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Model uncertainty in accelerator application simulations

Monte-Carlo nuclear reaction and transport codes are widely used to devise accelerator-based nuclear physics experiments; at the same time, many experiments are performed to validate the Monte-Carlo codes, which can be used for the design of full-scale nuclear power applications or the design of new benchmark experiments. Dedicated model benchmark studies investigate a broad range of nuclear reactions and quantities. Examples of these include isotope formation or secondary particle fluxes that result from the interactions of GeV-range hadrons with monoisotopic targets, which can be used to assess the respective systematic uncertainty of models. Such benchmark studies, as well as many nuclear application experiments and simulations carried out by various groups over the last few decades, enable us to draw methodological lessons. In this work, model uncertainty determined based on available experimental data allow us to identify the effects of practitioner expertise as well as the design of codes (user access to micro-scale parameters) on the range of uncertainties. We found that in cases when simulations are performed by code developers or users that are very experienced in performing simulations, the model to experiment quantity ratios generally agree with the limits determined by dedicated benchmark studies. In other cases, the ratios generally tend to be either smaller (underestimation of model error) or larger (overestimation of model error). A plausible explanation of the aforementioned effects is suggested.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

Experimental study for the feasibility of using hard x-rays for micro-XRF analysis of multilayered metals

Application of polycapillary optical systems to improve a spatial resolution for the μ-XRF analysis by focusing a primary x-ray beam and/or by collecting fluorescence emission is well known. The challenge is to optimize them in combination with x-ray source for exciting K-lines above 20 keV that could allow characterization of many materials composed by heavy elements. To pursue this goal, preliminary studies on possible polycapillary lens employment in thickness determination for multilayer metal materials will be presented in this work. In this paper, the results of first attempts of integrating PyMCA with Monte Carlo simulation code (XMI-MSIM) that takes into account the secondary fluorescence effects on quantitative analysis of homogeneous matrices, in particular, metal alloys, are presented.

36 MATERIALS SCIENCE↗

Using Tally Values to Add a Third Dimension when Viewing Planar Mesh Tallies - Paper 95

Mesh tallies show a calculated quantity, such as radiation flux or dose rate, in cells of a rectangular, cylindrical, or spherical mesh. Mesh tallies are useful when the spatial distribution of the quantity is of interest. They are usually displayed in two dimensional (2D) plots with the tally values represented by color coding. Geometry outlines and/or contour lines may be included. Three dimensional (3D) views of rectangular tallies often show a set of tally planes within the geometry. We describe here a new 3D mesh tally plot where the third dimension of a rectangular mesh tally plane is determined by the tally value in each mesh cell. We refer to such plots as elevated mesh tally plots. These plots may aid in analysis and presentation of the results of a transport calculation. The results shown here were calculated with the Monte Carlo transport code MCNPX2.7. The plots were made with the White Rock Science visualization and geometry editing program Moritz. The elevated mesh tally plots have been implemented within Moritz. The elevation method is, at present, only applicable to rectangular mesh tallies. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

Comparison of Transport Codes, HZETRN, HETC and FLUKA, Using 1977 GCR Solar Minimum Spectra

The HZETRN deterministic radiation transport code is one of several tools developed to analyze the effects of harmful galactic cosmic rays (GCR) and solar particle events (SPE) on mission planning, astronaut shielding and instrumentation. This paper is a comparison study involving the two Monte Carlo transport codes, HETC-HEDS and FLUKA, and the deterministic transport code, HZETRN. Each code is used to transport ions from the 1977 solar minimum GCR spectrum impinging upon a 20 g/cm2 Aluminum slab followed by a 30 g/cm2 water slab. This research is part of a systematic effort of verification and validation to quantify the accuracy of HZETRN and determine areas where it can be improved. Comparisons of dose and dose equivalent values at various depths in the water slab are presented in this report. This is followed by a comparison of the proton fluxes, and the forward, backward and total neutron fluxes at various depths in the water slab. Comparisons of the secondary light ion 2H, 3H, 3He and 4He fluxes are also examined.

Heinbockel, John H.↗

Enhancements in Continuous-Energy Monte Carlo Capabilities in SCALE

Monte Carlo tools in SCALE are commonly used in criticality safety calculations as well as sensitivity and uncertainty analysis, depletion, and criticality alarm system analyses. Recent improvements in the continuous-energy data generated by the AMPX code system and significant advancements in the continuous-energy treatment in the KENO Monte Carlo eigenvalue codes facilitate the use of SCALE Monte Carlo codes to model geometrically complex systems with enhanced solution fidelity. The addition of continuous-energy treatment to the SCALE Monaco code, which can be used with automatic variance reduction in the hybrid MAVRIC sequence, provides significant enhancements, especially for criticality alarm system modeling. This paper describes some of the advancements in continuous-energy Monte Carlo codes within the SCALE code system.

97 MATHEMATICS AND COMPUTING↗

The status of MCU-5

The software developed within the framework of the MCU project which implements the Monte Carlo method for solving the transport equations for various kinds of radiation is widely used for calculating nuclear reactors of different types. The final version of MCU-5 is described. The developed libraries of constants are briefly described, and the potentialities of the MCU-5 package modules and the executable codes compiled from them are characterized. Examples of important problems of reactor physics solved with the codes from the MCU family are presented.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Improving the efficiency of configurational-bias Monte Carlo: A density-guided method for generating bending angle trials for linear and branched molecules

A new method has been developed to generate bending angle trials to improve the acceptance rate and the speed of configurational-bias Monte Carlo. Whereas traditionally the trial geometries are generated from a uniform distribution, in this method we attempt to use the exact probability density function so that each geometry generated is likely to be accepted. In actual practice, due to the complexity of this probability density function, a numerical representation of this distribution function would be required. This numerical table can be generated a priori from the distribution function. This method has been tested on a united-atom model of alkanes including propane, 2-methylpropane, and 2,2-dimethylpropane, that are good representatives of both linear and branched molecules. It has been shown from these test cases that reasonable approximations can be made especially for the highly branched molecules to reduce drastically the dimensionality and correspondingly the amount of the tabulated data that is needed to be stored. Despite these approximations, the dependencies between the various geometrical variables can be still well considered, as evident from a nearly perfect acceptance rate achieved. For all cases, the bending angles were shown to be sampled correctly by this method with an acceptance rate of at least 96% for 2,2-dimethylpropane to more than 99% for propane. Since only one trial is required to be generated for each bending angle (instead of thousands of trials required by the conventional algorithm), this method can dramatically reduce the simulation time. The profiling results of our Monte Carlo simulation code show that trial generation, which used to be the most time consuming process, is no longer the time dominating component of the simulation.

2-2-DIMETHYLPROPANE↗

Analysis of Crystal Lattice Deformation by Ion Channeling

A model of dislocations has been developed for the use in Monte Carlo simulations of ion channeling spectra obtained for defected crystals. High resolution transmission electron microscopy micrographs show that the dominant type of defects in the majority of ion irradiated crystals are dislocations. The RBS/channeling spectrum is then composed of two components: one is due to direct scattering on randomly displaced atoms and the second one is related to beam defocussing on dislocations, which produce predominantly crystal lattice distortions, i.e. bent channels. In order to provide a correct analysis of backscattering spectra for the crystals containing dislocations we have modified the existing Monte Carlo simulation code "McChasy". A new version of the code has been developed by implementing dislocations on the basis of the Peierls-Nabarro model. Parameters of the model have been determined from the high resolution transmission electron microscopy data. The newly developed method has been used to study the Ar-ion bombarded SrTiO3 samples. The best fit to the Rutherford backscattering/channeling spectra has been obtained by optimizing the linear combination of two kinds of defects: displaced atoms and bent channels. The great virtue of the Monte Carlo simulation is that unlike a traditional dechanneling analysis it allows quantitative analysis of crystals containing a mixture of different types of defects.

Environmental Molecular Sciences Laboratory↗

Determining the effects of microsphere and surrounding material composition on {sup 90}Y dose kernels using egsnrc and mcnp5

Purpose: Recent advances in the imaging of {sup 90}Y using positron emission tomography (PET) and improved uncertainty in the branching ratio for the internal pair production component of {sup 90}Y decay allow for a more accurate determination of the activity distribution of {sup 90}Y microspheres within a patient. This improved activity distribution can be convolved with the dose kernel of {sup 90}Y to calculate the dose distribution within a patient. This work investigates the effects of microsphere and surrounding material composition on {sup 90}Y dose kernels using egsnrc and mcnp5 and compares the results of these two transport codes. Methods: Monte Carlo simulations were performed with egsnrc and mcnp5 to calculate the dose rate at multiple radial distances around various {sup 90}Y sources. Point source simulations were completed with mcnp5 to determine the optimal electron transport settings for this work. After determining the optimal settings, point source simulations were completed using egsnrc (user code edknrc) and mcnp5 in water and liver [as defined by the International Commission on Radiation Units and Measurements (ICRU) Report 44]. The results were compared to ICRU Report 72 reference data. Point source simulations were also completed in water with a density of 1.06 g{center_dot}cm{sup -3} to evaluate the effect of the density of the surrounding material. Glass and resin microsphere simulations were performed with average and maximum diameter and density values (based on values given in the literature) in water and in liver. The results were compared to point source simulation results using the same transport code and in the same surrounding material. All simulations had statistical uncertainties less than 1%. Results: The optimal transport settings in mcnp5 for this work included using the energy-and step-specific algorithm (DBCN 17J 2) and ESTEP set to 10. These settings were used for all subsequent simulations with mcnp5. The point source simulations in water for both egsnrc and mcnp5 were found to agree within 2% of the ICRU 72 reference data over the investigated range. Point source simulations in liver had large differences relative to ICRU 72, approaching -60% near the maximum range of {sup 90}Y. These differences are mostly attributed to the difference in density between water (1.0 g{center_dot}cm{sup -3}) and liver (1.06 g{center_dot}cm{sup -3}). Glass and resin microsphere simulations showed a slight decrease in the dose rate near the maximum range of {sup 90}Y relative to the point source simulations. The largest relative differences were approximately -4.2% and -2.8% for the glass and resin microspheres, respectively. Agreement between the egsnrc and mcnp5 simulations results was generally good. Conclusions: The presence of the microsphere material causes slight differences in the {sup 90}Y dose kernel compared to those calculated with point sources. Large differences were seen between simulations in water and those in liver. For the most accurate calculation of the dose distribution, the density of the patient's liver should be accounted for in the calculation of the dose kernel. Lastly, due to the need to determine the optimal transport settings with mcnp5, electron transport with this code should be used with caution.

61 RADIATION PROTECTION AND DOSIMETRY↗

Atomistic Method Applied to Computational Modeling of Surface Alloys

The formation of surface alloys is a growing research field that, in terms of the surface structure of multicomponent systems, defines the frontier both for experimental and theoretical techniques. Because of the impact that the formation of surface alloys has on surface properties, researchers need reliable methods to predict new surface alloys and to help interpret unknown structures. The structure of surface alloys and when, and even if, they form are largely unpredictable from the known properties of the participating elements. No unified theory or model to date can infer surface alloy structures from the constituents properties or their bulk alloy characteristics. In spite of these severe limitations, a growing catalogue of such systems has been developed during the last decade, and only recently are global theories being advanced to fully understand the phenomenon. None of the methods used in other areas of surface science can properly model even the already known cases. Aware of these limitations, the Computational Materials Group at the NASA Glenn Research Center at Lewis Field has developed a useful, computationally economical, and physically sound methodology to enable the systematic study of surface alloy formation in metals. This tool has been tested successfully on several known systems for which hard experimental evidence exists and has been used to predict ternary surface alloy formation (results to be published: Garces, J.E.; Bozzolo, G.; and Mosca, H.: Atomistic Modeling of Pd/Cu(100) Surface Alloy Formation. Surf. Sci., 2000 (in press); Mosca, H.; Garces J.E.; and Bozzolo, G.: Surface Ternary Alloys of (Cu,Au)/Ni(110). (Accepted for publication in Surf. Sci., 2000.); and Garces, J.E.; Bozzolo, G.; Mosca, H.; and Abel, P.: A New Approach for Atomistic Modeling of Pd/Cu(110) Surface Alloy Formation. (Submitted to Appl. Surf. Sci.)). Ternary alloy formation is a field yet to be fully explored experimentally. The computational tool, which is based on the BFS (Bozzolo, Ferrante, and Smith) method for the calculation of the energetics, consists of a small number of simple PCbased computer codes that deal with the different aspects of surface alloy formation. Two analysis modes are available within this package. The first mode provides an atom-by-atom description of real and virtual stages 1. during the process of surface alloying, based on the construction of catalogues of configurations where each configuration describes one possible atomic distribution. BFS analysis of this catalogue provides information on accessible states, possible ordering patterns, and details of island formation or film growth. More importantly, it provides insight into the evolution of the system. Software developed by the Computational Materials Group allows for the study of an arbitrary number of elements forming surface alloys, including an arbitrary number of surface atomic layers. The second mode involves large-scale temperature-dependent computer 2. simulations that use the BFS method for the energetics and provide information on the dynamic processes during surface alloying. These simulations require the implementation of Monte-Carlo-based codes with high efficiency within current workstation environments. This methodology capitalizes on the advantages of the BFS method: there are no restrictions on the number or type of elements or on the type of crystallographic structure considered. This removes any restrictions in the definition of the configuration catalogues used in the analytical calculations, thus allowing for the study of arbitrary ordering patterns, ultimately leading to the actual surface alloy structure. Moreover, the Monte Carlo numerical technique used for the large-scale simulations allows for a detailed visualization of the simulated process, the main advantage of this type of analysis being the ability to understand the underlying features that drive these processes. Because of the simplicity of the BFS method for e energetics used in these calculations, a detailed atom-by-atom analysis can be performed at any point in the simulation, providing necessary insight on the details of the process. The main objective of this research program is to develop a tool to guide experimenters in understanding and interpreting often unexpected results in alloy formation experiments. By reducing the computational effort without losing physical accuracy, we expect that powerful simulation tools will be developed in the immediate future, which will allow material scientists to easily visualize and analyze processes at a level not achievable experimentally.

Bozzolo, Guillermo H.↗

Kinetics calculation of fast periodic pulsed reactors using MCNP6

Fast periodic pulsed reactor is a type of reactor in which the fission bursts are formed entirely with external reactivity modulation with a specified time periodicity. This type of reactors could generate much larger intensity of neutron beams for experimental use, compared with the steady state reactors. In the design of fast periodic pulsed reactors, the time dependent simulation of the power pulse is majorly based on a point kinetic model, which is known to have limitations. A more accurate calculation method is desired for the design analyses of fast periodic pulsed reactors. Monte Carlo computer code MCNP6 is used for this task due to its three dimensional transport capability with a continuous energy library. Some new routines were added to simulate the rotation of the movable reflector parts in the time dependent calculation. Fast periodic pulsed reactor IBR-2M was utilized to validate the new routines. This reactor is periodically in prompt supercritical state, which lasts for similar to 400 mu s, during the equilibrium state. This generates long neutron fission chains, which requires tremendously large amount of computation time during Monte Carlo simulations. Russian Roulette was applied for these very long neutron chains in MCNP6 calculation, combined with other approaches to improve the efficiency of the simulations. In the power pulse of the IBR-2M at equilibrium state, there is some discrepancy between the experimental measurements and the calculated results using the point kinetics model. MCNP6 results matches better the experimental measurements, which shows the merit of using MCNP6 calculation relative to the point kinetics model.

Zhong, Z.↗

Test of a prototype neutron spectrometer based on diamond detectors in a fast reactor

A prototype of neutron spectrometer based on diamond detectors has been developed. This prototype consists of a {sup 6}Li neutron converter sandwiched between two CVD diamond crystals. The radiation hardness of the diamond crystals makes it suitable for applications in low power research reactors, while a low sensitivity to gamma rays and low leakage current of the detector permit to reach good energy resolution. A fast coincidence between two crystals is used to reject background. The detector was read out using two different electronic chains connected to it by a few meters of cable. The first chain was based on conventional charge-sensitive amplifiers, the other used a custom fast charge amplifier developed for this purpose. The prototype has been tested at various neutron sources and showed its practicability. In particular, the detector was calibrated in a TRIGA thermal reactor (LENA laboratory, University of Pavia) with neutron fluxes of 10{sup 8} n/cm{sup 2}s and at the 3 MeV D-D monochromatic neutron source named FNG (ENEA, Rome) with neutron fluxes of 10{sup 6} n/cm{sup 2}s. The neutron spectrum measurement was performed at the TAPIRO fast research reactor (ENEA, Casaccia) with fluxes of 10{sup 9} n/cm{sup 2}s. The obtained spectra were compared to Monte Carlo simulations, modeling detector response with MCNP and Geant4. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Photon-induced positron annihilation lifetime spectroscopy using ultrashort laser-Compton-scattered gamma-ray pulses

High-energy ultrashort gamma-ray pulses can be generated via laser Compton scattering with 90 Degree-Sign collisions at the UVSOR-II electron storage ring. As an applied study of ultrashort gamma-ray pulses, a new photon-induced positron annihilation lifetime spectroscopy approach has been developed. Ultrashort gamma-ray pulses with a maximum energy of 6.6 MeV and pulse width of 2.2 ps created positrons throughout bulk lead via pair production. Annihilation gamma rays were detected by a BaF{sub 2} scintillator mounted on a photomultiplier tube. A positron lifetime spectrum was obtained by measuring the time difference between the RF frequency of the electron storage ring and the detection time of the annihilation gamma rays. We calculated the response of the BaF{sub 2} scintillator and the time jitter caused by the variation in the total path length of the ultrashort gamma-ray pulses, annihilation gamma rays, and scintillation light using a Monte Carlo simulation code. The positron lifetime for bulk lead was successfully measured.

36 MATERIALS SCIENCE↗

Validation of MCNP5 for use in Calculating Temperature Coefficients of Reactivity for the SHINE System

Neutron transport calculations for the SHINE system are being done with MCNP, a general-purpose Monte Carlo NParticle code developed at Los Alamos National Laboratory (LANL). Current calculations are being done with MCNP5, version 1.6. This paper describes the methods used to calculate a tolerance band for temperature coefficients of reactivity (α{sub T}) calculated with MCNP5, v1.6 for the SHINE system. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Spatial and spectral effects in subcritical system pulsed experiments

Accurate neutronic models are needed for the interpretation of pulsed experiments in subcritical systems. In this work, the extent of spatial and spectral effects in the pulse propagation phenomena is investigated and the analysis is applied to the GUINEVERE experiment. The multigroup cross section data is generated by the Monte Carlo SERPENT code and the neutronic evolution following the source pulse is simulated by a kinetic diffusion code. The results presented show that important spatial and spectral aspects need to be properly accounted for and that a detailed energy approach may be needed to adequately capture the physical features of the system to the pulse injection. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Comparisons of neutrino event generators from an oscillation-experiment perspective

Monte Carlo generators are crucial to the analysis of high energy physics data, ideally giving a baseline comparison between the state-of-art theoretical models and experimental data. Presented here is a comparison between three of final state distributions from the GENIE, Neut, NUANCE, and NuWro neutrino Monte Carlo event generators. The final state distributions chosen for comparison are: the electromagnetic energy fraction in neutral current interactions, the energy of the leading π{sup 0} vs. the scattering angle for neutral current interactions, and the muon energy vs. scattering angle of ν{sub µ} charged current interactions.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

Study on in situ calibration for neutron flux monitor in the Large Helical Device based on Monte Carlo calculations

Neutron monitoring is important to manage safety of fusion experiment facilities because neutrons are generated in fusion reactions. Monte Carlo simulations play an important role in evaluating the influence of neutron scattering from various structures and correcting differences between deuterium plasma experiments and in situ calibration experiments. We evaluated these influences based on differences between the both experiments at Large Helical Device using Monte Carlo simulation code MCNP5. A difference between the both experiments in absolute detection efficiency of the fission chamber between O-ports is estimated to be the biggest of all monitors. We additionally evaluated correction coefficients for some neutron monitors.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗