ by up to two orders of magnitude. We calculate the 9 Be(n, γ) 10 Be integral cross section at fission neutron energies in an effort to resolve disagreements amongst the nuclear data libraries. Foil irradiation experiments were performed using the Flattop critical assembly at the National Criticality Experiments Research Center with either the highly enriched U or Pu cores, with target foil stacks placed at multiple locations to exploit different neutron energy profiles. Accelerator mass spectrometry was used to measure the 10 Be/ 9 Be ratio in irradiated Be foils, while all other activation products were quantified through gamma spectrometry. The experiments were simulated using the Monte Carlo N-Particle radiation transport code and combined with experimental results to determine the total neutron fluence, while the staysl-pnnl suite and fispact-ii code were used to validate the model and assess the systematic uncertainty. The new 9 Be(n, γ) 10 Be integral cross sections calculated in this work are 26.5 ± 2.2µb at 0.59 ± 0.07 MeV, 24 ± 3 µb at 0.98 ± 0.14 MeV, 21.7 ± 1.3 µb at 1.26 ± 0.11 MeV, 21.8 ± 1.4 µb at 1.32 ± 0.11 MeV, and 18.6 ± 1.1 µb at 1.46 ± 0.13 MeV. These results do not agree with integral cross sections from any of the nuclear data library evaluations. Discrepancies between the new integral cross sections reported here and the nuclear data libraries suggest a more complex cross-section structure in the MeV range which allows for more resonance contributions, and more work is needed to further constrain the evaluated cross sections.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2018

Monoenergetic photon-induced fission cross-section ratio measurements for U 235 , U 238 , and Pu 239 from 9.0 to 17.0 MeV

Background: High-accuracy data are needed to advance microscopic descriptions of the nuclear fission process, to improve the predictive precision of phenomenological models, and for applications in nuclear energy and homeland security. Purpose: In this work, the main goal is to provide high-accuracy cross-section data for photofission of 235 U, 238 U, and 239 Pu in the energy range from the fission threshold to the high-energy tail of the giant dipole resonance. These new data should contribute significantly to the reduction of the systematic uncertainty in evaluated photofission cross-section databases. Method: Cross-section ratios for photofission of the “big three” actinide nuclei were measured using a quasimonoenergetic photon beam and dual-fission chambers. The measurements were performed at the HIγS facility using a photon beam produced from Compton backscattering of free-electron laser light. The dual-fission chamber enabled simultaneous counting of fission events from two targets. This method allows for cross-section ratio measurements with very small systematic errors. The largest source of systematic error in the determination of the cross-section ratios is the uncertainty in the ratio of the target thicknesses. Results: We report photon-induced fission cross-section ratios for σ[ 235 U(γ,f) / 238 U(γ,f)] , σ[235U(γ,f) / 239 Pu(γ,f)] , and σ[ 238 U(γ,f) / 239 Pu(γ,f)] at photon energies from 9.0–17.0 MeV. More than 20 data points were measured for each ratio, and the systematic uncertainties are less than 2%. Conclusion: The present photofission cross-section ratio data sets are compared to ratios computed from previous measurements, to the corresponding neutron-induced fission cross-section ratios, and to ratios computed from evaluated databases. Finally, the data obtained in this work for σ[ 235 U(γ,f) / 238 U(γ,f)] are consistent with the existing data, while for σ[ 235 U(γ,f) / 239 Pu(γ,f)] and σ[ 238 U(γ,f) / 239 Pu(γ,f)] the present data are systematically lower.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2013

Methods for obtaining true particle size distributions from cross section measurements

Sectioning methods are frequently used to measure grain sizes in materials. These methods do not provide accurate grain sizes for two reasons. First, the sizes of features observed on random sections are always smaller than the true sizes of solid spherical shaped objects, as noted by Wicksell [1]. This is the case because the section very rarely passes through the center of solid spherical shaped objects randomly dispersed throughout a material. The sizes of features observed on random sections are inversely related to the distance of the center of the solid object from the section [1]. Second, on a plane section through the solid material, larger sized features are more frequently observed than smaller ones due to the larger probability for a section to come into contact with the larger sized portion of the spheres than the smaller sized portion. As a result, it is necessary to find a method that takes into account these reasons for inaccurate particle size measurements, while providing a correction factor for accurately determining true particle size measurements. I present a method for deducing true grain size distributions from those determined from specimen cross sections, either by measurement of equivalent grain diameters or linear intercepts.

36 MATERIALS SCIENCE↗
DOE OSTI2013

MC 2 -3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

The MC 2 -3 code is a Multigroup Cross section generation Code for fast reactor analysis, developed by improving the resonance self-shielding and spectrum calculation methods of MC 2 -2 and integrating the one-dimensional cell calculation capabilities of SDX. The code solves the consistent P1 multigroup transport equation using basic neutron data from ENDF/B data files to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (~2000) or hyperfine (~400,000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified isotopic temperatures. The pointwise cross sections are directly used in the hyperfine group calculation whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for two-dimensional whole-core problems to generate region-dependent broad-group cross sections. Multigroup cross sections are written in the ISOTXS format for a user-specified group structure. The code is executable on UNIX, Linux, and PC Windows systems, and its library includes all isotopes of the ENDF/BVII. 0 data.

20 FOSSIL-FUELED POWER PLANTS↗
NASA NTRS1978

Absolute elastic differential electron scattering cross sections in the intermediate energy region. IV - CO

Using a crossed electron beam-molecular beam scattering geometry and a relative-flow technique, ratios of elastic differential cross sections of CO to those of He have been measured at electron impact energies of 3, 5, 7.5, 9.9, 15, 20, 30, 50, 75, and 100 eV. At each energy, an angular range of 15 to 130 deg has been covered. These ratios have been multiplied by previously known He elastic differential cross sections to obtain elastic differential cross sections for CO. Since pure rotational excitations were not resolved, the elastic differential cross sections are a sum of elastic and pure rotational excitations at room temperature. From a knowledge of differential cross sections (DCS), integral and momentum transfer cross sections have been calculated. Both the DCS and integral cross sections are compared at 50, 75, and 100 eV to a recent two-potential theory of e-molecule scattering. Present results show that the isoelectronic molecules CO and N2 have very similar magnitudes and shapes of their differential cross sections.

Tanaka, H.↗
NASA NTRS1986

Absorption cross section measurements of oxygen in the wavelength region 195-241 nm of the Herzberg continuum

The continuous absorption cross section of oxygen in the region 205-241 nm is studied as a function of path length and oxygen pressure. The technique used to study the continuous absorption cross section is described. Cross section measurements of oxygen in the wavelength region 193-205 nm obtained by Cheung et al. (1984) are applied in this experiment. The measured cross section is analyzed in terms of a Herzberg continuum and a pressure-dependent continuum. The total measured continuum cross section, the cross section involving two molecules of O2, and the Herzberg continuum absorption cross section values are calculated. It is observed that the Herzberg continuum cross section of oxygen values measured at 1 nm intervals in the region 195-241 nm, increase from 6.3 x 10 to the -24th sq cm at 195 nm to a maximum of 6.6 x 10 to the -24th sq cm at 201 nm and then decrease to 0.85 x 10 to the -24th sq cm at 241 nm. The Herzberg values are compared with data from previous investigations and the values correlate well.

Cheung, A. S.-C.↗
NASA NTRS1989

Flexural fatigue life prediction of closed hat-section using materially nonlinear axial fatigue characteristics

Straight or curved hat-section members are often used as structural stiffeners in aircraft. For instance, they are employed as stiffeners for the dorsal skin as well as in the aerial refueling adjacent area structure in F-106 aircraft. The flanges of the hat-section are connected to the aircraft skin. Thus, the portion of the skin closing the hat-section interacts with the section itself when resisting the stresses due to service loads. The flexural fatigue life of such a closed section is estimated using materially nonlinear axial fatigue characteristics. It should be recognized that when a structural shape is subjected to bending, the fatigue life at the neutral axis is infinity since the normal stresses are zero at that location. Conversely, the fatigue life at the extreme fibers where the normal bending stresses are maximum can be expected to be finite. Thus, different fatigue life estimates can be visualized at various distances from the neural axis. The problem becomes compounded further when significant portions away from the neutral axis are stressed into plastic range. A theoretical analysis of the closed hat-section subjected to flexural cyclic loading is first conducted. The axial fatigue characteristics together with the related axial fatigue life formula and its inverted form given by Manson and Muralidharan are adopted for an aluminum alloy used in aircraft construction. A closed-form expression for predicting the flexural fatigue life is then derived for the closed hat-section including materially nonlinear action. A computer program is written to conduct a study of the variables such as the thicknesses of the hat-section and the skin, and the type of alloy used. The study has provided a fundamental understanding of the flexural fatigue life characteristics of a practical structural component used in aircraft when materially nonlinear action is present.

Razzaq, Zia↗
NASA NTRS1990

Testing of 3-D models in 2-D adaptive wall test sections

Considerable effort was devoted to the development of two-dimensional adaptive wall test sections. The development of fully three-dimensional test sections, however, has not had this same level of effort expended. The limitations on three-dimensional research are the result of the complexity of both the mechanical design and operational aspects. Typical of the mechanical complexities are lack of readily available model and flow visualization access, and problems of sealing between adjacent walls for multi-wall types of test sections. The operational aspects are complicated by the necessity for many measurements of the boundary conditions which can be very time consuming. In three-dimensional testing, it is impossible to remove all of the boundary interference. It will be necessary, therefore, to apply residual corrections to the data. The more complex the test section shape, the more complex the calculation of the residual corrections becomes. From both the viewpoint of test section complexity and of ease of residual corrections, a simplified test section geometry is desirable. To aid in the design of simplified test sections, several researchers investigated the capability of using two-dimensional adaptive wall test sections to perform three-dimensional testing. By use of this type of design, the inherent complications and limitations of the fully three-dimensional design are avoided. Strategies of adaptation used for this application are described, as well as, sample results and a discussion of some limitations of the technique.

Wedemeyer, Erich H.↗