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At least 1,747 records · Page 97

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DOE OSTI2018

Importance of treating correlations in the uncertainty quantification of radiation damage metrics - 25461

The radiation effects community embraces the importance of quantifying uncertainty in model predictions and the importance of propagating this uncertainty into the integral metrics used to validate models, but they are not always aware of the importance of addressing the energy- and reaction-dependent correlations in the underlying uncertainty contributors. This paper presents a rigorous high-fidelity Total Monte Carlo approach that addresses the correlation in the underlying uncertainty components and quantifies the role of both energy and reaction-dependent correlations in a sample application that addresses the damage metrics relevant to silicon semiconductors. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2018

Update of the nuclear criticality slide rule calculations: studies with common shielding materials - 25467

This paper presents preliminary results from the third phase of the current update to the nuclear criticality slide rule, in which MCNP, MAVRIC/Monaco, and COG have simulated the prompt neutron and photon dose from critical spheres of 4.95% enriched uranyl fluoride solution and 93.2% enriched uranium metal. New to this phase is the inclusion of various thicknesses of lead, steel, concrete, and water shielding. This phase also evaluated the effects of humidity on the unshielded configurations in the first phase, as well as the effects of changing the ground composition used in the first phase from concrete to dry soil. The small sample of results presented show that MCNP, MAVRIC/Monaco, and COG results are statistically the same or different by only a few percent. The impact of the shielding materials depends on their thickness, but doses decreased by nearly a factor of 100 are presented. In a few instances, buildup of gamma doses occurs due to neutron interactions in shielding materials. (authors)

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DOE OSTI2018

Updating NJOY's HEATR module for ENDF/B-VIII.0 - 25493

With ENDF/B-VIII.0, new formats were added for allowing tabulated fission energy release components in addition to the existing thermal point and polynomial representations. Within ENDF/B-VIII.0, this new format was adopted for U{sup 235}, U{sup 238} and Pu{sup 239}. The fission energy release is a major component in nuclear heating for fissionable nuclides. Within NJOY, all heating and damage related calculations are done in the HEATR module. The introduction of the tabulated fission energy release components requires changes in HEATR so that these new data can be used. At the same time, the calculation of heating from fission using the other available formats was reviewed so that fission heating can now be calculated in a consistent way if the same data is used, independent of the format. For thermal point evaluations, this leads to differences between the old and new implementation of the order of 0.1 % up to incident energies of 100 keV, which increases gradually up to a value of the order of 1 % above 10 MeV (it should be noted that thermal point evaluations are no longer used in ENDF/BVIII.0). Polynomial evaluations still yield the same results with respect to older versions of NJOY. (author)

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2018

Three-dimensional analysis using polygons in particle and heavy ion transport code system (PHITS) - 25504

Recently a function to treat tetrahedral-mesh geometry was introduced in Particle and Heavy Ion Transport code System (PHITS). Owing to this function, three-dimensional analysis using polygons became possible. The work flow of the three-dimensional analysis using a particular polygon object was demonstrated, converting the data into tetrahedral mesh with the tetrahedral-mesh generator TetGen, computing the radiation transport with PHITS, and finally analyzing with three-dimensional data analysis application ParaView. (author)

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DOE OSTI2018

Electron beams for radiation belt remediation - 25527

One consequence of a high altitude nuclear explosion (HANE) is the creation of an artificial radiation belt that can have severe consequences for space-based infrastructure in low Earth orbit (LEO). A HANE-generated radiation belt is produced when radioactive debris ions beta decay to produce an intense population of magnetically-trapped, high-energy electrons that can affect satellites through increased total dose rates and/or internal charging. Recent results from NASA's Van Allen Probes mission has shown how electromagnetic waves can scatter radiation belt electrons decreasing their intensity and effects. Radiation Belt Remediation (RBR) seeks to exploit those processes to rapidly deplete the intensity of either natural or HANE-produced radiation belts by artificially injecting waves to enhance electron scattering and loss rates. One promising option for producing the required waves is to use modulated medium energy electron beams as a type of 'wireless transmitter'. (author)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2018

A Monte Carlo importance-splitting analytic benchmark - 25539

This paper describes a procedure to create analytic benchmarks for Monte Carlo variance reduction techniques that can be used to predict and/or verify the technique's mean and variance. The process uses the solutions of the first and second history-score moment equations (HSMEs) to obtain the analytic solution for the Monte Carlo mean and variance. For this work, a simple demonstration of the procedure examines the behavior of uncollided particles subject to importance splitting. In addition, a forward approach is given for calculating the same parameters. Finally, the analytic solutions are compared to both Monte Carlo calculations that directly compute the mean and variance and deterministic calculations that solve the HSMEs. All analytic and numeric results agree within the associated Monte Carlo uncertainties. (authors)

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DOE OSTI2018

Performance assessment of alternative nested DXTRAN treatments - 25540

The current approach to creating DXTRAN particles with multiple variously nested DXTRAN regions is to create DXTRAN particles on all possible regions following an initiating event. However, the efficiency of this approach, from a figure of merit standpoint, versus alternative approaches has never been demonstrated. As such, this paper examines combinations of three alternative methods to project DXTRAN particles to variously nested DXTRAN regions. First, those methods are verified to be fair and implemented correctly with a simple test geometry. Next, the resulting figure of merits are compared with the current approach using both the simple test geometry and a more-realistic air transport problem. Because of the minimal efficiency improvements observed in this work, the opportunity for efficiency degradation, and the relative ease of understanding (and implementing) the current method, these authors recommend continuing with the current method of projecting to all possible DXTRAN region surfaces. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2018

Uncertainties in secondary angular distributions and their impact on shielding, criticality and kinetics analysis - 25556

The availability of the covariance data improved over the last decades. Use of covariance matrices of different origin (such as JEFF, JENDL, ENDF, TENDL, SCALE, etc.) and types (files MF31, MF33, MF34, MF35) in sensitivity and uncertainty analysis of vast list of cases ranging from fission to fusion and from criticality, kinetics and shielding to adjustment applications was already studied in the past. The present work focuses on the status and the availability of the covariance of the angular distributions of scattered neutrons (secondary angular distributions - SAD covariances). SAD uncertainties are usually neglected in classical uncertainty propagation codes, even though the methods and codes for calculating the sensitivities are available for decades and excellent agreement was demonstrated between different deterministic and Monte Carlo codes. Of particular concern is still the availability, quality and completeness of covariance matrices. Indeed, in most evaluations the SAD covariances are limited to P1 term, if available at all. Furthermore, some differences were observed between different evaluations, e.g. in the case of iron. (author)

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DOE OSTI2018

Implementation of Krylov acceleration method for 3D S{sub N} transport code AETIUS with unstructured tetrahedral mesh - 25557

We have been developing an in-house deterministic neutron and gamma transport solver, AETIUS (An Easy modeling Transport code using Unstructured tetrahedral mesh, Shared memory parallel). To achieve fast results, BiCGStab (one of the Krylov acceleration methods) was implemented as a within-group iteration solver. In this paper, two test problems (simple double cube problem and helical tube in the outer cube problem) were used to verify the acceleration performance. The number of iterations and the calculation time were demonstrated and compared with the source iteration. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2018

Electron transport algorithms in the Integrated Tiger Series (ITS) codes - 25576

We describe the three electron-transport algorithms that have been implemented in the ITS Monte Carlo codes. While the underlying cross-section data is similar, each uses a fundamentally unique method, which at a high level are best characterized as condensed history, multigroup, and single scatter. Through a set of comparisons with experimental data and some comparisons of purely numerical results, we discuss various attributes of each of the algorithms and show some of the defects that can affect results. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2018

ENDF/B-VIII.0: the 8. major release of the ENDF/B library containing CIELO-project evaluations, new standards and thermal scattering data - 25580

The ENDF/B-VIII.0 evaluated nuclear reaction data library was released on February 2, 2018. ENDF/B-VIII.0 incorporates many improvements including the new IAEA standards, a nearly completely new thermal neutron scattering sublibrary and the CIELO project evaluations for neutron reactions on {sup 16}O, {sup 56}Fe, {sup 235}U, {sup 238}U and {sup 239}Pu. These evaluations are a direct result of improvements in theory and simulation and benefit from recent experimental data obtained in the U.S. and Europe. Notable other advances include updated evaluated data for light nuclei, structural materials, actinides and charged-particle reactions, and specific updates to fission energy release, prompt fission neutron and γ-ray spectra. ENDF/B-VIII.0 has undergone substantial testing including a wide range of criticality, reaction rate, and shielding benchmarks. Our integral validation demonstrates that performance of the library is a noticeable improvement over the previous ENDF/B-VII.1 library. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2018

Chi-Nu measurements of the prompt fission neutron spectrum of {sup 235}U and {sup 239}Pu - 25581

Knowledge of the prompt fission neutron spectrum is crucial for understanding the behavior of energetic nuclear systems. Although this energy spectrum has been measured for the major actinides {sup 235}U and {sup 239}Pu for decades, the majority of the available literature data lack the coverage of both incident and outgoing neutron energies and the precision required to accurately guide evaluations and fission models. The Chi-Nu experiment is designed to measure the prompt fission neutron spectrum for a wide range of incident and outgoing neutron energies with attention to uncertainty estimation and assessment of systematic errors in order to provide reliable guidance for evaluations and models of nuclear fission. Preliminary results of the prompt fission neutron spectrum from the neutron-induced fission of {sup 235}U and {sup 239}Pu are presented here, with a focus on the presence of spectral features near multi-chance fission thresholds, pre-equilibrium neutrons in the deduced spectrum, and also on trends of the mean fission neutron energy as a function of the incident neutron energy. While the mean energies of from ENDF/B-VIII.0 are in good agreement with those reported for {sup 235}U spectra, significant disagreements exist above an incident neutron energy of approximately 13 MeV for {sup 239}Pu. Furthermore, in general, the physics features regarding multi-chance fission and pre-equilibrium neutron emission newly-introduced into ENDF/B-VIII.0 appear provide an improved description of the data as compared to ENDF/B-VII.1. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2018

Generating shield response functions using a source independent method - 25587

Generalized questions about shield materials are common and yet difficult to answer analytically. Most forms of quick analysis use simple attenuation. The drawback to such methods is the use of mono-energetic averaged cross sections. Averaging a cross section requires knowledge of the flux a priori. If the flux was known, then some more in-depth calculation was performed negating the need for quick scoping calculations. Source independent methods for assessing the transport medium provide a means of creating response functions that yield general statements about the effectiveness of a shielding material. The following paper uses MCNP 6.2 to generate source independent, energy dependent, response functions for the transport medium that allow for quick scoping of effectiveness of various neutron shields. The source independent method here leveraged FT SCX tally treatment, to assess the performance of various shields such that energy response is preserved. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2018

Applying a predictive modeling methodology to a radiation transport inverse problem - 25588

In this paper, the predictive modeling of coupled multiphysics systems (PM-CMPS) procedure, developed by Cacuci and embodied in the software package MULTIPRED, has been applied to determine unknown parameters of a radioactive system from its gamma-ray and neutron signature. PM-CMPS has been integrated in two different ways with the inverse transport code INVERSE to determine the parameters. The first uses PM-CMPS in the inverse mode to directly estimate the parameters while the second mode uses a traditional nonlinear optimization routine and has PMCMPS complete a final iteration to estimate the uncertainty in the parameter best estimates. A test problem has been created and used to test both methods for varying numbers and types of responses and parameters. The first method was capable of accurately determining parameters but had a number of drawbacks. The second method functioned well at determining the correct parameter values in cases with limited numbers of parameters. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2018

Testing of ENDF/B-VIII.0 in the GNDS format with LLNL transport codes - 25606

ENDF/B-VIII.0 was released in the Evaluated Nuclear Data Format and the Generalized Nuclear Data Structure format (GNDS), a new international standard for storing nuclear data developed under the auspices of the Nuclear Energy Agency Working Party on Evaluation Cooperation Subgroup 38. GNDS relies on modern coding practices, and offers a structured, flexible and physics-based standard for storing and exchanging nuclear data to address limitations in legacy nuclear data formats such as ENDF-6. GNDS supports the storage of evaluated nuclear data as well as processed data suitable for both deterministic (i.e., multi-group) and Monte Carlo transport codes. Lawrence Livermore National Laboratory has implemented a complete suite of codes to test GNDS data. Simulations for the ENDF/B-VIII.0 cross-section library in GNDS format are presented. Deterministic and Monte Carlo radiation transport simulations with the Ardra and Mercury codes respectively, are validated against reactivities in fast criticality safety benchmark assemblies. Finally, Mercury simulations of {sup 238}U(n,f), {sup 237}Np(n,f) and {sup 239}Pu(n,f) reaction rates in Godiva, Jezebel and Flattop are compared to MCNP results and measured data. (authors)

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DOE OSTI2018

Application of polynomial chaos expansion in inverse transport problems with neutron multiplication measurements and multiple unknowns - 25623

The polynomial chaos expansion technique is used to build surrogate models of the dependences of gamma-ray fluxes and neutron multiplication to unknown physical parameters in radiological source/shield systems. These surrogate models are used with the Differential Evolution Adaptive Metropolis (DREAM), a method to solve and quantify uncertainty in inverse transport problems. Measured data in the inverse problems includes both passive gamma rays and neutron multiplication. The polynomial chaos expansion approach is shown to increase the speed of DREAM by factors of greater than 60 while not degrading the accuracy of the solution. (authors)

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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 OSTI2018

Preliminary performance evaluation of P++ single event proton scattering algorithms on GPUS - 25688

A need has been realized for fast, accurate charged particle transport in a wide variety of domains. Utilizing hardware acceleration through GPUs now common in user desktops, speedups in simulation comparable to those found with HPC systems can be realized for a fraction of the price. Preliminary results on the P++ proton transport code developed by the Rensselaer Radiation Measurement and Dosimetry Group (RRMDG) are presented. The code has been validated for the 1 GeV to 1 MeV energy regime and is targeted to benefit application areas such as proton radiography, space radiation research, and clinical accelerator therapy. The code is notable for employing a single event, 'on the fly' cross section based transport algorithm rather than the traditional condensed history or step based method seen in many production Monte Carlo codes. Performance improvements for the P++ code have been introduced through MPI parallelization as well as a CUDA version of the transport algorithm capable of running on NVidia GPUs. (authors)

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