The effects of nuclear data library processing on Geant4 and MCNP simulations of the thermal neutron scattering law

NASA Astrophysics Data System (ADS)

Hartling, K.; Ciungu, B.; Li, G.; Bentoumi, G.; Sur, B.

2018-05-01

Monte Carlo codes such as MCNP and Geant4 rely on a combination of physics models and evaluated nuclear data files (ENDF) to simulate the transport of neutrons through various materials and geometries. The grid representation used to represent the final-state scattering energies and angles associated with neutron scattering interactions can significantly affect the predictions of these codes. In particular, the default thermal scattering libraries used by MCNP6.1 and Geant4.10.3 do not accurately reproduce the ENDF/B-VII.1 model in simulations of the double-differential cross section for thermal neutrons interacting with hydrogen nuclei in a thin layer of water. However, agreement between model and simulation can be achieved within the statistical error by re-processing ENDF/B-VII.I thermal scattering libraries with the NJOY code. The structure of the thermal scattering libraries and sampling algorithms in MCNP and Geant4 are also reviewed.

Release Notes for Whisper-1.1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Forrest B.; Rising, Michael Evan; Alwin, Jennifer Louise

Whisper is a statistical analysis package developed in 2014 to support nuclear criticality safety (NCS) validation [1-3]. It uses the sensitivity profile data for an application as computed by MCNP6 [4-6] along with covariance files [7,8] for the nuclear data to determine a baseline upper-subcritical-limit (USL) for the application. Whisper version 1.0 was first developed and used at LANL in 2014 [3]. During 2015-2016, Whisper was updated to version 1.1 [9] and is to be included with the upcoming release of MCNP6.2. This document describes the Whisper-1.1 package that will be included with the MCNP6.2 release during 2017. Specific detailsmore » are provided on the computer systems supported, the software quality assurance (SQA) procedures, installation, and testing. This document does not address other important topics, such as the importance of sensitivity-uncertainty (SU) methods to NCS validation, the theory underlying SU methodology, tutorials on the usage of MCNP-Whisper, practical approaches to using SU methodology to support and extend traditional validation, etc. There are over 50 documents included with Whisper-1.1 and available in the MCNP Reference Collection on the MCNP website (mcnp.lanl.gov) that address all of those topics and more. In this document, however, a complete bibliography of relevant MCNP-Whisper references is provided.« less

Anisn-Dort Neutron-Gamma Flux Intercomparison Exercise for a Simple Testing Model

NASA Astrophysics Data System (ADS)

Boehmer, B.; Konheiser, J.; Borodkin, G.; Brodkin, E.; Egorov, A.; Kozhevnikov, A.; Zaritsky, S.; Manturov, G.; Voloschenko, A.

2003-06-01

The ability of transport codes ANISN, DORT, ROZ-6, MCNP and TRAMO, as well as nuclear data libraries BUGLE-96, ABBN-93, VITAMIN-B6 and ENDF/B-6 to deliver consistent gamma and neutron flux results was tested in the calculation of a one-dimensional cylindrical model consisting of a homogeneous core and an outer zone with a single material. Model variants with H2O, Fe, Cr and Ni in the outer zones were investigated. The results are compared with MCNP-ENDF/B-6 results. Discrepancies are discussed. The specified test model is proposed as a computational benchmark for testing calculation codes and data libraries.

Comparison of EGS4 and MCNP Monte Carlo codes when calculating radiotherapy depth doses.

PubMed

Love, P A; Lewis, D G; Al-Affan, I A; Smith, C W

1998-05-01

The Monte Carlo codes EGS4 and MCNP have been compared when calculating radiotherapy depth doses in water. The aims of the work were to study (i) the differences between calculated depth doses in water for a range of monoenergetic photon energies and (ii) the relative efficiency of the two codes for different electron transport energy cut-offs. The depth doses from the two codes agree with each other within the statistical uncertainties of the calculations (1-2%). The relative depth doses also agree with data tabulated in the British Journal of Radiology Supplement 25. A discrepancy in the dose build-up region may by attributed to the different electron transport algorithims used by EGS4 and MCNP. This discrepancy is considerably reduced when the improved electron transport routines are used in the latest (4B) version of MCNP. Timing calculations show that EGS4 is at least 50% faster than MCNP for the geometries used in the simulations.

Lung Dosimetry for Radioiodine Treatment Planning in the Case of Diffuse Lung Metastases

PubMed Central

Song, Hong; He, Bin; Prideaux, Andrew; Du, Yong; Frey, Eric; Kasecamp, Wayne; Ladenson, Paul W.; Wahl, Richard L.; Sgouros, George

2010-01-01

The lungs are the most frequent sites of distant metastasis in differentiated thyroid carcinoma. Radioiodine treatment planning for these patients is usually performed following the Benua– Leeper method, which constrains the administered activity to 2.96 GBq (80 mCi) whole-body retention at 48 h after administration to prevent lung toxicity in the presence of iodine-avid lung metastases. This limit was derived from clinical experience, and a dosimetric analysis of lung and tumor absorbed dose would be useful to understand the implications of this limit on toxicity and tumor control. Because of highly nonuniform lung density and composition as well as the nonuniform activity distribution when the lungs contain tumor nodules, Monte Carlo dosimetry is required to estimate tumor and normal lung absorbed dose. Reassessment of this toxicity limit is also appropriate in light of the contemporary use of recombinant thyrotropin (thyroid-stimulating hormone) (rTSH) to prepare patients for radioiodine therapy. In this work we demonstrated the use of MCNP, a Monte Carlo electron and photon transport code, in a 3-dimensional (3D) imaging–based absorbed dose calculation for tumor and normal lungs. Methods A pediatric thyroid cancer patient with diffuse lung metastases was administered 37MBq of 131I after preparation with rTSH. SPECT/CT scans were performed over the chest at 27, 74, and 147 h after tracer administration. The time–activity curve for 131I in the lungs was derived from the whole-body planar imaging and compared with that obtained from the quantitative SPECT methods. Reconstructed and coregistered SPECT/CT images were converted into 3D density and activity probability maps suitable for MCNP4b input. Absorbed dose maps were calculated using electron and photon transport in MCNP4b. Administered activity was estimated on the basis of the maximum tolerated dose (MTD) of 27.25 Gy to the normal lungs. Computational efficiency of the MCNP4b code was studied with a simple segmentation approach. In addition, the Benua–Leeper method was used to estimate the recommended administered activity. The standard dosing plan was modified to account for the weight of this pediatric patient, where the 2.96-GBq (80 mCi) whole-body retention was scaled to 2.44 GBq (66 mCi) to give the same dose rate of 43.6 rad/h in the lungs at 48 h. Results Using the MCNP4b code, both the spatial dose distribution and a dose–volume histogram were obtained for the lungs. An administered activity of 1.72 GBq (46.4 mCi) delivered the putative MTD of 27.25 Gy to the lungs with a tumor absorbed dose of 63.7 Gy. Directly applying the Benua–Leeper method, an administered activity of 3.89 GBq (105.0 mCi) was obtained, resulting in tumor and lung absorbed doses of 144.2 and 61.6 Gy, respectively, when the MCNP-based dosimetry was applied. The voxel-by-voxel calculation time of 4,642.3 h for photon transport was reduced to 16.8 h when the activity maps were segmented into 20 regions. Conclusion MCNP4b–based, patient-specific 3D dosimetry is feasible and important in the dosimetry of thyroid cancer patients with avid lung metastases that exhibit prolonged retention in the lungs. PMID:17138741

MCNP Output Data Analysis with ROOT (MODAR)

NASA Astrophysics Data System (ADS)

Carasco, C.

2010-06-01

MCNP Output Data Analysis with ROOT (MODAR) is a tool based on CERN's ROOT software. MODAR has been designed to handle time-energy data issued by MCNP simulations of neutron inspection devices using the associated particle technique. MODAR exploits ROOT's Graphical User Interface and functionalities to visualize and process MCNP simulation results in a fast and user-friendly way. MODAR allows to take into account the detection system time resolution (which is not possible with MCNP) as well as detectors energy response function and counting statistics in a straightforward way. Program summaryProgram title: MODAR Catalogue identifier: AEGA_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEGA_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 155 373 No. of bytes in distributed program, including test data, etc.: 14 815 461 Distribution format: tar.gz Programming language: C++ Computer: Most Unix workstations and PC Operating system: Most Unix systems, Linux and windows, provided the ROOT package has been installed. Examples where tested under Suse Linux and Windows XP. RAM: Depends on the size of the MCNP output file. The example presented in the article, which involves three two-dimensional 139×740 bins histograms, allocates about 60 MB. These data are running under ROOT and include consumption by ROOT itself. Classification: 17.6 External routines: ROOT version 5.24.00 ( http://root.cern.ch/drupal/) Nature of problem: The output of an MCNP simulation is an ASCII file. The data processing is usually performed by copying and pasting the relevant parts of the ASCII file into Microsoft Excel. Such an approach is satisfactory when the quantity of data is small but is not efficient when the size of the simulated data is large, for example when time-energy correlations are studied in detail such as in problems involving the associated particle technique. In addition, since the finite time resolution of the simulated detector cannot be modeled with MCNP, systems in which time-energy correlation is crucial cannot be described in a satisfactory way. Finally, realistic particle energy deposit in detectors is calculated with MCNP in a two-step process involving type-5 then type-8 tallies. In the first step, the photon flux energy spectrum associated to a time region is selected and serves as a source energy distribution for the second step. Thus, several files must be manipulated before getting the result, which can be time consuming if one needs to study several time regions or different detectors performances. In the same way, modeling counting statistics obtained in a limited acquisition time requires several steps and can also be time consuming. Solution method: In order to overcome the previous limitations, the MODAR C++ code has been written to make use of CERN's ROOT data analysis software. MCNP output data are read from the MCNP output file with dedicated routines. Two-dimensional histograms are filled and can be handled efficiently within the ROOT framework. To keep a user friendly analysis tool, all processing and data display can be done by means of ROOT Graphical User Interface. Specific routines have been written to include detectors finite time resolution and energy response function as well as counting statistics in a straightforward way. Additional comments: The possibility of adding tallies has also been incorporated in MODAR in order to describe systems in which the signal from several detectors can be summed. Moreover, MODAR can be adapted to handle other problems involving two-dimensional data. Running time: The CPU time needed to smear a two-dimensional histogram depends on the size of the histogram. In the presented example, the time-energy smearing of one of the 139×740 two-dimensional histograms takes 3 minutes with a DELL computer equipped with INTEL Core 2.

Benchmark Evaluation of HTR-PROTEUS Pebble Bed Experimental Program

DOE PAGES

Bess, John D.; Montierth, Leland; Köberl, Oliver; ...

2014-10-09

Benchmark models were developed to evaluate 11 critical core configurations of the HTR-PROTEUS pebble bed experimental program. Various additional reactor physics measurements were performed as part of this program; currently only a total of 37 absorber rod worth measurements have been evaluated as acceptable benchmark experiments for Cores 4, 9, and 10. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the ²³⁵U enrichment of the fuel, impurities in the moderator pebbles, and the density and impurity content of the radial reflector. Calculations of k eff with MCNP5 and ENDF/B-VII.0 neutron nuclear data aremore » greater than the benchmark values but within 1% and also within the 3σ uncertainty, except for Core 4, which is the only randomly packed pebble configuration. Repeated calculations of k eff with MCNP6.1 and ENDF/B-VII.1 are lower than the benchmark values and within 1% (~3σ) except for Cores 5 and 9, which calculate lower than the benchmark eigenvalues within 4σ. The primary difference between the two nuclear data libraries is the adjustment of the absorption cross section of graphite. Simulations of the absorber rod worth measurements are within 3σ of the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

MCNP output data analysis with ROOT (MODAR)

NASA Astrophysics Data System (ADS)

Carasco, C.

2010-12-01

MCNP Output Data Analysis with ROOT (MODAR) is a tool based on CERN's ROOT software. MODAR has been designed to handle time-energy data issued by MCNP simulations of neutron inspection devices using the associated particle technique. MODAR exploits ROOT's Graphical User Interface and functionalities to visualize and process MCNP simulation results in a fast and user-friendly way. MODAR allows to take into account the detection system time resolution (which is not possible with MCNP) as well as detectors energy response function and counting statistics in a straightforward way. New version program summaryProgram title: MODAR Catalogue identifier: AEGA_v1_1 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEGA_v1_1.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 150 927 No. of bytes in distributed program, including test data, etc.: 4 981 633 Distribution format: tar.gz Programming language: C++ Computer: Most Unix workstations and PCs Operating system: Most Unix systems, Linux and windows, provided the ROOT package has been installed. Examples where tested under Suse Linux and Windows XP. RAM: Depends on the size of the MCNP output file. The example presented in the article, which involves three two dimensional 139×740 bins histograms, allocates about 60 MB. These data are running under ROOT and include consumption by ROOT itself. Classification: 17.6 Catalogue identifier of previous version: AEGA_v1_0 Journal reference of previous version: Comput. Phys. Comm. 181 (2010) 1161 External routines: ROOT version 5.24.00 ( http://root.cern.ch/drupal/) Does the new version supersede the previous version?: Yes Nature of problem: The output of a MCNP simulation is an ascii file. The data processing is usually performed by copying and pasting the relevant parts of the ascii file into Microsoft Excel. Such an approach is satisfactory when the quantity of data is small but is not efficient when the size of the simulated data is large, for example when time-energy correlations are studied in detail such as in problems involving the associated particle technique. In addition, since the finite time resolution of the simulated detector cannot be modeled with MCNP, systems in which time-energy correlation is crucial cannot be described in a satisfactory way. Finally, realistic particle energy deposit in detectors is calculated with MCNP in a two step process involving type-5 then type-8 tallies. In the first step, the photon flux energy spectrum associated to a time region is selected and serves as a source energy distribution for the second step. Thus, several files must be manipulated before getting the result, which can be time consuming if one needs to study several time regions or different detectors performances. In the same way, modeling counting statistics obtained in a limited acquisition time requires several steps and can also be time consuming. Solution method: In order to overcome the previous limitations, the MODAR C++ code has been written to make use of CERN's ROOT data analysis software. MCNP output data are read from the MCNP output file with dedicated routines. Two dimensional histograms are filled and can be handled efficiently within the ROOT framework. To keep a user friendly analysis tool, all processing and data display can be done by means of ROOT Graphical User Interface. Specific routines have been written to include detectors finite time resolution and energy response function as well as counting statistics in a straightforward way. Reasons for new version: For applications involving the Associate Particle Technique, a large number of gamma rays are produced by the fast neutrons interactions. To study the energy spectra, it is useful to identify the gamma-ray energy peaks in a straightforward way. Therefore, the possibility to show gamma rays corresponding to specific reactions has been added in MODAR. Summary of revisions: It is possible to use a gamma ray database to better identify in the energy spectra gamma ray peaks with their first and second escapes. Histograms can be scaled by the number of source particle to evaluate the number of counts that is expected without statistical uncertainties. Additional comments: The possibility of adding tallies has also been incorporated in MODAR in order to describe systems in which the signal from several detectors can be summed. Moreover, MODAR can be adapted to handle other problems involving two dimensional data. Running time: The CPU time needed to smear a two dimensional histogram depends on the size of the histogram. In the presented example, the time-energy smearing of one of the 139×740 two dimensional histograms takes 3 minutes with a DELL computer equipped with INTEL Core 2.

Monte Carlo calculations of thermal neutron capture in gadolinium: a comparison of GEANT4 and MCNP with measurements.

PubMed

Enger, Shirin A; Munck af Rosenschöld, Per; Rezaei, Arash; Lundqvist, Hans

2006-02-01

GEANT4 is a Monte Carlo code originally implemented for high-energy physics applications and is well known for particle transport at high energies. The capacity of GEANT4 to simulate neutron transport in the thermal energy region is not equally well known. The aim of this article is to compare MCNP, a code commonly used in low energy neutron transport calculations and GEANT4 with experimental results and select the suitable code for gadolinium neutron capture applications. To account for the thermal neutron scattering from chemically bound atoms [S(alpha,beta)] in biological materials a comparison of thermal neutron fluence in tissue-like poly(methylmethacrylate) phantom is made with MCNP4B, GEANT4 6.0 patch1, and measurements from the neutron capture therapy (NCT) facility at the Studsvik, Sweden. The fluence measurements agreed with MCNP calculated results considering S(alpha,beta). The location of the thermal neutron peak calculated with MCNP without S(alpha,beta) and GEANT4 is shifted by about 0.5 cm towards a shallower depth and is 25%-30% lower in amplitude. Dose distribution from the gadolinium neutron capture reaction is then simulated by MCNP and compared with measured data. The simulations made by MCNP agree well with experimental results. As long as thermal neutron scattering from chemically bound atoms are not included in GEANT4 it is not suitable for NCT applications.

Benchmarking the MCNP code for Monte Carlo modelling of an in vivo neutron activation analysis system.

PubMed

Natto, S A; Lewis, D G; Ryde, S J

1998-01-01

The Monte Carlo computer code MCNP (version 4A) has been used to develop a personal computer-based model of the Swansea in vivo neutron activation analysis (IVNAA) system. The model included specification of the neutron source (252Cf), collimators, reflectors and shielding. The MCNP model was 'benchmarked' against fast neutron and thermal neutron fluence data obtained experimentally from the IVNAA system. The Swansea system allows two irradiation geometries using 'short' and 'long' collimators, which provide alternative dose rates for IVNAA. The data presented here relate to the short collimator, although results of similar accuracy were obtained using the long collimator. The fast neutron fluence was measured in air at a series of depths inside the collimator. The measurements agreed with the MCNP simulation within the statistical uncertainty (5-10%) of the calculations. The thermal neutron fluence was measured and calculated inside the cuboidal water phantom. The depth of maximum thermal fluence was 3.2 cm (measured) and 3.0 cm (calculated). The width of the 50% thermal fluence level across the phantom at its mid-depth was found to be the same by both MCNP and experiment. This benchmarking exercise has given us a high degree of confidence in MCNP as a tool for the design of IVNAA systems.

Benchmarking comparison and validation of MCNP photon interaction data

NASA Astrophysics Data System (ADS)

Colling, Bethany; Kodeli, I.; Lilley, S.; Packer, L. W.

2017-09-01

The objective of the research was to test available photoatomic data libraries for fusion relevant applications, comparing against experimental and computational neutronics benchmarks. Photon flux and heating was compared using the photon interaction data libraries (mcplib 04p, 05t, 84p and 12p). Suitable benchmark experiments (iron and water) were selected from the SINBAD database and analysed to compare experimental values with MCNP calculations using mcplib 04p, 84p and 12p. In both the computational and experimental comparisons, the majority of results with the 04p, 84p and 12p photon data libraries were within 1σ of the mean MCNP statistical uncertainty. Larger differences were observed when comparing computational results with the 05t test photon library. The Doppler broadening sampling bug in MCNP-5 is shown to be corrected for fusion relevant problems through use of the 84p photon data library. The recommended libraries for fusion neutronics are 84p (or 04p) with MCNP6 and 84p if using MCNP-5.

A comparison between EGS4 and MCNP computer modeling of an in vivo X-ray fluorescence system.

PubMed

Al-Ghorabie, F H; Natto, S S; Al-Lyhiani, S H

2001-03-01

The Monte Carlo computer codes EGS4 and MCNP were used to develop a theoretical model of a 180 degrees geometry in vivo X-ray fluorescence system for the measurement of platinum concentration in head and neck tumors. The model included specification of the photon source, collimators, phantoms and detector. Theoretical results were compared and evaluated against X-ray fluorescence data obtained experimentally from an existing system developed by the Swansea In Vivo Analysis and Cancer Research Group. The EGS4 results agreed well with the MCNP results. However, agreement between the measured spectral shape obtained using the experimental X-ray fluorescence system and the simulated spectral shape obtained using the two Monte Carlo codes was relatively poor. The main reason for the disagreement between the results arises from the basic assumptions which the two codes used in their calculations. Both codes assume a "free" electron model for Compton interactions. This assumption will underestimate the results and invalidates any predicted and experimental spectra when compared with each other.

Experimental Criticality Benchmarks for SNAP 10A/2 Reactor Cores

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krass, A.W.

2005-12-19

This report describes computational benchmark models for nuclear criticality derived from descriptions of the Systems for Nuclear Auxiliary Power (SNAP) Critical Assembly (SCA)-4B experimental criticality program conducted by Atomics International during the early 1960's. The selected experimental configurations consist of fueled SNAP 10A/2-type reactor cores subject to varied conditions of water immersion and reflection under experimental control to measure neutron multiplication. SNAP 10A/2-type reactor cores are compact volumes fueled and moderated with the hydride of highly enriched uranium-zirconium alloy. Specifications for the materials and geometry needed to describe a given experimental configuration for a model using MCNP5 are provided. Themore » material and geometry specifications are adequate to permit user development of input for alternative nuclear safety codes, such as KENO. A total of 73 distinct experimental configurations are described.« less

MCNP-model for the OAEP Thai Research Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gallmeier, F.X.; Tang, J.S.; Primm, R.T. III

An MCNP input was prepared for the Thai Research Reactor, making extensive use of the MCNP geometry`s lattice feature that allows a flexible and easy rearrangement of the core components and the adjustment of the control elements. The geometry was checked for overdefined or undefined zones by two-dimensional plots of cuts through the core configuration with the MCNP geometry plotting capabilities, and by a three-dimensional view of the core configuration with the SABRINA code. Cross sections were defined for a hypothetical core of 67 standard fuel elements and 38 low-enriched uranium fuel elements--all filled with fresh fuel. Three test calculationsmore » were performed with the MCNP4B-code to obtain the multiplication factor for the cases with control elements fully inserted, fully withdrawn, and at a working position.« less

User Manual for Whisper-1.1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Forrest B.; Rising, Michael Evan; Alwin, Jennifer Louise

2017-01-26

Whisper is a statistical analysis package developed in 2014 to support nuclear criticality safety (NCS) validation [1-3]. It uses the sensitivity profile data for an application as computed by MCNP6 [4-6] along with covariance files [7,8] for the nuclear data to determine a baseline upper-subcritical-limit (USL) for the application. Whisper version 1.0 was first developed and used at LANL in 2014 [3]. During 2015- 2016, Whisper was updated to version 1.1 and is to be included with the upcoming release of MCNP6.2. This document describes the user input and options for running whisper-1.1, including 2 perl utility scripts that simplifymore » ordinary NCS work, whisper_mcnp.pl and whisper_usl.pl. For many detailed references on the theory, applications, nuclear data & covariances, SQA, verification-validation, adjointbased methods for sensitivity-uncertainty analysis, and more – see the Whisper – NCS Validation section of the MCNP Reference Collection at mcnp.lanl.gov. There are currently over 50 Whisper reference documents available.« less

Analysis of JSI TRIGA MARK II reactor physical parameters calculated with TRIPOLI and MCNP.

PubMed

Henry, R; Tiselj, I; Snoj, L

2015-03-01

New computational model of the JSI TRIGA Mark II research reactor was built for TRIPOLI computer code and compared with existing MCNP code model. The same modelling assumptions were used in order to check the differences of the mathematical models of both Monte Carlo codes. Differences between the TRIPOLI and MCNP predictions of keff were up to 100pcm. Further validation was performed with analyses of the normalized reaction rates and computations of kinetic parameters for various core configurations. Copyright © 2014 Elsevier Ltd. All rights reserved.

The X6XS. 0 cross section library for MCNP-4

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pruvost, N.L.; Seamon, R.E.; Rombaugh, C.T.

1991-06-01

This report documents the work done by X-6, HSE-6, and CTR Technical Services to produce a comprehensive working cross-section library for MCNP-4 suitable for SUN workstations and similar environments. The resulting library consists of a total of 436 files (one file for each ZAID). The library is 152 Megabytes in Type 1 format and 32 Megabytes in Type 2 format. Type 2 can be used when porting the library from one computer to another of the same make. Otherwise, Type 1 must be used to ensure portability between different computer systems. Instructions for installing the library and adding ZAIDs tomore » it are included here. Also included is a description of the steps necessary to install and test version 4 of MCNP. To improve readability of this report, certain commands and filenames are given in uppercase letters. The actual command or filename on the SUN workstation, however, must be specified in lowercase letters. Any questions regarding the data contained in the library should be directed to X-6 and any questions regarding the installation of the library and the testing that was performed should be directed to HSE-6. 9 refs., 7 tabs.« less

Calculation of self–shielding factor for neutron activation experiments using GEANT4 and MCNP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Romero–Barrientos, Jaime, E-mail: jaromero@ing.uchile.cl; Universidad de Chile, DFI, Facultad de Ciencias Físicas Y Matemáticas, Avenida Blanco Encalada 2008, Santiago; Molina, F.

2016-07-07

The neutron self–shielding factor G as a function of the neutron energy was obtained for 14 pure metallic samples in 1000 isolethargic energy bins from 1·10{sup −5}eV to 2·10{sup 7}eV using Monte Carlo simulations in GEANT4 and MCNP6. The comparison of these two Monte Carlo codes shows small differences in the final self–shielding factor mostly due to the different cross section databases that each program uses.

Lecture Notes on Criticality Safety Validation Using MCNP & Whisper

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Forrest B.; Rising, Michael Evan; Alwin, Jennifer Louise

Training classes for nuclear criticality safety, MCNP documentation. The need for, and problems surrounding, validation of computer codes and data area considered first. Then some background for MCNP & Whisper is given--best practices for Monte Carlo criticality calculations, neutron spectra, S(α,β) thermal neutron scattering data, nuclear data sensitivities, covariance data, and correlation coefficients. Whisper is computational software designed to assist the nuclear criticality safety analyst with validation studies with the Monte Carlo radiation transport package MCNP. Whisper's methodology (benchmark selection – C k's, weights; extreme value theory – bias, bias uncertainty; MOS for nuclear data uncertainty – GLLS) and usagemore » are discussed.« less

The development of a thermal hydraulic feedback mechanism with a quasi-fixed point iteration scheme for control rod position modeling for the TRIGSIMS-TH application

NASA Astrophysics Data System (ADS)

Karriem, Veronica V.

Nuclear reactor design incorporates the study and application of nuclear physics, nuclear thermal hydraulic and nuclear safety. Theoretical models and numerical methods implemented in computer programs are utilized to analyze and design nuclear reactors. The focus of this PhD study's is the development of an advanced high-fidelity multi-physics code system to perform reactor core analysis for design and safety evaluations of research TRIGA-type reactors. The fuel management and design code system TRIGSIMS was further developed to fulfill the function of a reactor design and analysis code system for the Pennsylvania State Breazeale Reactor (PSBR). TRIGSIMS, which is currently in use at the PSBR, is a fuel management tool, which incorporates the depletion code ORIGEN-S (part of SCALE system) and the Monte Carlo neutronics solver MCNP. The diffusion theory code ADMARC-H is used within TRIGSIMS to accelerate the MCNP calculations. It manages the data and fuel isotopic content and stores it for future burnup calculations. The contribution of this work is the development of an improved version of TRIGSIMS, named TRIGSIMS-TH. TRIGSIMS-TH incorporates a thermal hydraulic module based on the advanced sub-channel code COBRA-TF (CTF). CTF provides the temperature feedback needed in the multi-physics calculations as well as the thermal hydraulics modeling capability of the reactor core. The temperature feedback model is using the CTF-provided local moderator and fuel temperatures for the cross-section modeling for ADMARC-H and MCNP calculations. To perform efficient critical control rod calculations, a methodology for applying a control rod position was implemented in TRIGSIMS-TH, making this code system a modeling and design tool for future core loadings. The new TRIGSIMS-TH is a computer program that interlinks various other functional reactor analysis tools. It consists of the MCNP5, ADMARC-H, ORIGEN-S, and CTF. CTF was coupled with both MCNP and ADMARC-H to provide the heterogeneous temperature distribution throughout the core. Each of these codes is written in its own computer language performing its function and outputs a set of data. TRIGSIMS-TH provides an effective use and data manipulation and transfer between different codes. With the implementation of feedback and control- rod-position modeling methodologies, the TRIGSIMS-TH calculations are more accurate and in a better agreement with measured data. The PSBR is unique in many ways and there are no "off-the-shelf" codes, which can model this design in its entirety. In particular, PSBR has an open core design, which is cooled by natural convection. Combining several codes into a unique system brings many challenges. It also requires substantial knowledge of both operation and core design of the PSBR. This reactor is in operation decades and there is a fair amount of studies and developments in both PSBR thermal hydraulics and neutronics. Measured data is also available for various core loadings and can be used for validation activities. The previous studies and developments in PSBR modeling also aids as a guide to assess the findings of the work herein. In order to incorporate new methods and codes into exiting TRIGSIMS, a re-evaluation of various components of the code was performed to assure the accuracy and efficiency of the existing CTF/MCNP5/ADMARC-H multi-physics coupling. A new set of ADMARC-H diffusion coefficients and cross sections was generated using the SERPENT code. This was needed as the previous data was not generated with thermal hydraulic feedback and the ARO position was used as the critical rod position. The B4C was re-evaluated for this update. The data exchange between ADMARC-H and MCNP5 was modified. The basic core model is given a flexibility to allow for various changes within the core model, and this feature was implemented in TRIGSIMS-TH. The PSBR core in the new code model can be expanded and changed. This allows the new code to be used as a modeling tool for design and analyses of future code loadings.

Neutronics Investigations for the Lower Part of a Westinghouse SVEA-96+ Assembly

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murphy, M.F.; Luethi, A.; Seiler, R.

2002-05-15

Accurate critical experiments have been performed for the validation of total fission (F{sub tot}) and {sup 238}U-capture (C{sub 8}) reaction rate distributions obtained with CASMO-4, HELIOS, BOXER, and MCNP4B for the lower axial region of a real Westinghouse SVEA-96+ fuel assembly. The assembly comprised fresh fuel with an average {sup 235}U enrichment of 4.02 wt%, a maximum enrichment of 4.74 wt%, 14 burnable-absorber fuel pins, and full-density water moderation. The experimental configuration investigated was core 1A of the LWR-PROTEUS Phase I project, where 61 different fuel pins, representing {approx}64% of the assembly, were gamma-scanned individually. Calculated (C) and measured (E)more » values have been compared in terms of C/E distributions. For F{sub tot}, the standard deviations are 1.2% for HELIOS, 0.9% for CASMO-4, 0.8% for MCNP4B, and 1.7% for BOXER. Standard deviations of 1.1% for HELIOS, CASMO-4, and MCNP4B and 1.2% for BOXER were obtained in the case of C{sub 8}. Despite the high degree of accuracy observed on the average, it was found that the five burnable-absorber fuel pins investigated showed a noticeable underprediction of F{sub tot}, quite systematically, for the deterministic codes evaluated (average C/E for the burnable-absorber fuel pins in the range 0.974 to 0.988, depending on the code)« less

Analytic computation of average energy of neutrons inducing fission

DOE Office of Scientific and Technical Information (OSTI.GOV)

Clark, Alexander Rich

2016-08-12

The objective of this report is to describe how I analytically computed the average energy of neutrons that induce fission in the bare BeRP ball. The motivation of this report is to resolve a discrepancy between the average energy computed via the FMULT and F4/FM cards in MCNP6 by comparison to the analytic results.

CAD-Based Shielding Analysis for ITER Port Diagnostics

NASA Astrophysics Data System (ADS)

Serikov, Arkady; Fischer, Ulrich; Anthoine, David; Bertalot, Luciano; De Bock, Maartin; O'Connor, Richard; Juarez, Rafael; Krasilnikov, Vitaly

2017-09-01

Radiation shielding analysis conducted in support of design development of the contemporary diagnostic systems integrated inside the ITER ports is relied on the use of CAD models. This paper presents the CAD-based MCNP Monte Carlo radiation transport and activation analyses for the Diagnostic Upper and Equatorial Port Plugs (UPP #3 and EPP #8, #17). The creation process of the complicated 3D MCNP models of the diagnostics systems was substantially accelerated by application of the CAD-to-MCNP converter programs MCAM and McCad. High performance computing resources of the Helios supercomputer allowed to speed-up the MCNP parallel transport calculations with the MPI/OpenMP interface. The found shielding solutions could be universal, reducing ports R&D costs. The shield block behind the Tritium and Deposit Monitor (TDM) optical box was added to study its influence on Shut-Down Dose Rate (SDDR) in Port Interspace (PI) of EPP#17. Influence of neutron streaming along the Lost Alpha Monitor (LAM) on the neutron energy spectra calculated in the Tangential Neutron Spectrometer (TNS) of EPP#8. For the UPP#3 with Charge eXchange Recombination Spectroscopy (CXRS-core), an excessive neutron streaming along the CXRS shutter, which should be prevented in further design iteration.

SU-C-209-05: Monte Carlo Model of a Prototype Backscatter X-Ray (BSX) Imager for Projective and Selective Object-Plane Imaging

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rolison, L; Samant, S; Baciak, J

Purpose: To develop a Monte Carlo N-Particle (MCNP) model for the validation of a prototype backscatter x-ray (BSX) imager, and optimization of BSX technology for medical applications, including selective object-plane imaging. Methods: BSX is an emerging technology that represents an alternative to conventional computed tomography (CT) and projective digital radiography (DR). It employs detectors located on the same side as the incident x-ray source, making use of backscatter and avoiding ring geometry to enclose the imaging object. Current BSX imagers suffer from low spatial resolution. A MCNP model was designed to replicate a BSX prototype used for flaw detection inmore » industrial materials. This prototype consisted of a 1.5mm diameter 60kVp pencil beam surrounded by a ring of four 5.0cm diameter NaI scintillation detectors. The imaging phantom consisted of a 2.9cm thick aluminum plate with five 0.6cm diameter holes drilled halfway. The experimental image was created using a raster scanning motion (in 1.5mm increments). Results: A qualitative comparison between the physical and simulated images showed very good agreement with 1.5mm spatial resolution in plane perpendicular to incident x-ray beam. The MCNP model developed the concept of radiography by selective plane detection (RSPD) for BSX, whereby specific object planes can be imaged by varying kVp. 10keV increments in mean x-ray energy yielded 4mm thick slice resolution in the phantom. Image resolution in the MCNP model can be further increased by increasing the number of detectors, and decreasing raster step size. Conclusion: MCNP modelling was used to validate a prototype BSX imager and introduce the RSPD concept, allowing for selective object-plane imaging. There was very good visual agreement between the experimental and MCNP imaging. Beyond optimizing system parameters for the existing prototype, new geometries can be investigated for volumetric image acquisition in medical applications. This material is based upon work supported under an Integrated University Program Graduate Fellowship sponsored by the Department of Energy Office of Nuclear Energy.« less

Possible Improvements to MCNP6 and its CEM/LAQGSM Event-Generators

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mashnik, Stepan Georgievich

2015-08-04

This report is intended to the MCNP6 developers and sponsors of MCNP6. It presents a set of suggested possible future improvements to MCNP6 and to its CEM03.03 and LAQGSM03.03 event-generators. A few suggested modifications of MCNP6 are quite simple, aimed at avoiding possible problems with running MCNP6 on various computers, i.e., these changes are not expected to change or improve any results, but should make the use of MCNP6 easier; such changes are expected to require limited man-power resources. On the other hand, several other suggested improvements require a serious further development of nuclear reaction models, are expected to improvemore » significantly the predictive power of MCNP6 for a number of nuclear reactions; but, such developments require several years of work by real experts on nuclear reactions.« less

Radiation shielding evaluation of the BNCT treatment room at THOR: a TORT-coupled MCNP Monte Carlo simulation study.

PubMed

Chen, A Y; Liu, Y-W H; Sheu, R J

2008-01-01

This study investigates the radiation shielding design of the treatment room for boron neutron capture therapy at Tsing Hua Open-pool Reactor using "TORT-coupled MCNP" method. With this method, the computational efficiency is improved significantly by two to three orders of magnitude compared to the analog Monte Carlo MCNP calculation. This makes the calculation feasible using a single CPU in less than 1 day. Further optimization of the photon weight windows leads to additional 50-75% improvement in the overall computational efficiency.

Benchmark of PENELOPE code for low-energy photon transport: dose comparisons with MCNP4 and EGS4.

PubMed

Ye, Sung-Joon; Brezovich, Ivan A; Pareek, Prem; Naqvi, Shahid A

2004-02-07

The expanding clinical use of low-energy photon emitting 125I and 103Pd seeds in recent years has led to renewed interest in their dosimetric properties. Numerous papers pointed out that higher accuracy could be obtained in Monte Carlo simulations by utilizing newer libraries for the low-energy photon cross-sections, such as XCOM and EPDL97. The recently developed PENELOPE 2001 Monte Carlo code is user friendly and incorporates photon cross-section data from the EPDL97. The code has been verified for clinical dosimetry of high-energy electron and photon beams, but has not yet been tested at low energies. In the present work, we have benchmarked the PENELOPE code for 10-150 keV photons. We computed radial dose distributions from 0 to 10 cm in water at photon energies of 10-150 keV using both PENELOPE and MCNP4C with either DLC-146 or DLC-200 cross-section libraries, assuming a point source located at the centre of a 30 cm diameter and 20 cm length cylinder. Throughout the energy range of simulated photons (except for 10 keV), PENELOPE agreed within statistical uncertainties (at worst +/- 5%) with MCNP/DLC-146 in the entire region of 1-10 cm and with published EGS4 data up to 5 cm. The dose at 1 cm (or dose rate constant) of PENELOPE agreed with MCNP/DLC-146 and EGS4 data within approximately +/- 2% in the range of 20-150 keV, while MCNP/DLC-200 produced values up to 9% lower in the range of 20-100 keV than PENELOPE or the other codes. However, the differences among the four datasets became negligible above 100 keV.

Recalculation with SEACAB of the activation by spent fuel neutrons and residual dose originated in the racks replaced at Cofrentes NPP

NASA Astrophysics Data System (ADS)

Ortego, Pedro; Rodriguez, Alain; Töre, Candan; Compadre, José Luis de Diego; Quesada, Baltasar Rodriguez; Moreno, Raul Orive

2017-09-01

In order to increase the storage capacity of the East Spent Fuel Pool at the Cofrentes NPP, located in Valencia province, Spain, the existing storage stainless steel racks were replaced by a new design of compact borated stainless steel racks allowing a 65% increase in fuel storing capacity. Calculation of the activation of the used racks was successfully performed with the use of MCNP4B code. Additionally the dose rate at contact with a row of racks in standing position and behind a wall of shielding material has been calculated using MCNP4B code as well. These results allowed a preliminary definition of the burnker required for the storage of racks. Recently the activity in the racks has been recalculated with SEACAB system which combines the mesh tally of MCNP codes with the activation code ACAB, applying the rigorous two-step method (R2S) developed at home, benchmarked with FNG irradiation experiments and usually applied in fusion calculations for ITER project.

Using MCNP6 to Estimate Fission Neutron Properties of a Reflected Plutonium Sphere

DOE Office of Scientific and Technical Information (OSTI.GOV)

Clark, Alexander Rich; Nelson, Mark Andrew; Hutchinson, Jesson D.

The purpose of this project was to determine the fission multiplicity distribution, p(v), for the Beryllium Reflected Plutonium (BeRP) ball and to determine whether or not it changed appreciably for various High Density Polyethylene (HDPE) reflected configurations. The motivation for this project was to determine whether or not the average number of neutrons emitted per fission, v, changed significantly enough to reduce the discrepancy between MCNP6 and Robba, Dowdy, Atwater (RDA) point kinetic model estimates of multiplication. The energy spectrum of neutrons that induced fissions in the BeRP ball, NIF (E), was also computed in order to determine the averagemore » energy of neutrons inducing fissions, NIF . p(v) was computed using the FMULT card, NIF (E) and NIF were computed using an F4 tally with an FM tally modifier (F4/FM) card, and the multiplication factor, k eff, was computed using the KCODE card. Although NIF (E) changed significantly between bare and HDPE reflected configurations of the BeRP ball, the change in p(v), and thus the change in v, was insignificant. This is likely due to a difference between the way that NIF is computed using the FMULT and F4/FM cards. The F4/FM card indicated that NIF (E) was essentially Watt-fission distributed for a bare configuration and highly thermalized for all HDPE reflected configurations, while the FMULT card returned an average energy between 1 and 2 MeV for all configurations, which would indicate that the spectrum is Watt-fission distributed, regardless of the amount of HDPE reflector. The spectrum computed with the F4/FM cards is more physically meaningful and so the discrepancy between it and the FMULT card result is being investigated. It is hoped that resolving the discrepancy between the FMULT and F4/FM card estimates of NIF(E) will provide better v estimates that will lead to RDA multiplication estimates that are in better agreement with MCNP6 simulations.« less

Evaluation of RAPID for a UNF cask benchmark problem

NASA Astrophysics Data System (ADS)

Mascolino, Valerio; Haghighat, Alireza; Roskoff, Nathan J.

2017-09-01

This paper examines the accuracy and performance of the RAPID (Real-time Analysis for Particle transport and In-situ Detection) code system for the simulation of a used nuclear fuel (UNF) cask. RAPID is capable of determining eigenvalue, subcritical multiplication, and pin-wise, axially-dependent fission density throughout a UNF cask. We study the source convergence based on the analysis of the different parameters used in an eigenvalue calculation in the MCNP Monte Carlo code. For this study, we consider a single assembly surrounded by absorbing plates with reflective boundary conditions. Based on the best combination of eigenvalue parameters, a reference MCNP solution for the single assembly is obtained. RAPID results are in excellent agreement with the reference MCNP solutions, while requiring significantly less computation time (i.e., minutes vs. days). A similar set of eigenvalue parameters is used to obtain a reference MCNP solution for the whole UNF cask. Because of time limitation, the MCNP results near the cask boundaries have significant uncertainties. Except for these, the RAPID results are in excellent agreement with the MCNP predictions, and its computation time is significantly lower, 35 second on 1 core versus 9.5 days on 16 cores.

Sensitivity-Uncertainty Based Nuclear Criticality Safety Validation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Forrest B.

2016-09-20

These are slides from a seminar given to the University of Mexico Nuclear Engineering Department. Whisper is a statistical analysis package developed to support nuclear criticality safety validation. It uses the sensitivity profile data for an application as computed by MCNP6 along with covariance files for the nuclear data to determine a baseline upper-subcritical-limit for the application. Whisper and its associated benchmark files are developed and maintained as part of MCNP6, and will be distributed with all future releases of MCNP6. Although sensitivity-uncertainty methods for NCS validation have been under development for 20 years, continuous-energy Monte Carlo codes such asmore » MCNP could not determine the required adjoint-weighted tallies for sensitivity profiles. The recent introduction of the iterated fission probability method into MCNP led to the rapid development of sensitivity analysis capabilities for MCNP6 and the development of Whisper. Sensitivity-uncertainty based methods represent the future for NCS validation – making full use of today’s computer power to codify past approaches based largely on expert judgment. Validation results are defensible, auditable, and repeatable as needed with different assumptions and process models. The new methods can supplement, support, and extend traditional validation approaches.« less

Automated variance reduction for MCNP using deterministic methods.

PubMed

Sweezy, J; Brown, F; Booth, T; Chiaramonte, J; Preeg, B

2005-01-01

In order to reduce the user's time and the computer time needed to solve deep penetration problems, an automated variance reduction capability has been developed for the MCNP Monte Carlo transport code. This new variance reduction capability developed for MCNP5 employs the PARTISN multigroup discrete ordinates code to generate mesh-based weight windows. The technique of using deterministic methods to generate importance maps has been widely used to increase the efficiency of deep penetration Monte Carlo calculations. The application of this method in MCNP uses the existing mesh-based weight window feature to translate the MCNP geometry into geometry suitable for PARTISN. The adjoint flux, which is calculated with PARTISN, is used to generate mesh-based weight windows for MCNP. Additionally, the MCNP source energy spectrum can be biased based on the adjoint energy spectrum at the source location. This method can also use angle-dependent weight windows.

Monte Carlo Modeling of the Initial Radiation Emitted by a Nuclear Device in the National Capital Region

DTIC Science & Technology

2013-07-01

also simulated in the models. Data was derived from calculations using the three-dimensional Monte Carlo radiation transport code MCNP (Monte Carlo N...32  B.  MCNP PHYSICS OPTIONS ......................................................................................... 33  C.  HAZUS...input deck’) for the MCNP , Monte Carlo N-Particle, radiation transport code. MCNP is a general-purpose code designed to simulate neutron, photon

Montgomery Point Lock and Dam, White River, Arkansas

DTIC Science & Technology

2016-01-01

ER D C/ CH L TR -1 6- 1 Monitoring Completed Navigation Projects (MCNP) Program Montgomery Point Lock and Dam, White River, Arkansas Co...Navigation Projects (MCNP) Program ERDC/CHL TR-16-1 January 2016 Montgomery Point Lock and Dam, White River, Arkansas Allen Hammack, Michael Winkler, and...20314-1000 Under MCNP Work Unit: Montgomery Point Lock and Dam, White River, Arkansas ERDC/CHL TR-16-1 ii Abstract Montgomery Point Lock and

Verification of MCNP6.2 for Nuclear Criticality Safety Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Forrest B.; Rising, Michael Evan; Alwin, Jennifer Louise

2017-05-10

Several suites of verification/validation benchmark problems were run in early 2017 to verify that the new production release of MCNP6.2 performs correctly for nuclear criticality safety applications (NCS). MCNP6.2 results for several NCS validation suites were compared to the results from MCNP6.1 [1] and MCNP6.1.1 [2]. MCNP6.1 is the production version of MCNP® released in 2013, and MCNP6.1.1 is the update released in 2014. MCNP6.2 includes all of the standard features for NCS calculations that have been available for the past 15 years, along with new features for sensitivity-uncertainty based methods for NCS validation [3]. Results from the benchmark suitesmore » were compared with results from previous verification testing [4-8]. Criticality safety analysts should consider testing MCNP6.2 on their particular problems and validation suites. No further development of MCNP5 is planned. MCNP6.1 is now 4 years old, and MCNP6.1.1 is now 3 years old. In general, released versions of MCNP are supported only for about 5 years, due to resource limitations. All future MCNP improvements, bug fixes, user support, and new capabilities are targeted only to MCNP6.2 and beyond.« less

Bias estimates used in lieu of validation of fission products and minor actinides in MCNP K eff calculations for PWR burnup credit casks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mueller, Don E.; Marshall, William J.; Wagner, John C.

The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (k eff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the biasmore » due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of k eff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.« less

Dosimetric characterization of the M−15 high‐dose‐rate Iridium−192 brachytherapy source using the AAPM and ESTRO formalism

PubMed Central

Thanh, Minh‐Tri Ho; Munro, John J.

2015-01-01

The Source Production & Equipment Co. (SPEC) model M−15 is a new Iridium−192 brachytherapy source model intended for use as a temporary high‐dose‐rate (HDR) brachytherapy source for the Nucletron microSelectron Classic afterloading system. The purpose of this study is to characterize this HDR source for clinical application by obtaining a complete set of Monte Carlo calculated dosimetric parameters for the M‐15, as recommended by AAPM and ESTRO, for isotopes with average energies greater than 50 keV. This was accomplished by using the MCNP6 Monte Carlo code to simulate the resulting source dosimetry at various points within a pseudoinfinite water phantom. These dosimetric values next were converted into the AAPM and ESTRO dosimetry parameters and the respective statistical uncertainty in each parameter also calculated and presented. The M−15 source was modeled in an MCNP6 Monte Carlo environment using the physical source specifications provided by the manufacturer. Iridium−192 photons were uniformly generated inside the iridium core of the model M−15 with photon and secondary electron transport replicated using photoatomic cross‐sectional tables supplied with MCNP6. Simulations were performed for both water and air/vacuum computer models with a total of 4×109 sources photon history for each simulation and the in‐air photon spectrum filtered to remove low‐energy photons below δ=10%keV. Dosimetric data, including D(r,θ),gL(r),F(r,θ),Φan(r), and φ¯an, and their statistical uncertainty were calculated from the output of an MCNP model consisting of an M−15 source placed at the center of a spherical water phantom of 100 cm diameter. The air kerma strength in free space, SK, and dose rate constant, Λ, also was computed from a MCNP model with M−15 Iridium−192 source, was centered at the origin of an evacuated phantom in which a critical volume containing air at STP was added 100 cm from the source center. The reference dose rate, D˙(r0,θ0)≡D˙(1cm,π/2), is found to be 4.038±0.064 cGy mCi−1 h−1. The air kerma strength, SK, is reported to be 3.632±0.086 cGy cm2 mCi−1 g−1, and the dose rate constant, Λ, is calculated to be 1.112±0.029 cGy h−1 U−1. The normalized dose rate, radial dose function, and anisotropy function with their uncertainties were computed and are represented in both tabular and graphical format in the report. A dosimetric study was performed of the new M−15 Iridium−192 HDR brachytherapy source using the MCNP6 radiation transport code. Dosimetric parameters, including the dose‐rate constant, radial dose function, and anisotropy function, were calculated in accordance with the updated AAPM and ESTRO dosimetric parameters for brachytherapy sources of average energy greater than 50 keV. These data therefore may be applied toward the development of a treatment planning program and for clinical use of the source. PACS numbers: 87.56.bg, 87.53.Jw PMID:26103489

Multi-threading performance of Geant4, MCNP6, and PHITS Monte Carlo codes for tetrahedral-mesh geometry.

PubMed

Han, Min Cheol; Yeom, Yeon Soo; Lee, Hyun Su; Shin, Bangho; Kim, Chan Hyeong; Furuta, Takuya

2018-05-04

In this study, the multi-threading performance of the Geant4, MCNP6, and PHITS codes was evaluated as a function of the number of threads (N) and the complexity of the tetrahedral-mesh phantom. For this, three tetrahedral-mesh phantoms of varying complexity (simple, moderately complex, and highly complex) were prepared and implemented in the three different Monte Carlo codes, in photon and neutron transport simulations. Subsequently, for each case, the initialization time, calculation time, and memory usage were measured as a function of the number of threads used in the simulation. It was found that for all codes, the initialization time significantly increased with the complexity of the phantom, but not with the number of threads. Geant4 exhibited much longer initialization time than the other codes, especially for the complex phantom (MRCP). The improvement of computation speed due to the use of a multi-threaded code was calculated as the speed-up factor, the ratio of the computation speed on a multi-threaded code to the computation speed on a single-threaded code. Geant4 showed the best multi-threading performance among the codes considered in this study, with the speed-up factor almost linearly increasing with the number of threads, reaching ~30 when N  =  40. PHITS and MCNP6 showed a much smaller increase of the speed-up factor with the number of threads. For PHITS, the speed-up factors were low when N  =  40. For MCNP6, the increase of the speed-up factors was better, but they were still less than ~10 when N  =  40. As for memory usage, Geant4 was found to use more memory than the other codes. In addition, compared to that of the other codes, the memory usage of Geant4 more rapidly increased with the number of threads, reaching as high as ~74 GB when N  =  40 for the complex phantom (MRCP). It is notable that compared to that of the other codes, the memory usage of PHITS was much lower, regardless of both the complexity of the phantom and the number of threads, hardly increasing with the number of threads for the MRCP.

Calculated organ doses for Mayak production association central hall using ICRP and MCNP.

PubMed

Choe, Dong-Ok; Shelkey, Brenda N; Wilde, Justin L; Walk, Heidi A; Slaughter, David M

2003-03-01

As part of an ongoing dose reconstruction project, equivalent organ dose rates from photons and neutrons were estimated using the energy spectra measured in the central hall above the graphite reactor core located in the Russian Mayak Production Association facility. Reconstruction of the work environment was necessary due to the lack of personal dosimeter data for neutrons in the time period prior to 1987. A typical worker scenario for the central hall was developed for the Monte Carlo Neutron Photon-4B (MCNP) code. The resultant equivalent dose rates for neutrons and photons were compared with the equivalent dose rates derived from calculations using the conversion coefficients in the International Commission on Radiological Protection Publications 51 and 74 in order to validate the model scenario for this Russian facility. The MCNP results were in good agreement with the results of the ICRP publications indicating the modeling scenario was consistent with actual work conditions given the spectra provided. The MCNP code will allow for additional orientations to accurately reflect source locations.

ZEPrompt: An Algorithm for Rapid Estimation of Building Attenuation for Prompt Radiation from a Nuclear Detonation

DTIC Science & Technology

2014-01-01

and 50 kT, to within 30% of first-principles code ( MCNP ) for complicated cities and 10% for simpler cities. 15. SUBJECT TERMS Radiation Transport...Use of MCNP for Dose Calculations .................................................................... 3 2.3 MCNP Open-Field Absorbed Dose...Calculations .................................................. 4 2.4 The MCNP Urban Model

LIBMAKER

DOE Office of Scientific and Technical Information (OSTI.GOV)

2015-08-01

Version 00 COG LibMaker contains various utilities to convert common data formats into a format usable by the COG - Multi-particle Monte Carlo Code System package, (C00777MNYCP01). Utilities included: ACEtoCOG - ACE formatted neutron data: Currently ENDFB7R0.BNL, ENDFB7R1.BNL, JEFF3.1, JEFF3.1.1, JEFF3.1.2, MCNP.50c, MCNP.51c, MCNP.55c, MCNP.66c, and MCNP.70c. ACEUtoCOG - ACEU formatted photonuclear data: Currently PN.MCNP.30c and PN.MCNP.70u. ACTLtoCOG - Creates a COG library from ENDL formatted activation data COG library. EDDLtoCOG - Creates a COG library from ENDL formatted LLNL deuteron data. ENDLtoCOG - Creates a COG library from ENDL formatted LLNL neutron data. EPDLtoCOG - Creates a COG librarymore » from ENDL formatted LLNL photon data. LEX - Creates a COG dictionary file. SAB.ACEtoCOG - Creates a COG library from ACE formatted S(a,b) data. SABtoCOG - Creates a COG library from ENDF6 formatted S(a,b) data. URRtoCOG - Creates a COG library from ACE formatted probability table data. This package also includes library checking and bit swapping capability.« less

Comparisons between MCNP, EGS4 and experiment for clinical electron beams.

PubMed

Jeraj, R; Keall, P J; Ostwald, P M

1999-03-01

Understanding the limitations of Monte Carlo codes is essential in order to avoid systematic errors in simulations, and to suggest further improvement of the codes. MCNP and EGS4, Monte Carlo codes commonly used in medical physics, were compared and evaluated against electron depth dose data and experimental backscatter results obtained using clinical radiotherapy beams. Different physical models and algorithms used in the codes give significantly different depth dose curves and electron backscattering factors. The default version of MCNP calculates electron depth dose curves which are too penetrating. The MCNP results agree better with experiment if the ITS-style energy-indexing algorithm is used. EGS4 underpredicts electron backscattering for high-Z materials. The results slightly improve if optimal PRESTA-I parameters are used. MCNP simulates backscattering well even for high-Z materials. To conclude the comparison, a timing study was performed. EGS4 is generally faster than MCNP and use of a large number of scoring voxels dramatically slows down the MCNP calculation. However, use of a large number of geometry voxels in MCNP only slightly affects the speed of the calculation.

Listing of Available ACE Data Tables

DOE Office of Scientific and Technical Information (OSTI.GOV)

Conlin, Jeremy Lloyd

This document is divided into multiple sections. Section 2 lists some of the more frequently used ENDF/B reaction types that can be used with the FM input card. The remaining sections (described below) contain tables showing the available ACE data tables for various types of data. These ACE data libraries are distributed by the Radiation Safety Information Computational Center (RSICC) with MCNP6.

Investigation on gamma and neutron radiation shielding parameters for BaO/SrO‒Bi2O3‒B2O3 glasses

NASA Astrophysics Data System (ADS)

Sayyed, M. I.; Lakshminarayana, G.; Dong, M. G.; Ersundu, M. Çelikbilek; Ersundu, A. E.; Kityk, I. V.

2018-04-01

In this work, mass attenuation coefficients (μ/ρ), effective atomic number (Zeff), electron density (Ne), mean free path (MFP), and half-value layer (HVL) of 20 BaO/SrO‒(x) Bi2O3‒(80‒x) B2O3 glasses (where x=10, 20, 30, 40, 50 and 60 mol%) were calculated using WinXCom program and MCNP5 code. The obtained (μ/ρ) results using both MCNP5 code and WinXCom program were in good agreement. It is found that the addition of Bi2O3 leads to increase the Zeff values in both BaO/SrO‒Bi2O3‒B2O3 glass systems. However, the Zeff values of the BaO‒Bi2O3‒B2O3 glass system are higher than those of the SrO‒Bi2O3‒B2O3 glasses. The fast neutrons effective removal cross sections (ΣR) for 20 SrO‒40 Bi2O3‒40 B2O3 glass is the highest among all studied glasses. The calculated half-value layer values were compared with different glass systems and it was found that the shielding properties of the selected glasses are comparable or even better than other glass systems such as phosphate glasses.

The MCNP6 Analytic Criticality Benchmark Suite

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Forrest B.

2016-06-16

Analytical benchmarks provide an invaluable tool for verifying computer codes used to simulate neutron transport. Several collections of analytical benchmark problems [1-4] are used routinely in the verification of production Monte Carlo codes such as MCNP® [5,6]. Verification of a computer code is a necessary prerequisite to the more complex validation process. The verification process confirms that a code performs its intended functions correctly. The validation process involves determining the absolute accuracy of code results vs. nature. In typical validations, results are computed for a set of benchmark experiments using a particular methodology (code, cross-section data with uncertainties, and modeling)more » and compared to the measured results from the set of benchmark experiments. The validation process determines bias, bias uncertainty, and possibly additional margins. Verification is generally performed by the code developers, while validation is generally performed by code users for a particular application space. The VERIFICATION_KEFF suite of criticality problems [1,2] was originally a set of 75 criticality problems found in the literature for which exact analytical solutions are available. Even though the spatial and energy detail is necessarily limited in analytical benchmarks, typically to a few regions or energy groups, the exact solutions obtained can be used to verify that the basic algorithms, mathematics, and methods used in complex production codes perform correctly. The present work has focused on revisiting this benchmark suite. A thorough review of the problems resulted in discarding some of them as not suitable for MCNP benchmarking. For the remaining problems, many of them were reformulated to permit execution in either multigroup mode or in the normal continuous-energy mode for MCNP. Execution of the benchmarks in continuous-energy mode provides a significant advance to MCNP verification methods.« less

A Monte Carlo model system for core analysis and epithermal neutron beam design at the Washington State University Radiation Center

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burns, T.D. Jr.

1996-05-01

The Monte Carlo Model System (MCMS) for the Washington State University (WSU) Radiation Center provides a means through which core criticality and power distributions can be calculated, as well as providing a method for neutron and photon transport necessary for BNCT epithermal neutron beam design. The computational code used in this Model System is MCNP4A. The geometric capability of this Monte Carlo code allows the WSU system to be modeled very accurately. A working knowledge of the MCNP4A neutron transport code increases the flexibility of the Model System and is recommended, however, the eigenvalue/power density problems can be run withmore » little direct knowledge of MCNP4A. Neutron and photon particle transport require more experience with the MCNP4A code. The Model System consists of two coupled subsystems; the Core Analysis and Source Plane Generator Model (CASP), and the BeamPort Shell Particle Transport Model (BSPT). The CASP Model incorporates the S({alpha}, {beta}) thermal treatment, and is run as a criticality problem yielding, the system eigenvalue (k{sub eff}), the core power distribution, and an implicit surface source for subsequent particle transport in the BSPT Model. The BSPT Model uses the source plane generated by a CASP run to transport particles through the thermal column beamport. The user can create filter arrangements in the beamport and then calculate characteristics necessary for assessing the BNCT potential of the given filter want. Examples of the characteristics to be calculated are: neutron fluxes, neutron currents, fast neutron KERMAs and gamma KERMAs. The MCMS is a useful tool for the WSU system. Those unfamiliar with the MCNP4A code can use the MCMS transparently for core analysis, while more experienced users will find the particle transport capabilities very powerful for BNCT filter design.« less

Shielding properties of 80TeO2-5TiO2-(15-x) WO3-xAnOm glasses using WinXCom and MCNP5 code

NASA Astrophysics Data System (ADS)

Dong, M. G.; El-Mallawany, R.; Sayyed, M. I.; Tekin, H. O.

2017-12-01

Gamma ray shielding properties of 80TeO2-5TiO2-(15-x) WO3-xAnOm glasses, where AnOm is Nb2O5 = 0.01, 5, Nd2O3 = 3, 5 and Er2O3 = 5 mol% have been achieved. Shielding parameters; mass attenuation coefficients, half value layers, and macroscopic effective removal cross section for fast neutrons have been computed by using WinXCom program and MCNP5 Monte Carlo code. In addition, by using Geometric Progression method (G-P), exposure buildup factor values were also calculated. Variations of shielding parameters are discussed for the effect of REO addition into the glasses and photon energy.

G4DARI: Geant4/GATE based Monte Carlo simulation interface for dosimetry calculation in radiotherapy.

PubMed

Slimani, Faiçal A A; Hamdi, Mahdjoub; Bentourkia, M'hamed

2018-05-01

Monte Carlo (MC) simulation is widely recognized as an important technique to study the physics of particle interactions in nuclear medicine and radiation therapy. There are different codes dedicated to dosimetry applications and widely used today in research or in clinical application, such as MCNP, EGSnrc and Geant4. However, such codes made the physics easier but the programming remains a tedious task even for physicists familiar with computer programming. In this paper we report the development of a new interface GEANT4 Dose And Radiation Interactions (G4DARI) based on GEANT4 for absorbed dose calculation and for particle tracking in humans, small animals and complex phantoms. The calculation of the absorbed dose is performed based on 3D CT human or animal images in DICOM format, from images of phantoms or from solid volumes which can be made from any pure or composite material to be specified by its molecular formula. G4DARI offers menus to the user and tabs to be filled with values or chemical formulas. The interface is described and as application, we show results obtained in a lung tumor in a digital mouse irradiated with seven energy beams, and in a patient with glioblastoma irradiated with five photon beams. In conclusion, G4DARI can be easily used by any researcher without the need to be familiar with computer programming, and it will be freely available as an application package. Copyright © 2018 Elsevier Ltd. All rights reserved.

Tally and geometry definition influence on the computing time in radiotherapy treatment planning with MCNP Monte Carlo code.

PubMed

Juste, B; Miro, R; Gallardo, S; Santos, A; Verdu, G

2006-01-01

The present work has simulated the photon and electron transport in a Theratron 780 (MDS Nordion) (60)Co radiotherapy unit, using the Monte Carlo transport code, MCNP (Monte Carlo N-Particle), version 5. In order to become computationally more efficient in view of taking part in the practical field of radiotherapy treatment planning, this work is focused mainly on the analysis of dose results and on the required computing time of different tallies applied in the model to speed up calculations.

MCNP capabilities for nuclear well logging calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Forster, R.A.; Little, R.C.; Briesmeister, J.F.

The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. This paper discusses how the general-purpose continuous-energy Monte Carlo code MCNP ({und M}onte {und C}arlo {und n}eutron {und p}hoton), part of the LARTCS, provides a computational predictive capability for many applications of interest to the nuclear well logging community. The generalized three-dimensional geometry of MCNP is well suited for borehole-tool models. SABRINA, another component of the LARTCS, is a graphics code that can be used to interactively create a complex MCNP geometry. Users can define many source and tallymore » characteristics with standard MCNP features. The time-dependent capability of the code is essential when modeling pulsed sources. Problems with neutrons, photons, and electrons as either single particle or coupled particles can be calculated with MCNP. The physics of neutron and photon transport and interactions is modeled in detail using the latest available cross-section data.« less

Calculation of the effective dose from natural radioactivity in soil using MCNP code.

PubMed

Krstic, D; Nikezic, D

2010-01-01

Effective dose delivered by photon emitted from natural radioactivity in soil was calculated in this work. Calculations have been done for the most common natural radionuclides in soil (238)U, (232)Th series and (40)K. A ORNL human phantoms and the Monte Carlo transport code MCNP-4B were employed to calculate the energy deposited in all organs. The effective dose was calculated according to ICRP 74 recommendations. Conversion factors of effective dose per air kerma were determined. Results obtained here were compared with other authors. Copyright 2009 Elsevier Ltd. All rights reserved.

Evaluation of computational models and cross sections used by MCNP6 for simulation of characteristic X-ray emission from thick targets bombarded by kiloelectronvolt electrons

NASA Astrophysics Data System (ADS)

Poškus, A.

2016-09-01

This paper evaluates the accuracy of the single-event (SE) and condensed-history (CH) models of electron transport in MCNP6.1 when simulating characteristic Kα, total K (=Kα + Kβ) and Lα X-ray emission from thick targets bombarded by electrons with energies from 5 keV to 30 keV. It is shown that the MCNP6.1 implementation of the CH model for the K-shell impact ionization leads to underestimation of the K yield by 40% or more for the elements with atomic numbers Z < 15 and overestimation of the Kα yield by more than 40% for the elements with Z > 25. The Lα yields are underestimated by more than an order of magnitude in CH mode, because MCNP6.1 neglects X-ray emission caused by electron-impact ionization of L, M and higher shells in CH mode (the Lα yields calculated in CH mode reflect only X-ray fluorescence, which is mainly caused by photoelectric absorption of bremsstrahlung photons). The X-ray yields calculated by MCNP6.1 in SE mode (using ENDF/B-VII.1 library data) are more accurate: the differences of the calculated and experimental K yields are within the experimental uncertainties for the elements C, Al and Si, and the calculated Kα yields are typically underestimated by (20-30)% for the elements with Z > 25, whereas the Lα yields are underestimated by (60-70)% for the elements with Z > 49. It is also shown that agreement of the experimental X-ray yields with those calculated in SE mode is additionally improved by replacing the ENDF/B inner-shell electron-impact ionization cross sections with the set of cross sections obtained from the distorted-wave Born approximation (DWBA), which are also used in the PENELOPE code system. The latter replacement causes a decrease of the average relative difference of the experimental X-ray yields and the simulation results obtained in SE mode to approximately 10%, which is similar to accuracy achieved with PENELOPE. This confirms that the DWBA inner-shell impact ionization cross sections are significantly more accurate than the corresponding ENDF/B cross sections when energy of incident electrons is of the order of the binding energy.

Monte Carlo simulation of x-ray spectra in diagnostic radiology and mammography using MCNP4C

NASA Astrophysics Data System (ADS)

Ay, M. R.; Shahriari, M.; Sarkar, S.; Adib, M.; Zaidi, H.

2004-11-01

The general purpose Monte Carlo N-particle radiation transport computer code (MCNP4C) was used for the simulation of x-ray spectra in diagnostic radiology and mammography. The electrons were transported until they slow down and stop in the target. Both bremsstrahlung and characteristic x-ray production were considered in this work. We focus on the simulation of various target/filter combinations to investigate the effect of tube voltage, target material and filter thickness on x-ray spectra in the diagnostic radiology and mammography energy ranges. The simulated x-ray spectra were compared with experimental measurements and spectra calculated by IPEM report number 78. In addition, the anode heel effect and off-axis x-ray spectra were assessed for different anode angles and target materials and the results were compared with EGS4-based Monte Carlo simulations and measured data. Quantitative evaluation of the differences between our Monte Carlo simulated and comparison spectra was performed using student's t-test statistical analysis. Generally, there is a good agreement between the simulated x-ray and comparison spectra, although there are systematic differences between the simulated and reference spectra especially in the K-characteristic x-rays intensity. Nevertheless, no statistically significant differences have been observed between IPEM spectra and the simulated spectra. It has been shown that the difference between MCNP simulated spectra and IPEM spectra in the low energy range is the result of the overestimation of characteristic photons following the normalization procedure. The transmission curves produced by MCNP4C have good agreement with the IPEM report especially for tube voltages of 50 kV and 80 kV. The systematic discrepancy for higher tube voltages is the result of systematic differences between the corresponding spectra.

Comparison of penumbra regions produced by ancient Gamma knife model C and Gamma ART 6000 using Monte Carlo MCNP6 simulation.

PubMed

Banaee, Nooshin; Asgari, Sepideh; Nedaie, Hassan Ali

2018-07-01

The accuracy of penumbral measurements in radiotherapy is pivotal because dose planning computers require accurate data to adequately modeling the beams, which in turn are used to calculate patient dose distributions. Gamma knife is a non-invasive intracranial technique based on principles of the Leksell stereotactic system for open deep brain surgeries, invented and developed by Professor Lars Leksell. The aim of this study is to compare the penumbra widths of Leksell Gamma Knife model C and Gamma ART 6000. Initially, the structure of both systems were simulated by using Monte Carlo MCNP6 code and after validating the accuracy of simulation, beam profiles of different collimators were plotted. MCNP6 beam profile calculations showed that the penumbra values of Leksell Gamma knife model C and Gamma ART 6000 for 18, 14, 8 and 4 mm collimators are 9.7, 7.9, 4.3, 2.6 and 8.2, 6.9, 3.6, 2.4, respectively. The results of this study showed that since Gamma ART 6000 has larger solid angle in comparison with Gamma Knife model C, it produces better beam profile penumbras than Gamma Knife model C in the direct plane. Copyright © 2017 Elsevier Ltd. All rights reserved.

Comparison of Monte Carlo simulation of gamma ray attenuation coefficients of amino acids with XCOM program and experimental data

NASA Astrophysics Data System (ADS)

Elbashir, B. O.; Dong, M. G.; Sayyed, M. I.; Issa, Shams A. M.; Matori, K. A.; Zaid, M. H. M.

2018-06-01

The mass attenuation coefficients (μ/ρ), effective atomic numbers (Zeff) and electron densities (Ne) of some amino acids obtained experimentally by the other researchers have been calculated using MCNP5 simulations in the energy range 0.122-1.330 MeV. The simulated values of μ/ρ, Zeff, and Ne were compared with the previous experimental work for the amino acids samples and a good agreement was noticed. Moreover, the values of mean free path (MFP) for the samples were calculated using MCNP5 program and compared with the theoretical results obtained by XCOM. The investigation of μ/ρ, Zeff, Ne and MFP values of amino acids using MCNP5 simulations at various photon energies when compared with the XCOM values and previous experimental data for the amino acids samples revealed that MCNP5 code provides accurate photon interaction parameters for amino acids.

Comparison of CdZnTe neutron detector models using MCNP6 and Geant4

NASA Astrophysics Data System (ADS)

Wilson, Emma; Anderson, Mike; Prendergasty, David; Cheneler, David

2018-01-01

The production of accurate detector models is of high importance in the development and use of detectors. Initially, MCNP and Geant were developed to specialise in neutral particle models and accelerator models, respectively; there is now a greater overlap of the capabilities of both, and it is therefore useful to produce comparative models to evaluate detector characteristics. In a collaboration between Lancaster University, UK, and Innovative Physics Ltd., UK, models have been developed in both MCNP6 and Geant4 of Cadmium Zinc Telluride (CdZnTe) detectors developed by Innovative Physics Ltd. Herein, a comparison is made of the relative strengths of MCNP6 and Geant4 for modelling neutron flux and secondary γ-ray emission. Given the increasing overlap of the modelling capabilities of MCNP6 and Geant4, it is worthwhile to comment on differences in results for simulations which have similarities in terms of geometries and source configurations.

Validation of the MCNP computational model for neutron flux distribution with the neutron activation analysis measurement

NASA Astrophysics Data System (ADS)

Tiyapun, K.; Chimtin, M.; Munsorn, S.; Somchit, S.

2015-05-01

The objective of this work is to demonstrate the method for validating the predication of the calculation methods for neutron flux distribution in the irradiation tubes of TRIGA research reactor (TRR-1/M1) using the MCNP computer code model. The reaction rate using in the experiment includes 27Al(n, α)24Na and 197Au(n, γ)198Au reactions. Aluminium (99.9 wt%) and gold (0.1 wt%) foils and the gold foils covered with cadmium were irradiated in 9 locations in the core referred to as CT, C8, C12, F3, F12, F22, F29, G5, and G33. The experimental results were compared to the calculations performed using MCNP which consisted of the detailed geometrical model of the reactor core. The results from the experimental and calculated normalized reaction rates in the reactor core are in good agreement for both reactions showing that the material and geometrical properties of the reactor core are modelled very well. The results indicated that the difference between the experimental measurements and the calculation of the reactor core using the MCNP geometrical model was below 10%. In conclusion the MCNP computational model which was used to calculate the neutron flux and reaction rate distribution in the reactor core can be used for others reactor core parameters including neutron spectra calculation, dose rate calculation, power peaking factors calculation and optimization of research reactor utilization in the future with the confidence in the accuracy and reliability of the calculation.

IDACstar: A MCNP Application to Perform Realistic Dose Estimations from Internal or External Contamination of Radiopharmaceuticals.

PubMed

Ören, Ünal; Hiller, Mauritius; Andersson, M

2017-04-28

A Monte Carlo-based stand-alone program, IDACstar (Internal Dose Assessment by Computer), was developed, dedicated to perform radiation dose calculations using complex voxel simulations. To test the program, two irradiation situations were simulated, one hypothetical contamination case with 600 MBq of 99mTc and one extravasation case involving 370 MBq of 18F-FDG. The effective dose was estimated to be 0.042 mSv for the contamination case and 4.5 mSv for the extravasation case. IDACstar has demonstrated that dosimetry results from contamination or extravasation cases can be acquired with great ease. An effective tool for radiation protection applications is provided with IDACstar allowing physicists at nuclear medicine departments to easily quantify the radiation risk of stochastic effects when a radiation accident has occurred. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Assessment of background hydrogen by the Monte Carlo computer code MCNP-4A during measurements of total body nitrogen.

PubMed

Ryde, S J; al-Agel, F A; Evans, C J; Hancock, D A

2000-05-01

The use of a hydrogen internal standard to enable the estimation of absolute mass during measurement of total body nitrogen by in vivo neutron activation is an established technique. Central to the technique is a determination of the H prompt gamma ray counts arising from the subject. In practice, interference counts from other sources--e.g., neutron shielding--are included. This study reports use of the Monte Carlo computer code, MCNP-4A, to investigate the interference counts arising from shielding both with and without a phantom containing a urea solution. Over a range of phantom size (depth 5 to 30 cm, width 20 to 40 cm), the counts arising from shielding increased by between 4% and 32% compared with the counts without a phantom. For any given depth, the counts increased approximately linearly with width. For any given width, there was little increase for depths exceeding 15 centimeters. The shielding counts comprised between 15% and 26% of those arising from the urea phantom. These results, although specific to the Swansea apparatus, suggest that extraneous hydrogen counts can be considerable and depend strongly on the subject's size.

MCNP6 Simulation of Light and Medium Nuclei Fragmentation at Intermediate Energies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mashnik, Stepan Georgievich; Kerby, Leslie Marie

2015-05-22

MCNP6, the latest and most advanced LANL Monte Carlo transport code, representing a merger of MCNP5 and MCNPX, is actually much more than the sum of those two computer codes; MCNP6 is available to the public via RSICC at Oak Ridge, TN, USA. In the present work, MCNP6 was validated and verified (V&V) against different experimental data on intermediate-energy fragmentation reactions, and results by several other codes, using mainly the latest modifications of the Cascade-Exciton Model (CEM) and of the Los Alamos version of the Quark-Gluon String Model (LAQGSM) event generators CEM03.03 and LAQGSM03.03. It was found that MCNP6 usingmore » CEM03.03 and LAQGSM03.03 describes well fragmentation reactions induced on light and medium target nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below, and can serve as a reliable simulation tool for different applications, like cosmic-ray-induced single event upsets (SEU’s), radiation protection, and cancer therapy with proton and ion beams, to name just a few. Future improvements of the predicting capabilities of MCNP6 for such reactions are possible, and are discussed in this work.« less

SIGACE Code for Generating High-Temperature ACE Files; Validation and Benchmarking

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sharma, Amit R.; Ganesan, S.; Trkov, A.

2005-05-24

A code named SIGACE has been developed as a tool for MCNP users within the scope of a research contract awarded by the Nuclear Data Section of the International Atomic Energy Agency (IAEA) (Ref: 302-F4-IND-11566 B5-IND-29641). A new recipe has been evolved for generating high-temperature ACE files for use with the MCNP code. Under this scheme the low-temperature ACE file is first converted to an ENDF formatted file using the ACELST code and then Doppler broadened, essentially limited to the data in the resolved resonance region, to any desired higher temperature using SIGMA1. The SIGACE code then generates a high-temperaturemore » ACE file for use with the MCNP code. A thinning routine has also been introduced in the SIGACE code for reducing the size of the ACE files. The SIGACE code and the recipe for generating ACE files at higher temperatures has been applied to the SEFOR fast reactor benchmark problem (sodium-cooled fast reactor benchmark described in ENDF-202/BNL-19302, 1974 document). The calculated Doppler coefficient is in good agreement with the experimental value. A similar calculation using ACE files generated directly with the NJOY system also agrees with our SIGACE computed results. The SIGACE code and the recipe is further applied to study the numerical benchmark configuration of selected idealized PWR pin cell configurations with five different fuel enrichments as reported by Mosteller and Eisenhart. The SIGACE code that has been tested with several FENDL/MC files will be available, free of cost, upon request, from the Nuclear Data Section of the IAEA.« less

Implementation of a tree algorithm in MCNP code for nuclear well logging applications.

PubMed

Li, Fusheng; Han, Xiaogang

2012-07-01

The goal of this paper is to develop some modeling capabilities that are missing in the current MCNP code. Those missing capabilities can greatly help for some certain nuclear tools designs, such as a nuclear lithology/mineralogy spectroscopy tool. The new capabilities to be developed in this paper include the following: zone tally, neutron interaction tally, gamma rays index tally and enhanced pulse-height tally. The patched MCNP code also can be used to compute neutron slowing-down length and thermal neutron diffusion length. Copyright © 2011 Elsevier Ltd. All rights reserved.

SABRINA: an interactive three-dimensional geometry-mnodeling program for MCNP

DOE Office of Scientific and Technical Information (OSTI.GOV)

West, J.T. III

SABRINA is a fully interactive three-dimensional geometry-modeling program for MCNP, a Los Alamos Monte Carlo code for neutron and photon transport. In SABRINA, a user constructs either body geometry or surface geometry models and debugs spatial descriptions for the resulting objects. This enhanced capability significantly reduces effort in constructing and debugging complicated three-dimensional geometry models for Monte Carlo analysis. 2 refs., 33 figs.

A comparison of the COG and MCNP codes in computational neutron capture therapy modeling, Part II: gadolinium neutron capture therapy models and therapeutic effects.

PubMed

Wangerin, K; Culbertson, C N; Jevremovic, T

2005-08-01

The goal of this study was to evaluate the COG Monte Carlo radiation transport code, developed and tested by Lawrence Livermore National Laboratory, for gadolinium neutron capture therapy (GdNCT) related modeling. The validity of COG NCT model has been established for this model, and here the calculation was extended to analyze the effect of various gadolinium concentrations on dose distribution and cell-kill effect of the GdNCT modality and to determine the optimum therapeutic conditions for treating brain cancers. The computational results were compared with the widely used MCNP code. The differences between the COG and MCNP predictions were generally small and suggest that the COG code can be applied to similar research problems in NCT. Results for this study also showed that a concentration of 100 ppm gadolinium in the tumor was most beneficial when using an epithermal neutron beam.

Li-Ion Batteries for Forensic Neutron Dosimetry

DTIC Science & Technology

2016-03-01

capture via lithium ions is tritium requires extraction from the battery such that it can be measured. This research program provides a method for...RMD) The following is a list of papers: 1. Amy Kaczmarowski, “Use of Lithium Ion Batteries for Nuclear Forensic Applications”, Undergraduate...2013. 3. Keith E. Holbert, Amy Kaczmarowski, Tyler Stannard, Erik B. Johnson, “MCNP Estimation of Trace Elements in Lithium - Ion Batteries Subjected

Comparison of TG-43 dosimetric parameters of brachytherapy sources obtained by three different versions of MCNP codes.

PubMed

Zaker, Neda; Zehtabian, Mehdi; Sina, Sedigheh; Koontz, Craig; Meigooni, Ali S

2016-03-08

Monte Carlo simulations are widely used for calculation of the dosimetric parameters of brachytherapy sources. MCNP4C2, MCNP5, MCNPX, EGS4, EGSnrc, PTRAN, and GEANT4 are among the most commonly used codes in this field. Each of these codes utilizes a cross-sectional library for the purpose of simulating different elements and materials with complex chemical compositions. The accuracies of the final outcomes of these simulations are very sensitive to the accuracies of the cross-sectional libraries. Several investigators have shown that inaccuracies of some of the cross section files have led to errors in 125I and 103Pd parameters. The purpose of this study is to compare the dosimetric parameters of sample brachytherapy sources, calculated with three different versions of the MCNP code - MCNP4C, MCNP5, and MCNPX. In these simulations for each source type, the source and phantom geometries, as well as the number of the photons, were kept identical, thus eliminating the possible uncertainties. The results of these investigations indicate that for low-energy sources such as 125I and 103Pd there are discrepancies in gL(r) values. Discrepancies up to 21.7% and 28% are observed between MCNP4C and other codes at a distance of 6 cm for 103Pd and 10 cm for 125I from the source, respectively. However, for higher energy sources, the discrepancies in gL(r) values are less than 1.1% for 192Ir and less than 1.2% for 137Cs between the three codes.

Monte Carlo dose calculations in homogeneous media and at interfaces: a comparison between GEPTS, EGSnrc, MCNP, and measurements.

PubMed

Chibani, Omar; Li, X Allen

2002-05-01

Three Monte Carlo photon/electron transport codes (GEPTS, EGSnrc, and MCNP) are bench-marked against dose measurements in homogeneous (both low- and high-Z) media as well as at interfaces. A brief overview on physical models used by each code for photon and electron (positron) transport is given. Absolute calorimetric dose measurements for 0.5 and 1 MeV electron beams incident on homogeneous and multilayer media are compared with the predictions of the three codes. Comparison with dose measurements in two-layer media exposed to a 60Co gamma source is also performed. In addition, comparisons between the codes (including the EGS4 code) are done for (a) 0.05 to 10 MeV electron beams and positron point sources in lead, (b) high-energy photons (10 and 20 MeV) irradiating a multilayer phantom (water/steel/air), and (c) simulation of a 90Sr/90Y brachytherapy source. A good agreement is observed between the calorimetric electron dose measurements and predictions of GEPTS and EGSnrc in both homogeneous and multilayer media. MCNP outputs are found to be dependent on the energy-indexing method (Default/ITS style). This dependence is significant in homogeneous media as well as at interfaces. MCNP(ITS) fits more closely the experimental data than MCNP(DEF), except for the case of Be. At low energy (0.05 and 0.1 MeV), MCNP(ITS) dose distributions in lead show higher maximums in comparison with GEPTS and EGSnrc. EGS4 produces too penetrating electron-dose distributions in high-Z media, especially at low energy (<0.1 MeV). For positrons, differences between GEPTS and EGSnrc are observed in lead because GEPTS distinguishes positrons from electrons for both elastic multiple scattering and bremsstrahlung emission models. For the 60Co source, a quite good agreement between calculations and measurements is observed with regards to the experimental uncertainty. For the other cases (10 and 20 MeV photon sources and the 90Sr/90Y beta source), a good agreement is found between the three codes. In conclusion, differences between GEPTS and EGSnrc results are found to be very small for almost all media and energies studied. MCNP results depend significantly on the electron energy-indexing method.

Calculation of conversion coefficients for clinical photon spectra using the MCNP code.

PubMed

Lima, M A F; Silva, A X; Crispim, V R

2004-01-01

In this work, the MCNP4B code has been employed to calculate conversion coefficients from air kerma to the ambient dose equivalent, H*(10)/Ka, for monoenergetic photon energies from 10 keV to 50 MeV, assuming the kerma approximation. Also estimated are the H*(10)/Ka for photon beams produced by linear accelerators, such as Clinac-4 and Clinac-2500, after transmission through primary barriers of radiotherapy treatment rooms. The results for the conversion coefficients for monoenergetic photon energies, with statistical uncertainty <2%, are compared with those in ICRP publication 74 and good agreements were obtained. The conversion coefficients calculated for real clinic spectra transmitted through walls of concrete of 1, 1.5 and 2 m thick, are in the range of 1.06-1.12 Sv Gy(-1).

Calculations of the thermal and fast neutron fluxes in the Syrian miniature neutron source reactor using the MCNP-4C code.

PubMed

Khattab, K; Sulieman, I

2009-04-01

The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ((197)Au (n, gamma) (198)Au and (59)Co (n, gamma) (60)Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.

Addressing Fission Product Validation in MCNP Burnup Credit Criticality Calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mueller, Don; Bowen, Douglas G; Marshall, William BJ J

2015-01-01

The US Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation issued Interim Staff Guidance (ISG) 8, Revision 3 in September 2012. This ISG provides guidance for NRC staff members’ review of burnup credit (BUC) analyses supporting transport and dry storage of pressurized water reactor spent nuclear fuel (SNF) in casks. The ISG includes guidance for addressing validation of criticality (k eff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MAs). Based on previous work documented in NRC Regulatory Guide (NUREG) Contractor Report (CR)-7109, the ISG recommends that NRC staff members acceptmore » the use of either 1.5 or 3% of the FP&MA worth—in addition to bias and bias uncertainty resulting from validation of k eff calculations for the major actinides in SNF—to conservatively account for the bias and bias uncertainty associated with the specified unvalidated FP&MAs. The ISG recommends (1) use of 1.5% of the FP&MA worth if a modern version of SCALE and its nuclear data are used and (2) 3% of the FP&MA worth for well qualified, industry standard code systems other than SCALE with the Evaluated Nuclear Data Files, Part B (ENDF/B),-V, ENDF/B-VI, or ENDF/B-VII cross sections libraries. The work presented in this paper provides a basis for extending the use of the 1.5% of the FP&MA worth bias to BUC criticality calculations performed using the Monte Carlo N-Particle (MCNP) code. The extended use of the 1.5% FP&MA worth bias is shown to be acceptable by comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII–based nuclear data. The comparison supports use of the 1.5% FP&MA worth bias when the MCNP code is used for criticality calculations, provided that the cask design is similar to the hypothetical generic BUC-32 cask model and that the credited FP&MA worth is no more than 0.1 Δk eff (ISG-8, Rev. 3, Recommendation 4).« less

Comparison of TG‐43 dosimetric parameters of brachytherapy sources obtained by three different versions of MCNP codes

PubMed Central

Zaker, Neda; Sina, Sedigheh; Koontz, Craig; Meigooni1, Ali S.

2016-01-01

Monte Carlo simulations are widely used for calculation of the dosimetric parameters of brachytherapy sources. MCNP4C2, MCNP5, MCNPX, EGS4, EGSnrc, PTRAN, and GEANT4 are among the most commonly used codes in this field. Each of these codes utilizes a cross‐sectional library for the purpose of simulating different elements and materials with complex chemical compositions. The accuracies of the final outcomes of these simulations are very sensitive to the accuracies of the cross‐sectional libraries. Several investigators have shown that inaccuracies of some of the cross section files have led to errors in  125I and  103Pd parameters. The purpose of this study is to compare the dosimetric parameters of sample brachytherapy sources, calculated with three different versions of the MCNP code — MCNP4C, MCNP5, and MCNPX. In these simulations for each source type, the source and phantom geometries, as well as the number of the photons, were kept identical, thus eliminating the possible uncertainties. The results of these investigations indicate that for low‐energy sources such as  125I and  103Pd there are discrepancies in gL(r) values. Discrepancies up to 21.7% and 28% are observed between MCNP4C and other codes at a distance of 6 cm for  103Pd and 10 cm for  125I from the source, respectively. However, for higher energy sources, the discrepancies in gL(r) values are less than 1.1% for  192Ir and less than 1.2% for  137Cs between the three codes. PACS number(s): 87.56.bg PMID:27074460

Modeling Sodium Iodide Detector Response Using Parametric Equations

DTIC Science & Technology

2013-03-22

MCNP particle current and pulse height tally functions, backscattering photons are quantified as a function of material thickness and energy...source – detector – scattering medium arrangements were modeled in MCNP using the pulse height tally functions, integrated over a 70 keV – 360 keV energy...15  4.1  MCNP

Simulation of the Mg(Ar) ionization chamber currents by different Monte Carlo codes in benchmark gamma fields

NASA Astrophysics Data System (ADS)

Lin, Yi-Chun; Liu, Yuan-Hao; Nievaart, Sander; Chen, Yen-Fu; Wu, Shu-Wei; Chou, Wen-Tsae; Jiang, Shiang-Huei

2011-10-01

High energy photon (over 10 MeV) and neutron beams adopted in radiobiology and radiotherapy always produce mixed neutron/gamma-ray fields. The Mg(Ar) ionization chambers are commonly applied to determine the gamma-ray dose because of its neutron insensitive characteristic. Nowadays, many perturbation corrections for accurate dose estimation and lots of treatment planning systems are based on Monte Carlo technique. The Monte Carlo codes EGSnrc, FLUKA, GEANT4, MCNP5, and MCNPX were used to evaluate energy dependent response functions of the Exradin M2 Mg(Ar) ionization chamber to a parallel photon beam with mono-energies from 20 keV to 20 MeV. For the sake of validation, measurements were carefully performed in well-defined (a) primary M-100 X-ray calibration field, (b) primary 60Co calibration beam, (c) 6-MV, and (d) 10-MV therapeutic beams in hospital. At energy region below 100 keV, MCNP5 and MCNPX both had lower responses than other codes. For energies above 1 MeV, the MCNP ITS-mode greatly resembled other three codes and the differences were within 5%. Comparing to the measured currents, MCNP5 and MCNPX using ITS-mode had perfect agreement with the 60Co, and 10-MV beams. But at X-ray energy region, the derivations reached 17%. This work shows us a better insight into the performance of different Monte Carlo codes in photon-electron transport calculation. Regarding the application of the mixed field dosimetry like BNCT, MCNP with ITS-mode is recognized as the most suitable tool by this work.

Preliminary Experiments with a Triple-Layer Phoswich Detector for Radioxenon Detection

DTIC Science & Technology

2008-09-01

Figure 7b; with a significant attenuation which was predicted by our MCNP modeling (Farsoni et al., 2007). The 81 keV peak in the NaI spectrum has a...analysis technique and confirmed our previous MCNP modeling. Our future work includes use of commercially available radioxenon gas (133Xe) to test

Determination of neutron flux distribution in an Am-Be irradiator using the MCNP.

PubMed

Shtejer-Diaz, K; Zamboni, C B; Zahn, G S; Zevallos-Chávez, J Y

2003-10-01

A neutron irradiator has been assembled at IPEN facilities to perform qualitative-quantitative analysis of many materials using thermal and fast neutrons outside the nuclear reactor premises. To establish the prototype specifications, the neutron flux distribution and the absorbed dose rates were calculated using the MCNP computer code. These theoretical predictions then allow one to discuss the optimum irradiator design and its performance.

Monte Carlo simulation of random, porous (foam) structures for neutron detection

NASA Astrophysics Data System (ADS)

Reichenberger, Michael A.; Fronk, Ryan G.; Shultis, J. Kenneth; Roberts, Jeremy A.; Edwards, Nathaniel S.; Stevenson, Sarah R.; Tiner, Christopher N.; McGregor, Douglas S.

2017-01-01

Porous media incorporating highly neutron-sensitive materials are of interest for use in the development of neutron detectors. Previous studies have shown experimentally the feasibility of 6LiF-saturated, multi-layered detectors; however, the random geometry of porous materials has limited the effectiveness of simulation efforts. The results of scatterless neutron transport and subsequent charged reaction product ion energy deposition are reported here using a novel Monte Carlo method and compared to results obtained by MCNP6. This new Dynamic Path Generation (DPG) Monte Carlo method was developed in order to overcome the complexities of modeling a random porous geometry in MCNP6. The DPG method is then applied to determine the optimal coating thickness for 10B4C-coated reticulated vitreous-carbon (RVC) foams. The optimal coating thickness for 4.1275 cm-thick 10B4C-coated reticulated vitreous carbon foams with porosities of 5, 10, 20, 30, 45, and 80 pores per inch (PPI) were determined for ionizing gas pressures of 1.0 and 2.8 atm. A simulated, maximum, intrinsic thermal-neutron detection efficiency of 62.8±0.25% was predicted for an 80 PPI RVC foam with a 0.2 μm thick coating of 10B4C, for a lower level discriminator setting of 75 keV and an argon pressure of 2.8 atm.

Effects of the Application of the New Nuclear Data Library ENDF/B to the Criticality Analysis of AP1000

NASA Astrophysics Data System (ADS)

Kuntoro, Iman; Sembiring, T. M.; Susilo, Jati; Deswandri; Sunaryo, G. R.

2018-02-01

Calculations of criticality of the AP1000 core due to the use of new edition of nuclear data library namely ENDF/B-VII and ENDF/B-VII.1 have been done. This work is aimed to know the accuracy of ENDF/B-VII.1 compared to ENDF/B-VII and ENDF/B-VI.8. in determining the criticality parameter of AP1000. Analysis ws imposed to core at cold zero power (CZP) conditions. The calculations have been carried out by means of MCNP computer code for 3 dimension geometry. The results show that criticality parameter namely effective multiplication factor of the AP1000 core are higher than that ones resulted from ENDF/B-VI.8 with relative differences of 0.39% for application of ENDF/B-VII and of 0.34% for application of ENDF/B-VII.1.

TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.

PubMed

Kurosawa, Masahiko

2005-01-01

For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martin, William R.; Lee, John C.; baxter, Alan

Information and measured data from the intial Fort St. Vrain (FSV) high temperature gas reactor core is used to develop a benchmark configuration to validate computational methods for analysis of a full-core, commercial HTR configuration. Large uncertainties in the geometry and composition data for the FSV fuel and core are identified, including: (1) the relative numbers of fuel particles for the four particle types, (2) the distribution of fuel kernel diameters for the four particle types, (3) the Th:U ratio in the initial FSV core, (4) and the buffer thickness for the fissile and fertile particles. Sensitivity studies were performedmore » to assess each of these uncertainties. A number of methods were developed to assist in these studies, including: (1) the automation of MCNP5 input files for FSV using Python scripts, (2) a simple method to verify isotopic loadings in MCNP5 input files, (3) an automated procedure to conduct a coupled MCNP5-RELAP5 analysis for a full-core FSV configuration with thermal-hydraulic feedback, and (4) a methodology for sampling kernel diameters from arbitrary power law and Gaussian PDFs that preserved fuel loading and packing factor constraints. A reference FSV fuel configuration was developed based on having a single diameter kernel for each of the four particle types, preserving known uranium and thorium loadings and packing factor (58%). Three fuel models were developed, based on representing the fuel as a mixture of kernels with two diameters, four diameters, or a continuous range of diameters. The fuel particles were put into a fuel compact using either a lattice-bsed approach or a stochastic packing methodology from RPI, and simulated with MCNP5. The results of the sensitivity studies indicated that the uncertainties in the relative numbers and sizes of fissile and fertile kernels were not important nor were the distributions of kernel diameters within their diameter ranges. The uncertainty in the Th:U ratio in the intial FSV core was found to be important with a crude study. The uncertainty in the TRISO buffer thickness was estimated to be unimportant but the study was not conclusive. FSV fuel compacts and a regular FSV fuel element were analyzed with MCNP5 and compared with predictions using a modified version of HELIOS that is capable of analyzing TRISO fuel configurations. The HELIOS analyses were performed by SSP. The eigenvalue discrepancies between HELIOS and MCNP5 are currently on the order of 1% but these are still being evaluated. Full-core FSV configurations were developed for two initial critical configurations - a cold, clean critical loading and a critical configuration at 70% power. MCNP5 predictions are compared to experimental data and the results are mixed. Analyses were also done for the pulsed neutron experiments that were conducted by GA for the initial FSV core. MCNP5 was used to model these experiments and reasonable agreement with measured results has been observed.« less

Some Experimental and Monte Carlo Investigations of the Plastic Scintillators for the Current Mode Measurements at Pulsed Neutron Sources

NASA Astrophysics Data System (ADS)

Rogov, A.; Pepyolyshev, Yu.; Carta, M.; d'Angelo, A.

Scintillation detector (SD) is widely used in neutron and gamma-spectrometry in a count mode. The organic scintillators for the count mode of the detector operation are investigated rather well. Usually, they are applied for measurement of amplitude and time distributions of pulses caused by single interaction events of neutrons or gamma's with scintillator material. But in a large area of scientific research scintillation detectors can alternatively be used on a current mode by recording the average current from the detector. For example,the measurements of the neutron pulse shape at the pulsed reactors or another pulsed neutron sources. So as to get a rather large volume of experimental data at pulsed neutron sources, it is necessary to use the current mode detector for registration of fast neutrons. Many parameters of the SD are changed with a transition from an accounting mode to current one. For example, the detector efficiency is different in counting and current modes. Many effects connected with time accuracy become substantial. Besides, for the registration of solely fast neutrons, as must be in many measurements, in the mixed radiation field of the pulsed neutron sources, SD efficiency has to be determined with a gamma-radiation shield present. Here is no calculations or experimental data on SD current mode operation up to now. The response functions of the detectors can be either measured in high-precision reference fields or calculated by a computer simulation. We have used the MCNP code [1] and carried out some experiments for investigation of the plastic performances in a current mode. There are numerous programs performing simulating similar to the MCNP code. For example, for neutrons there are [2-4], for photons - [5-8]. However, all known codes to use (SCINFUL, NRESP4, SANDYL, EGS49) have more stringent restrictions on the source, geometry and detector characteristics. In MCNP code a lot of these restrictions are absent and you need only to write special additions for proton and electron recoil and transfer energy to light output. These code modifications allow taking into account all processes in organic scintillator influence the light yield.

MCNP5 CALCULATIONS REPLICATING ARH-600 NITRATE DATA

DOE Office of Scientific and Technical Information (OSTI.GOV)

FINFROCK SH

This report serves to extend the previous document: 'MCNP Calculations Replicating ARH-600 Data' by replicating the nitrate curves found in ARH-600. This report includes the MCNP models used, the calculated critical dimension for each analyzed parameter set, and the resulting data libraries for use with the CritView code. As with the ARH-600 data, this report is not meant to replace the analysis of the fissile systems by qualified criticality personnel. The M CNP data is presented without accounting for the statistical uncertainty (although this is typically less than 0.001) or bias and, as such, the application of a reasonable safetymore » margin is required. The data that follows pertains to the uranyl nitrate and plutonium nitrate spheres, infinite cylinders, and infinite slabs of varying isotopic composition, reflector thickness, and molarity. Each of the cases was modeled in MCNP (version 5.1.40), using the ENDF/B-VI cross section set. Given a molarity, isotopic composition, and reflector thickness, the fissile concentration and diameter (or thicknesses in the case of the slab geometries) were varied. The diameter for which k-effective equals 1.00 for a given concentration could then be calculated and graphed. These graphs are included in this report. The pages that follow describe the regions modeled, formulas for calculating the various parameters, a list of cross-sections used in the calculations, a description of the automation routine and data, and finally the data output. The data of most interest are the critical dimensions of the various systems analyzed. This is presented graphically, and in table format, in Appendix B. Appendix C provides a text listing of the same data in a format that is compatible with the CritView code. Appendices D and E provide listing of example Template files and MCNP input files (these are discussed further in Section 4). Appendix F is a complete listing of all of the output data (i.e., all of the analyzed dimensions and the resulting k{sub eff} values).« less

Development and Validation of a Monte Carlo Simulation Tool for Multi-Pinhole SPECT

PubMed Central

Mok, Greta S. P.; Du, Yong; Wang, Yuchuan; Frey, Eric C.; Tsui, Benjamin M. W.

2011-01-01

Purpose In this work, we developed and validated a Monte Carlo simulation (MCS) tool for investigation and evaluation of multi-pinhole (MPH) SPECT imaging. Procedures This tool was based on a combination of the SimSET and MCNP codes. Photon attenuation and scatter in the object, as well as penetration and scatter through the collimator detector, are modeled in this tool. It allows accurate and efficient simulation of MPH SPECT with focused pinhole apertures and user-specified photon energy, aperture material, and imaging geometry. The MCS method was validated by comparing the point response function (PRF), detection efficiency (DE), and image profiles obtained from point sources and phantom experiments. A prototype single-pinhole collimator and focused four- and five-pinhole collimators fitted on a small animal imager were used for the experimental validations. We have also compared computational speed among various simulation tools for MPH SPECT, including SimSET-MCNP, MCNP, SimSET-GATE, and GATE for simulating projections of a hot sphere phantom. Results We found good agreement between the MCS and experimental results for PRF, DE, and image profiles, indicating the validity of the simulation method. The relative computational speeds for SimSET-MCNP, MCNP, SimSET-GATE, and GATE are 1: 2.73: 3.54: 7.34, respectively, for 120-view simulations. We also demonstrated the application of this MCS tool in small animal imaging by generating a set of low-noise MPH projection data of a 3D digital mouse whole body phantom. Conclusions The new method is useful for studying MPH collimator designs, data acquisition protocols, image reconstructions, and compensation techniques. It also has great potential to be applied for modeling the collimator-detector response with penetration and scatter effects for MPH in the quantitative reconstruction method. PMID:19779896

SU-F-T-140: Assessment of the Proton Boron Fusion Reaction for Practical Radiation Therapy Applications Using MCNP6

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adam, D; Bednarz, B

Purpose: The proton boron fusion reaction is a reaction that describes the creation of three alpha particles as the result of the interaction of a proton incident upon a 11B target. Theoretically, the proton boron fusion reaction is a desirable reaction for radiation therapy applications in that, with the appropriate boron delivery agent, it could potentially combine the localized dose delivery protons exhibit (Bragg peak) and the local deposition of high LET alpha particles in cancerous sites. Previous efforts have shown significant dose enhancement using the proton boron fusion reaction; the overarching purpose of this work is an attempt tomore » validate previous Monte Carlo results of the proton boron fusion reaction. Methods: The proton boron fusion reaction, 11B(p, 3α), is investigated using MCNP6 to assess the viability for potential use in radiation therapy. Simple simulations of a proton pencil beam incident upon both a water phantom and a water phantom with an axial region containing 100ppm boron were modeled using MCNP6 in order to determine the extent of the impact boron had upon the calculated energy deposition. Results: The maximum dose increase calculated was 0.026% for the incident 250 MeV proton beam scenario. The MCNP simulations performed demonstrated that the proton boron fusion reaction rate at clinically relevant boron concentrations was too small in order to have any measurable impact on the absorbed dose. Conclusion: For all MCNP6 simulations conducted, the increase of absorbed dose of a simple water phantom due to the 11B(p, 3α) reaction was found to be inconsequential. In addition, it was determined that there are no good evaluations of the 11B(p, 3α) reaction for use in MCNPX/6 and further work should be conducted in cross section evaluations in order to definitively evaluate the feasibility of the proton boron fusion reaction for use in radiation therapy applications.« less

MCNP-REN - A Monte Carlo Tool for Neutron Detector Design Without Using the Point Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Abhold, M.E.; Baker, M.C.

1999-07-25

The development of neutron detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model fails to accurately predict detector response in common applications. For this reason, the general Monte Carlo N-Particle code (MCNP) was modified to simulate the pulse streams that would be generated by a neutron detector and normally analyzed by a shift register. This modified code, MCNP - Random Exponentially Distributed Neutron Source (MCNP-REN), along with the Time Analysis Program (TAP) predict neutron detector response without using the pointmore » reactor model, making it unnecessary for the user to decide whether or not the assumptions of the point model are met for their application. MCNP-REN is capable of simulating standard neutron coincidence counting as well as neutron multiplicity counting. Measurements of MOX fresh fuel made using the Underwater Coincidence Counter (UWCC) as well as measurements of HEU reactor fuel using the active neutron Research Reactor Fuel Counter (RRFC) are compared with calculations. The method used in MCNP-REN is demonstrated to be fundamentally sound and shown to eliminate the need to use the point model for detector performance predictions.« less

The viability of ADVANTG deterministic method for synthetic radiography generation

NASA Astrophysics Data System (ADS)

Bingham, Andrew; Lee, Hyoung K.

2018-07-01

Fast simulation techniques to generate synthetic radiographic images of high resolution are helpful when new radiation imaging systems are designed. However, the standard stochastic approach requires lengthy run time with poorer statistics at higher resolution. The investigation of the viability of a deterministic approach to synthetic radiography image generation was explored. The aim was to analyze a computational time decrease over the stochastic method. ADVANTG was compared to MCNP in multiple scenarios including a small radiography system prototype, to simulate high resolution radiography images. By using ADVANTG deterministic code to simulate radiography images the computational time was found to decrease 10 to 13 times compared to the MCNP stochastic approach while retaining image quality.

Simulation of the GCR spectrum in the Mars curiosity rover's RAD detector using MCNP6

NASA Astrophysics Data System (ADS)

Ratliff, Hunter N.; Smith, Michael B. R.; Heilbronn, Lawrence

2017-08-01

The paper presents results from MCNP6 simulations of galactic cosmic ray (GCR) propagation down through the Martian atmosphere to the surface and comparison with RAD measurements made there. This effort is part of a collaborative modeling workshop for space radiation hosted by Southwest Research Institute (SwRI). All modeling teams were tasked with simulating the galactic cosmic ray (GCR) spectrum through the Martian atmosphere and the Radiation Assessment Detector (RAD) on-board the Curiosity rover. The detector had two separate particle acceptance angles, 4π and 30 ° off zenith. All ions with Z = 1 through Z = 28 were tracked in both scenarios while some additional secondary particles were only tracked in the 4π cases. The MCNP6 4π absorbed dose rate was 307.3 ± 1.3 μGy/day while RAD measured 233 μGy/day. Using the ICRP-60 dose equivalent conversion factors built into MCNP6, the simulated 4π dose equivalent rate was found to be 473.1 ± 2.4 μSv/day while RAD reported 710 μSv/day.

Validation of the WIMSD4M cross-section generation code with benchmark results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Deen, J.R.; Woodruff, W.L.; Leal, L.E.

1995-01-01

The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment Research and Test Reactor (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the WIMSD4M cross-section librariesmore » for reactor modeling of fresh water moderated cores. The results of calculations performed with multigroup cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory (ORNL) unreflected HEU critical spheres, the TRX LEU critical experiments, and calculations of a modified Los Alamos HEU D{sub 2}O moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented.« less

Validation of the WIMSD4M cross-section generation code with benchmark results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leal, L.C.; Deen, J.R.; Woodruff, W.L.

1995-02-01

The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment for Research and Test (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the procedure to generatemore » cross-section libraries for reactor analyses and calculations utilizing the WIMSD4M code. To do so, the results of calculations performed with group cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory(ORNL) unreflected critical spheres, the TRX critical experiments, and calculations of a modified Los Alamos highly-enriched heavy-water moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burke, Timothy P.; Martz, Roger L.; Kiedrowski, Brian C.

New unstructured mesh capabilities in MCNP6 (developmental version during summer 2012) show potential for conducting multi-physics analyses by coupling MCNP to a finite element solver such as Abaqus/CAE[2]. Before these new capabilities can be utilized, the ability of MCNP to accurately estimate eigenvalues and pin powers using an unstructured mesh must first be verified. Previous work to verify the unstructured mesh capabilities in MCNP was accomplished using the Godiva sphere [1], and this work attempts to build on that. To accomplish this, a criticality benchmark and a fuel assembly benchmark were used for calculations in MCNP using both the Constructivemore » Solid Geometry (CSG) native to MCNP and the unstructured mesh geometry generated using Abaqus/CAE. The Big Ten criticality benchmark [3] was modeled due to its geometry being similar to that of a reactor fuel pin. The C5G7 3-D Mixed Oxide (MOX) Fuel Assembly Benchmark [4] was modeled to test the unstructured mesh capabilities on a reactor-type problem.« less

Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR)

NASA Astrophysics Data System (ADS)

Brovchenko, Mariya; Dechenaux, Benjamin; Burn, Kenneth W.; Console Camprini, Patrizio; Duhamel, Isabelle; Peron, Arthur

2017-09-01

The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR). The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

Monte Carlo Techniques for Nuclear Systems - Theory Lectures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Forrest B.

These are lecture notes for a Monte Carlo class given at the University of New Mexico. The following topics are covered: course information; nuclear eng. review & MC; random numbers and sampling; computational geometry; collision physics; tallies and statistics; eigenvalue calculations I; eigenvalue calculations II; eigenvalue calculations III; variance reduction; parallel Monte Carlo; parameter studies; fission matrix and higher eigenmodes; doppler broadening; Monte Carlo depletion; HTGR modeling; coupled MC and T/H calculations; fission energy deposition. Solving particle transport problems with the Monte Carlo method is simple - just simulate the particle behavior. The devil is in the details, however. Thesemore » lectures provide a balanced approach to the theory and practice of Monte Carlo simulation codes. The first lectures provide an overview of Monte Carlo simulation methods, covering the transport equation, random sampling, computational geometry, collision physics, and statistics. The next lectures focus on the state-of-the-art in Monte Carlo criticality simulations, covering the theory of eigenvalue calculations, convergence analysis, dominance ratio calculations, bias in Keff and tallies, bias in uncertainties, a case study of a realistic calculation, and Wielandt acceleration techniques. The remaining lectures cover advanced topics, including HTGR modeling and stochastic geometry, temperature dependence, fission energy deposition, depletion calculations, parallel calculations, and parameter studies. This portion of the class focuses on using MCNP to perform criticality calculations for reactor physics and criticality safety applications. It is an intermediate level class, intended for those with at least some familiarity with MCNP. Class examples provide hands-on experience at running the code, plotting both geometry and results, and understanding the code output. The class includes lectures & hands-on computer use for a variety of Monte Carlo calculations. Beginning MCNP users are encouraged to review LA-UR-09-00380, "Criticality Calculations with MCNP: A Primer (3nd Edition)" (available at http:// mcnp.lanl.gov under "Reference Collection") prior to the class. No Monte Carlo class can be complete without having students write their own simple Monte Carlo routines for basic random sampling, use of the random number generator, and simplified particle transport simulation.« less

Geometry creation for MCNP by Sabrina and XSM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Riper, K.A.

The Monte Carlo N-Particle transport code MCNP is based on a surface description of 3-dimensional geometry. Cells are defined in terms of boolean operations on signed quadratic surfaces. MCNP geometry is entered as a card image file containing coefficients of the surface equations and a list of surfaces and operators describing cells. Several programs are available to assist in creation of the geometry specification, among them Sabrina and the new ``Smart Editor`` code XSM. We briefly describe geometry creation in Sabrina and then discuss XSM in detail. XSM is under development; our discussion is based on the state of XSMmore » as of January 1, 1994.« less

Simulations of neutron transport at low energy: a comparison between GEANT and MCNP.

PubMed

Colonna, N; Altieri, S

2002-06-01

The use of the simulation tool GEANT for neutron transport at energies below 20 MeV is discussed, in particular with regard to shielding and dose calculations. The reliability of the GEANT/MICAP package for neutron transport in a wide energy range has been verified by comparing the results of simulations performed with this package in a wide energy range with the prediction of MCNP-4B, a code commonly used for neutron transport at low energy. A reasonable agreement between the results of the two codes is found for the neutron flux through a slab of material (iron and ordinary concrete), as well as for the dose released in soft tissue by neutrons. These results justify the use of the GEANT/MICAP code for neutron transport in a wide range of applications, including health physics problems.

An evaluation of a manganese bath system having a new geometry through MCNP modelling.

PubMed

Khabaz, Rahim

2012-12-01

In this study, an approximate symmetric cylindrical manganese bath system with equal diameter and height was appraised using a Monte Carlo simulation. For nine sizes of the tank filled with MnSO(4).H(2)O solution of three different concentrations, the necessary correction factors involved in the absolute measurement of neutron emission rate were determined by a detailed modelling of the MCNP4C code with the ENDF/B-VII.0 neutron cross section data library. The results obtained were also used to determine the optimum dimensions of the bath for each concentration of solution in the calibration of (241)Am-Be and (252)Cf sources. Also, the amount of gamma radiation produced as a result of (n,γ) the reaction with the nuclei of the manganese sulphate solution that escaped from the boundary of each tank was evaluated. This gamma can be important for the background in NaI(Tl) detectors and issues concerned with radiation protection.

An investigation of voxel geometries for MCNP-based radiation dose calculations.

PubMed

Zhang, Juying; Bednarz, Bryan; Xu, X George

2006-11-01

Voxelized geometry such as those obtained from medical images is increasingly used in Monte Carlo calculations of absorbed doses. One useful application of calculated absorbed dose is the determination of fluence-to-dose conversion factors for different organs. However, confusion still exists about how such a geometry is defined and how the energy deposition is best computed, especially involving a popular code, MCNP5. This study investigated two different types of geometries in the MCNP5 code, cell and lattice definitions. A 10 cm x 10 cm x 10 cm test phantom, which contained an embedded 2 cm x 2 cm x 2 cm target at its center, was considered. A planar source emitting parallel photons was also considered in the study. The results revealed that MCNP5 does not calculate total target volume for multi-voxel geometries. Therefore, tallies which involve total target volume must be divided by the user by the total number of voxels to obtain a correct dose result. Also, using planar source areas greater than the phantom size results in the same fluence-to-dose conversion factor.

Confirmation of a realistic reactor model for BNCT dosimetry at the TRIGA Mainz

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ziegner, Markus, E-mail: Markus.Ziegner.fl@ait.ac.at; Schmitz, Tobias; Hampel, Gabriele

2014-11-01

Purpose: In order to build up a reliable dose monitoring system for boron neutron capture therapy (BNCT) applications at the TRIGA reactor in Mainz, a computer model for the entire reactor was established, simulating the radiation field by means of the Monte Carlo method. The impact of different source definition techniques was compared and the model was validated by experimental fluence and dose determinations. Methods: The depletion calculation code ORIGEN2 was used to compute the burn-up and relevant material composition of each burned fuel element from the day of first reactor operation to its current core. The material composition ofmore » the current core was used in a MCNP5 model of the initial core developed earlier. To perform calculations for the region outside the reactor core, the model was expanded to include the thermal column and compared with the previously established ATTILA model. Subsequently, the computational model is simplified in order to reduce the calculation time. Both simulation models are validated by experiments with different setups using alanine dosimetry and gold activation measurements with two different types of phantoms. Results: The MCNP5 simulated neutron spectrum and source strength are found to be in good agreement with the previous ATTILA model whereas the photon production is much lower. Both MCNP5 simulation models predict all experimental dose values with an accuracy of about 5%. The simulations reveal that a Teflon environment favorably reduces the gamma dose component as compared to a polymethyl methacrylate phantom. Conclusions: A computer model for BNCT dosimetry was established, allowing the prediction of dosimetric quantities without further calibration and within a reasonable computation time for clinical applications. The good agreement between the MCNP5 simulations and experiments demonstrates that the ATTILA model overestimates the gamma dose contribution. The detailed model can be used for the planning of structural modifications in the thermal column irradiation channel or the use of different irradiation sites than the thermal column, e.g., the beam tubes.« less

Comparison of the thermal neutron scattering treatment in MCNP6 and GEANT4 codes

NASA Astrophysics Data System (ADS)

Tran, H. N.; Marchix, A.; Letourneau, A.; Darpentigny, J.; Menelle, A.; Ott, F.; Schwindling, J.; Chauvin, N.

2018-06-01

To ensure the reliability of simulation tools, verification and comparison should be made regularly. This paper describes the work performed in order to compare the neutron transport treatment in MCNP6.1 and GEANT4-10.3 in the thermal energy range. This work focuses on the thermal neutron scattering processes for several potential materials which would be involved in the neutron source designs of Compact Accelerator-based Neutrons Sources (CANS), such as beryllium metal, beryllium oxide, polyethylene, graphite, para-hydrogen, light water, heavy water, aluminium and iron. Both thermal scattering law and free gas model, coming from the evaluated data library ENDF/B-VII, were considered. It was observed that the GEANT4.10.03-patch2 version was not able to account properly the coherent elastic process occurring in crystal lattice. This bug is treated in this work and it should be included in the next release of the code. Cross section sampling and integral tests have been performed for both simulation codes showing a fair agreement between the two codes for most of the materials except for iron and aluminium.

Comparative Dosimetric Estimates of a 25 keV Electron Micro-beam with three Monte Carlo Codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mainardi, Enrico; Donahue, Richard J.; Blakely, Eleanor A.

2002-09-11

The calculations presented compare the different performances of the three Monte Carlo codes PENELOPE-1999, MCNP-4C and PITS, for the evaluation of Dose profiles from a 25 keV electron micro-beam traversing individual cells. The overall model of a cell is a water cylinder equivalent for the three codes but with a different internal scoring geometry: hollow cylinders for PENELOPE and MCNP, whereas spheres are used for the PITS code. A cylindrical cell geometry with scoring volumes with the shape of hollow cylinders was initially selected for PENELOPE and MCNP because of its superior simulation of the actual shape and dimensions ofmore » a cell and for its improved computer-time efficiency if compared to spherical internal volumes. Some of the transfer points and energy transfer that constitute a radiation track may actually fall in the space between spheres, that would be outside the spherical scoring volume. This internal geometry, along with the PENELOPE algorithm, drastically reduced the computer time when using this code if comparing with event-by-event Monte Carlo codes like PITS. This preliminary work has been important to address dosimetric estimates at low electron energies. It demonstrates that codes like PENELOPE can be used for Dose evaluation, even with such small geometries and energies involved, which are far below the normal use for which the code was created. Further work (initiated in Summer 2002) is still needed however, to create a user-code for PENELOPE that allows uniform comparison of exact cell geometries, integral volumes and also microdosimetric scoring quantities, a field where track-structure codes like PITS, written for this purpose, are believed to be superior.« less

A new graphical user interface for fast construction of computation phantoms and MCNP calculations: application to calibration of in vivo measurement systems.

PubMed

Borisov, N; Franck, D; de Carlan, L; Laval, L

2002-08-01

The paper reports on a new utility for development of computational phantoms for Monte Carlo calculations and data analysis for in vivo measurements of radionuclides deposited in tissues. The individual properties of each worker can be acquired for a rather precise geometric representation of his (her) anatomy, which is particularly important for low energy gamma ray emitting sources such as thorium, uranium, plutonium and other actinides. The software discussed here enables automatic creation of an MCNP input data file based on scanning data. The utility includes segmentation of images obtained with either computed tomography or magnetic resonance imaging by distinguishing tissues according to their signal (brightness) and specification of the source and detector. In addition, a coupling of individual voxels within the tissue is used to reduce the memory demand and to increase the calculational speed. The utility was tested for low energy emitters in plastic and biological tissues as well as for computed tomography and magnetic resonance imaging scanning information.

Development of a patient-specific dosimetry estimation system in nuclear medicine examination

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lin, H. H.; Dong, S. L.; Yang, H. J.

2011-07-01

The purpose of this study is to develop a patient-specific dosimetry estimation system in nuclear medicine examination using a SimSET-based Monte Carlo code. We added a dose deposition routine to store the deposited energy of the photons during their flights in SimSET and developed a user-friendly interface for reading PET and CT images. Dose calculated on ORNL phantom was used to validate the accuracy of this system. The S values for {sup 99m}Tc, {sup 18}F and {sup 131}I obtained by the system were compared to those from the MCNP4C code and OLINDA. The ratios of S values computed by thismore » system to those obtained with OLINDA for various organs were ranged from 0.93 to 1.18, which are comparable to that obtained from MCNP4C code (0.94 to 1.20). The average ratios of S value were 0.99{+-}0.04, 1.03{+-}0.05, and 1.00{+-}0.07 for isotopes {sup 131}I, {sup 18}F, and {sup 99m}Tc, respectively. The simulation time of SimSET was two times faster than MCNP4C's for various isotopes. A 3D dose calculation was also performed on a patient data set with PET/CT examination using this system. Results from the patient data showed that the estimated S values using this system differed slightly from those of OLINDA for ORNL phantom. In conclusion, this system can generate patient-specific dose distribution and display the isodose curves on top of the anatomic structure through a friendly graphic user interface. It may also provide a useful tool to establish an appropriate dose-reduction strategy to patients in nuclear medicine environments. (authors)« less

Comparison of GATE/GEANT4 with EGSnrc and MCNP for electron dose calculations at energies between 15 keV and 20 MeV.

PubMed

Maigne, L; Perrot, Y; Schaart, D R; Donnarieix, D; Breton, V

2011-02-07

The GATE Monte Carlo simulation platform based on the GEANT4 toolkit has come into widespread use for simulating positron emission tomography (PET) and single photon emission computed tomography (SPECT) imaging devices. Here, we explore its use for calculating electron dose distributions in water. Mono-energetic electron dose point kernels and pencil beam kernels in water are calculated for different energies between 15 keV and 20 MeV by means of GATE 6.0, which makes use of the GEANT4 version 9.2 Standard Electromagnetic Physics Package. The results are compared to the well-validated codes EGSnrc and MCNP4C. It is shown that recent improvements made to the GEANT4/GATE software result in significantly better agreement with the other codes. We furthermore illustrate several issues of general interest to GATE and GEANT4 users who wish to perform accurate simulations involving electrons. Provided that the electron step size is sufficiently restricted, GATE 6.0 and EGSnrc dose point kernels are shown to agree to within less than 3% of the maximum dose between 50 keV and 4 MeV, while pencil beam kernels are found to agree to within less than 4% of the maximum dose between 15 keV and 20 MeV.

Dose mapping using MCNP code and experiment for SVST-Co-60/B irradiator in Vietnam.

PubMed

Tran, Van Hung; Tran, Khac An

2010-06-01

By using MCNP code and ethanol-chlorobenzene (ECB) dosimeters the simulations and measurements of absorbed dose distribution in a tote-box of the Cobalt-60 irradiator, SVST-Co60/B at VINAGAMMA have been done. Based on the results Dose Uniformity Ratios (DUR), positions and values of minimum and maximum dose extremes in a tote-box, and efficiency of the irradiator for the different dummy densities have been gained. There is a good agreement between simulation and experimental results in comparison and they have valuable meanings for operation of the irradiator. Copyright 2010 Elsevier Ltd. All rights reserved.

Neutron Radiography and Computed Tomography at Oak Ridge National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Raine, Dudley A. III; Hubbard, Camden R.; Whaley, Paul M.

1997-12-31

The capability to perform neutron radiography and computed tomography is being developed at Oak Ridge National Laboratory. The facility will be located at the High Flux Isotope Reactor (HFIR), which has the highest steady state neutron flux of any reactor in the world. The Monte Carlo N-Particle transport code (MCNP), versions 4A and 4B, has been used extensively in the design phase of the facility to predict and optimize the operating characteristics, and to ensure the safety of personnel working in and around the blockhouse. Neutrons are quite penetrating in most engineering materials and can be useful to detect internalmore » flaws and features. Hydrogen atoms, such as in a hydrocarbon fuel, lubricant or a metal hydride, are relatively opaque to neutron transmission. Thus, neutron based tomography or radiography is ideal to image their presence. The source flux also provides unparalleled flexibility for future upgrades, including real time radiography where dynamic processes can be observed. A novel tomography detector has been designed using optical fibers and digital technology to provide a large dynamic range for reconstructions. Film radiography is also available for high resolution imaging applications. This paper summarizes the results of the design phase of this facility and the potential benefits to science and industry.« less

Simulation of the GCR spectrum in the Mars curiosity rover's RAD detector using MCNP6.

PubMed

Ratliff, Hunter N; Smith, Michael B R; Heilbronn, Lawrence

2017-08-01

The paper presents results from MCNP6 simulations of galactic cosmic ray (GCR) propagation down through the Martian atmosphere to the surface and comparison with RAD measurements made there. This effort is part of a collaborative modeling workshop for space radiation hosted by Southwest Research Institute (SwRI). All modeling teams were tasked with simulating the galactic cosmic ray (GCR) spectrum through the Martian atmosphere and the Radiation Assessment Detector (RAD) on-board the Curiosity rover. The detector had two separate particle acceptance angles, 4π and 30 ° off zenith. All ions with Z = 1 through Z = 28 were tracked in both scenarios while some additional secondary particles were only tracked in the 4π cases. The MCNP6 4π absorbed dose rate was 307.3 ± 1.3 µGy/day while RAD measured 233 µGy/day. Using the ICRP-60 dose equivalent conversion factors built into MCNP6, the simulated 4π dose equivalent rate was found to be 473.1 ± 2.4 µSv/day while RAD reported 710 µSv/day. Copyright © 2017 The Committee on Space Research (COSPAR). Published by Elsevier Ltd. All rights reserved.

Development of an Efficient Approach to Perform Neutronics Simulations for Plutonium-238 Production

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chandler, David; Ellis, Ronald James

Conversion of 238Pu decay heat into usable electricity is imperative to power National Aeronautics and Space Administration (NASA) deep space exploration missions; however, the current stockpile of 238Pu is diminishing and the quality is less than ideal. In response, the US Department of Energy and NASA have undertaken a program to reestablish a domestic 238Pu production program and a technology demonstration sub-project has been initiated. Neutronics simulations for 238Pu production play a vital role in this project because the results guide reactor safety-basis, target design and optimization, and post-irradiation examination activities. A new, efficient neutronics simulation tool written in Pythonmore » was developed to evaluate, with the highest fidelity possible with approved tools, the time-dependent nuclide evolution and heat deposition rates in 238Pu production targets irradiated in the High Flux Isotope Reactor (HFIR). The Python Activation and Heat Deposition Script (PAHDS) was developed specifically for experiment analysis in HFIR and couples the MCNP5 and SCALE 6.1.3 software quality assured tools to take advantage of an existing high-fidelity MCNP HFIR model, the most up-to-date ORIGEN code, and the most up-to-date nuclear data. Three cycle simulations were performed with PAHDS implementing ENDF/B-VII.0, ENDF/B-VII.1, and the Hybrid Library GPD-Rev0 cross-section libraries. The 238Pu production results were benchmarked against VESTA-obtained results and the impact of various cross-section libraries on the calculated metrics were assessed.« less

MCNP6 Status

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goorley, John T.

2012-06-25

We, the development teams for MCNP, NJOY, and parts of ENDF, would like to invite you to a proposed 3 day workshop October 30, 31 and November 1 2012, to be held at Los Alamos National Laboratory. At this workshop, we will review new and developing missions that MCNP6 and the underlying nuclear data are being asked to address. LANL will also present its internal plans to address these missions and recent advances in these three capabilities and we will be interested to hear your input on these topics. Additionally we are interested in hearing from you additional technical advances,more » missions, concerns, and other issues that we should be considering for both short term (1-3 years) and long term (4-6 years)? What are the additional existing capabilities and methods that we should be investigating? The goal of the workshop is to refine priorities for mcnp6 transport methods, algorithms, physics, data and processing as they relate to the intersection of MCNP, NJOY and ENDF.« less

Neutron Fluence And DPA Rate Analysis In Pebble-Bed HTR Reactor Vessel Using MCNP

NASA Astrophysics Data System (ADS)

Hamzah, Amir; Suwoto; Rohanda, Anis; Adrial, Hery; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

In the Pebble-bed HTR reactor, the distance between the core and the reactor vessel is very close and the media inside are carbon and He gas. Neutron moderation capability of graphite material is theoretically lower than that of water-moderated reactors. Thus, it is estimated much more the fast neutrons will reach the reactor vessel. The fast neutron collisions with the atoms in the reactor vessel will result in radiation damage and could be reducing the vessel life. The purpose of this study was to obtain the magnitude of neutron fluence in the Pebble-bed HTR reactor vessel. Neutron fluence calculations in the pebble-bed HTR reactor vessel were performed using the MCNP computer program. By determining the tally position, it can be calculated flux, spectrum and neutron fluence in the position of Pebble-bed HTR reactor vessel. The calculations results of total neutron flux and fast neutron flux in the reactor vessel of 1.82x108 n/cm2/s and 1.79x108 n/cm2/s respectively. The fast neutron fluence in the reactor vessel is 3.4x1017 n/cm2 for 60 years reactor operation. Radiation damage in stainless steel material caused by high-energy neutrons (> 1.0 MeV) will occur when it has reached the neutron flux level of 1.0x1024 n/cm2. The neutron fluence results show that there is no radiation damage in the Pebble-bed HTR reactor vessel, so it is predicted that it will be safe to operate at least for 60 years.

Verification of MCNP simulation of neutron flux parameters at TRIGA MK II reactor of Malaysia.

PubMed

Yavar, A R; Khalafi, H; Kasesaz, Y; Sarmani, S; Yahaya, R; Wood, A K; Khoo, K S

2012-10-01

A 3-D model for 1 MW TRIGA Mark II research reactor was simulated. Neutron flux parameters were calculated using MCNP-4C code and were compared with experimental results obtained by k(0)-INAA and absolute method. The average values of φ(th),φ(epi), and φ(fast) by MCNP code were (2.19±0.03)×10(12) cm(-2)s(-1), (1.26±0.02)×10(11) cm(-2)s(-1) and (3.33±0.02)×10(10) cm(-2)s(-1), respectively. These average values were consistent with the experimental results obtained by k(0)-INAA. The findings show a good agreement between MCNP code results and experimental results. Copyright © 2012 Elsevier Ltd. All rights reserved.

Hybrid Skyshine Calculations for Complex Neutron and Gamma-Ray Sources

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shultis, J. Kenneth

2000-10-15

A two-step hybrid method is described for computationally efficient estimation of neutron and gamma-ray skyshine doses far from a shielded source. First, the energy and angular dependence of radiation escaping into the atmosphere from a source containment is determined by a detailed transport model such as MCNP. Then, an effective point source with this energy and angular dependence is used in the integral line-beam method to transport the radiation through the atmosphere up to 2500 m from the source. An example spent-fuel storage cask is analyzed with this hybrid method and compared to detailed MCNP skyshine calculations.

Determining the mass attenuation coefficient, effective atomic number, and electron density of raw wood and binderless particleboards of Rhizophora spp. by using Monte Carlo simulation

NASA Astrophysics Data System (ADS)

Marashdeh, Mohammad W.; Al-Hamarneh, Ibrahim F.; Abdel Munem, Eid M.; Tajuddin, A. A.; Ariffin, Alawiah; Al-Omari, Saleh

Rhizophora spp. wood has the potential to serve as a solid water or tissue equivalent phantom for photon and electron beam dosimetry. In this study, the effective atomic number (Zeff) and effective electron density (Neff) of raw wood and binderless Rhizophora spp. particleboards in four different particle sizes were determined in the 10-60 keV energy region. The mass attenuation coefficients used in the calculations were obtained using the Monte Carlo N-Particle (MCNP5) simulation code. The MCNP5 calculations of the attenuation parameters for the Rhizophora spp. samples were plotted graphically against photon energy and discussed in terms of their relative differences compared with those of water and breast tissue. Moreover, the validity of the MCNP5 code was examined by comparing the calculated attenuation parameters with the theoretical values obtained by the XCOM program based on the mixture rule. The results indicated that the MCNP5 process can be followed to determine the attenuation of gamma rays with several photon energies in other materials.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Turner, A.; Davis, A.; University of Wisconsin-Madison, Madison, WI 53706

CCFE perform Monte-Carlo transport simulations on large and complex tokamak models such as ITER. Such simulations are challenging since streaming and deep penetration effects are equally important. In order to make such simulations tractable, both variance reduction (VR) techniques and parallel computing are used. It has been found that the application of VR techniques in such models significantly reduces the efficiency of parallel computation due to 'long histories'. VR in MCNP can be accomplished using energy-dependent weight windows. The weight window represents an 'average behaviour' of particles, and large deviations in the arriving weight of a particle give rise tomore » extreme amounts of splitting being performed and a long history. When running on parallel clusters, a long history can have a detrimental effect on the parallel efficiency - if one process is computing the long history, the other CPUs complete their batch of histories and wait idle. Furthermore some long histories have been found to be effectively intractable. To combat this effect, CCFE has developed an adaptation of MCNP which dynamically adjusts the WW where a large weight deviation is encountered. The method effectively 'de-optimises' the WW, reducing the VR performance but this is offset by a significant increase in parallel efficiency. Testing with a simple geometry has shown the method does not bias the result. This 'long history method' has enabled CCFE to significantly improve the performance of MCNP calculations for ITER on parallel clusters, and will be beneficial for any geometry combining streaming and deep penetration effects. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Huml, O.

The objective of this work was to determine the neutron flux density distribution in various places of the training reactor VR-1 Sparrow. This experiment was performed on the new core design C1, composed of the new low-enriched uranium fuel cells IRT-4M (19.7 %). This fuel replaced the old high-enriched uranium fuel IRT-3M (36 %) within the framework of the RERTR Program in September 2005. The measurement used the neutron activation analysis method with gold wires. The principle of this method consists in neutron capture in a nucleus of the material forming the activation detector. This capture can change the nucleusmore » in a radioisotope, whose activity can be measured. The absorption cross-section values were evaluated by MCNP computer code. The gold wires were irradiated in seven different positions in the core C1. All irradiations were performed at reactor power level 1E8 (1 kW{sub therm}). The activity of segments of irradiated wires was measured by special automatic device called 'Drat' (Wire in English). (author)« less

Effect of electron transport properties on unipolar CdZnTe radiation detectors: LUND, SpectrumPlus, and Coplanar Grid

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ralph B. James

2000-01-07

Device simulations of (1) the laterally-contacted-unipolar-nuclear detector (LUND), (2) the SpectrumPlus, (3) and the coplanar grid made of Cd{sub 0.9}Zn{sub 0.1}Te (CZT) were performed for {sup 137}Cs irradiation by 662.15 keV gamma-rays. Realistic and controlled simulations of the gamma-ray interactions with the CZT material were done using the MCNP4B2 Monte Carlo program, and the detector responses were simulated using the Sandia three-dimensional multielectrode simulation program (SandTMSP). The simulations were done for the best and the worst expected carrier nobilities and lifetimes of currently commercially available CZT materials for radiation detector applications. For the simulated unipolar devices, the active device volumesmore » were relatively large and the energy resolutions were fairly good, but these performance characteristics were found to be very sensitive to the materials properties. The internal electric fields, the weighting potentials, and the charge induced efficiency maps were calculated to give insights into the operation of these devices.« less

Ford Motor Company NDE facility shielding design.

PubMed

Metzger, Robert L; Van Riper, Kenneth A; Jones, Martin H

2005-01-01

Ford Motor Company proposed the construction of a large non-destructive evaluation laboratory for radiography of automotive power train components. The authors were commissioned to design the shielding and to survey the completed facility for compliance with radiation doses for occupationally and non-occupationally exposed personnel. The two X-ray sources are Varian Linatron 3000 accelerators operating at 9-11 MV. One performs computed tomography of automotive transmissions, while the other does real-time radiography of operating engines and transmissions. The shield thickness for the primary barrier and all secondary barriers were determined by point-kernel techniques. Point-kernel techniques did not work well for skyshine calculations and locations where multiple sources (e.g. tube head leakage and various scatter fields) impacted doses. Shielding for these areas was determined using transport calculations. A number of MCNP [Briesmeister, J. F. MCNPCA general Monte Carlo N-particle transport code version 4B. Los Alamos National Laboratory Manual (1997)] calculations focused on skyshine estimates and the office areas. Measurements on the operational facility confirmed the shielding calculations.

26 CFR 1.50B-4 - Partnerships.

Code of Federal Regulations, 2010 CFR

2010-04-01

... 26 Internal Revenue 1 2010-04-01 2010-04-01 true Partnerships. 1.50B-4 Section 1.50B-4 Internal Revenue INTERNAL REVENUE SERVICE, DEPARTMENT OF THE TREASURY INCOME TAX INCOME TAXES Rules for Computing Credit for Expenses of Work Incentive Programs § 1.50B-4 Partnerships. (a) General rule—(1) In general...

Monte Carlo method for calculating the radiation skyshine produced by electron accelerators

NASA Astrophysics Data System (ADS)

Kong, Chaocheng; Li, Quanfeng; Chen, Huaibi; Du, Taibin; Cheng, Cheng; Tang, Chuanxiang; Zhu, Li; Zhang, Hui; Pei, Zhigang; Ming, Shenjin

2005-06-01

Using the MCNP4C Monte Carlo code, the X-ray skyshine produced by 9 MeV, 15 MeV and 21 MeV electron linear accelerators were calculated respectively with a new two-step method combined with the split and roulette variance reduction technique. Results of the Monte Carlo simulation, the empirical formulas used for skyshine calculation and the dose measurements were analyzed and compared. In conclusion, the skyshine dose measurements agreed reasonably with the results computed by the Monte Carlo method, but deviated from computational results given by empirical formulas. The effect on skyshine dose caused by different structures of accelerator head is also discussed in this paper.

Consolidated Cab Display (CCD) System, Project Planning Document (PPD),

DTIC Science & Technology

1981-02-01

1980 1981 I 2 3 4 5 6 7 8 9 10 1112 1 2 3 4 5 6 7 8 9 1011112 1 2 31 12. Software Documentation a. Overall Computer Program Description ( OCPD ) b...Approve OCPD c. Computer Program Functional Specifications (CPFS) d. Data Base Table Design Specification (DBTDS) e. Software Interface Control Document...Parts List Master Pattern and Plan View Reproducible Drawings Instruction Book Training Aids/Materials b. Software: OCPD CPFS SI CD PDS DBTDS SDD

Design of boron carbide-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP.

PubMed

Abrefah, R G; Sogbadji, R B M; Ampomah-Amoako, E; Birikorang, S A; Odoi, H C; Nyarko, B J B

2011-01-01

The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbide-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one boron carbide covered outer irradiation channel; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model, which has a boron carbide-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the boron carbide shielded channel was made. The final effective criticality of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new boron carbide designed model was recorded as 1.00282. Also, a final prompt neutron lifetime of 1.5245 × 10(-4)s was recorded for the new boron carbide designed model while a value of 1.5571 × 10(-7)s was recorded for the original MCNP design of the GHARR-1. Copyright © 2010 Elsevier Ltd. All rights reserved.

Strategies for Dealing with the Defense Budget

DTIC Science & Technology

1983-08-17

changes were computed and are shown in Tables B-3 and B-4, on pages B-66 and B-67. B.6 ACQUISITION PROGRAM TURBULENCE The purpose of this section is to...planning. The following brief overviev / of the NAVMAT study illustrates this cause of program turbulence. Figure C-13 shows the NAVMAT analytical...Inflation, Industrial Base, Life-Cycle Costs, Material Acquisition, Material Balance, Multi-year Contracting/procurement, Planning, Programming and

SABRINA - An interactive geometry modeler for MCNP (Monte Carlo Neutron Photon)

DOE Office of Scientific and Technical Information (OSTI.GOV)

West, J.T.; Murphy, J.

SABRINA is an interactive three-dimensional geometry modeler developed to produce complicated models for the Los Alamos Monte Carlo Neutron Photon program MCNP. SABRINA produces line drawings and color-shaded drawings for a wide variety of interactive graphics terminals. It is used as a geometry preprocessor in model development and as a Monte Carlo particle-track postprocessor in the visualization of complicated particle transport problem. SABRINA is written in Fortran 77 and is based on the Los Alamos Common Graphics System, CGS. 5 refs., 2 figs.

Covariance Data File Formats for Whisper-1.0 & Whisper-1.1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Forrest B.; Rising, Michael Evan

2017-01-09

Whisper is a statistical analysis package developed in 2014 to support nuclear criticality safety (NCS) validation. It uses the sensitivity profile data for an application as computed by MCNP6 along with covariance files for the nuclear data to determine a baseline upper-subcritical-limit (USL) for the application. Whisper version 1.0 was first developed and used at LANL in 2014. During 2015-2016, Whisper was updated to version 1.1 and is to be included with the upcoming release of MCNP6.2. This report describes the file formats used for the covariance data in both Whisper-1.0 and Whisper-1.1.

Benchmark study for total enery electrons in thick slabs

NASA Technical Reports Server (NTRS)

Jun, I.

2002-01-01

The total energy deposition profiles when highenergy electrons impinge on a thick slab of elemental aluminum, copper, and tungsten have been computed using representative Monte Carlo codes (NOVICE, TIGER, MCNP), and compared in this paper.

Dosimetric comparison of Monte Carlo codes (EGS4, MCNP, MCNPX) considering external and internal exposures of the Zubal phantom to electron and photon sources.

PubMed

Chiavassa, S; Lemosquet, A; Aubineau-Lanièce, I; de Carlan, L; Clairand, I; Ferrer, L; Bardiès, M; Franck, D; Zankl, M

2005-01-01

This paper aims at comparing dosimetric assessments performed with three Monte Carlo codes: EGS4, MCNP4c2 and MCNPX2.5e, using a realistic voxel phantom, namely the Zubal phantom, in two configurations of exposure. The first one deals with an external irradiation corresponding to the example of a radiological accident. The results are obtained using the EGS4 and the MCNP4c2 codes and expressed in terms of the mean absorbed dose (in Gy per source particle) for brain, lungs, liver and spleen. The second one deals with an internal exposure corresponding to the treatment of a medullary thyroid cancer by 131I-labelled radiopharmaceutical. The results are obtained by EGS4 and MCNPX2.5e and compared in terms of S-values (expressed in mGy per kBq and per hour) for liver, kidney, whole body and thyroid. The results of these two studies are presented and differences between the codes are analysed and discussed.

Improved radial dose function estimation using current version MCNP Monte-Carlo simulation: Model 6711 and ISC3500 125I brachytherapy sources.

PubMed

Duggan, Dennis M

2004-12-01

Improved cross-sections in a new version of the Monte-Carlo N-particle (MCNP) code may eliminate discrepancies between radial dose functions (as defined by American Association of Physicists in Medicine Task Group 43) derived from Monte-Carlo simulations of low-energy photon-emitting brachytherapy sources and those from measurements on the same sources with thermoluminescent dosimeters. This is demonstrated for two 125I brachytherapy seed models, the Implant Sciences Model ISC3500 (I-Plant) and the Amersham Health Model 6711, by simulating their radial dose functions with two versions of MCNP, 4c2 and 5.

Validation of MCNP: SPERT-D and BORAX-V fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crawford, C.; Palmer, B.

1992-11-01

This report discusses critical experiments involving SPERT-D{sup 1,2} fuel elements and BORAX-V{sup 3-8} fuel which have been modeled and calculations performed with MCNP. MCNP is a Monte Carlo based transport code. For this study continuous-energy nuclear data from the ENDF/B-V cross section library was used. The SPERT-D experiments consisted of various arrays of fuel elements moderated and reflected with either water or a uranyl nitrate solution. Some SPERT-D experiments used cadmium as a fixed neutron poison, while others were poisoned with various concentrations of boron in the moderating/reflecting solution. ne BORAX-V experiments were arrays of either boiling fuel rod assembliesmore » or superheater assemblies, both types of arrays were moderated and reflected with water. In one boiling fuel experiment, two fuel rods were replaced with borated stainless steel poison rods.« less

Validation of MCNP: SPERT-D and BORAX-V fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crawford, C.; Palmer, B.

1992-11-01

This report discusses critical experiments involving SPERT-D[sup 1,2] fuel elements and BORAX-V[sup 3-8] fuel which have been modeled and calculations performed with MCNP. MCNP is a Monte Carlo based transport code. For this study continuous-energy nuclear data from the ENDF/B-V cross section library was used. The SPERT-D experiments consisted of various arrays of fuel elements moderated and reflected with either water or a uranyl nitrate solution. Some SPERT-D experiments used cadmium as a fixed neutron poison, while others were poisoned with various concentrations of boron in the moderating/reflecting solution. ne BORAX-V experiments were arrays of either boiling fuel rod assembliesmore » or superheater assemblies, both types of arrays were moderated and reflected with water. In one boiling fuel experiment, two fuel rods were replaced with borated stainless steel poison rods.« less

The use of tetrahedral mesh geometries in Monte Carlo simulation of applicator based brachytherapy dose distributions

NASA Astrophysics Data System (ADS)

Paiva Fonseca, Gabriel; Landry, Guillaume; White, Shane; D'Amours, Michel; Yoriyaz, Hélio; Beaulieu, Luc; Reniers, Brigitte; Verhaegen, Frank

2014-10-01

Accounting for brachytherapy applicator attenuation is part of the recommendations from the recent report of AAPM Task Group 186. To do so, model based dose calculation algorithms require accurate modelling of the applicator geometry. This can be non-trivial in the case of irregularly shaped applicators such as the Fletcher Williamson gynaecological applicator or balloon applicators with possibly irregular shapes employed in accelerated partial breast irradiation (APBI) performed using electronic brachytherapy sources (EBS). While many of these applicators can be modelled using constructive solid geometry (CSG), the latter may be difficult and time-consuming. Alternatively, these complex geometries can be modelled using tessellated geometries such as tetrahedral meshes (mesh geometries (MG)). Recent versions of Monte Carlo (MC) codes Geant4 and MCNP6 allow for the use of MG. The goal of this work was to model a series of applicators relevant to brachytherapy using MG. Applicators designed for 192Ir sources and 50 kV EBS were studied; a shielded vaginal applicator, a shielded Fletcher Williamson applicator and an APBI balloon applicator. All applicators were modelled in Geant4 and MCNP6 using MG and CSG for dose calculations. CSG derived dose distributions were considered as reference and used to validate MG models by comparing dose distribution ratios. In general agreement within 1% for the dose calculations was observed for all applicators between MG and CSG and between codes when considering volumes inside the 25% isodose surface. When compared to CSG, MG required longer computation times by a factor of at least 2 for MC simulations using the same code. MCNP6 calculation times were more than ten times shorter than Geant4 in some cases. In conclusion we presented methods allowing for high fidelity modelling with results equivalent to CSG. To the best of our knowledge MG offers the most accurate representation of an irregular APBI balloon applicator.

Use of Transportable Radiation Detection Instruments to Assess Internal Contamination from Intakes of Radionuclides Part II: Calibration Factors and ICAT Computer Program.

PubMed

Anigstein, Robert; Olsher, Richard H; Loomis, Donald A; Ansari, Armin

2016-12-01

The detonation of a radiological dispersion device or other radiological incidents could result in widespread releases of radioactive materials and intakes of radionuclides by affected individuals. Transportable radiation monitoring instruments could be used to measure radiation from gamma-emitting radionuclides in the body for triaging individuals and assigning priorities to their bioassay samples for in vitro assessments. The present study derived sets of calibration factors for four instruments: the Ludlum Model 44-2 gamma scintillator, a survey meter containing a 2.54 × 2.54-cm NaI(Tl) crystal; the Captus 3000 thyroid uptake probe, which contains a 5.08 × 5.08-cm NaI(Tl) crystal; the Transportable Portal Monitor Model TPM-903B, which contains two 3.81 × 7.62 × 182.9-cm polyvinyltoluene plastic scintillators; and a generic instrument, such as an ionization chamber, that measures exposure rates. The calibration factors enable these instruments to be used for assessing inhaled or ingested intakes of any of four radionuclides: Co, I, Cs, and Ir. The derivations used biokinetic models embodied in the DCAL computer software system developed by the Oak Ridge National Laboratory and Monte Carlo simulations using the MCNPX radiation transport code. The three physical instruments were represented by MCNP models that were developed previously. The affected individuals comprised children of five ages who were represented by the revised Oak Ridge National Laboratory pediatric phantoms, and adult men and adult women represented by the Adult Reference Computational Phantoms described in Publication 110 of the International Commission on Radiological Protection. These calibration factors can be used to calculate intakes; the intakes can be converted to committed doses by the use of tabulated dose coefficients. These calibration factors also constitute input data to the ICAT computer program, an interactive Microsoft Windows-based software package that estimates intakes of radionuclides and cumulative and committed effective doses, based on measurements made with these instruments. This program constitutes a convenient tool for assessing intakes and doses without consulting tabulated calibration factors and dose coefficients.

USE OF TRANSPORTABLE RADIATION DETECTION INSTRUMENTS TO ASSESS INTERNAL CONTAMINATION FROM INTAKES OF RADIONUCLIDES PART II: CALIBRATION FACTORS AND ICAT COMPUTER PROGRAM

PubMed Central

Anigstein, Robert; Olsher, Richard H.; Loomis, Donald A.; Ansari, Armin

2017-01-01

The detonation of a radiological dispersion device or other radiological incidents could result in widespread releases of radioactive materials and intakes of radionuclides by affected individuals. Transportable radiation monitoring instruments could be used to measure radiation from gamma-emitting radionuclides in the body for triaging individuals and assigning priorities to their bioassay samples for in vitro assessments. The present study derived sets of calibration factors for four instruments: the Ludlum Model 44-2 gamma scintillator, a survey meter containing a 2.54 × 2.54-cm NaI(Tl) crystal; the Captus 3000 thyroid uptake probe, which contains a 5.08 × 5.08-cm NaI(Tl) crystal; the Transportable Portal Monitor Model TPM-903B, which contains two 3.81 × 7.62 × 182.9-cm polyvinyltoluene plastic scintillators; and a generic instrument, such as an ionization chamber, that measures exposure rates. The calibration factors enable these instruments to be used for assessing inhaled or ingested intakes of any of four radionuclides: 60Co, 131I, 137Cs, and 192Ir. The derivations used biokinetic models embodied in the DCAL computer software system developed by the Oak Ridge National Laboratory and Monte Carlo simulations using the MCNPX radiation transport code. The three physical instruments were represented by MCNP models that were developed previously. The affected individuals comprised children of five ages who were represented by the revised Oak Ridge National Laboratory pediatric phantoms, and adult men and adult women represented by the Adult Reference Computational Phantoms described in Publication 110 of the International Commission on Radiological Protection. These calibration factors can be used to calculate intakes; the intakes can be converted to committed doses by the use of tabulated dose coefficients. These calibration factors also constitute input data to the ICAT computer program, an interactive Microsoft Windows-based software package that estimates intakes of radionuclides and cumulative and committed effective doses, based on measurements made with these instruments. This program constitutes a convenient tool for assessing intakes and doses without consulting tabulated calibration factors and dose coefficients. PMID:27798478

Testing of ENDF71x: A new ACE-formatted neutron data library based on ENDF/B-VII.1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gardiner, S. J.; Conlin, J. L.; Kiedrowski, B. C.

The ENDF71x library [1] is the most thoroughly tested set of ACE-format data tables ever released by the Nuclear Data Team at Los Alamos National Laboratory (LANL). It is based on ENDF/B-VII. 1, the most recently released set of evaluated nuclear data files produced by the US Cross Section Evaluation Working Group (CSEWG). A variety of techniques were used to test and verify the ENDF7 1x library before its public release. These include the use of automated checking codes written by members of the Nuclear Data Team, visual inspections of key neutron data, MCNP6 calculations designed to test data formore » every included combination of isotope and temperature as comprehensively as possible, and direct comparisons between ENDF71x and previous ACE library releases. Visual inspection of some of the most important neutron data revealed energy balance problems and unphysical discontinuities in the cross sections for some nuclides. Doppler broadening of the total cross sections with increasing temperature was found to be qualitatively correct. Test calculations performed using MCNP prompted two modifications to the MCNP6 source code and also exposed bad secondary neutron yields for {sup 231,233}Pa that are present in both ENDF/B-VII.1 and ENDF/B-VII.0. A comparison of ENDF71x with its predecessor ACE library, ENDF70, showed that dramatic changes have been made in the neutron cross section data for a number of isotopes between ENDF/B-VII.0 and ENDF/B-VII.1. Based on the results of these verification tests and the validation tests performed by Kahler, et al. [2], the ENDF71x library is recommended for use in all Monte Carlo applications. (authors)« less

MONTE CARLO SIMULATIONS OF PERIODIC PULSED REACTOR WITH MOVING GEOMETRY PARTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cao, Yan; Gohar, Yousry

2015-11-01

In a periodic pulsed reactor, the reactor state varies periodically from slightly subcritical to slightly prompt supercritical for producing periodic power pulses. Such periodic state change is accomplished by a periodic movement of specific reactor parts, such as control rods or reflector sections. The analysis of such reactor is difficult to perform with the current reactor physics computer programs. Based on past experience, the utilization of the point kinetics approximations gives considerable errors in predicting the magnitude and the shape of the power pulse if the reactor has significantly different neutron life times in different zones. To accurately simulate themore » dynamics of this type of reactor, a Monte Carlo procedure using the transfer function TRCL/TR of the MCNP/MCNPX computer programs is utilized to model the movable reactor parts. In this paper, two algorithms simulating the geometry part movements during a neutron history tracking have been developed. Several test cases have been developed to evaluate these procedures. The numerical test cases have shown that the developed algorithms can be utilized to simulate the reactor dynamics with movable geometry parts.« less

Inter-comparison of Dose Distributions Calculated by FLUKA, GEANT4, MCNP, and PHITS for Proton Therapy

NASA Astrophysics Data System (ADS)

Yang, Zi-Yi; Tsai, Pi-En; Lee, Shao-Chun; Liu, Yen-Chiang; Chen, Chin-Cheng; Sato, Tatsuhiko; Sheu, Rong-Jiun

2017-09-01

The dose distributions from proton pencil beam scanning were calculated by FLUKA, GEANT4, MCNP, and PHITS, in order to investigate their applicability in proton radiotherapy. The first studied case was the integrated depth dose curves (IDDCs), respectively from a 100 and a 226-MeV proton pencil beam impinging a water phantom. The calculated IDDCs agree with each other as long as each code employs 75 eV for the ionization potential of water. The second case considered a similar condition of the first case but with proton energies in a Gaussian distribution. The comparison to the measurement indicates the inter-code differences might not only due to different stopping power but also the nuclear physics models. How the physics parameter setting affect the computation time was also discussed. In the third case, the applicability of each code for pencil beam scanning was confirmed by delivering a uniform volumetric dose distribution based on the treatment plan, and the results showed general agreement between each codes, the treatment plan, and the measurement, except that some deviations were found in the penumbra region. This study has demonstrated that the selected codes are all capable of performing dose calculations for therapeutic scanning proton beams with proper physics settings.

Monte Carlo simulations and benchmark measurements on the response of TE(TE) and Mg(Ar) ionization chambers in photon, electron and neutron beams

NASA Astrophysics Data System (ADS)

Lin, Yi-Chun; Huang, Tseng-Te; Liu, Yuan-Hao; Chen, Wei-Lin; Chen, Yen-Fu; Wu, Shu-Wei; Nievaart, Sander; Jiang, Shiang-Huei

2015-06-01

The paired ionization chambers (ICs) technique is commonly employed to determine neutron and photon doses in radiology or radiotherapy neutron beams, where neutron dose shows very strong dependence on the accuracy of accompanying high energy photon dose. During the dose derivation, it is an important issue to evaluate the photon and electron response functions of two commercially available ionization chambers, denoted as TE(TE) and Mg(Ar), used in our reactor based epithermal neutron beam. Nowadays, most perturbation corrections for accurate dose determination and many treatment planning systems are based on the Monte Carlo technique. We used general purposed Monte Carlo codes, MCNP5, EGSnrc, FLUKA or GEANT4 for benchmark verifications among them and carefully measured values for a precise estimation of chamber current from absorbed dose rate of cavity gas. Also, energy dependent response functions of two chambers were calculated in a parallel beam with mono-energies from 20 keV to 20 MeV photons and electrons by using the optimal simple spherical and detailed IC models. The measurements were performed in the well-defined (a) four primary M-80, M-100, M120 and M150 X-ray calibration fields, (b) primary 60Co calibration beam, (c) 6 MV and 10 MV photon, (d) 6 MeV and 18 MeV electron LINACs in hospital and (e) BNCT clinical trials neutron beam. For the TE(TE) chamber, all codes were almost identical over the whole photon energy range. In the Mg(Ar) chamber, MCNP5 showed lower response than other codes for photon energy region below 0.1 MeV and presented similar response above 0.2 MeV (agreed within 5% in the simple spherical model). With the increase of electron energy, the response difference between MCNP5 and other codes became larger in both chambers. Compared with the measured currents, MCNP5 had the difference from the measurement data within 5% for the 60Co, 6 MV, 10 MV, 6 MeV and 18 MeV LINACs beams. But for the Mg(Ar) chamber, the derivations reached 7.8-16.5% below 120 kVp X-ray beams. In this study, we were especially interested in BNCT doses where low energy photon contribution is less to ignore, MCNP model is recognized as the most suitable to simulate wide photon-electron and neutron energy distributed responses of the paired ICs. Also, MCNP provides the best prediction of BNCT source adjustment by the detector's neutron and photon responses.

Cellular dosimetry of (111)In using monte carlo N-particle computer code: comparison with analytic methods and correlation with in vitro cytotoxicity.

PubMed

Cai, Zhongli; Pignol, Jean-Philippe; Chan, Conrad; Reilly, Raymond M

2010-03-01

Our objective was to compare Monte Carlo N-particle (MCNP) self- and cross-doses from (111)In to the nucleus of breast cancer cells with doses calculated by reported analytic methods (Goddu et al. and Farragi et al.). A further objective was to determine whether the MCNP-predicted surviving fraction (SF) of breast cancer cells exposed in vitro to (111)In-labeled diethylenetriaminepentaacetic acid human epidermal growth factor ((111)In-DTPA-hEGF) could accurately predict the experimentally determined values. MCNP was used to simulate the transport of electrons emitted by (111)In from the cell surface, cytoplasm, or nucleus. The doses to the nucleus per decay (S values) were calculated for single cells, closely packed monolayer cells, or cell clusters. The cell and nucleus dimensions of 6 breast cancer cell lines were measured, and cell line-specific S values were calculated. For self-doses, MCNP S values of nucleus to nucleus agreed very well with those of Goddu et al. (ratio of S values using analytic methods vs. MCNP = 0.962-0.995) and Faraggi et al. (ratio = 1.011-1.024). MCNP S values of cytoplasm and cell surface to nucleus compared fairly well with the reported values (ratio = 0.662-1.534 for Goddu et al.; 0.944-1.129 for Faraggi et al.). For cross doses, the S values to the nucleus were independent of (111)In subcellular distribution but increased with cluster size. S values for monolayer cells were significantly different from those of single cells and cell clusters. The MCNP-predicted SF for monolayer MDA-MB-468, MDA-MB-231, and MCF-7 cells agreed with the experimental data (relative error of 3.1%, -1.0%, and 1.7%). The single-cell and cell cluster models were less accurate in predicting the SF. For MDA-MB-468 cells, relative error was 8.1% using the single-cell model and -54% to -67% using the cell cluster model. Individual cell-line dimensions had large effects on S values and were needed to estimate doses and SF accurately. MCNP simulation compared well with the reported analytic methods in the calculation of subcellular S values for single cells and cell clusters. Application of a monolayer model was most accurate in predicting the SF of breast cancer cells exposed in vitro to (111)In-DTPA-hEGF.

Adjoint acceleration of Monte Carlo simulations using TORT/MCNP coupling approach: a case study on the shielding improvement for the cyclotron room of the Buddhist Tzu Chi General Hospital.

PubMed

Sheu, R J; Sheu, R D; Jiang, S H; Kao, C H

2005-01-01

Full-scale Monte Carlo simulations of the cyclotron room of the Buddhist Tzu Chi General Hospital were carried out to improve the original inadequate maze design. Variance reduction techniques are indispensable in this study to facilitate the simulations for testing a variety of configurations of shielding modification. The TORT/MCNP manual coupling approach based on the Consistent Adjoint Driven Importance Sampling (CADIS) methodology has been used throughout this study. The CADIS utilises the source and transport biasing in a consistent manner. With this method, the computational efficiency was increased significantly by more than two orders of magnitude and the statistical convergence was also improved compared to the unbiased Monte Carlo run. This paper describes the shielding problem encountered, the procedure for coupling the TORT and MCNP codes to accelerate the calculations and the calculation results for the original and improved shielding designs. In order to verify the calculation results and seek additional accelerations, sensitivity studies on the space-dependent and energy-dependent parameters were also conducted.

Conversion coefficients for determination of dispersed photon dose during radiotherapy: NRUrad input code for MCNP.

PubMed

Shahmohammadi Beni, Mehrdad; Ng, C Y P; Krstic, D; Nikezic, D; Yu, K N

2017-01-01

Radiotherapy is a common cancer treatment module, where a certain amount of dose will be delivered to the targeted organ. This is achieved usually by photons generated by linear accelerator units. However, radiation scattering within the patient's body and the surrounding environment will lead to dose dispersion to healthy tissues which are not targets of the primary radiation. Determination of the dispersed dose would be important for assessing the risk and biological consequences in different organs or tissues. In the present work, the concept of conversion coefficient (F) of the dispersed dose was developed, in which F = (Dd/Dt), where Dd was the dispersed dose in a non-targeted tissue and Dt is the absorbed dose in the targeted tissue. To quantify Dd and Dt, a comprehensive model was developed using the Monte Carlo N-Particle (MCNP) package to simulate the linear accelerator head, the human phantom, the treatment couch and the radiotherapy treatment room. The present work also demonstrated the feasibility and power of parallel computing through the use of the Message Passing Interface (MPI) version of MCNP5.

Conversion coefficients for determination of dispersed photon dose during radiotherapy: NRUrad input code for MCNP

PubMed Central

Krstic, D.; Nikezic, D.

2017-01-01

Radiotherapy is a common cancer treatment module, where a certain amount of dose will be delivered to the targeted organ. This is achieved usually by photons generated by linear accelerator units. However, radiation scattering within the patient’s body and the surrounding environment will lead to dose dispersion to healthy tissues which are not targets of the primary radiation. Determination of the dispersed dose would be important for assessing the risk and biological consequences in different organs or tissues. In the present work, the concept of conversion coefficient (F) of the dispersed dose was developed, in which F = (Dd/Dt), where Dd was the dispersed dose in a non-targeted tissue and Dt is the absorbed dose in the targeted tissue. To quantify Dd and Dt, a comprehensive model was developed using the Monte Carlo N-Particle (MCNP) package to simulate the linear accelerator head, the human phantom, the treatment couch and the radiotherapy treatment room. The present work also demonstrated the feasibility and power of parallel computing through the use of the Message Passing Interface (MPI) version of MCNP5. PMID:28362837

Recent advances and future prospects for Monte Carlo

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Forrest B

2010-01-01

The history of Monte Carlo methods is closely linked to that of computers: The first known Monte Carlo program was written in 1947 for the ENIAC; a pre-release of the first Fortran compiler was used for Monte Carlo In 1957; Monte Carlo codes were adapted to vector computers in the 1980s, clusters and parallel computers in the 1990s, and teraflop systems in the 2000s. Recent advances include hierarchical parallelism, combining threaded calculations on multicore processors with message-passing among different nodes. With the advances In computmg, Monte Carlo codes have evolved with new capabilities and new ways of use. Production codesmore » such as MCNP, MVP, MONK, TRIPOLI and SCALE are now 20-30 years old (or more) and are very rich in advanced featUres. The former 'method of last resort' has now become the first choice for many applications. Calculations are now routinely performed on office computers, not just on supercomputers. Current research and development efforts are investigating the use of Monte Carlo methods on FPGAs. GPUs, and many-core processors. Other far-reaching research is exploring ways to adapt Monte Carlo methods to future exaflop systems that may have 1M or more concurrent computational processes.« less

The incorporation of plotting capability into the Unified Subsonic Supersonic Aerodynamic Analysis program, version B

NASA Technical Reports Server (NTRS)

Winter, O. A.

1980-01-01

The B01 version of the United Subsonic Supersonic Aerodynamic Analysis program is the result of numerous modifications and additions made to the B00 version. These modifications and additions affect the program input, its computational options, the code readability, and the overlay structure. The following are described: (1) the revised input; (2) the plotting overlay programs which were also modified, and their associated subroutines, (3) the auxillary files used by the program, the revised output data; and (4) the program overlay structure.

Simulation of image detectors in radiology for determination of scatter-to-primary ratios using Monte Carlo radiation transport code MCNP/MCNPX.

PubMed

Smans, Kristien; Zoetelief, Johannes; Verbrugge, Beatrijs; Haeck, Wim; Struelens, Lara; Vanhavere, Filip; Bosmans, Hilde

2010-05-01

The purpose of this study was to compare and validate three methods to simulate radiographic image detectors with the Monte Carlo software MCNP/MCNPX in a time efficient way. The first detector model was the standard semideterministic radiography tally, which has been used in previous image simulation studies. Next to the radiography tally two alternative stochastic detector models were developed: A perfect energy integrating detector and a detector based on the energy absorbed in the detector material. Validation of three image detector models was performed by comparing calculated scatter-to-primary ratios (SPRs) with the published and experimentally acquired SPR values. For mammographic applications, SPRs computed with the radiography tally were up to 44% larger than the published results, while the SPRs computed with the perfect energy integrating detectors and the blur-free absorbed energy detector model were, on the average, 0.3% (ranging from -3% to 3%) and 0.4% (ranging from -5% to 5%) lower, respectively. For general radiography applications, the radiography tally overestimated the measured SPR by as much as 46%. The SPRs calculated with the perfect energy integrating detectors were, on the average, 4.7% (ranging from -5.3% to -4%) lower than the measured SPRs, whereas for the blur-free absorbed energy detector model, the calculated SPRs were, on the average, 1.3% (ranging from -0.1% to 2.4%) larger than the measured SPRs. For mammographic applications, both the perfect energy integrating detector model and the blur-free energy absorbing detector model can be used to simulate image detectors, whereas for conventional x-ray imaging using higher energies, the blur-free energy absorbing detector model is the most appropriate image detector model. The radiography tally overestimates the scattered part and should therefore not be used to simulate radiographic image detectors.

Plutonium Critical Mass Curve Comparison to Mass at Upper Subcritical Limit (USL) Using Whisper

DOE Office of Scientific and Technical Information (OSTI.GOV)

Alwin, Jennifer Louise; Zhang, Ning

Whisper is computational software designed to assist the nuclear criticality safety analyst with validation studies with the MCNP ® Monte Carlo radiation transport package. Standard approaches to validation rely on the selection of benchmarks based upon expert judgment. Whisper uses sensitivity/uncertainty (S/U) methods to select relevant benchmarks to a particular application or set of applications being analyzed. Using these benchmarks, Whisper computes a calculational margin. Whisper attempts to quantify the margin of subcriticality (MOS) from errors in software and uncertainties in nuclear data. The combination of the Whisper-derived calculational margin and MOS comprise the baseline upper subcritical limit (USL), tomore » which an additional margin may be applied by the nuclear criticality safety analyst as appropriate to ensure subcriticality. A series of critical mass curves for plutonium, similar to those found in Figure 31 of LA-10860-MS, have been generated using MCNP6.1.1 and the iterative parameter study software, WORM_Solver. The baseline USL for each of the data points of the curves was then computed using Whisper 1.1. The USL was then used to determine the equivalent mass for plutonium metal-water system. ANSI/ANS-8.1 states that it is acceptable to use handbook data, such as the data directly from the LA-10860-MS, as it is already considered validated (Section 4.3 4) “Use of subcritical limit data provided in ANSI/ANS standards or accepted reference publications does not require further validation.”). This paper attempts to take a novel approach to visualize traditional critical mass curves and allows comparison with the amount of mass for which the k eff is equal to the USL (calculational margin + margin of subcriticality). However, the intent is to plot the critical mass data along with USL, not to suggest that already accepted handbook data should have new and more rigorous requirements for validation.« less

Elaborate SMART MCNP Modelling Using ANSYS and Its Applications

NASA Astrophysics Data System (ADS)

Song, Jaehoon; Surh, Han-bum; Kim, Seung-jin; Koo, Bonsueng

2017-09-01

An MCNP 3-dimensional model can be widely used to evaluate various design parameters such as a core design or shielding design. Conventionally, a simplified 3-dimensional MCNP model is applied to calculate these parameters because of the cumbersomeness of modelling by hand. ANSYS has a function for converting the CAD `stp' format into an MCNP input in the geometry part. Using ANSYS and a 3- dimensional CAD file, a very detailed and sophisticated MCNP 3-dimensional model can be generated. The MCNP model is applied to evaluate the assembly weighting factor at the ex-core detector of SMART, and the result is compared with a simplified MCNP SMART model and assembly weighting factor calculated by DORT, which is a deterministic Sn code.

TRIPOLI-4® - MCNP5 ITER A-lite neutronic model benchmarking

NASA Astrophysics Data System (ADS)

Jaboulay, J.-C.; Cayla, P.-Y.; Fausser, C.; Lee, Y.-K.; Trama, J.-C.; Li-Puma, A.

2014-06-01

The aim of this paper is to present the capability of TRIPOLI-4®, the CEA Monte Carlo code, to model a large-scale fusion reactor with complex neutron source and geometry. In the past, numerous benchmarks were conducted for TRIPOLI-4® assessment on fusion applications. Experiments (KANT, OKTAVIAN, FNG) analysis and numerical benchmarks (between TRIPOLI-4® and MCNP5) on the HCLL DEMO2007 and ITER models were carried out successively. In this previous ITER benchmark, nevertheless, only the neutron wall loading was analyzed, its main purpose was to present MCAM (the FDS Team CAD import tool) extension for TRIPOLI-4®. Starting from this work a more extended benchmark has been performed about the estimation of neutron flux, nuclear heating in the shielding blankets and tritium production rate in the European TBMs (HCLL and HCPB) and it is presented in this paper. The methodology to build the TRIPOLI-4® A-lite model is based on MCAM and the MCNP A-lite model (version 4.1). Simplified TBMs (from KIT) have been integrated in the equatorial-port. Comparisons of neutron wall loading, flux, nuclear heating and tritium production rate show a good agreement between the two codes. Discrepancies are mainly included in the Monte Carlo codes statistical error.

MCNP Version 6.2 Release Notes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Werner, Christopher John; Bull, Jeffrey S.; Solomon, C. J.

Monte Carlo N-Particle or MCNP ® is a general-purpose Monte Carlo radiation-transport code designed to track many particle types over broad ranges of energies. This MCNP Version 6.2 follows the MCNP6.1.1 beta version and has been released in order to provide the radiation transport community with the latest feature developments and bug fixes for MCNP. Since the last release of MCNP major work has been conducted to improve the code base, add features, and provide tools to facilitate ease of use of MCNP version 6.2 as well as the analysis of results. These release notes serve as a general guidemore » for the new/improved physics, source, data, tallies, unstructured mesh, code enhancements and tools. For more detailed information on each of the topics, please refer to the appropriate references or the user manual which can be found at http://mcnp.lanl.gov. This release of MCNP version 6.2 contains 39 new features in addition to 172 bug fixes and code enhancements. There are still some 33 known issues the user should familiarize themselves with (see Appendix).« less

Effect of a dual-purpose cask payload increment of spent fuel assemblies from VVER 1000 Bushehr Nuclear Power Plant on basket criticality.

PubMed

Rezaeian, M; Kamali, J

2017-01-01

Dual-purpose casks can be utilized for dry interim storage and transportation of the highly radioactive spent fuel assemblies (SFAs) of Bushehr Nuclear Power Plant (NPP). Criticality safety analysis was carried out using the MCNP code for the cask containing 12, 18, or 19 SFAs. The basket materials of borated stainless steel and Boral (Al-B 4 C) were investigated, and the minimum required receptacle pitch of the basket was determined. Copyright © 2016 Elsevier Ltd. All rights reserved.

Computer Operations Study of Reservoir Operations for Six Mississippi River Headwaters Dams. Appendix A.

DTIC Science & Technology

1982-06-01

p*A C.._ _ __ _ _ A, d.tibutiou is unhimta 4 iit 84~ L0 TABLE OF CONTENTS APPENDIX SCOPE OF WORK B MERGE AND COST PROGRAM DOCUMENTATION C FATSCO... PROGRAM TO COMPUTE TIME SERIES FREQUENCY RELATIONSHIPS D HEC-DSS - TIME SERIES DATA FILE MANAGEMENT SYSTEM E PLAN 1 -TIM SERIES DATA PLOTS AND ANNUAL...University of Minnesota, utilized an early version of the Hydrologic Engineering * Center’s (HEC) EEC-5c Computer Program . EEC is a Corps of Engineers

Mathematical simulations of photon interactions using Monte Carlo analysis to evaluate the uncertainty associated with in vivo K X-ray fluorescence measurements of stable lead in bone

NASA Astrophysics Data System (ADS)

Lodwick, Camille J.

This research utilized Monte Carlo N-Particle version 4C (MCNP4C) to simulate K X-ray fluorescent (K XRF) measurements of stable lead in bone. Simulations were performed to investigate the effects that overlying tissue thickness, bone-calcium content, and shape of the calibration standard have on detector response in XRF measurements at the human tibia. Additional simulations of a knee phantom considered uncertainty associated with rotation about the patella during XRF measurements. Simulations tallied the distribution of energy deposited in a high-purity germanium detector originating from collimated 88 keV 109Cd photons in backscatter geometry. Benchmark measurements were performed on simple and anthropometric XRF calibration phantoms of the human leg and knee developed at the University of Cincinnati with materials proven to exhibit radiological characteristics equivalent to human tissue and bone. Initial benchmark comparisons revealed that MCNP4C limits coherent scatter of photons to six inverse angstroms of momentum transfer and a Modified MCNP4C was developed to circumvent the limitation. Subsequent benchmark measurements demonstrated that Modified MCNP4C adequately models photon interactions associated with in vivo K XRF of lead in bone. Further simulations of a simple leg geometry possessing tissue thicknesses from 0 to 10 mm revealed increasing overlying tissue thickness from 5 to 10 mm reduced predicted lead concentrations an average 1.15% per 1 mm increase in tissue thickness (p < 0.0001). An anthropometric leg phantom was mathematically defined in MCNP to more accurately reflect the human form. A simulated one percent increase in calcium content (by mass) of the anthropometric leg phantom's cortical bone demonstrated to significantly reduce the K XRF normalized ratio by 4.5% (p < 0.0001). Comparison of the simple and anthropometric calibration phantoms also suggested that cylindrical calibration standards can underestimate lead content of a human leg up to 4%. The patellar bone structure in which the fluorescent photons originate was found to vary dramatically with measurement angle. The relative contribution of lead signal from the patella declined from 65% to 27% when rotated 30°. However, rotation of the source-detector about the patella from 0 to 45° demonstrated no significant effect on the net K XRF response at the knee.

Monte Carlo simulation of x-ray buildup factors of lead and its applications in shielding of diagnostic x-ray facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kharrati, Hedi; Agrebi, Amel; Karaoui, Mohamed-Karim

2007-04-15

X-ray buildup factors of lead in broad beam geometry for energies from 15 to 150 keV are determined using the general purpose Monte Carlo N-particle radiation transport computer code (MCNP4C). The obtained buildup factors data are fitted to a modified three parameter Archer et al. model for ease in calculating the broad beam transmission with computer at any tube potentials/filters combinations in diagnostic energies range. An example for their use to compute the broad beam transmission at 70, 100, 120, and 140 kVp is given. The calculated broad beam transmission is compared to data derived from literature, presenting good agreement.more » Therefore, the combination of the buildup factors data as determined and a mathematical model to generate x-ray spectra provide a computationally based solution to broad beam transmission for lead barriers in shielding x-ray facilities.« less

Instructions for the use of the CIVM-Jet 4C finite-strain computer code to calculate the transient structural responses of partial and/or complete arbitrarily-curved rings subjected to fragment impact

NASA Technical Reports Server (NTRS)

Rodal, J. J. A.; French, S. E.; Witmer, E. A.; Stagliano, T. R.

1979-01-01

The CIVM-JET 4C computer program for the 'finite strain' analysis of 2 d transient structural responses of complete or partial rings and beams subjected to fragment impact stored on tape as a series of individual files. Which subroutines are found in these files are described in detail. All references to the CIVM-JET 4C program are made assuming that the user has a copy of NASA CR-134907 (ASRL TR 154-9) which serves as a user's guide to (1) the CIVM-JET 4B computer code and (2) the CIVM-JET 4C computer code 'with the use of the modified input instructions' attached hereto.

Neutronics Analyses of the Minimum Original HEU TREAT Core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kontogeorgakos, D.; Connaway, H.; Yesilyurt, G.

2014-04-01

This work was performed to support the feasibility study on the potential conversion of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory from the use of high-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by the GTRI Reactor Conversion staff at the Argonne National Laboratory (ANL). The objective of this study was to validate the MCNP model of the TREAT reactor with the well-documented measurements which were taken during the start-up and early operation of TREAT. Furthermore, the effect of carbon graphitization was also addressed. The graphitization level was assumedmore » to be 100% (ANL/GTRI/TM-13/4). For this purpose, a set of experiments was chosen to validate the TREAT MCNP model, involving the approach to criticality procedure, in-core neutron flux measurements with foils, and isothermal temperature coefficient and temperature distribution measurements. The results of this study extended the knowledge base for the TREAT MCNP calculations and established the credibility of the MCNP model to be used in the core conversion feasibility analysis.« less

On the effect of updated MCNP photon cross section data on the simulated response of the HPA TLD.

PubMed

Eakins, Jonathan

2009-02-01

The relative response of the new Health Protection Agency thermoluminescence dosimeter (TLD) has been calculated for Narrow Series X-ray distribution and (137)Cs photon sources using the Monte Carlo code MCNP5, and the results compared with those obtained during its design stage using the predecessor code, MCNP4c2. The results agreed at intermediate energies (approximately 0.1 MeV to (137)Cs), but differed at low energies (<0.1 MeV) by up to approximately 10%. This disparity has been ascribed to differences in the default photon interaction data used by the two codes, and derives ultimately from the effect on absorbed dose of the recent updates to the photoelectric cross sections. The sources of these data have been reviewed.

Calculation of absorbed dose and biological effectiveness from photonuclear reactions in a bremsstrahlung beam of end point 50 MeV.

PubMed

Gudowska, I; Brahme, A; Andreo, P; Gudowski, W; Kierkegaard, J

1999-09-01

The absorbed dose due to photonuclear reactions in soft tissue, lung, breast, adipose tissue and cortical bone has been evaluated for a scanned bremsstrahlung beam of end point 50 MeV from a racetrack accelerator. The Monte Carlo code MCNP4B was used to determine the photon source spectrum from the bremsstrahlung target and to simulate the transport of photons through the treatment head and the patient. Photonuclear particle production in tissue was calculated numerically using the energy distributions of photons derived from the Monte Carlo simulations. The transport of photoneutrons in the patient and the photoneutron absorbed dose to tissue were determined using MCNP4B; the absorbed dose due to charged photonuclear particles was calculated numerically assuming total energy absorption in tissue voxels of 1 cm3. The photonuclear absorbed dose to soft tissue, lung, breast and adipose tissue is about (0.11-0.12)+/-0.05% of the maximum photon dose at a depth of 5.5 cm. The absorbed dose to cortical bone is about 45% larger than that to soft tissue. If the contributions from all photoparticles (n, p, 3He and 4He particles and recoils of the residual nuclei) produced in the soft tissue and the accelerator, and from positron radiation and gammas due to induced radioactivity and excited states of the nuclei, are taken into account the total photonuclear absorbed dose delivered to soft tissue is about 0.15+/-0.08% of the maximum photon dose. It has been estimated that the RBE of the photon beam of 50 MV acceleration potential is approximately 2% higher than that of conventional 60Co radiation.

Measurement and simulation of thermal neutron flux distribution in the RTP core

NASA Astrophysics Data System (ADS)

Rabir, Mohamad Hairie B.; Jalal Bayar, Abi Muttaqin B.; Hamzah, Na'im Syauqi B.; Mustafa, Muhammad Khairul Ariff B.; Karim, Julia Bt. Abdul; Zin, Muhammad Rawi B. Mohamed; Ismail, Yahya B.; Hussain, Mohd Huzair B.; Mat Husin, Mat Zin B.; Dan, Roslan B. Md; Ismail, Ahmad Razali B.; Husain, Nurfazila Bt.; Jalil Khan, Zareen Khan B. Abdul; Yakin, Shaiful Rizaide B. Mohd; Saad, Mohamad Fauzi B.; Masood, Zarina Bt.

2018-01-01

The in-core thermal neutron flux distribution was determined using measurement and simulation methods for the Malaysian’s PUSPATI TRIGA Reactor (RTP). In this work, online thermal neutron flux measurement using Self Powered Neutron Detector (SPND) has been performed to verify and validate the computational methods for neutron flux calculation in RTP calculations. The experimental results were used as a validation to the calculations performed with Monte Carlo code MCNP. The detail in-core neutron flux distributions were estimated using MCNP mesh tally method. The neutron flux mapping obtained revealed the heterogeneous configuration of the core. Based on the measurement and simulation, the thermal flux profile peaked at the centre of the core and gradually decreased towards the outer side of the core. The results show a good agreement (relatively) between calculation and measurement where both show the same radial thermal flux profile inside the core: MCNP model over estimation with maximum discrepancy around 20% higher compared to SPND measurement. As our model also predicts well the neutron flux distribution in the core it can be used for the characterization of the full core, that is neutron flux and spectra calculation, dose rate calculations, reaction rate calculations, etc.

SABRINA: an interactive solid geometry modeling program for Monte Carlo

DOE Office of Scientific and Technical Information (OSTI.GOV)

West, J.T.

SABRINA is a fully interactive three-dimensional geometry modeling program for MCNP. In SABRINA, a user interactively constructs either body geometry, or surface geometry models, and interactively debugs spatial descriptions for the resulting objects. This enhanced capability significantly reduces the effort in constructing and debugging complicated three-dimensional geometry models for Monte Carlo Analysis.

Development of Multi-physics (Multiphase CFD + MCNP) simulation for generic solution vessel power calculation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Seung Jun; Buechler, Cynthia Eileen

The current study aims to predict the steady state power of a generic solution vessel and to develop a corresponding heat transfer coefficient correlation for a Moly99 production facility by conducting a fully coupled multi-physics simulation. A prediction of steady state power for the current application is inherently interconnected between thermal hydraulic characteristics (i.e. Multiphase computational fluid dynamics solved by ANSYS-Fluent 17.2) and the corresponding neutronic behavior (i.e. particle transport solved by MCNP6.2) in the solution vessel. Thus, the development of a coupling methodology is vital to understand the system behavior at a variety of system design and postulated operatingmore » scenarios. In this study, we report on the k-effective (keff) calculation for the baseline solution vessel configuration with a selected solution concentration using MCNP K-code modeling. The associated correlation of thermal properties (e.g. density, viscosity, thermal conductivity, specific heat) at the selected solution concentration are developed based on existing experimental measurements in the open literature. The numerical coupling methodology between multiphase CFD and MCNP is successfully demonstrated, and the detailed coupling procedure is documented. In addition, improved coupling methods capturing realistic physics in the solution vessel thermal-neutronic dynamics are proposed and tested further (i.e. dynamic height adjustment, mull-cell approach). As a key outcome of the current study, a multi-physics coupling methodology between MCFD and MCNP is demonstrated and tested for four different operating conditions. Those different operating conditions are determined based on the neutron source strength at a fixed geometry condition. The steady state powers for the generic solution vessel at various operating conditions are reported, and a generalized correlation of the heat transfer coefficient for the current application is discussed. The assessment of multi-physics methodology and preliminary results from various coupled calculations (power prediction and heat transfer coefficient) can be further utilized for the system code validation and generic solution vessel design improvement.« less

Monte Carlo determination of the conversion coefficients Hp(3)/Ka in a right cylinder phantom with 'PENELOPE' code. Comparison with 'MCNP' simulations.

PubMed

Daures, J; Gouriou, J; Bordy, J M

2011-03-01

This work has been performed within the frame of the European Union ORAMED project (Optimisation of RAdiation protection for MEDical staff). The main goal of the project is to improve standards of protection for medical staff for procedures resulting in potentially high exposures and to develop methodologies for better assessing and for reducing, exposures to medical staff. The Work Package WP2 is involved in the development of practical eye-lens dosimetry in interventional radiology. This study is complementary of the part of the ENEA report concerning the calculations with the MCNP-4C code of the conversion factors related to the operational quantity H(p)(3). In this study, a set of energy- and angular-dependent conversion coefficients (H(p)(3)/K(a)), in the newly proposed square cylindrical phantom made of ICRU tissue, have been calculated with the Monte-Carlo code PENELOPE and MCNP5. The H(p)(3) values have been determined in terms of absorbed dose, according to the definition of this quantity, and also with the kerma approximation as formerly reported in ICRU reports. At a low-photon energy (up to 1 MeV), the two results obtained with the two methods are consistent. Nevertheless, large differences are showed at a higher energy. This is mainly due to the lack of electronic equilibrium, especially for small angle incidences. The values of the conversion coefficients obtained with the MCNP-4C code published by ENEA quite agree with the kerma approximation calculations obtained with PENELOPE. We also performed the same calculations with the code MCNP5 with two types of tallies: F6 for kerma approximation and *F8 for estimating the absorbed dose that is, as known, due to secondary electrons. PENELOPE and MCNP5 results agree for the kerma approximation and for the absorbed dose calculation of H(p)(3) and prove that, for photon energies larger than 1 MeV, the transport of the secondary electrons has to be taken into account.

32 CFR 806b.50 - Computer matching.

Code of Federal Regulations, 2013 CFR

2013-07-01

... 32 National Defense 6 2013-07-01 2013-07-01 false Computer matching. 806b.50 Section 806b.50... PROGRAM Disclosing Records to Third Parties § 806b.50 Computer matching. Computer matching programs... on forms used in applying for benefits. Coordinate computer matching statements on forms with Air...

32 CFR 806b.50 - Computer matching.

Code of Federal Regulations, 2014 CFR

2014-07-01

... 32 National Defense 6 2014-07-01 2014-07-01 false Computer matching. 806b.50 Section 806b.50... PROGRAM Disclosing Records to Third Parties § 806b.50 Computer matching. Computer matching programs... on forms used in applying for benefits. Coordinate computer matching statements on forms with Air...

32 CFR 806b.50 - Computer matching.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 32 National Defense 6 2012-07-01 2012-07-01 false Computer matching. 806b.50 Section 806b.50... PROGRAM Disclosing Records to Third Parties § 806b.50 Computer matching. Computer matching programs... on forms used in applying for benefits. Coordinate computer matching statements on forms with Air...

32 CFR 806b.50 - Computer matching.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 32 National Defense 6 2010-07-01 2010-07-01 false Computer matching. 806b.50 Section 806b.50... PROGRAM Disclosing Records to Third Parties § 806b.50 Computer matching. Computer matching programs... on forms used in applying for benefits. Coordinate computer matching statements on forms with Air...

32 CFR 806b.50 - Computer matching.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 32 National Defense 6 2011-07-01 2011-07-01 false Computer matching. 806b.50 Section 806b.50... PROGRAM Disclosing Records to Third Parties § 806b.50 Computer matching. Computer matching programs... on forms used in applying for benefits. Coordinate computer matching statements on forms with Air...

Stress analyses of B-52 pylon hooks

NASA Technical Reports Server (NTRS)

Ko, W. L.; Schuster, L. S.

1985-01-01

The NASTRAN finite element computer program was used in the two dimensional stress analysis of B-52 carrier aircraft pylon hooks: (1) old rear hook (which failed), (2) new rear hook (improved geometry), (3) new DAST rear hook (derated geometry), and (4) front hook. NASTRAN model meshes were generated by the aid of PATRAN-G computer program. Brittle limit loads for all the four hooks were established. The critical stress level calculated from NASTRAN agrees reasonably well with the values predicted from the fracture mechanics for the failed old rear hook.

Testing the Delayed Gamma Capability in MCNP6

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weldon, Robert A.; Fensin, Michael L.; McKinney, Gregg W.

The mission of the Domestic Nuclear Detection Office is to quickly and reliably detect unauthorized attempts to import or transport special nuclear material for use against the United States. Developing detection equipment to meet this objective requires accurate simulation of both the detectable signature and detection mechanism. A delayed particle capability was initially added to MCNPX 2.6.A in 2005 to sample the radioactive fission product parents and emit decay particles resulting from the decay chain. To meet the objectives of detection scenario modeling, the capability was designed to sample a particular time for emitting particular multiplicity of a particular energy.more » Because the sampling process of selecting both time and energy is interdependent, to linearize the time and emission sampling, atom densities are computed at several discrete time steps, and the time-integrated production is computed by multiplying the atom density by the decay constant and time step size to produce a cumulative distribution function for sampling the emission time, energy, and multiplicity. The delayed particle capability was initially given a time-bin structure to help reasonably reproduce, from a qualitative sense, a fission benchmark by Beddingfield, which examined the delayed gamma emission. This original benchmark was only qualitative and did not contain the magnitudes of the actual measured data but did contain relative graphical representation of the spectra. A better benchmark with measured data was later provided by Hunt, Mozin, Reedy, Selpel, and Tobin at the Idaho Accelerator Center; however, because of the complexity of the benchmark setup, sizable systematic errors were expected in the modeling, and initial results compared to MCNPX 2.7.0 showed errors outside of statistical fluctuation. Presented in this paper is a more simplified approach to benchmarking, utilizing closed form analytic solutions to the granddaughter equations for particular sets of decay systems. We examine five different decay chains (two-stage decay to stable) and show the predictability of the MCNP6 delayed gamma feature. Results do show that while the default delayed gamma calculations available in the MCNP6 1.0 release can give accurate results for some isotopes (e.g., 137Ba), the percent differences between the closed form analytic solutions and the MCNP6 calculations were often >40% ( 28Mg, 28Al, 42K, 47Ca, 47Sc, 60Co). With the MCNP6 1.1 Beta release, the tenth entry on the DBCN card allows improved calculation within <5% as compared to the closed form analytic solutions for immediate parent emissions and transient equilibrium systems. While the tenth entry on the DBCN card for MCNP6 1.1 gives much better results for transient equilibrium systems and parent emissions in general, it does little to improve daughter emissions of secular equilibrium systems. Finally, hypotheses were presented as to why daughter emissions of secular equilibrium systems might be mispredicted in some cases and not in others.« less

On the development of radiation tolerant surveillance camera from consumer-grade components

NASA Astrophysics Data System (ADS)

Klemen, Ambrožič; Luka, Snoj; Lars, Öhlin; Jan, Gunnarsson; Niklas, Barringer

2017-09-01

In this paper an overview on the process of designing a radiation tolerant surveillance camera from consumer grade components and commercially available particle shielding materials is given. This involves utilization of Monte-Carlo particle transport code MCNP6 and ENDF/B-VII.0 nuclear data libraries, as well as testing the physical electrical systems against γ radiation, utilizing JSI TRIGA mk. II fuel elements as a γ-ray sources. A new, aluminum, 20 cm × 20 cm × 30 cm irradiation facility with electrical power and signal wire guide-tube to the reactor platform, was designed and constructed and used for irradiation of large electronic and optical components assemblies with activated fuel elements. Electronic components to be used in the camera were tested against γ-radiation in an independent manner, to determine their radiation tolerance. Several camera designs were proposed and simulated using MCNP, to determine incident particle and dose attenuation factors. Data obtained from the measurements and MCNP simulations will be used to finalize the design of 3 surveillance camera models, with different radiation tolerances.

Balancing Materiel Readiness Risks and Concurrency in Weapon System Acquisition: A Handbook for Program Managers

DTIC Science & Technology

1984-07-15

ftViCCii UWNC COMMAND MIX CM AFT DCP OUTUNC moo ffEOUCST FOR PROGRAM DECISION DRAFT DCP AFSC REVIEW RECOWMEM CATIONS OHI*OC Arse wioc...CS.P3 F16. El*. P» MCA Exhibit 4-6b. EMBEDDED COMPUTER HARDWARE vs. SOFTWARE Exhibit 4-6c. DoD EMBEDDED COMPUTER MARKET 31.J1...the mix of stores carried by that vehicle 6. Anticipated combat tactics employed by the carrying or launching vehicle and its maneuvering

Advances in the computation of the Sjöstrand, Rossi, and Feynman distributions

DOE PAGES

Talamo, A.; Gohar, Y.; Gabrielli, F.; ...

2017-02-01

This study illustrates recent computational advances in the application of the Sjöstrand (area), Rossi, and Feynman methods to estimate the effective multiplication factor of a subcritical system driven by an external neutron source. The methodologies introduced in this study have been validated with the experimental results from the KUKA facility of Japan by Monte Carlo (MCNP6 and MCNPX) and deterministic (ERANOS, VARIANT, and PARTISN) codes. When the assembly is driven by a pulsed neutron source generated by a particle accelerator and delayed neutrons are at equilibrium, the Sjöstrand method becomes extremely fast if the integral of the reaction rate frommore » a single pulse is split into two parts. These two integrals distinguish between the neutron counts during and after the pulse period. To conclude, when the facility is driven by a spontaneous fission neutron source, the timestamps of the detector neutron counts can be obtained up to the nanosecond precision using MCNP6, which allows obtaining the Rossi and Feynman distributions.« less

Shielding analysis of the Microtron MT-25 bunker using the MCNP-4C code and NCRP Report 51.

PubMed

Casanova, A O; López, N; Gelen, A; Guevara, M V Manso; Díaz, O; Cimino, L; D'Alessandro, K; Melo, J C

2004-01-01

A cyclic electron accelerator Microtron MT-25 will be installed in Havana, Cuba. Electrons, neutrons and gamma radiation up to 25 MeV can be produced in the MT-25. A detailed shielding analysis for the bunker is carried out using two ways: the NCRP-51 Report and the Monte Carlo Method (MCNP-4C Code). The walls and ceiling thicknesses are estimated with dose constraints of 0.5 and 20 mSv y(-1), respectively, and an area occupancy factor of 1/16. Both results are compared and a preliminary bunker design is shown. Copyright 2004 Oxford University Press

MAGIC Computer Simulation. Volume 1: User Manual

DTIC Science & Technology

1970-07-01

vulnerability and MAGIC programs. A three-digit code is assigned to each component of the target, such as armor, gun tube; and a two-digit code is assigned to...A review of the subject Magic Computer Simulation User and Analyst Manuals has been conducted based upon a request received from the US Army...1970 4. TITLE AND SUBTITLE MAGIC Computer Simulation 5a. CONTRACT NUMBER 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) 5d. PROJECT

Review of heavy charged particle transport in MCNP6.2

NASA Astrophysics Data System (ADS)

Zieb, K.; Hughes, H. G.; James, M. R.; Xu, X. G.

2018-04-01

The release of version 6.2 of the MCNP6 radiation transport code is imminent. To complement the newest release, a summary of the heavy charged particle physics models used in the 1 MeV to 1 GeV energy regime is presented. Several changes have been introduced into the charged particle physics models since the merger of the MCNP5 and MCNPX codes into MCNP6. This paper discusses the default models used in MCNP6 for continuous energy loss, energy straggling, and angular scattering of heavy charged particles. Explanations of the physics models' theories are included as well.

Review of Heavy Charged Particle Transport in MCNP6.2

DOE PAGES

Zieb, Kristofer James Ekhart; Hughes, Henry Grady III; Xu, X. George; ...

2018-01-05

The release of version 6.2 of the MCNP6 radiation transport code is imminent. To complement the newest release, a summary of the heavy charged particle physics models used in the 1 MeV to 1 GeV energy regime is presented. Several changes have been introduced into the charged particle physics models since the merger of the MCNP5 and MCNPX codes into MCNP6. Here, this article discusses the default models used in MCNP6 for continuous energy loss, energy straggling, and angular scattering of heavy charged particles. Explanations of the physics models’ theories are included as well.

MC2-3 / DIF3D Analysis for the ZPPR-15 Doppler and Sodium Void Worth Measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Micheal A.; Lell, Richard M.; Lee, Changho

This manuscript covers validation efforts for our deterministic codes at Argonne National Laboratory. The experimental results come from the ZPPR-15 work in 1985-1986 which was focused on the accuracy of physics data for the integral fast reactor concept. Results for six loadings are studied in this document and focus on Doppler sample worths and sodium void worths. The ZPPR-15 loadings are modeled using the MC2-3/DIF3D codes developed and maintained at ANL and the MCNP code from LANL. The deterministic models are generated by processing the as-built geometry information, i.e. MCNP input, and generating MC2-3 cross section generation instructions and amore » drawer homogenized equivalence problem. The Doppler reactivity worth measurements are small heated samples which insert very small amounts of reactivity into the system (< 2 pcm). The results generated by the MC2-3/DIF3D codes were excellent for ZPPR-15A and ZPPR-15B and good for ZPPR-15D, compared to the MCNP solutions. In all cases, notable improvements were made over the analysis techniques applied to the same problems in 1987. The sodium void worths from MC2-3/DIF3D were quite good at 37.5 pcm while MCNP result was 33 pcm and the measured result was 31.5 pcm. Copyright © (2015) by the American Nuclear Society All rights reserved.« less

Experimental characterization of the AFIT neutron facility. Master's thesis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lessard, O.J.

1993-09-01

AFIT's Neutron Facility was characterized for room-return neutrons using a (252)Cf source and a Bonner sphere spectrometer with three experimental models, the shadow shield, the Eisenhauer, Schwartz, and Johnson (ESJ), and the polynomial models. The free-field fluences at one meter from the ESJ and polynomial models were compared to the equivalent value from the accepted experimental shadow shield model to determine the suitability of the models in the AFIT facility. The polynomial model behaved erratically, as expected, while the ESJ model compared to within 4.8% of the shadow shield model results for the four Bonner sphere calibration. The ratio ofmore » total fluence to free-field fluence at one meter for the ESJ model was then compared to the equivalent ratio obtained by a Monte Cario Neutron-Photon transport code (MCNP), an accepted computational model. The ESJ model compared to within 6.2% of the MCNP results. AFIT's fluence ratios were compared to equivalent ratios reported by three other neutron facilities which verified that AFIT's results fit previously published trends based on room volumes. The ESJ model appeared adequate for health physics applications and was chosen was chosen for calibration of the AFIT facility. Neutron Detector, Bonner Sphere, Neutron Dosimetry, Room Characterization.« less

An MCNP-based model for the evaluation of the photoneutron dose in high energy medical electron accelerators.

PubMed

Carinou, Eleutheria; Stamatelatos, Ion Evangelos; Kamenopoulou, Vassiliki; Georgolopoulou, Paraskevi; Sandilos, Panayotis

The development of a computational model for the treatment head of a medical electron accelerator (Elekta/Philips SL-18) by the Monte Carlo code mcnp-4C2 is discussed. The model includes the major components of the accelerator head and a pmma phantom representing the patient body. Calculations were performed for a 14 MeV electron beam impinging on the accelerator target and a 10 cmx10 cm beam area at the isocentre. The model was used in order to predict the neutron ambient dose equivalent at the isocentre level and moreover the neutron absorbed dose distribution within the phantom. Calculations were validated against experimental measurements performed by gold foil activation detectors. The results of this study indicated that the equivalent dose at tissues or organs adjacent to the treatment field due to photoneutrons could be up to 10% of the total peripheral dose, for the specific accelerator characteristics examined. Therefore, photoneutrons should be taken into account when accurate dose calculations are required to sensitive tissues that are adjacent to the therapeutic X-ray beam. The method described can be extended to other accelerators and collimation configurations as well, upon specification of treatment head component dimensions, composition and nominal accelerating potential.

Validation of the MCNP6 electron-photon transport algorithm: multiple-scattering of 13- and 20-MeV electrons in thin foils

NASA Astrophysics Data System (ADS)

Dixon, David A.; Hughes, H. Grady

2017-09-01

This paper presents a validation test comparing angular distributions from an electron multiple-scattering experiment with those generated using the MCNP6 Monte Carlo code system. In this experiment, a 13- and 20-MeV electron pencil beam is deflected by thin foils with atomic numbers from 4 to 79. To determine the angular distribution, the fluence is measured down range of the scattering foil at various radii orthogonal to the beam line. The characteristic angle (the angle for which the max of the distribution is reduced by 1/e) is then determined from the angular distribution and compared with experiment. Multiple scattering foils tested herein include beryllium, carbon, aluminum, copper, and gold. For the default electron-photon transport settings, the calculated characteristic angle was statistically distinguishable from measurement and generally broader than the measured distributions. The average relative difference ranged from 5.8% to 12.2% over all of the foils, source energies, and physics settings tested. This validation illuminated a deficiency in the computation of the underlying angular distributions that is well understood. As a result, code enhancements were made to stabilize the angular distributions in the presence of very small substeps. However, the enhancement only marginally improved results indicating that additional algorithmic details should be studied.

V&V of MCNP 6.1.1 Beta Against Intermediate and High-Energy Experimental Data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mashnik, Stepan G

This report presents a set of validation and verification (V&V) MCNP 6.1.1 beta results calculated in parallel, with MPI, obtained using its event generators at intermediate and high-energies compared against various experimental data. It also contains several examples of results using the models at energies below 150 MeV, down to 10 MeV, where data libraries are normally used. This report can be considered as the forth part of a set of MCNP6 Testing Primers, after its first, LA-UR-11-05129, and second, LA-UR-11-05627, and third, LA-UR-26944, publications, but is devoted to V&V with the latest, 1.1 beta version of MCNP6. The MCNP6more » test-problems discussed here are presented in the /VALIDATION_CEM/and/VALIDATION_LAQGSM/subdirectories in the MCNP6/Testing/directory. README files that contain short descriptions of every input file, the experiment, the quantity of interest that the experiment measures and its description in the MCNP6 output files, and the publication reference of that experiment are presented for every test problem. Templates for plotting the corresponding results with xmgrace as well as pdf files with figures representing the final results of our V&V efforts are presented. Several technical “bugs” in MCNP 6.1.1 beta were discovered during our current V&V of MCNP6 while running it in parallel with MPI using its event generators. These “bugs” are to be fixed in the following version of MCNP6. Our results show that MCNP 6.1.1 beta using its CEM03.03, LAQGSM03.03, Bertini, and INCL+ABLA, event generators describes, as a rule, reasonably well different intermediate- and high-energy measured data. This primer isn’t meant to be read from cover to cover. Readers may skip some sections and go directly to any test problem in which they are interested.« less

Acceleration of MCNP calculations for small pipes configurations by using Weigth Windows Importance cards created by the SN-3D ATTILA

NASA Astrophysics Data System (ADS)

Castanier, Eric; Paterne, Loic; Louis, Céline

2017-09-01

In the nuclear engineering, you have to manage time and precision. Especially in shielding design, you have to be more accurate and efficient to reduce cost (shielding thickness optimization), and for this, you use 3D codes. In this paper, we want to see if we can easily applicate the CADIS methods for design shielding of small pipes which go through large concrete walls. We assess the impact of the WW generated by the 3D-deterministic code ATTILA versus WW directly generated by MCNP (iterative and manual process). The comparison is based on the quality of the convergence (estimated relative error (σ), Variance of Variance (VOV) and Figure of Merit (FOM)), on time (computer time + modelling) and on the implement for the engineer.

Path Toward a Unifid Geometry for Radiation Transport

NASA Technical Reports Server (NTRS)

Lee, Kerry; Barzilla, Janet; Davis, Andrew; Zachmann

2014-01-01

The Direct Accelerated Geometry for Radiation Analysis and Design (DAGRAD) element of the RadWorks Project under Advanced Exploration Systems (AES) within the Space Technology Mission Directorate (STMD) of NASA will enable new designs and concepts of operation for radiation risk assessment, mitigation and protection. This element is designed to produce a solution that will allow NASA to calculate the transport of space radiation through complex computer-aided design (CAD) models using the state-of-the-art analytic and Monte Carlo radiation transport codes. Due to the inherent hazard of astronaut and spacecraft exposure to ionizing radiation in low-Earth orbit (LEO) or in deep space, risk analyses must be performed for all crew vehicles and habitats. Incorporating these analyses into the design process can minimize the mass needed solely for radiation protection. Transport of the radiation fields as they pass through shielding and body materials can be simulated using Monte Carlo techniques or described by the Boltzmann equation, which is obtained by balancing changes in particle fluxes as they traverse a small volume of material with the gains and losses caused by atomic and nuclear collisions. Deterministic codes that solve the Boltzmann transport equation, such as HZETRN [high charge and energy transport code developed by NASA Langley Research Center (LaRC)], are generally computationally faster than Monte Carlo codes such as FLUKA, GEANT4, MCNP(X) or PHITS; however, they are currently limited to transport in one dimension, which poorly represents the secondary light ion and neutron radiation fields. NASA currently uses HZETRN space radiation transport software, both because it is computationally efficient and because proven methods have been developed for using this software to analyze complex geometries. Although Monte Carlo codes describe the relevant physics in a fully three-dimensional manner, their computational costs have thus far prevented their widespread use for analysis of complex CAD models, leading to the creation and maintenance of toolkit-specific simplistic geometry models. The work presented here builds on the Direct Accelerated Geometry Monte Carlo (DAGMC) toolkit developed for use with the Monte Carlo N-Particle (MCNP) transport code. The workflow for achieving radiation transport on CAD models using MCNP and FLUKA has been demonstrated and the results of analyses on realistic spacecraft/habitats will be presented. Future work is planned that will further automate this process and enable the use of multiple radiation transport codes on identical geometry models imported from CAD. This effort will enhance the modeling tools used by NASA to accurately evaluate the astronaut space radiation risk and accurately determine the protection provided by as-designed exploration mission vehicles and habitats

Path Toward a Unified Geometry for Radiation Transport

NASA Astrophysics Data System (ADS)

Lee, Kerry

The Direct Accelerated Geometry for Radiation Analysis and Design (DAGRAD) element of the RadWorks Project under Advanced Exploration Systems (AES) within the Space Technology Mission Directorate (STMD) of NASA will enable new designs and concepts of operation for radiation risk assessment, mitigation and protection. This element is designed to produce a solution that will allow NASA to calculate the transport of space radiation through complex CAD models using the state-of-the-art analytic and Monte Carlo radiation transport codes. Due to the inherent hazard of astronaut and spacecraft exposure to ionizing radiation in low-Earth orbit (LEO) or in deep space, risk analyses must be performed for all crew vehicles and habitats. Incorporating these analyses into the design process can minimize the mass needed solely for radiation protection. Transport of the radiation fields as they pass through shielding and body materials can be simulated using Monte Carlo techniques or described by the Boltzmann equation, which is obtained by balancing changes in particle fluxes as they traverse a small volume of material with the gains and losses caused by atomic and nuclear collisions. Deterministic codes that solve the Boltzmann transport equation, such as HZETRN (high charge and energy transport code developed by NASA LaRC), are generally computationally faster than Monte Carlo codes such as FLUKA, GEANT4, MCNP(X) or PHITS; however, they are currently limited to transport in one dimension, which poorly represents the secondary light ion and neutron radiation fields. NASA currently uses HZETRN space radiation transport software, both because it is computationally efficient and because proven methods have been developed for using this software to analyze complex geometries. Although Monte Carlo codes describe the relevant physics in a fully three-dimensional manner, their computational costs have thus far prevented their widespread use for analysis of complex CAD models, leading to the creation and maintenance of toolkit specific simplistic geometry models. The work presented here builds on the Direct Accelerated Geometry Monte Carlo (DAGMC) toolkit developed for use with the Monte Carlo N-Particle (MCNP) transport code. The work-flow for doing radiation transport on CAD models using MCNP and FLUKA has been demonstrated and the results of analyses on realistic spacecraft/habitats will be presented. Future work is planned that will further automate this process and enable the use of multiple radiation transport codes on identical geometry models imported from CAD. This effort will enhance the modeling tools used by NASA to accurately evaluate the astronaut space radiation risk and accurately determine the protection provided by as-designed exploration mission vehicles and habitats.

Geant4 Modifications for Accurate Fission Simulations

NASA Astrophysics Data System (ADS)

Tan, Jiawei; Bendahan, Joseph

Monte Carlo is one of the methods to simulate the generation and transport of radiation through matter. The most widely used radiation simulation codes are MCNP and Geant4. The simulation of fission production and transport by MCNP has been thoroughly benchmarked. There is an increasing number of users that prefer using Geant4 due to the flexibility of adding features. However, it has been found that Geant4 does not have the proper fission-production cross sections and does not produce the correct fission products. To achieve accurate results for studies in fissionable material applications, Geant4 was modified to correct these inaccuracies and to add new capabilities. The fission model developed by the Lawrence Livermore National Laboratory was integrated into the neutron-fission modeling package. The photofission simulation capability was enabled using the same neutron-fission library under the assumption that nuclei fission in the same way, independent of the excitation source. The modified fission code provides the correct multiplicity of prompt neutrons and gamma rays, and produces delayed gamma rays and neutrons with time and energy dependencies that are consistent with ENDF/B-VII. The delayed neutrons are now directly produced by a custom package that bypasses the fragment cascade model. The modifications were made for U-235, U-238 and Pu-239 isotopes; however, the new framework allows adding new isotopes easily. The SLAC nuclear data library is used for simulation of isotopes with an atomic number above 92 because it is not available in Geant4. Results of the modified Geant4.10.1 package of neutron-fission and photofission for prompt and delayed radiation are compared with ENDFB-VII and with results produced with the original package.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Perret, Gregory

The critical decay constant (B/A), delayed neutron fraction (B) and generation time (A) of the Minerve reactor were measured by the Paul Scherrer Institut (PSI) and the Commissariat a l'Energie Atomique (CEA) in September 2014 using the Feynman-alpha and Power Spectral Density neutron noise measurement techniques. Three slightly subcritical configuration were measured using two 1-g {sup 235}U fission chambers. This paper reports on the results obtained by PSI in the near critical configuration (-2g). The most reliable and precise results were obtained with the Cross-Power Spectral Density technique: B = 708.4±9.2 pcm, B/A = 79.0±0.6 s{sup -1} and A 89.7±1.4more » micros. Predictions of the same kinetic parameters were obtained with MCNP5-v1.6 and the JEFF-3.1 and ENDF/B-VII.1 nuclear data libraries. On average the predictions for B and B/A overestimate the experimental results by 5% and 11%, respectively. The discrepancy is suspected to come from either a corruption of the data or from the inadequacy of the point kinetic equations to interpret the measurements in the Minerve driven system. (authors)« less

Improvements in Routing for Packet-Switched Networks

DTIC Science & Technology

1975-02-18

PROGRAM FOR COMPUTER SIMULATION . . 90 B.l Flow Diagram of Adaptive Routine 90 B.2 Progiam ARPSIM 93 B.3 Explanation of Variables...equa. 90 APPENDIX B ADAPTIVE ROUTING PROGRAM FOR COMPUTER SIMULA HON The computer simulation for adaptive routing was initially run on a DDP-24 small...TRANSMIT OVER AVAILABLE LINKS MESSAGES IN QUEUE COMPUTE Ni NUMBER OF ARRIVALS AT EACH NODE i AT TIME T Fig. Bla - Flow Diagram of Program Routine 92

Image based Monte Carlo Modeling for Computational Phantom

NASA Astrophysics Data System (ADS)

Cheng, Mengyun; Wang, Wen; Zhao, Kai; Fan, Yanchang; Long, Pengcheng; Wu, Yican

2014-06-01

The evaluation on the effects of ionizing radiation and the risk of radiation exposure on human body has been becoming one of the most important issues for radiation protection and radiotherapy fields, which is helpful to avoid unnecessary radiation and decrease harm to human body. In order to accurately evaluate the dose on human body, it is necessary to construct more realistic computational phantom. However, manual description and verfication of the models for Monte carlo(MC)simulation are very tedious, error-prone and time-consuming. In addiation, it is difficult to locate and fix the geometry error, and difficult to describe material information and assign it to cells. MCAM (CAD/Image-based Automatic Modeling Program for Neutronics and Radiation Transport Simulation) was developed as an interface program to achieve both CAD- and image-based automatic modeling by FDS Team (Advanced Nuclear Energy Research Team, http://www.fds.org.cn). The advanced version (Version 6) of MCAM can achieve automatic conversion from CT/segmented sectioned images to computational phantoms such as MCNP models. Imaged-based automatic modeling program(MCAM6.0) has been tested by several medical images and sectioned images. And it has been applied in the construction of Rad-HUMAN. Following manual segmentation and 3D reconstruction, a whole-body computational phantom of Chinese adult female called Rad-HUMAN was created by using MCAM6.0 from sectioned images of a Chinese visible human dataset. Rad-HUMAN contains 46 organs/tissues, which faithfully represented the average anatomical characteristics of the Chinese female. The dose conversion coefficients(Dt/Ka) from kerma free-in-air to absorbed dose of Rad-HUMAN were calculated. Rad-HUMAN can be applied to predict and evaluate dose distributions in the Treatment Plan System (TPS), as well as radiation exposure for human body in radiation protection.

FY2017 Report on NISC Measurements and Detector Simulations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrews, Madison Theresa; Meierbachtol, Krista Cruse; Jordan, Tyler Alexander

FY17 work focused on automation, both of the measurement analysis and comparison of simulations. The experimental apparatus was relocated and weeks of continuous measurements of the spontaneous fission source 252Cf was performed. Programs were developed to automate the conversion of measurements into ROOT data framework files with a simple terminal input. The complete analysis of the measurement (which includes energy calibration and the identification of correlated counts) can now be completed with a documented process which involves one simple execution line as well. Finally, the hurdles of slow MCNP simulations resulting in low simulation statistics have been overcome with themore » generation of multi-run suites which make use of the highperformance computing resources at LANL. Preliminary comparisons of measurements and simulations have been performed and will be the focus of FY18 work.« less

Neutron skyshine calculations with the integral line-beam method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gui, A.A.; Shultis, J.K.; Faw, R.E.

1997-10-01

Recently developed line- and conical-beam response functions are used to calculate neutron skyshine doses for four idealized source geometries. These calculations, which can serve as benchmarks, are compared with MCNP calculations, and the excellent agreement indicates that the integral conical- and line-beam method is an effective alternative to more computationally expensive transport calculations.

MCNP6 Fission Multiplicity with FMULT Card

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wilcox, Trevor; Fensin, Michael Lorne; Hendricks, John S.

With the merger of MCNPX and MCNP5 into MCNP6, MCNP6 now provides all the capabilities of both codes allowing the user to access all the fission multiplicity data sets. Detailed in this paper is: (1) the new FMULT card capabilities for accessing these different data sets; (2) benchmark calculations, as compared to experiment, detailing the results of selecting these separate data sets for thermal neutron induced fission on U-235.

Production of energetic light fragments in extensions of the CEM and LAQGSM event generators of the Monte Carlo transport code MCNP6 [Production of energetic light fragments in CEM, LAQGSM, and MCNP6

DOE PAGES

Mashnik, Stepan Georgievich; Kerby, Leslie Marie; Gudima, Konstantin K.; ...

2017-03-23

We extend the cascade-exciton model (CEM), and the Los Alamos version of the quark-gluon string model (LAQGSM), event generators of the Monte Carlo N-particle transport code version 6 (MCNP6), to describe production of energetic light fragments (LF) heavier than 4He from various nuclear reactions induced by particles and nuclei at energies up to about 1 TeV/nucleon. In these models, energetic LF can be produced via Fermi breakup, preequilibrium emission, and coalescence of cascade particles. Initially, we study several variations of the Fermi breakup model and choose the best option for these models. Then, we extend the modified exciton model (MEM)more » used by these codes to account for a possibility of multiple emission of up to 66 types of particles and LF (up to 28Mg) at the preequilibrium stage of reactions. Then, we expand the coalescence model to allow coalescence of LF from nucleons emitted at the intranuclear cascade stage of reactions and from lighter clusters, up to fragments with mass numbers A ≤ 7, in the case of CEM, and A ≤ 12, in the case of LAQGSM. Next, we modify MCNP6 to allow calculating and outputting spectra of LF and heavier products with arbitrary mass and charge numbers. The improved version of CEM is implemented into MCNP6. Lastly, we test the improved versions of CEM, LAQGSM, and MCNP6 on a variety of measured nuclear reactions. The modified codes give an improved description of energetic LF from particle- and nucleus-induced reactions; showing a good agreement with a variety of available experimental data. They have an improved predictive power compared to the previous versions and can be used as reliable tools in simulating applications involving such types of reactions.« less

Production of energetic light fragments in extensions of the CEM and LAQGSM event generators of the Monte Carlo transport code MCNP6 [Production of energetic light fragments in CEM, LAQGSM, and MCNP6

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mashnik, Stepan Georgievich; Kerby, Leslie Marie; Gudima, Konstantin K.

We extend the cascade-exciton model (CEM), and the Los Alamos version of the quark-gluon string model (LAQGSM), event generators of the Monte Carlo N-particle transport code version 6 (MCNP6), to describe production of energetic light fragments (LF) heavier than 4He from various nuclear reactions induced by particles and nuclei at energies up to about 1 TeV/nucleon. In these models, energetic LF can be produced via Fermi breakup, preequilibrium emission, and coalescence of cascade particles. Initially, we study several variations of the Fermi breakup model and choose the best option for these models. Then, we extend the modified exciton model (MEM)more » used by these codes to account for a possibility of multiple emission of up to 66 types of particles and LF (up to 28Mg) at the preequilibrium stage of reactions. Then, we expand the coalescence model to allow coalescence of LF from nucleons emitted at the intranuclear cascade stage of reactions and from lighter clusters, up to fragments with mass numbers A ≤ 7, in the case of CEM, and A ≤ 12, in the case of LAQGSM. Next, we modify MCNP6 to allow calculating and outputting spectra of LF and heavier products with arbitrary mass and charge numbers. The improved version of CEM is implemented into MCNP6. Lastly, we test the improved versions of CEM, LAQGSM, and MCNP6 on a variety of measured nuclear reactions. The modified codes give an improved description of energetic LF from particle- and nucleus-induced reactions; showing a good agreement with a variety of available experimental data. They have an improved predictive power compared to the previous versions and can be used as reliable tools in simulating applications involving such types of reactions.« less

User's guide to computer program CIVM-JET 4B to calculate the transient structural responses of partial and/or complete structural rings to engine-rotor-fragment impact

NASA Technical Reports Server (NTRS)

Stagliano, T. R.; Spilker, R. L.; Witmer, E. A.

1976-01-01

A user-oriented computer program CIVM-JET 4B is described to predict the large-deflection elastic-plastic structural responses of fragment impacted single-layer: (a) partial-ring fragment containment or deflector structure or (b) complete-ring fragment containment structure. These two types of structures may be either free or supported in various ways. Supports accommodated include: (1) point supports such as pinned-fixed, ideally-clamped, or supported by a structural branch simulating mounting-bracket structure and (2) elastic foundation support distributed over selected regions of the structure. The initial geometry of each partial or complete ring may be circular or arbitrarily curved; uniform or variable thicknesses of the structure are accommodated. The structural material is assumed to be initially isotropic; strain hardening and strain rate effects are taken into account.

Language-Agnostic Reproducible Data Analysis Using Literate Programming.

PubMed

Vassilev, Boris; Louhimo, Riku; Ikonen, Elina; Hautaniemi, Sampsa

2016-01-01

A modern biomedical research project can easily contain hundreds of analysis steps and lack of reproducibility of the analyses has been recognized as a severe issue. While thorough documentation enables reproducibility, the number of analysis programs used can be so large that in reality reproducibility cannot be easily achieved. Literate programming is an approach to present computer programs to human readers. The code is rearranged to follow the logic of the program, and to explain that logic in a natural language. The code executed by the computer is extracted from the literate source code. As such, literate programming is an ideal formalism for systematizing analysis steps in biomedical research. We have developed the reproducible computing tool Lir (literate, reproducible computing) that allows a tool-agnostic approach to biomedical data analysis. We demonstrate the utility of Lir by applying it to a case study. Our aim was to investigate the role of endosomal trafficking regulators to the progression of breast cancer. In this analysis, a variety of tools were combined to interpret the available data: a relational database, standard command-line tools, and a statistical computing environment. The analysis revealed that the lipid transport related genes LAPTM4B and NDRG1 are coamplified in breast cancer patients, and identified genes potentially cooperating with LAPTM4B in breast cancer progression. Our case study demonstrates that with Lir, an array of tools can be combined in the same data analysis to improve efficiency, reproducibility, and ease of understanding. Lir is an open-source software available at github.com/borisvassilev/lir.

Language-Agnostic Reproducible Data Analysis Using Literate Programming

PubMed Central

Vassilev, Boris; Louhimo, Riku; Ikonen, Elina; Hautaniemi, Sampsa

2016-01-01

A modern biomedical research project can easily contain hundreds of analysis steps and lack of reproducibility of the analyses has been recognized as a severe issue. While thorough documentation enables reproducibility, the number of analysis programs used can be so large that in reality reproducibility cannot be easily achieved. Literate programming is an approach to present computer programs to human readers. The code is rearranged to follow the logic of the program, and to explain that logic in a natural language. The code executed by the computer is extracted from the literate source code. As such, literate programming is an ideal formalism for systematizing analysis steps in biomedical research. We have developed the reproducible computing tool Lir (literate, reproducible computing) that allows a tool-agnostic approach to biomedical data analysis. We demonstrate the utility of Lir by applying it to a case study. Our aim was to investigate the role of endosomal trafficking regulators to the progression of breast cancer. In this analysis, a variety of tools were combined to interpret the available data: a relational database, standard command-line tools, and a statistical computing environment. The analysis revealed that the lipid transport related genes LAPTM4B and NDRG1 are coamplified in breast cancer patients, and identified genes potentially cooperating with LAPTM4B in breast cancer progression. Our case study demonstrates that with Lir, an array of tools can be combined in the same data analysis to improve efficiency, reproducibility, and ease of understanding. Lir is an open-source software available at github.com/borisvassilev/lir. PMID:27711123

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Der Marck, S. C.

Three nuclear data libraries have been tested extensively using criticality safety benchmark calculations. The three libraries are the new release of the US library ENDF/B-VII.1 (2011), the new release of the Japanese library JENDL-4.0 (2011), and the OECD/NEA library JEFF-3.1 (2006). All calculations were performed with the continuous-energy Monte Carlo code MCNP (version 4C3, as well as version 6-beta1). Around 2000 benchmark cases from the International Handbook of Criticality Safety Benchmark Experiments (ICSBEP) were used. The results were analyzed per ICSBEP category, and per element. Overall, the three libraries show similar performance on most criticality safety benchmarks. The largest differencesmore » are probably caused by elements such as Be, C, Fe, Zr, W. (authors)« less

Joint Services Electronics Program Annual Progress Report.

DTIC Science & Technology

1985-11-01

one symbol memory) adaptive lHuffman codes were performed, and the compression achieved was compared with that of Ziv - Lempel coding. As was expected...MATERIALS 8 4. Information Systems 9 4.1 REAL TIME STATISTICAL DATA PROCESSING 9 -. 4.2 DATA COMPRESSION for COMPUTER DATA STRUCTURES 9 5. PhD...a. Real Time Statistical Data Processing (T. Kailatb) b. Data Compression for Computer Data Structures (J. Gill) Acces Fo NTIS CRA&I I " DTIC TAB

An Analysis of Radiation Penetration through the U-Shaped Cast Concrete Joints of Concrete Shielding in the Multipurpose Gamma Irradiator of BATAN

NASA Astrophysics Data System (ADS)

Ardiyati, Tanti; Rozali, Bang; Kasmudin

2018-02-01

An analysis of radiation penetration through the U-shaped joints of cast concrete shielding in BATAN’s multipurpose gamma irradiator has been carried out. The analysis has been performed by calculating the radiation penetration through the U-shaped joints of the concrete shielding using MCNP computer code. The U-shaped joints were a new design in massive concrete construction in Indonesia and, in its actual application, it is joined by a bonding agent. In the MCNP simulation model, eight detectors were located close to the observed irradiation room walls of the concrete shielding. The simulation results indicated that the radiation levels outside the concrete shielding was less than the permissible limit of 2.5 μSv/h so that the workers could safely access electrical room, control room, water treatment facility and outside irradiation room. The radiation penetration decreased as the density of material increased.

MCNP HPGe detector benchmark with previously validated Cyltran model.

PubMed

Hau, I D; Russ, W R; Bronson, F

2009-05-01

An exact copy of the detector model generated for Cyltran was reproduced as an MCNP input file and the detection efficiency was calculated similarly with the methodology used in previous experimental measurements and simulation of a 280 cm(3) HPGe detector. Below 1000 keV the MCNP data correlated to the Cyltran results within 0.5% while above this energy the difference between MCNP and Cyltran increased to about 6% at 4800 keV, depending on the electron cut-off energy.

A quantitative three-dimensional dose attenuation analysis around Fletcher-Suit-Delclos due to stainless steel tube for high-dose-rate brachytherapy by Monte Carlo calculations.

PubMed

Parsai, E Ishmael; Zhang, Zhengdong; Feldmeier, John J

2009-01-01

The commercially available brachytherapy treatment-planning systems today, usually neglects the attenuation effect from stainless steel (SS) tube when Fletcher-Suit-Delclos (FSD) is used in treatment of cervical and endometrial cancers. This could lead to potential inaccuracies in computing dwell times and dose distribution. A more accurate analysis quantifying the level of attenuation for high-dose-rate (HDR) iridium 192 radionuclide ((192)Ir) source is presented through Monte Carlo simulation verified by measurement. In this investigation a general Monte Carlo N-Particles (MCNP) transport code was used to construct a typical geometry of FSD through simulation and compare the doses delivered to point A in Manchester System with and without the SS tubing. A quantitative assessment of inaccuracies in delivered dose vs. the computed dose is presented. In addition, this investigation expanded to examine the attenuation-corrected radial and anisotropy dose functions in a form parallel to the updated AAPM Task Group No. 43 Report (AAPM TG-43) formalism. This will delineate quantitatively the inaccuracies in dose distributions in three-dimensional space. The changes in dose deposition and distribution caused by increased attenuation coefficient resulted from presence of SS are quantified using MCNP Monte Carlo simulations in coupled photon/electron transport. The source geometry was that of the Vari Source wire model VS2000. The FSD was that of the Varian medical system. In this model, the bending angles of tandem and colpostats are 15 degrees and 120 degrees , respectively. We assigned 10 dwell positions to the tandem and 4 dwell positions to right and left colpostats or ovoids to represent a typical treatment case. Typical dose delivered to point A was determined according to Manchester dosimetry system. Based on our computations, the reduction of dose to point A was shown to be at least 3%. So this effect presented by SS-FSD systems on patient dose is of concern.

MCNP and GADRAS Comparisons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Klasky, Marc Louis; Myers, Steven Charles; James, Michael R.

To facilitate the timely execution of System Threat Reviews (STRs) for DNDO, and also to develop a methodology for performing STRs, LANL performed comparisons of several radiation transport codes (MCNP, GADRAS, and Gamma-Designer) that have been previously utilized to compute radiation signatures. While each of these codes has strengths, it is of paramount interest to determine the limitations of each of the respective codes and also to identify the most time efficient means by which to produce computational results, given the large number of parametric cases that are anticipated in performing STR's. These comparisons serve to identify regions of applicabilitymore » for each code and provide estimates of uncertainty that may be anticipated. Furthermore, while performing these comparisons, examination of the sensitivity of the results to modeling assumptions was also examined. These investigations serve to enable the creation of the LANL methodology for performing STRs. Given the wide variety of radiation test sources, scenarios, and detectors, LANL calculated comparisons of the following parameters: decay data, multiplicity, device (n,γ) leakages, and radiation transport through representative scenes and shielding. This investigation was performed to understand potential limitations utilizing specific codes for different aspects of the STR challenges.« less

Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

NASA Astrophysics Data System (ADS)

Krása, Antonín; Kochetkov, Anatoly; Baeten, Peter; Vittiglio, Guido; Wagemans, Jan; Bécares, Vicente

2017-09-01

VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector). Discrepancies between experiments and Monte Carlo calculations (MCNP5) of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler) depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2) are presented.

A Monte-Carlo Benchmark of TRIPOLI-4® and MCNP on ITER neutronics

NASA Astrophysics Data System (ADS)

Blanchet, David; Pénéliau, Yannick; Eschbach, Romain; Fontaine, Bruno; Cantone, Bruno; Ferlet, Marc; Gauthier, Eric; Guillon, Christophe; Letellier, Laurent; Proust, Maxime; Mota, Fernando; Palermo, Iole; Rios, Luis; Guern, Frédéric Le; Kocan, Martin; Reichle, Roger

2017-09-01

Radiation protection and shielding studies are often based on the extensive use of 3D Monte-Carlo neutron and photon transport simulations. ITER organization hence recommends the use of MCNP-5 code (version 1.60), in association with the FENDL-2.1 neutron cross section data library, specifically dedicated to fusion applications. The MCNP reference model of the ITER tokamak, the `C-lite', is being continuously developed and improved. This article proposes to develop an alternative model, equivalent to the 'C-lite', but for the Monte-Carlo code TRIPOLI-4®. A benchmark study is defined to test this new model. Since one of the most critical areas for ITER neutronics analysis concerns the assessment of radiation levels and Shutdown Dose Rates (SDDR) behind the Equatorial Port Plugs (EPP), the benchmark is conducted to compare the neutron flux through the EPP. This problem is quite challenging with regard to the complex geometry and considering the important neutron flux attenuation ranging from 1014 down to 108 n•cm-2•s-1. Such code-to-code comparison provides independent validation of the Monte-Carlo simulations, improving the confidence in neutronic results.

Delta-ray Production in MCNP 6.2.0

NASA Astrophysics Data System (ADS)

Anderson, C.; McKinney, G.; Tutt, J.; James, M.

Secondary electrons in the form of delta-rays, also referred to as knock-on electrons, have been a feature of MCNP for electron and positron transport for over 20 years. While MCNP6 now includes transport for a suite of heavy-ions and charged particles from its integration with MCNPX, the production of delta-rays was still limited to electron and positron transport. In the newest release of MCNP6, version 6.2.0, delta-ray production has now been extended for all energetic charged particles. The basis of this production is the analytical formulation from Rossi and ICRU Report 37. This paper discusses the MCNP6 heavy charged-particle implementation and provides production results for several benchmark/test problems.

SU-F-BRA-12: End-User Oriented Tools and Procedures for Testing Brachytherapy TPSs Employing MBDCAs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peppa, V; Pappas, E; Lahanas, V

2015-06-15

Purpose: To develop user-oriented tools for commissioning and dosimetry testing of {sup 192}Ir brachytherapy treatment planning systems (TPSs) employing model based dose calculation algorithms (MBDCAs). Methods: A software tool (BrachyGuide) has been developed for the automatic generation of MCNP6 input files from any CT based plan exported in DICOM RT format from Elekta and Varian TPSs. BrachyGuide also facilitates the evaluation of imported Monte Carlo (MC) and TPS dose distributions in terms of % dose differences and gamma index (CT overlaid colormaps or relative frequency plots) as well as DVHs and related indices. For users not equipped to perform MC,more » a set of computational models was prepared in DICOM format, accompanied by treatment plans and corresponding MCNP6 generated reference data. BrachyGuide can then be used to compare institutional and reference data as per TG186. The model set includes a water sphere with the MBDCA WG {sup 192}Ir source placed centrically and in two eccentric positions, a water sphere with cubic bone and lung inhomogeneities and a five source dwells plan, and a patient equivalent model with an Accelerated Partial Breast Irradiation (APBI) plan. Results: The tools developed were used for the dosimetry testing of the Acuros and ACE MBDCAs implemented in BrachyVision v.13 and Oncentra Brachy v.4.5, respectively. Findings were consistent with previous results in the literature. Besides points close to the source dwells, Acuros was found to agree within type A uncertainties with the reference MC results. Differences greater than MC type A uncertainty were observed for ACE at distances >5cm from the source dwells and in bone. Conclusion: The tools developed are efficient for brachytherapy MBDCA planning commissioning and testing. Since they are appropriate for distribution over the web, they will be put at the AAPM WG MBDCA’s disposal. Research co-financed by the ESF and Greek funds. NSRF operational Program: Education and Lifelong Learning Investing in Knowledge Society-Aristeia. Varian Medical Systems and Nucletron, an Elekta company provided access to TPSs for research purposes. Miss Peppa was supported by IKY-fellowships of excellence for postgraduate studies in Greece,Siemens Program.« less

Spent nuclear fuel assembly inspection using neutron computed tomography

NASA Astrophysics Data System (ADS)

Pope, Chad Lee

The research presented here focuses on spent nuclear fuel assembly inspection using neutron computed tomography. Experimental measurements involving neutron beam transmission through a spent nuclear fuel assembly serve as benchmark measurements for an MCNP simulation model. Comparison of measured results to simulation results shows good agreement. Generation of tomography images from MCNP tally results was accomplished using adapted versions of built in MATLAB algorithms. Multiple fuel assembly models were examined to provide a broad set of conclusions. Tomography images revealing assembly geometric information including the fuel element lattice structure and missing elements can be obtained using high energy neutrons. A projection difference technique was developed which reveals the substitution of unirradiated fuel elements for irradiated fuel elements, using high energy neutrons. More subtle material differences such as altering the burnup of individual elements can be identified with lower energy neutrons provided the scattered neutron contribution to the image is limited. The research results show that neutron computed tomography can be used to inspect spent nuclear fuel assemblies for the purpose of identifying anomalies such as missing elements or substituted elements. The ability to identify anomalies in spent fuel assemblies can be used to deter diversion of material by increasing the risk of early detection as well as improve reprocessing facility operations by confirming the spent fuel configuration is as expected or allowing segregation if anomalies are detected.

National Dam Safety Program. Nelson Dam (Inventory Number VA 12501), James River Basin, Nelson County, Virginia. Phase I Inspection Report.

DTIC Science & Technology

1981-06-01

during tropical storm Camille. 5.4 Flood Potential: The 100-Year Flood, 1/2 PMF, and PH? were developed by use of the HEC-l computer program (Reference 2...Appendix V) and routed through the reservoir using the NWS-Dambreak computer program (Reference 3, Appendix V). Clark’s Tc and R coefficients for...AD-AO" 330 ARMY ENGINEER DISTRICT NORFOLK VA F/6 13/13 NATIONAL DAM SAFETY PROGRAM . NELSON DAM (INVENTORY NUMBER VA 12--ETC(U) JUN 81 B 0 TARANUNCL

User’s Guide for SHIPINT - A Computer Program to Compute Two Ship Interaction in Waves

DTIC Science & Technology

1996-08-01

P500693.PDF [Page: 1 of 84] Image Cover Sheet CLASSIFICATION SYSTEM NUMBER 500693 UNCLASSIFIED I llllll 111111111111111111111111111111111 TITLE...Halifax, Nova Scotia, Canada B3J 2X4 1 ---------· CONTRACTOR REPORT I I I I I iii;,"’: · 1 Defence Research Establishment Atlantic Canada...SUMMARY 1 Introduction 2 Coordinate Systems and Two Ship Motions 3 Flow Chart 4 Input Data File Description 4.1 shipint.in ........ . 4.2 paneLa.in

Status Report on the MCNP 2020 Initiative

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Forrest B.; Rising, Michael Evan

2017-10-02

The discussion below provides a status report on the MCNP 2020 initiative. It includes discussion of the history of MCNP 2020, accomplishments during 2013-17, priorities for near-term development, other related efforts, a brief summary, and a list of references for the plans and work accomplished.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Forrest B.

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.

DEAN: A Program for Dynamic Engine Analysis.

DTIC Science & Technology

1985-01-01

hardware and memory limitations. DIGTEM (ref. 4), a recently written code allows steady-state as well as transient calculations to be performed. DIGTEM has...Computer Program for Generating Dynamic Turbofan Engine Models ( DIGTEM )," NASA TM-83446. 5. Carnahan, B., Luther, H.A., and Wilkes, J.O., Applied Numerical

Monte Carlo N-particle simulation of neutron-based sterilisation of anthrax contamination

PubMed Central

Liu, B; Xu, J; Liu, T; Ouyang, X

2012-01-01

Objective To simulate the neutron-based sterilisation of anthrax contamination by Monte Carlo N-particle (MCNP) 4C code. Methods Neutrons are elementary particles that have no charge. They are 20 times more effective than electrons or γ-rays in killing anthrax spores on surfaces and inside closed containers. Neutrons emitted from a 252Cf neutron source are in the 100 keV to 2 MeV energy range. A 2.5 MeV D–D neutron generator can create neutrons at up to 1013 n s−1 with current technology. All these enable an effective and low-cost method of killing anthrax spores. Results There is no effect on neutron energy deposition on the anthrax sample when using a reflector that is thicker than its saturation thickness. Among all three reflecting materials tested in the MCNP simulation, paraffin is the best because it has the thinnest saturation thickness and is easy to machine. The MCNP radiation dose and fluence simulation calculation also showed that the MCNP-simulated neutron fluence that is needed to kill the anthrax spores agrees with previous analytical estimations very well. Conclusion The MCNP simulation indicates that a 10 min neutron irradiation from a 0.5 g 252Cf neutron source or a 1 min neutron irradiation from a 2.5 MeV D–D neutron generator may kill all anthrax spores in a sample. This is a promising result because a 2.5 MeV D–D neutron generator output >1013 n s−1 should be attainable in the near future. This indicates that we could use a D–D neutron generator to sterilise anthrax contamination within several seconds. PMID:22573293

Monte Carlo N-particle simulation of neutron-based sterilisation of anthrax contamination.

PubMed

Liu, B; Xu, J; Liu, T; Ouyang, X

2012-10-01

To simulate the neutron-based sterilisation of anthrax contamination by Monte Carlo N-particle (MCNP) 4C code. Neutrons are elementary particles that have no charge. They are 20 times more effective than electrons or γ-rays in killing anthrax spores on surfaces and inside closed containers. Neutrons emitted from a (252)Cf neutron source are in the 100 keV to 2 MeV energy range. A 2.5 MeV D-D neutron generator can create neutrons at up to 10(13) n s(-1) with current technology. All these enable an effective and low-cost method of killing anthrax spores. There is no effect on neutron energy deposition on the anthrax sample when using a reflector that is thicker than its saturation thickness. Among all three reflecting materials tested in the MCNP simulation, paraffin is the best because it has the thinnest saturation thickness and is easy to machine. The MCNP radiation dose and fluence simulation calculation also showed that the MCNP-simulated neutron fluence that is needed to kill the anthrax spores agrees with previous analytical estimations very well. The MCNP simulation indicates that a 10 min neutron irradiation from a 0.5 g (252)Cf neutron source or a 1 min neutron irradiation from a 2.5 MeV D-D neutron generator may kill all anthrax spores in a sample. This is a promising result because a 2.5 MeV D-D neutron generator output >10(13) n s(-1) should be attainable in the near future. This indicates that we could use a D-D neutron generator to sterilise anthrax contamination within several seconds.

How to Build MCNP 6.2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bull, Jeffrey S.

This presentation describes how to build MCNP 6.2. MCNP®* 6.2 can be compiled on Macs, PCs, and most Linux systems. It can also be built for parallel execution using both OpenMP and Messing Passing Interface (MPI) methods. MCNP6 requires Fortran, C, and C++ compilers to build the code.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zieb, Kristofer James Ekhart; Hughes, Henry Grady III; Xu, X. George

The release of version 6.2 of the MCNP6 radiation transport code is imminent. To complement the newest release, a summary of the heavy charged particle physics models used in the 1 MeV to 1 GeV energy regime is presented. Several changes have been introduced into the charged particle physics models since the merger of the MCNP5 and MCNPX codes into MCNP6. Here, this article discusses the default models used in MCNP6 for continuous energy loss, energy straggling, and angular scattering of heavy charged particles. Explanations of the physics models’ theories are included as well.

Delta-ray Production in MCNP 6.2.0

DOE PAGES

Anderson, Casey Alan; McKinney, Gregg Walter; Tutt, James Robert; ...

2017-10-26

Secondary electrons in the form of delta-rays, also referred to as knock-on electrons, have been a feature of MCNP for electron and positron transport for over 20 years. While MCNP6 now includes transport for a suite of heavy-ions and charged particles from its integration with MCNPX, the production of delta-rays was still limited to electron and positron transport. In the newest release of MCNP6, version 6.2.0, delta-ray production has now been extended for all energetic charged particles. The basis of this production is the analytical formulation from Rossi and ICRU Report 37. As a result, this paper discusses the MCNP6more » heavy charged-particle implementation and provides production results for several benchmark/test problems.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marck, Steven C. van der, E-mail: vandermarck@nrg.eu

Recent releases of three major world nuclear reaction data libraries, ENDF/B-VII.1, JENDL-4.0, and JEFF-3.1.1, have been tested extensively using benchmark calculations. The calculations were performed with the latest release of the continuous energy Monte Carlo neutronics code MCNP, i.e. MCNP6. Three types of benchmarks were used, viz. criticality safety benchmarks, (fusion) shielding benchmarks, and reference systems for which the effective delayed neutron fraction is reported. For criticality safety, more than 2000 benchmarks from the International Handbook of Criticality Safety Benchmark Experiments were used. Benchmarks from all categories were used, ranging from low-enriched uranium, compound fuel, thermal spectrum ones (LEU-COMP-THERM), tomore » mixed uranium-plutonium, metallic fuel, fast spectrum ones (MIX-MET-FAST). For fusion shielding many benchmarks were based on IAEA specifications for the Oktavian experiments (for Al, Co, Cr, Cu, LiF, Mn, Mo, Si, Ti, W, Zr), Fusion Neutronics Source in Japan (for Be, C, N, O, Fe, Pb), and Pulsed Sphere experiments at Lawrence Livermore National Laboratory (for {sup 6}Li, {sup 7}Li, Be, C, N, O, Mg, Al, Ti, Fe, Pb, D2O, H2O, concrete, polyethylene and teflon). The new functionality in MCNP6 to calculate the effective delayed neutron fraction was tested by comparison with more than thirty measurements in widely varying systems. Among these were measurements in the Tank Critical Assembly (TCA in Japan) and IPEN/MB-01 (Brazil), both with a thermal spectrum, two cores in Masurca (France) and three cores in the Fast Critical Assembly (FCA, Japan), all with fast spectra. The performance of the three libraries, in combination with MCNP6, is shown to be good. The results for the LEU-COMP-THERM category are on average very close to the benchmark value. Also for most other categories the results are satisfactory. Deviations from the benchmark values do occur in certain benchmark series, or in isolated cases within benchmark series. Such instances can often be related to nuclear data for specific non-fissile elements, such as C, Fe, or Gd. Indications are that the intermediate and mixed spectrum cases are less well described. The results for the shielding benchmarks are generally good, with very similar results for the three libraries in the majority of cases. Nevertheless there are, in certain cases, strong deviations between calculated and benchmark values, such as for Co and Mg. Also, the results show discrepancies at certain energies or angles for e.g. C, N, O, Mo, and W. The functionality of MCNP6 to calculate the effective delayed neutron fraction yields very good results for all three libraries.« less

Results on the neutron energy distribution measurements at the RECH-1 Chilean nuclear reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aguilera, P., E-mail: paguilera87@gmail.com; Romero-Barrientos, J.; Universidad de Chile, Dpto. de Física, Facultad de Ciencias, Las Palmeras 3425, Nuñoa, Santiago

2016-07-07

Neutron activations experiments has been perform at the RECH-1 Chilean Nuclear Reactor to measure its neutron flux energy distribution. Samples of pure elements was activated to obtain the saturation activities for each reaction. Using - ray spectroscopy we identify and measure the activity of the reaction product nuclei, obtaining the saturation activities of 20 reactions. GEANT4 and MCNP was used to compute the self shielding factor to correct the cross section for each element. With the Expectation-Maximization algorithm (EM) we were able to unfold the neutron flux energy distribution at dry tube position, near the RECH-1 core. In this work,more » we present the unfolding results using the EM algorithm.« less

Monte Carlo calculations of the incineration of plutonium and minor actinides of laser fusion inertial confinement fusion fission energy (LIFE) engine

NASA Astrophysics Data System (ADS)

Adem, ACIR; Eşref, BAYSAL

2018-07-01

In this paper, neutronic analysis in a laser fusion inertial confinement fusion fission energy (LIFE) engine fuelled plutonium and minor actinides using a MCNP codes was investigated. LIFE engine fuel zone contained 10 vol% TRISO particles and 90 vol% natural lithium coolant mixture. TRISO fuel compositions have Mod①: reactor grade plutonium (RG-Pu), Mod②: weapon grade plutonium (WG-Pu) and Mod③: minor actinides (MAs). Tritium breeding ratios (TBR) were computed as 1.52, 1.62 and 1.46 for Mod①, Mod② and Mod③, respectively. The operation period was computed as ∼21 years when the reference TBR > 1.05 for a self-sustained reactor for all investigated cases. Blanket energy multiplication values (M) were calculated as 4.18, 4.95 and 3.75 for Mod①, Mod② and Mod③, respectively. The burnup (BU) values were obtained as ∼1230, ∼1550 and ∼1060 GWd tM–1, respectively. As a result, the higher BU were provided with using TRISO particles for all cases in LIFE engine.

D-T Neutron Skyshine Experiments at JAERI/FNS

NASA Astrophysics Data System (ADS)

Nishitani, Takeo; Ochiai, Kentaro; Yoshida, Shigeo; Tanaka, Ryohei; Wakisaka, Masashi; Nakao, Makoto; Sato, Satoshi; Yamauchi, Michinori; Hori, Jun-Ichi; Takahashi, Akito; Kaneko, Jun-Ichi; Sawamura, Teruko

The D-T neutron skyshine experiments have been carried out at the Fusion Neutronics Source (FNS) of JAERI with the neutron yield of ˜1.7×1011n/s. The concrete thickness of the roof and the wall of a FNS target room are 1.15 and 2 m, respectively. The FNS skyshine port with a size of 0.9 × 0.9 m2 was open during the experimental period.The radiation dose rate outside the target room was measured as far as about 550 m away from the D-T target point with a spherical rem-counter. The highest neutron dose was about 0.5 μSv/hr at a distance of 30 m from the D-T target point and the dose rate was attenuated to 0.002 μSv/hr at a distance of 550 m. The measured neutron dose distribution was analyzed with Monte Carlo code MCNP-4B and a simple line source model. The MCNP calculation overestimates the neutron dose in the distance range larger than 250 m. The neutron spectra were evaluated with a 3He detector with different thickness of polyethylene neutron moderators. Secondary gamma-rays were measured with high purity Ge detectors and NaI scintillation detectors.

MAGIC Computer Simulation. Volume 2: Analyst Manual, Part 1

DTIC Science & Technology

1971-05-01

A review of the subject Magic Computer Simulation User and Analyst Manuals has been conducted based upon a request received from the US Army...1971 4. TITLE AND SUBTITLE MAGIC Computer Simulation Analyst Manual Part 1 5a. CONTRACT NUMBER 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6...14. ABSTRACT The MAGIC computer simulation generates target description data consisting of item-by-item listings of the target’s components and air

Computational analysis of the dose rates at JSI TRIGA reactor irradiation facilities.

PubMed

Ambrožič, K; Žerovnik, G; Snoj, L

2017-12-01

The JSI TRIGA Mark II, IJS research reactor is equipped with numerous irradiation positions, where samples can be irradiated by neutrons and γ-rays. Irradiation position selection is based on its properties, such as physical size and accessibility, as well as neutron and γ-ray spectra, flux and dose intensities. This paper presents an overview on the neutron and γ-ray fluxes, spectra and dose intensities calculations using Monte Carlo MCNP software and ENDF/B-VII.0 nuclear data libraries. The dose-rates are presented in terms of ambient dose equivalents, air kerma, and silicon dose equivalent. At full reactor power the neutron ambient dose equivalent ranges from 5.5×10 3 Svh -1 to 6×10 6 Svh -1 , silicon dose equivalent from 6×10 2 Gy/h si to 3×10 5 Gy/h si , and neutron air kerma from 4.3×10 3 Gyh -1 to 2×10 5 Gyh -1 . Ratio of fast (1MeV

Benchmarking of MCNP for calculating dose rates at an interim storage facility for nuclear waste.

PubMed

Heuel-Fabianek, Burkhard; Hille, Ralf

2005-01-01

During the operation of research facilities at Research Centre Jülich, Germany, nuclear waste is stored in drums and other vessels in an interim storage building on-site, which has a concrete shielding at the side walls. Owing to the lack of a well-defined source, measured gamma spectra were unfolded to determine the photon flux on the surface of the containers. The dose rate simulation, including the effects of skyshine, using the Monte Carlo transport code MCNP is compared with the measured dosimetric data at some locations in the vicinity of the interim storage building. The MCNP data for direct radiation confirm the data calculated using a point-kernel method. However, a comparison of the modelled dose rates for direct radiation and skyshine with the measured data demonstrate the need for a more precise definition of the source. Both the measured and the modelled dose rates verified the fact that the legal limits (<1 mSv a(-1)) are met in the area outside the perimeter fence of the storage building to which members of the public have access. Using container surface data (gamma spectra) to define the source may be a useful tool for practical calculations and additionally for benchmarking of computer codes if the discussed critical aspects with respect to the source can be addressed adequately.

Preliminary Shielding Analysis for HCCB TBM Transport

NASA Astrophysics Data System (ADS)

Miao, Peng; Zhao, Fengchao; Cao, Qixiang; Zhang, Guoshu; Feng, Kaiming

2015-09-01

A preliminary shielding analysis on the transport of the Chinese helium cooled ceramic breeder test blanket module (HCCB TBM) from France back to China after being irradiated in ITER is presented in this contribution. Emphasis was placed on irradiation safety during transport. The dose rate calculated by MCNP/4C for the conceptual package design satisfies the relevant dose limits from IAEA that the dose rate 3 m away from the surface of the package containing low specific activity III materials should be less than 10 mSv/h. The change with location and the time evolution of dose rates after shutdown have also been studied. This will be helpful for devising the detailed transport plan of HCCB TBM back to China in the near future. supported by the Major State Basic Research Development Program of China (973 Program) (No. 2013GB108000)

Multiple Detector Optimization for Hidden Radiation Source Detection

DTIC Science & Technology

2015-03-26

important in achieving operationally useful methods for optimizing detector emplacement, the 2-D attenuation model approach promises to speed up the...process of hidden source detection significantly. The model focused on detection of the full energy peak of a radiation source. Methods to optimize... radioisotope identification is possible without using a computationally intensive stochastic model such as the Monte Carlo n-Particle (MCNP) code

Monte Carlo N Particle code - Dose distribution of clinical electron beams in inhomogeneous phantoms

PubMed Central

Nedaie, H. A.; Mosleh-Shirazi, M. A.; Allahverdi, M.

2013-01-01

Electron dose distributions calculated using the currently available analytical methods can be associated with large uncertainties. The Monte Carlo method is the most accurate method for dose calculation in electron beams. Most of the clinical electron beam simulation studies have been performed using non- MCNP [Monte Carlo N Particle] codes. Given the differences between Monte Carlo codes, this work aims to evaluate the accuracy of MCNP4C-simulated electron dose distributions in a homogenous phantom and around inhomogeneities. Different types of phantoms ranging in complexity were used; namely, a homogeneous water phantom and phantoms made of polymethyl methacrylate slabs containing different-sized, low- and high-density inserts of heterogeneous materials. Electron beams with 8 and 15 MeV nominal energy generated by an Elekta Synergy linear accelerator were investigated. Measurements were performed for a 10 cm × 10 cm applicator at a source-to-surface distance of 100 cm. Individual parts of the beam-defining system were introduced into the simulation one at a time in order to show their effect on depth doses. In contrast to the first scattering foil, the secondary scattering foil, X and Y jaws and applicator provide up to 5% of the dose. A 2%/2 mm agreement between MCNP and measurements was found in the homogenous phantom, and in the presence of heterogeneities in the range of 1-3%, being generally within 2% of the measurements for both energies in a "complex" phantom. A full-component simulation is necessary in order to obtain a realistic model of the beam. The MCNP4C results agree well with the measured electron dose distributions. PMID:23533162

Specific absorbed fractions of electrons and photons for Rad-HUMAN phantom using Monte Carlo method

NASA Astrophysics Data System (ADS)

Wang, Wen; Cheng, Meng-Yun; Long, Peng-Cheng; Hu, Li-Qin

2015-07-01

The specific absorbed fractions (SAF) for self- and cross-irradiation are effective tools for the internal dose estimation of inhalation and ingestion intakes of radionuclides. A set of SAFs of photons and electrons were calculated using the Rad-HUMAN phantom, which is a computational voxel phantom of a Chinese adult female that was created using the color photographic image of the Chinese Visible Human (CVH) data set by the FDS Team. The model can represent most Chinese adult female anatomical characteristics and can be taken as an individual phantom to investigate the difference of internal dose with Caucasians. In this study, the emission of mono-energetic photons and electrons of 10 keV to 4 MeV energy were calculated using the Monte Carlo particle transport calculation code MCNP. Results were compared with the values from ICRP reference and ORNL models. The results showed that SAF from the Rad-HUMAN have similar trends but are larger than those from the other two models. The differences were due to the racial and anatomical differences in organ mass and inter-organ distance. The SAFs based on the Rad-HUMAN phantom provide an accurate and reliable data for internal radiation dose calculations for Chinese females. Supported by Strategic Priority Research Program of Chinese Academy of Sciences (XDA03040000), National Natural Science Foundation of China (910266004, 11305205, 11305203) and National Special Program for ITER (2014GB112001)

Verification of BWR Turbine Skyshine Dose with the MCNP5 Code Based on an Experiment Made at SHIMANE Nuclear Power Station

NASA Astrophysics Data System (ADS)

Tayama, Ryuichi; Wakasugi, Kenichi; Kawanaka, Ikunori; Kadota, Yoshinobu; Murakami, Yasuhiro

We measured the skyshine dose from turbine buildings at Shimane Nuclear Power Station Unit 1 (NS-1) and Unit 2 (NS-2), and then compared it with the dose calculated with the Monte Carlo transport code MCNP5. The skyshine dose values calculated with the MCNP5 code agreed with the experimental data within a factor of 2.8, when the roof of the turbine building was precisely modeled. We concluded that our MCNP5 calculation was valid for BWR turbine skyshine dose evaluation.

Comparison study of photon attenuation characteristics of Lead-Boron Polyethylene by MCNP code, XCOM and experimental data

NASA Astrophysics Data System (ADS)

Zhang, Lei; Jia, Mingchun; Gong, Junjun; Xia, Wenming

2017-08-01

The linear attenuation coefficient, mass attenuation coefficient and mean free path of various Lead-Boron Polyethylene (PbBPE) samples which can be used as the photon shielding materials in marine reactor have been simulated using the Monte Carlo N-Particle (MCNP)-5 code. The MCNP simulation results are in good agreement with the XCOM values and the reported experimental data for source Cesium-137 and Cobalt-60. Thus, this method based on MCNP can be used to simulate the photon attenuation characteristics of various types of PbBPE materials.

MCNP simulation of the dose distribution in liver cancer treatment for BNC therapy

NASA Astrophysics Data System (ADS)

Krstic, Dragana; Jovanovic, Zoran; Markovic, Vladimir; Nikezic, Dragoslav; Urosevic, Vlade

2014-10-01

The Boron Neutron Capture Therapy ( BNCT) is based on selective uptake of boron in tumour tissue compared to the surrounding normal tissue. Infusion of compounds with boron is followed by irradiation with neutrons. Neutron capture on 10B, which gives rise to an alpha particle and recoiled 7Li ion, enables the therapeutic dose to be delivered to tumour tissue while healthy tissue can be spared. Here, therapeutic abilities of BNCT were studied for possible treatment of liver cancer using thermal and epithermal neutron beam. For neutron transport MCNP software was used and doses in organs of interest in ORNL phantom were evaluated. Phantom organs were filled with voxels in order to obtain depth-dose distributions in them. The result suggests that BNCT using an epithermal neutron beam could be applied for liver cancer treatment.

Four pi calibration and modeling of a bare germanium detector in a cylindrical field source

NASA Astrophysics Data System (ADS)

Dewberry, R. A.; Young, J. E.

2012-05-01

In this paper we describe a 4π cylindrical field acquisition configuration surrounding a bare (unshielded, uncollimated) high purity germanium detector. We perform an efficiency calibration with a flexible planar source and model the configuration in the 4π cylindrical field. We then use exact calculus to model the flux on the cylindrical sides and end faces of the detector. We demonstrate that the model accurately represents the experimental detection efficiency compared to that of a point source and to Monte Carlo N-particle (MCNP) calculations of the flux. The model sums over the entire source surface area and the entire detector surface area including both faces and the detector's cylindrical sides. Agreement between the model and both experiment and the MCNP calculation is within 8%.

MCNP6.1 simulations for low-energy atomic relaxation: Code-to-code comparison with GATEv7.2, PENELOPE2014, and EGSnrc

NASA Astrophysics Data System (ADS)

Jung, Seongmoon; Sung, Wonmo; Lee, Jaegi; Ye, Sung-Joon

2018-01-01

Emerging radiological applications of gold nanoparticles demand low-energy electron/photon transport calculations including details of an atomic relaxation process. Recently, MCNP® version 6.1 (MCNP6.1) has been released with extended cross-sections for low-energy electron/photon, subshell photoelectric cross-sections, and more detailed atomic relaxation data than the previous versions. With this new feature, the atomic relaxation process of MCNP6.1 has not been fully tested yet with its new physics library (eprdata12) that is based on the Evaluated Atomic Data Library (EADL). In this study, MCNP6.1 was compared with GATEv7.2, PENELOPE2014, and EGSnrc that have been often used to simulate low-energy atomic relaxation processes. The simulations were performed to acquire both photon and electron spectra produced by interactions of 15 keV electrons or photons with a 10-nm-thick gold nano-slab. The photon-induced fluorescence X-rays from MCNP6.1 fairly agreed with those from GATEv7.2 and PENELOPE2014, while the electron-induced fluorescence X-rays of the four codes showed more or less discrepancies. A coincidence was observed in the photon-induced Auger electrons simulated by MCNP6.1 and GATEv7.2. A recent release of MCNP6.1 with eprdata12 can be used to simulate the photon-induced atomic relaxation.

Thermal Neutron Point Source Imaging using a Rotating Modulation Collimator (RMC)

DTIC Science & Technology

2010-03-01

Source Details.........................................................................................37 3.5 Simulation of RMC in MCNP ...passed through the masks at each rotation angle. ................................. 42 19. Figure 19: MCNP Generate Modulation Profile for Cadmium. The...Cadmium. The multi-energetic neutron source simulation from MCNP is used for this plot. The energy is values are shown per energy bin. The

Shuttle cryogenics supply system. Optimization study. Volume 5 B-4: Programmers manual for space shuttle orbit injection analysis (SOPSA)

NASA Technical Reports Server (NTRS)

1973-01-01

A computer program for space shuttle orbit injection propulsion system analysis (SOPSA) is described to show the operational characteristics and the computer system requirements. The program was developed as an analytical tool to aid in the preliminary design of propellant feed systems for the space shuttle orbiter main engines. The primary purpose of the program is to evaluate the propellant tank ullage pressure requirements imposed by the need to accelerate propellants rapidly during the engine start sequence. The SOPSA program will generate parametric feed system pressure histories and weight data for a range of nominal feedline sizes.

77 FR 12848 - Medicare Program; Solicitation of Independent Accrediting Organizations To Participate in the...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-02

...)(4)(B) (excluding x-ray, ultrasound, and fluoroscopy), as specified by the Secretary in consultation... imaging services as ``imaging and computer-assisted imaging services, including x-ray, ultrasound...

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adigun, Babatunde John; Fensin, Michael Lorne; Galloway, Jack D.

Our burnup study examined the effect of a predicted critical control rod position on the nuclide predictability of several axial and radial locations within a 4×4 graphite moderated gas cooled reactor fuel cluster geometry. To achieve this, a control rod position estimator (CRPE) tool was developed within the framework of the linkage code Monteburns between the transport code MCNP and depletion code CINDER90, and four methodologies were proposed within the tool for maintaining criticality. Two of the proposed methods used an inverse multiplication approach - where the amount of fissile material in a set configuration is slowly altered until criticalitymore » is attained - in estimating the critical control rod position. Another method carried out several MCNP criticality calculations at different control rod positions, then used a linear fit to estimate the critical rod position. The final method used a second-order polynomial fit of several MCNP criticality calculations at different control rod positions to guess the critical rod position. The results showed that consistency in prediction of power densities as well as uranium and plutonium isotopics was mutual among methods within the CRPE tool that predicted critical position consistently well. Finall, while the CRPE tool is currently limited to manipulating a single control rod, future work could be geared toward implementing additional criticality search methodologies along with additional features.« less

Coupled multi-group neutron photon transport for the simulation of high-resolution gamma-ray spectroscopy applications

NASA Astrophysics Data System (ADS)

Burns, Kimberly Ann

The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples. In these applications, high-resolution gamma-ray spectrometers are used to preserve as much information as possible about the emitted photon flux, which consists of both continuum and characteristic gamma rays with discrete energies. Monte Carlo transport is the most commonly used modeling tool for this type of problem, but computational times for many problems can be prohibitive. This work explores the use of coupled Monte Carlo-deterministic methods for the simulation of neutron-induced photons for high-resolution gamma-ray spectroscopy applications. RAdiation Detection Scenario Analysis Toolbox (RADSAT), a code which couples deterministic and Monte Carlo transport to perform radiation detection scenario analysis in three dimensions [1], was used as the building block for the methods derived in this work. RADSAT was capable of performing coupled deterministic-Monte Carlo simulations for gamma-only and neutron-only problems. The purpose of this work was to develop the methodology necessary to perform coupled neutron-photon calculations and add this capability to RADSAT. Performing coupled neutron-photon calculations requires four main steps: the deterministic neutron transport calculation, the neutron-induced photon spectrum calculation, the deterministic photon transport calculation, and the Monte Carlo detector response calculation. The necessary requirements for each of these steps were determined. A major challenge in utilizing multigroup deterministic transport methods for neutron-photon problems was maintaining the discrete neutron-induced photon signatures throughout the simulation. Existing coupled neutron-photon cross-section libraries and the methods used to produce neutron-induced photons were unsuitable for high-resolution gamma-ray spectroscopy applications. Central to this work was the development of a method for generating multigroup neutron-photon cross-sections in a way that separates the discrete and continuum photon emissions so the neutron-induced photon signatures were preserved. The RADSAT-NG cross-section library was developed as a specialized multigroup neutron-photon cross-section set for the simulation of high-resolution gamma-ray spectroscopy applications. The methodology and cross sections were tested using code-to-code comparison with MCNP5 [2] and NJOY [3]. A simple benchmark geometry was used for all cases compared with MCNP. The geometry consists of a cubical sample with a 252Cf neutron source on one side and a HPGe gamma-ray spectrometer on the opposing side. Different materials were examined in the cubical sample: polyethylene (C2H4), P, N, O, and Fe. The cross sections for each of the materials were compared to cross sections collapsed using NJOY. Comparisons of the volume-averaged neutron flux within the sample, volume-averaged photon flux within the detector, and high-purity gamma-ray spectrometer response (only for polyethylene) were completed using RADSAT and MCNP. The code-to-code comparisons show promising results for the coupled Monte Carlo-deterministic method. The RADSAT-NG cross-section production method showed good agreement with NJOY for all materials considered although some additional work is needed in the resonance region and in the first and last energy bin. Some cross section discrepancies existed in the lowest and highest energy bin, but the overall shape and magnitude of the two methods agreed. For the volume-averaged photon flux within the detector, typically the five most intense lines agree to within approximately 5% of the MCNP calculated flux for all of materials considered. The agreement in the code-to-code comparisons cases demonstrates a proof-of-concept of the method for use in RADSAT for coupled neutron-photon problems in high-resolution gamma-ray spectroscopy applications. One of the primary motivators for using the coupled method over pure Monte Carlo method is the potential for significantly lower computational times. For the code-to-code comparison cases, the run times for RADSAT were approximately 25--500 times shorter than for MCNP, as shown in Table 1. This was assuming a 40 mCi 252Cf neutron source and 600 seconds of "real-world" measurement time. The only variance reduction technique implemented in the MCNP calculation was forward biasing of the source toward the sample target. Improved MCNP runtimes could be achieved with the addition of more advanced variance reduction techniques.

Design of a boron neutron capture enhanced fast neutron therapy assembly

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Zhonglu

The use of boron neutron capture to boost tumor dose in fast neutron therapy has been investigated at several fast neutron therapy centers worldwide. This treatment is termed boron neutron capture enhanced fast neutron therapy (BNCEFNT). It is a combination of boron neutron capture therapy (BNCT) and fast neutron therapy (FNT). It is believed that BNCEFNT may be useful in the treatment of some radioresistant brain tumors, such as glioblastoma multiform (GBM). A boron neutron capture enhanced fast neutron therapy assembly has been designed for the Fermilab Neutron Therapy Facility (NTF). This assembly uses a tungsten filter and collimator nearmore » the patient's head, with a graphite reflector surrounding the head to significantly increase the dose due to boron neutron capture reactions. The assembly was designed using Monte Carlo radiation transport code MCNP version 5 for a standard 20x20 cm 2 treatment beam. The calculated boron dose enhancement at 5.7-cm depth in a water-filled head phantom in the assembly with a 5x5 cm 2 collimation was 21.9% per 100-ppm 10B for a 5.0-cm tungsten filter and 29.8% for a 8.5-cm tungsten filter. The corresponding dose rate for the 5.0-cm and 8.5-cm thick filters were 0.221 and 0.127 Gy/min, respectively; about 48.5% and 27.9% of the dose rate of the standard 10x10 cm 2 fast neutron treatment beam. To validate the design calculations, a simplified BNCEFNT assembly was built using four lead bricks to form a 5x5 cm 2 collimator. Five 1.0-cm thick 20x20 cm 2 tungsten plates were used to obtain different filter thicknesses and graphite bricks/blocks were used to form a reflector. Measurements of the dose enhancement of the simplified assembly in a water-filled head phantom were performed using a pair of tissue-equivalent ion chambers. One of the ion chambers is loaded with 1000-ppm natural boron (184-ppm 10B) to measure dose due to boron neutron capture. The measured dose enhancement at 5.0-cm depth in the head phantom for the 5.0-cm thick tungsten filter is (16.6 ± 1.8)%, which agrees well with the MCNP simulation of the simplified BNCEFNT assembly, (16.4 ± 0.5)%. The error in the calculated dose enhancement only considers the statistical uncertainties. The total dose rate measured at 5.0-cm depth using the non-borated ion chamber is (0.765 ± 0.076) Gy/MU, about 61% of the fast neutron standard dose rate (1.255Gy/MU) at 5.0-cm depth for the standard 10x10 cm 2 treatment beam. The increased doses to other organs due to the use of the BNCEFNT assembly were calculated using MCNP5 and a MIRD phantom. The activities of the activation products produced in the BNCEFNT assembly after neutron beam delivery were computed. The photon ambient dose rate due to the radioactive activation products was also estimated.« less

Advanced Software Techniques for Data Management Systems. Volume 2: Space Shuttle Flight Executive System: Functional Design

NASA Technical Reports Server (NTRS)

Pepe, J. T.

1972-01-01

A functional design of software executive system for the space shuttle avionics computer is presented. Three primary functions of the executive are emphasized in the design: task management, I/O management, and configuration management. The executive system organization is based on the applications software and configuration requirements established during the Phase B definition of the Space Shuttle program. Although the primary features of the executive system architecture were derived from Phase B requirements, it was specified for implementation with the IBM 4 Pi EP aerospace computer and is expected to be incorporated into a breadboard data management computer system at NASA Manned Spacecraft Center's Information system division. The executive system was structured for internal operation on the IBM 4 Pi EP system with its external configuration and applications software assumed to the characteristic of the centralized quad-redundant avionics systems defined in Phase B.

The MCNP-DSP code for calculations of time and frequency analysis parameters for subcritical systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Valentine, T.E.; Mihalczo, J.T.

1995-12-31

This paper describes a modified version of the MCNP code, the MCNP-DSP. Variance reduction features were disabled to have strictly analog particle tracking in order to follow fluctuating processes more accurately. Some of the neutron and photon physics routines were modified to better represent the production of particles. Other modifications are discussed.

Computational Model of D-Region Ion Production Caused by Energetic Electron Precipitations Based on General Monte Carlo Transport Calculations

NASA Astrophysics Data System (ADS)

Kouznetsov, A.; Cully, C. M.

2017-12-01

During enhanced magnetic activities, large ejections of energetic electrons from radiation belts are deposited in the upper polar atmosphere where they play important roles in its physical and chemical processes, including VLF signals subionospheric propagation. Electron deposition can affect D-Region ionization, which are estimated based on ionization rates derived from energy depositions. We present a model of D-region ion production caused by an arbitrary (in energy and pitch angle) distribution of fast (10 keV - 1 MeV) electrons. The model relies on a set of pre-calculated results obtained using a general Monte Carlo approach with the latest version of the MCNP6 (Monte Carlo N-Particle) code for the explicit electron tracking in magnetic fields. By expressing those results using the ionization yield functions, the pre-calculated results are extended to cover arbitrary magnetic field inclinations and atmospheric density profiles, allowing ionization rate altitude profile computations in the range of 20 and 200 km at any geographic point of interest and date/time by adopting results from an external atmospheric density model (e.g. NRLMSISE-00). The pre-calculated MCNP6 results are stored in a CDF (Common Data Format) file, and IDL routines library is written to provide an end-user interface to the model.

MCNP6 simulation of radiographs generated from megaelectron volt X-rays for characterizing a computed tomography system

NASA Astrophysics Data System (ADS)

Dooraghi, Alex A.; Tringe, Joseph W.

2018-04-01

To evaluate conventional munition, we simulated an x-ray computed tomography (CT) system for generating radiographs from nominal x-ray energies of 6 or 9 megaelectron volts (MeV). CT simulations, informed by measured data, allow for optimization of both system design and acquisition techniques necessary to enhance image quality. MCNP6 radiographic simulation tools were used to model ideal detector responses (DR) that assume either (1) a detector response proportional to photon flux (N) or (2) a detector response proportional to energy flux (E). As scatter may become significant with MeV x-ray systems, simulations were performed with and without the inclusion of object scatter. Simulations were compared against measurements of a cylindrical munition component principally composed of HMX, tungsten and aluminum encased in carbon fiber. Simulations and measurements used a 6 MeV peak energy x-ray spectrum filtered with 3.175 mm of tantalum. A detector response proportional to energy which includes object scatter agrees to within 0.6 % of the measured line integral of the linear attenuation coefficient. Exclusion of scatter increases the difference between measurement and simulation to 5 %. A detector response proportional to photon flux agrees to within 20 % when object scatter is included in the simulation and 27 % when object scatter is excluded.

32 CFR 32.51 - Monitoring and reporting program performance.

Code of Federal Regulations, 2012 CFR

2012-07-01

... readily quantified, such quantitative data should be related to cost data for computation of unit costs. However, unit costs are generally inappropriate for research (see § 32.21 (a) and (b)(4)). (2) Reasons why...

32 CFR 32.51 - Monitoring and reporting program performance.

Code of Federal Regulations, 2011 CFR

2011-07-01

... readily quantified, such quantitative data should be related to cost data for computation of unit costs. However, unit costs are generally inappropriate for research (see § 32.21 (a) and (b)(4)). (2) Reasons why...

32 CFR 32.51 - Monitoring and reporting program performance.

Code of Federal Regulations, 2013 CFR

2013-07-01

... readily quantified, such quantitative data should be related to cost data for computation of unit costs. However, unit costs are generally inappropriate for research (see § 32.21 (a) and (b)(4)). (2) Reasons why...

32 CFR 32.51 - Monitoring and reporting program performance.

Code of Federal Regulations, 2014 CFR

2014-07-01

... readily quantified, such quantitative data should be related to cost data for computation of unit costs. However, unit costs are generally inappropriate for research (see § 32.21 (a) and (b)(4)). (2) Reasons why...

32 CFR 32.51 - Monitoring and reporting program performance.

Code of Federal Regulations, 2010 CFR

2010-07-01

... readily quantified, such quantitative data should be related to cost data for computation of unit costs. However, unit costs are generally inappropriate for research (see § 32.21 (a) and (b)(4)). (2) Reasons why...

Strategic Control Algorithm Development : Volume 4B. Computer Program Report (Concluded)

DOT National Transportation Integrated Search

1974-08-01

A description of the strategic algorithm evaluation model is presented, both at the user and programmer levels. The model representation of an airport configuration, environmental considerations, the strategic control algorithm logic, and the airplan...

Numerical analysis of stiffened shells of revolution. Volume 3: Users' manual for STARS-2B, 2V, shell theory automated for rotational structures, 2 (buckling, vibrations), digital computer programs

NASA Technical Reports Server (NTRS)

Svalbonas, V.

1973-01-01

The User's manual for the shell theory automated for rotational structures (STARS) 2B and 2V (buckling, vibrations) is presented. Several features of the program are: (1) arbitrary branching of the shell meridians, (2) arbitrary boundary conditions, (3) minimum input requirements to describe a complex, practical shell of revolution structure, and (4) accurate analysis capability using a minimum number of degrees of freedom.

An MCNP-based model of a medical linear accelerator x-ray photon beam.

PubMed

Ajaj, F A; Ghassal, N M

2003-09-01

The major components in the x-ray photon beam path of the treatment head of the VARIAN Clinac 2300 EX medical linear accelerator were modeled and simulated using the Monte Carlo N-Particle radiation transport computer code (MCNP). Simulated components include x-ray target, primary conical collimator, x-ray beam flattening filter and secondary collimators. X-ray photon energy spectra and angular distributions were calculated using the model. The x-ray beam emerging from the secondary collimators were scored by considering the total x-ray spectra from the target as the source of x-rays at the target position. The depth dose distribution and dose profiles at different depths and field sizes have been calculated at a nominal operating potential of 6 MV and found to be within acceptable limits. It is concluded that accurate specification of the component dimensions, composition and nominal accelerating potential gives a good assessment of the x-ray energy spectra.

Analysis of MCNP simulated gamma spectra of CdTe detectors for boron neutron capture therapy.

PubMed

Winkler, Alexander; Koivunoro, Hanna; Savolainen, Sauli

2017-06-01

The next step in the boron neutron capture therapy (BNCT) is the real time imaging of the boron concentration in healthy and tumor tissue. Monte Carlo simulations are employed to predict the detector response required to realize single-photon emission computed tomography in BNCT, but have failed to correctly resemble measured data for cadmium telluride detectors. In this study we have tested the gamma production cross-section data tables of commonly used libraries in the Monte Carlo code MCNP in comparison to measurements. The cross section data table TENDL-2008-ACE is reproducing measured data best, whilst the commonly used ENDL92 and other studied libraries do not include correct tables for the gamma production from the cadmium neutron capture reaction that is occurring inside the detector. Furthermore, we have discussed the size of the annihilation peaks of spectra obtained by cadmium telluride and germanium detectors. Copyright © 2017 Elsevier Ltd. All rights reserved.

MCNP-based computational model for the Leksell gamma knife.

PubMed

Trnka, Jiri; Novotny, Josef; Kluson, Jaroslav

2007-01-01

We have focused on the usage of MCNP code for calculation of Gamma Knife radiation field parameters with a homogenous polystyrene phantom. We have investigated several parameters of the Leksell Gamma Knife radiation field and compared the results with other studies based on EGS4 and PENELOPE code as well as the Leksell Gamma Knife treatment planning system Leksell GammaPlan (LGP). The current model describes all 201 radiation beams together and simulates all the sources in the same time. Within each beam, it considers the technical construction of the source, the source holder, collimator system, the spherical phantom, and surrounding material. We have calculated output factors for various sizes of scoring volumes, relative dose distributions along basic planes including linear dose profiles, integral doses in various volumes, and differential dose volume histograms. All the parameters have been calculated for each collimator size and for the isocentric configuration of the phantom. We have found the calculated output factors to be in agreement with other authors' works except the case of 4 mm collimator size, where averaging over the scoring volume and statistical uncertainties strongly influences the calculated results. In general, all the results are dependent on the choice of the scoring volume. The calculated linear dose profiles and relative dose distributions also match independent studies and the Leksell GammaPlan, but care must be taken about the fluctuations within the plateau, which can influence the normalization, and accuracy in determining the isocenter position, which is important for comparing different dose profiles. The calculated differential dose volume histograms and integral doses have been compared with data provided by the Leksell GammaPlan. The dose volume histograms are in good agreement as well as integral doses calculated in small calculation matrix volumes. However, deviations in integral doses up to 50% can be observed for large volumes such as for the total skull volume. The differences observed in treatment of scattered radiation between the MC method and the LGP may be important in this case. We have also studied the influence of differential direction sampling of primary photons and have found that, due to the anisotropic sampling, doses around the isocenter deviate from each other by up to 6%. With caution about the details of the calculation settings, it is possible to employ the MCNP Monte Carlo code for independent verification of the Leksell Gamma Knife radiation field properties.

Background-Source Cosmic-Photon Elevation Scaling and Cosmic-Neutron/Photon Date Scaling in MCNP6

DOE PAGES

Tutt, James Robert; Anderson, Casey Alan; McKinney, Gregg Walter

2017-10-26

Here, cosmic neutron and photon fluxes are known to scale exponentially with elevation. Consequently, cosmic neutron elevation scaling was implemented for use with the background-source option shortly after its introduction into MCNP6, whereby the neutron flux weight factor was adjusted by the elevation scaling factor when the user-specified elevation differed from the selected background.dat grid-point elevation. At the same time, an elevation scaling factor was suggested for the cosmic photon flux, however, cosmic photon elevation scaling is complicated by the fact that the photon background consists of two components: cosmic and terrestrial. Previous versions of the background.dat file did notmore » provide any way to separate these components. With Rel. 4 of this file in 2015, two new columns were added that provide the energy grid and differential cosmic photon flux separately from the total photon flux. Here we show that the cosmic photon flux component can now be scaled independently and combined with the terrestrial component to form the total photon flux at a user-specified elevation in MCNP6.« less

Background-Source Cosmic-Photon Elevation Scaling and Cosmic-Neutron/Photon Date Scaling in MCNP6

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tutt, James Robert; Anderson, Casey Alan; McKinney, Gregg Walter

Here, cosmic neutron and photon fluxes are known to scale exponentially with elevation. Consequently, cosmic neutron elevation scaling was implemented for use with the background-source option shortly after its introduction into MCNP6, whereby the neutron flux weight factor was adjusted by the elevation scaling factor when the user-specified elevation differed from the selected background.dat grid-point elevation. At the same time, an elevation scaling factor was suggested for the cosmic photon flux, however, cosmic photon elevation scaling is complicated by the fact that the photon background consists of two components: cosmic and terrestrial. Previous versions of the background.dat file did notmore » provide any way to separate these components. With Rel. 4 of this file in 2015, two new columns were added that provide the energy grid and differential cosmic photon flux separately from the total photon flux. Here we show that the cosmic photon flux component can now be scaled independently and combined with the terrestrial component to form the total photon flux at a user-specified elevation in MCNP6.« less

Extensions of the MCNP5 and TRIPOLI4 Monte Carlo Codes for Transient Reactor Analysis

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard; Sjenitzer, Bart L.

2014-06-01

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.

Characterization of gamma rays existing in the NMIJ standard neutron field.

PubMed

Harano, H; Matsumoto, T; Ito, Y; Uritani, A; Kudo, K

2004-01-01

Our laboratory provides national standards on fast neutron fluence. Neutron fields are always accompanied by gamma rays produced in neutron sources and surroundings. We have characterised these gamma rays in the 5.0 MeV standard neutron field. Gamma ray measurement was performed using an NE213 liquid scintillator. Pulse shape discrimination was incorporated to separate the events induced by gamma rays from those by neutrons. The measured gamma ray spectra were unfolded with the HEPRO program package to obtain the spectral fluences using the response matrix prepared with the EGS4 code. Corrections were made for the gamma rays produced by neutrons in the detector assembly using the MCNP4C code. The effective dose equivalents were estimated to be of the order of 25 microSv at the neutron fluence of 10(7) neutrons cm(-2).

Dosimetric verification of the anisotropic analytical algorithm in lung equivalent heterogeneities with and without bone equivalent heterogeneities

PubMed Central

Ono, Kaoru; Endo, Satoru; Tanaka, Kenichi; Hoshi, Masaharu; Hirokawa, Yutaka

2010-01-01

Purpose: In this study, the authors evaluated the accuracy of dose calculations performed by the convolution∕superposition based anisotropic analytical algorithm (AAA) in lung equivalent heterogeneities with and without bone equivalent heterogeneities. Methods: Calculations of PDDs using the AAA and Monte Carlo simulations (MCNP4C) were compared to ionization chamber measurements with a heterogeneous phantom consisting of lung equivalent and bone equivalent materials. Both 6 and 10 MV photon beams of 4×4 and 10×10 cm2 field sizes were used for the simulations. Furthermore, changes of energy spectrum with depth for the heterogeneous phantom using MCNP were calculated. Results: The ionization chamber measurements and MCNP calculations in a lung equivalent phantom were in good agreement, having an average deviation of only 0.64±0.45%. For both 6 and 10 MV beams, the average deviation was less than 2% for the 4×4 and 10×10 cm2 fields in the water-lung equivalent phantom and the 4×4 cm2 field in the water-lung-bone equivalent phantom. Maximum deviations for the 10×10 cm2 field in the lung equivalent phantom before and after the bone slab were 5.0% and 4.1%, respectively. The Monte Carlo simulation demonstrated an increase of the low-energy photon component in these regions, more for the 10×10 cm2 field compared to the 4×4 cm2 field. Conclusions: The low-energy photon by Monte Carlo simulation component increases sharply in larger fields when there is a significant presence of bone equivalent heterogeneities. This leads to great changes in the build-up and build-down at the interfaces of different density materials. The AAA calculation modeling of the effect is not deemed to be sufficiently accurate. PMID:20879604

Parallel Flux Tensor Analysis for Efficient Moving Object Detection

DTIC Science & Technology

2011-07-01

computing as well as parallelization to enable real time performance in analyzing complex video [3, 4 ]. There are a number of challenging computer vision... 4 . TITLE AND SUBTITLE Parallel Flux Tensor Analysis for Efficient Moving Object Detection 5a. CONTRACT NUMBER 5b. GRANT NUMBER 5c. PROGRAM ELEMENT...We use the trace of the flux tensor matrix, referred to as Tr JF , that is defined below, Tr JF = ∫ Ω W (x− y)(I2xt(y) + I2yt(y) + I2tt(y))dy ( 4 ) as

Computer Program for Estimating Stability Derivatives of Missile Configurations - Users Manual.

DTIC Science & Technology

1982-08-01

D.C., 1959, NACA Report 1307. 6. Adams, Gaynor J. and Dugan, Duane W., Td rDQJ.Iin Rol ld Rlling Moen PU Ig Di fferen tia 1 kgIni~daence~ f=~ Slende...1,.b.8,.2 L)AIA IMN O,!,2.,,.,b.,$ b.994 DA 1.$, 1 I Ib, . I b . I , 31. 2, 1 .;?, 137,134 DATA IMN2/..1.,34.b DATA rCL)C2 /* 3Uv ,31v, 4$, 72v.9bvl

Advanced foil activation techniques for the measurement of within-pin distributions of the 63Cu(n,γ) 64Cu reaction rate in nuclear fuel

NASA Astrophysics Data System (ADS)

Macku, K.; Jatuff, F.; Murphy, M. F.; Joneja, O. P.; Bischofberger, R.; Chawla, R.

2006-06-01

Different foil activation techniques have been used for measuring spatial distributions of the 63Cu(n,γ) 64Cu reaction within two pins of a SVEA-96 Optima2 boiling water reactor fuel assembly, at the critical facility PROTEUS. This reaction is of interest because its 1/v cross-section gives it a good representation of the 235U fission rate. Initially, radial capture rate profiles were measured with mechanically punched copper foils. More detailed profiles were then determined by using a 0.2 mm copper wire spiral (˜200 μm resolution), as well as 5-, 10-, and 20-ring UV-lithography, electroplating, and molding (UV-LIGA) foils (up to a 100 μm resolution). For azimuthal measurements, apart from manually cut activation foils (into 8 sectors), 8- and 12-sector LIGA foils were used. The highly versatile LIGA foils have the additional advantage of being very easily separated into individual pieces after irradiation without the use of punches or other cutting tools. In order to account for the invasive character of the foil activation techniques, corrections to account for sample perturbations and for self-shielding effects were determined via simplified Monte Carlo (MCNP4C) modeling of the experimental setup. The final results from the various measurements of 63Cu(n,γ) 64Cu within-pin distributions have been compared with MCNP computations employing a detailed model of the full SVEA Optima2 fuel assembly.

Neutron and photon shielding benchmark calculations by MCNP on the LR-0 experimental facility.

PubMed

Hordósy, G

2005-01-01

In the framework of the REDOS project, the space-energy distribution of the neutron and photon flux has been calculated over the pressure vessel simulator thickness of the LR-0 experimental reactor, Rez, Czech Republic. The results calculated by the Monte Carlo code MCNP4C are compared with the measurements performed in the Nuclear Research Institute, Rez. The spectra have been measured at the barrel, in front of, inside and behind the pressure vessel in different configurations. The neutron measurements were performed in the energy range 0.1-10 MeV. This work has been done in the frame of the 5th Frame Work Programme of the European Community 1998-2002.

DPEMC: A Monte Carlo for double diffraction

NASA Astrophysics Data System (ADS)

Boonekamp, M.; Kúcs, T.

2005-05-01

We extend the POMWIG Monte Carlo generator developed by B. Cox and J. Forshaw, to include new models of central production through inclusive and exclusive double Pomeron exchange in proton-proton collisions. Double photon exchange processes are described as well, both in proton-proton and heavy-ion collisions. In all contexts, various models have been implemented, allowing for comparisons and uncertainty evaluation and enabling detailed experimental simulations. Program summaryTitle of the program:DPEMC, version 2.4 Catalogue identifier: ADVF Program summary URL:http://cpc.cs.qub.ac.uk/summaries/ADVF Program obtainable from: CPC Program Library, Queen's University of Belfast, N. Ireland Computer: any computer with the FORTRAN 77 compiler under the UNIX or Linux operating systems Operating system: UNIX; Linux Programming language used: FORTRAN 77 High speed storage required:<25 MB No. of lines in distributed program, including test data, etc.: 71 399 No. of bytes in distributed program, including test data, etc.: 639 950 Distribution format: tar.gz Nature of the physical problem: Proton diffraction at hadron colliders can manifest itself in many forms, and a variety of models exist that attempt to describe it [A. Bialas, P.V. Landshoff, Phys. Lett. B 256 (1991) 540; A. Bialas, W. Szeremeta, Phys. Lett. B 296 (1992) 191; A. Bialas, R.A. Janik, Z. Phys. C 62 (1994) 487; M. Boonekamp, R. Peschanski, C. Royon, Phys. Rev. Lett. 87 (2001) 251806; Nucl. Phys. B 669 (2003) 277; R. Enberg, G. Ingelman, A. Kissavos, N. Timneanu, Phys. Rev. Lett. 89 (2002) 081801; R. Enberg, G. Ingelman, L. Motyka, Phys. Lett. B 524 (2002) 273; R. Enberg, G. Ingelman, N. Timneanu, Phys. Rev. D 67 (2003) 011301; B. Cox, J. Forshaw, Comput. Phys. Comm. 144 (2002) 104; B. Cox, J. Forshaw, B. Heinemann, Phys. Lett. B 540 (2002) 26; V. Khoze, A. Martin, M. Ryskin, Phys. Lett. B 401 (1997) 330; Eur. Phys. J. C 14 (2000) 525; Eur. Phys. J. C 19 (2001) 477; Erratum, Eur. Phys. J. C 20 (2001) 599; Eur. Phys. J. C 23 (2002) 311]. This program implements some of the more significant ones, enabling the simulation of central particle production through color singlet exchange between interacting protons or antiprotons. Method of solution: The Monte Carlo method is used to simulate all elementary 2→2 and 2→1 processes available in HERWIG. The color singlet exchanges implemented in DPEMC are implemented as functions reweighting the photon flux already present in HERWIG. Restriction on the complexity of the problem: The program relying extensively on HERWIG, the limitations are the same as in [G. Marchesini, B.R. Webber, G. Abbiendi, I.G. Knowles, M.H. Seymour, L. Stanco, Comput. Phys. Comm. 67 (1992) 465; G. Corcella, I.G. Knowles, G. Marchesini, S. Moretti, K. Odagiri, P. Richardson, M. Seymour, B. Webber, JHEP 0101 (2001) 010]. Typical running time: Approximate times on a 800 MHz Pentium III: 5-20 min per 10 000 unweighted events, depending on the process under consideration.

Development of PIMAL: Mathematical Phantom with Moving Arms and Legs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Akkurt, Hatice; Eckerman, Keith F.

2007-05-01

The computational model of the human anatomy (phantom) has gone through many revisions since its initial development in the 1970s. The computational phantom model currently used by the Nuclear Regulatory Commission (NRC) is based on a model published in 1974. Hence, the phantom model used by the NRC staff was missing some organs (e.g., neck, esophagus) and tissues. Further, locations of some organs were inappropriate (e.g., thyroid).Moreover, all the computational phantoms were assumed to be in the vertical-upright position. However, many occupational radiation exposures occur with the worker in other positions. In the first phase of this work, updates onmore » the computational phantom models were reviewed and a revised phantom model, which includes the updates for the relevant organs and compositions, was identified. This revised model was adopted as the starting point for this development work, and hence a series of radiation transport computations, using the Monte Carlo code MCNP5, was performed. The computational results were compared against values reported by the International Commission on Radiation Protection (ICRP) in Publication 74. For some of the organs (e.g., thyroid), there were discrepancies between the computed values and the results reported in ICRP-74. The reasons behind these discrepancies have been investigated and are discussed in this report.Additionally, sensitivity computations were performed to determine the sensitivity of the organ doses for certain parameters, including composition and cross sections used in the simulations. To assess the dose for more realistic exposure configurations, the phantom model was revised to enable flexible positioning of the arms and legs. Furthermore, to reduce the user time for analyses, a graphical user interface (GUI) was developed. The GUI can be used to visualize the positioning of the arms and legs as desired posture is achieved to generate the input file, invoke the computations, and extract the organ dose values from the MCNP5 output file. In this report, the main features of the phantom model with moving arms and legs and user interface are described.« less

Statistical Tools for Determining Fitness to Fly

DTIC Science & Technology

1981-09-01

program. (a) Number of Cards in file: 13 (b) Layout of Card 1: iIi Field Length a•e. Variable 1 8 Real EFAIL : Average # of failures for size of control...Method Compute Survival Probability and Frequency Tables 4-4 END 25P FLUW CHARTS 27 QSTART Input EFAIL ,CYEAR,NVAR,NAV,XINC BB~i i=1,3 name (1) i=1,4 Call

Maintaining a Critical Spectra within Monteburns for a Gas-Cooled Reactor Array by Way of Control Rod Manipulation

DOE PAGES

Adigun, Babatunde John; Fensin, Michael Lorne; Galloway, Jack D.; ...

2016-10-01

Our burnup study examined the effect of a predicted critical control rod position on the nuclide predictability of several axial and radial locations within a 4×4 graphite moderated gas cooled reactor fuel cluster geometry. To achieve this, a control rod position estimator (CRPE) tool was developed within the framework of the linkage code Monteburns between the transport code MCNP and depletion code CINDER90, and four methodologies were proposed within the tool for maintaining criticality. Two of the proposed methods used an inverse multiplication approach - where the amount of fissile material in a set configuration is slowly altered until criticalitymore » is attained - in estimating the critical control rod position. Another method carried out several MCNP criticality calculations at different control rod positions, then used a linear fit to estimate the critical rod position. The final method used a second-order polynomial fit of several MCNP criticality calculations at different control rod positions to guess the critical rod position. The results showed that consistency in prediction of power densities as well as uranium and plutonium isotopics was mutual among methods within the CRPE tool that predicted critical position consistently well. Finall, while the CRPE tool is currently limited to manipulating a single control rod, future work could be geared toward implementing additional criticality search methodologies along with additional features.« less

Verification and Validation of Monte Carlo n-Particle Code 6 (MCNP6) with Neutron Protection Factor Measurements of an Iron Box

DTIC Science & Technology

2014-03-27

VERIFICATION AND VALIDATION OF MONTE CARLO N- PARTICLE CODE 6 (MCNP6) WITH NEUTRON PROTECTION FACTOR... PARTICLE CODE 6 (MCNP6) WITH NEUTRON PROTECTION FACTOR MEASUREMENTS OF AN IRON BOX THESIS Presented to the Faculty Department of Engineering...STATEMENT A. APPROVED FOR PUBLIC RELEASE; DISTRIBUTION UNLIMITED iv AFIT-ENP-14-M-05 VERIFICATION AND VALIDATION OF MONTE CARLO N- PARTICLE CODE 6

Characterization and Performance Evaluation of an HPXe Detector for Nuclear Explosion Monitoring Applications

DTIC Science & Technology

2007-09-01

performance of the detector, and to compare the performance with sodium iodide and germanium detectors. Monte Carlo ( MCNP ) simulation was used to...aluminum ~50% more efficient), and to estimate optimum shield dimensions for an HPXe based nuclear explosion monitor. MCNP modeling was also used to...detector were calculated with MCNP by using input activity levels as measured in routine NEM runs at Pacific Northwest National Laboratory (PNNL

Modeling of Radioxenon Production and Release Pathways

DTIC Science & Technology

2010-09-01

MCNP is utilized to model the neutron transport, while ORIGEN 2.2 is utilized to calculate the production and decay of fission products, activation...products, and transuranics resulting from the calculated neutron flux profile. MONTEBURNS is a pearl script that couples the MCNP and ORIGEN 2.2...core enriched to 93.80% 239Pu was used. MCNP was used to determine the thickness of soil or rock necessary to accurately model the attenuation and

Detection And Mapping (DAM) package. Volume 4B: Software System Manual, part 2

NASA Technical Reports Server (NTRS)

Schlosser, E. H.

1980-01-01

Computer programs, graphic devices, and an integrated set of manual procedures designed for efficient production of precisely registered and formatted maps from digital data are presented. The software can be used on any Univac 1100 series computer. The software includes pre-defined spectral limits for use in classifying and mapping surface water for LANDSAT-1, LANDSAT-2, and LANDSAT-3.

Development of monitoring method of spatial neutron distribution in neutrons-gamma rays mixed field using imaging plate for NCT--depression of the field.

PubMed

Tanaka, Kenichi; Endo, Satoru; Hoshi, Masaharu; Takada, Jun

2011-12-01

The degree of depression in the neutron field caused by neutron absorption in the materials of an imaging plate (IP) was investigated using MCNP-4C. Consequently, the IP doped with Gd, which reproduced the distribution of (157)Gd(n,γ)(158)Gd reaction rate in the previous study, depresses the relative distribution by about 50%. The depression for the IP in which Gd is replaced with similar amount of B atoms was estimated to be about 10%. The signal intensity for this IP is estimated to be at a similar level with that for Gd-doped IP. Copyright © 2011 Elsevier Ltd. All rights reserved.

Impact of thorium based molten salt reactor on the closure of the nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Jaradat, Safwan Qasim Mohammad

Molten salt reactor (MSR) is one of six reactors selected by the Generation IV International Forum (GIF). The liquid fluoride thorium reactor (LFTR) is a MSR concept based on thorium fuel cycle. LFTR uses liquid fluoride salts as a nuclear fuel. It uses 232Th and 233U as the fertile and fissile materials, respectively. Fluoride salt of these nuclides is dissolved in a mixed carrier salt of lithium and beryllium (FLiBe). The objective of this research was to complete feasibility studies of a small commercial thermal LFTR. The focus was on neutronic calculations in order to prescribe core design parameter such as core size, fuel block pitch (p), fuel channel radius, fuel path, reflector thickness, fuel salt composition, and power. In order to achieve this objective, the applicability of Monte Carlo N-Particle Transport Code (MCNP) to MSR modeling was verified. Then, a prescription for conceptual small thermal reactor LFTR and relevant calculations were performed using MCNP to determine the main neutronic parameters of the core reactor. The MCNP code was used to study the reactor physics characteristics for the FUJI-U3 reactor. The results were then compared with the results obtained from the original FUJI-U3 using the reactor physics code SRAC95 and the burnup analysis code ORIPHY2. The results were comparable with each other. Based on the results, MCNP was found to be a reliable code to model a small thermal LFTR and study all the related reactor physics characteristics. The results of this study were promising and successful in demonstrating a prefatory small commercial LFTR design. The outcome of using a small core reactor with a diameter/height of 280/260 cm that would operate for more than five years at a power level of 150 MWth was studied. The fuel system 7LiF - BeF2 - ThF4 - UF4 with a (233U/ 232Th) = 2.01 % was the candidate fuel for this reactor core.

Monte Carlo simulation for Neptun 10 PC medical linear accelerator and calculations of output factor for electron beam

PubMed Central

Bahreyni Toossi, Mohammad Taghi; Momennezhad, Mehdi; Hashemi, Seyed Mohammad

2012-01-01

Aim Exact knowledge of dosimetric parameters is an essential pre-requisite of an effective treatment in radiotherapy. In order to fulfill this consideration, different techniques have been used, one of which is Monte Carlo simulation. Materials and methods This study used the MCNP-4Cb to simulate electron beams from Neptun 10 PC medical linear accelerator. Output factors for 6, 8 and 10 MeV electrons applied to eleven different conventional fields were both measured and calculated. Results The measurements were carried out by a Wellhofler-Scanditronix dose scanning system. Our findings revealed that output factors acquired by MCNP-4C simulation and the corresponding values obtained by direct measurements are in a very good agreement. Conclusion In general, very good consistency of simulated and measured results is a good proof that the goal of this work has been accomplished. PMID:24377010

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hiller, Mauritius M.; Veinot, Kenneth G.; Easterly, Clay E.

In this study, methods are addressed to reduce the computational time to compute organ-dose rate coefficients using Monte Carlo techniques. Several variance reduction techniques are compared including the reciprocity method, importance sampling, weight windows and the use of the ADVANTG software package. For low-energy photons, the runtime was reduced by a factor of 10 5 when using the reciprocity method for kerma computation for immersion of a phantom in contaminated water. This is particularly significant since impractically long simulation times are required to achieve reasonable statistical uncertainties in organ dose for low-energy photons in this source medium and geometry. Althoughmore » the MCNP Monte Carlo code is used in this paper, the reciprocity technique can be used equally well with other Monte Carlo codes.« less

Evaluation of the new electron-transport algorithm in MCNP6.1 for the simulation of dose point kernel in water

NASA Astrophysics Data System (ADS)

Antoni, Rodolphe; Bourgois, Laurent

2017-12-01

In this work, the calculation of specific dose distribution in water is evaluated in MCNP6.1 with the regular condensed history algorithm the "detailed electron energy-loss straggling logic" and the new electrons transport algorithm proposed the "single event algorithm". Dose Point Kernel (DPK) is calculated with monoenergetic electrons of 50, 100, 500, 1000 and 3000 keV for different scoring cells dimensions. A comparison between MCNP6 results and well-validated codes for electron-dosimetry, i.e., EGSnrc or Penelope, is performed. When the detailed electron energy-loss straggling logic is used with default setting (down to the cut-off energy 1 keV), we infer that the depth of the dose peak increases with decreasing thickness of the scoring cell, largely due to combined step-size and boundary crossing artifacts. This finding is less prominent for 500 keV, 1 MeV and 3 MeV dose profile. With an appropriate number of sub-steps (ESTEP value in MCNP6), the dose-peak shift is almost complete absent to 50 keV and 100 keV electrons. However, the dose-peak is more prominent compared to EGSnrc and the absorbed dose tends to be underestimated at greater depths, meaning that boundaries crossing artifact are still occurring while step-size artifacts are greatly reduced. When the single-event mode is used for the whole transport, we observe the good agreement of reference and calculated profile for 50 and 100 keV electrons. Remaining artifacts are fully vanished, showing a possible transport treatment for energies less than a hundred of keV and accordance with reference for whatever scoring cell dimension, even if the single event method initially intended to support electron transport at energies below 1 keV. Conversely, results for 500 keV, 1 MeV and 3 MeV undergo a dramatic discrepancy with reference curves. These poor results and so the current unreliability of the method is for a part due to inappropriate elastic cross section treatment from the ENDF/B-VI.8 library in those energy ranges. Accordingly, special care has to be taken in setting choice for calculating electron dose distribution with MCNP6, in particular with regards to dosimetry or nuclear medicine applications.

Computer Program Development Specification for Ada Integrated Environment: KAPSE (Kernel Ada Programming Support Environment)/Database, Type B5, B5-AIE(1).KAPSE(1).

DTIC Science & Technology

1982-11-12

File 1/0 Prgram Invocation Other Access M and Control Services KAPSE/Host Interface most Operating System Peripherals/ 01 su ?eetworks 6282318-2 Figure 3...3.2.4.3.8.5 Transitory Windows The TRANSITORY flag is used to prevent permanent dependence on temporary windows created simply for focusing on a part of the...KAPSE/Tool interfaces in terms of these low-level host-independent interfaces. In addition, the KAPSE/Host interface packages prevent the application

Distributed multitasking ITS with PVM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fan, W.C.; Halbleib, J.A. Sr.

1995-12-31

Advances in computer hardware and communication software have made it possible to perform parallel-processing computing on a collection of desktop workstations. For many applications, multitasking on a cluster of high-performance workstations has achieved performance comparable to or better than that on a traditional supercomputer. From the point of view of cost-effectiveness, it also allows users to exploit available but unused computational resources and thus achieve a higher performance-to-cost ratio. Monte Carlo calculations are inherently parallelizable because the individual particle trajectories can be generated independently with minimum need for interprocessor communication. Furthermore, the number of particle histories that can be generatedmore » in a given amount of wall-clock time is nearly proportional to the number of processors in the cluster. This is an important fact because the inherent statistical uncertainty in any Monte Carlo result decreases as the number of histories increases. For these reasons, researchers have expended considerable effort to take advantage of different parallel architectures for a variety of Monte Carlo radiation transport codes, often with excellent results. The initial interest in this work was sparked by the multitasking capability of the MCNP code on a cluster of workstations using the Parallel Virtual Machine (PVM) software. On a 16-machine IBM RS/6000 cluster, it has been demonstrated that MCNP runs ten times as fast as on a single-processor CRAY YMP. In this paper, we summarize the implementation of a similar multitasking capability for the coupled electronphoton transport code system, the Integrated TIGER Series (ITS), and the evaluation of two load-balancing schemes for homogeneous and heterogeneous networks.« less

Distributed multitasking ITS with PVM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fan, W.C.; Halbleib, J.A. Sr.

1995-02-01

Advances of computer hardware and communication software have made it possible to perform parallel-processing computing on a collection of desktop workstations. For many applications, multitasking on a cluster of high-performance workstations has achieved performance comparable or better than that on a traditional supercomputer. From the point of view of cost-effectiveness, it also allows users to exploit available but unused computational resources, and thus achieve a higher performance-to-cost ratio. Monte Carlo calculations are inherently parallelizable because the individual particle trajectories can be generated independently with minimum need for interprocessor communication. Furthermore, the number of particle histories that can be generated inmore » a given amount of wall-clock time is nearly proportional to the number of processors in the cluster. This is an important fact because the inherent statistical uncertainty in any Monte Carlo result decreases as the number of histories increases. For these reasons, researchers have expended considerable effort to take advantage of different parallel architectures for a variety of Monte Carlo radiation transport codes, often with excellent results. The initial interest in this work was sparked by the multitasking capability of MCNP on a cluster of workstations using the Parallel Virtual Machine (PVM) software. On a 16-machine IBM RS/6000 cluster, it has been demonstrated that MCNP runs ten times as fast as on a single-processor CRAY YMP. In this paper, we summarize the implementation of a similar multitasking capability for the coupled electron/photon transport code system, the Integrated TIGER Series (ITS), and the evaluation of two load balancing schemes for homogeneous and heterogeneous networks.« less

Computational Assessment of Naturally Occurring Neutron and Photon Background Radiation Produced by Extraterrestrial Sources

DOE PAGES

Miller, Thomas Martin; de Wet, Wouter C.; Patton, Bruce W.

2015-10-28

In this study, a computational assessment of the variation in terrestrial neutron and photon background from extraterrestrial sources is presented. The motivation of this assessment is to evaluate the practicality of developing a tool or database to estimate background in real time (or near–real time) during an experimental measurement or to even predict the background for future measurements. The extraterrestrial source focused on during this assessment is naturally occurring galactic cosmic rays (GCRs). The MCNP6 transport code was used to perform the computational assessment. However, the GCR source available in MCNP6 was not used. Rather, models developed and maintained bymore » NASA were used to generate the GCR sources. The largest variation in both neutron and photon background spectra was found to be caused by changes in elevation on Earth's surface, which can be as large as an order of magnitude. All other perturbations produced background variations on the order of a factor of 3 or less. The most interesting finding was that ~80% and 50% of terrestrial background neutrons and photons, respectively, are generated by interactions in Earth's surface and other naturally occurring and man-made objects near a detector of particles from extraterrestrial sources and their progeny created in Earth's atmosphere. In conclusion, this assessment shows that it will be difficult to estimate the terrestrial background from extraterrestrial sources without a good understanding of a detector's surroundings. Therefore, estimating or predicting background during a measurement environment like a mobile random search will be difficult.« less

NOTE: Development of modified voxel phantoms for the numerical dosimetric reconstruction of radiological accidents involving external sources: implementation in SESAME tool

NASA Astrophysics Data System (ADS)

Courageot, Estelle; Sayah, Rima; Huet, Christelle

2010-05-01

Estimating the dose distribution in a victim's body is a relevant indicator in assessing biological damage from exposure in the event of a radiological accident caused by an external source. When the dose distribution is evaluated with a numerical anthropomorphic model, the posture and morphology of the victim have to be reproduced as realistically as possible. Several years ago, IRSN developed a specific software application, called the simulation of external source accident with medical images (SESAME), for the dosimetric reconstruction of radiological accidents by numerical simulation. This tool combines voxel geometry and the MCNP(X) Monte Carlo computer code for radiation-material interaction. This note presents a new functionality in this software that enables the modelling of a victim's posture and morphology based on non-uniform rational B-spline (NURBS) surfaces. The procedure for constructing the modified voxel phantoms is described, along with a numerical validation of this new functionality using a voxel phantom of the RANDO tissue-equivalent physical model.

Development of modified voxel phantoms for the numerical dosimetric reconstruction of radiological accidents involving external sources: implementation in SESAME tool.

PubMed

Courageot, Estelle; Sayah, Rima; Huet, Christelle

2010-05-07

Estimating the dose distribution in a victim's body is a relevant indicator in assessing biological damage from exposure in the event of a radiological accident caused by an external source. When the dose distribution is evaluated with a numerical anthropomorphic model, the posture and morphology of the victim have to be reproduced as realistically as possible. Several years ago, IRSN developed a specific software application, called the simulation of external source accident with medical images (SESAME), for the dosimetric reconstruction of radiological accidents by numerical simulation. This tool combines voxel geometry and the MCNP(X) Monte Carlo computer code for radiation-material interaction. This note presents a new functionality in this software that enables the modelling of a victim's posture and morphology based on non-uniform rational B-spline (NURBS) surfaces. The procedure for constructing the modified voxel phantoms is described, along with a numerical validation of this new functionality using a voxel phantom of the RANDO tissue-equivalent physical model.

Computer-aided roll pass design in rolling of airfoil shapes

NASA Technical Reports Server (NTRS)

Akgerman, N.; Lahoti, G. D.; Altan, T.

1980-01-01

This paper describes two computer-aided design (CAD) programs developed for modeling the shape rolling process for airfoil sections. The first program, SHPROL, uses a modular upper-bound method of analysis and predicts the lateral spread, elongation, and roll torque. The second program, ROLPAS, predicts the stresses, roll separating force, the roll torque and the details of metal flow by simulating the rolling process, using the slab method of analysis. ROLPAS is an interactive program; it offers graphic display capabilities and allows the user to interact with the computer via a keyboard, CRT, and a light pen. The accuracy of the computerized models was evaluated by (a) rolling a selected airfoil shape at room temperature from 1018 steel and isothermally at high temperature from Ti-6Al-4V, and (b) comparing the experimental results with computer predictions. The comparisons indicated that the CAD systems, described here, are useful for practical engineering purposes and can be utilized in roll pass design and analysis for airfoil and similar shapes.

Summer 2016

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mendoza, Paul Michael

2016-08-31

The project goals seek to develop applications in order to automate MCNP criticality benchmark execution; create a dataset containing static benchmark information; combine MCNP output with benchmark information; and fit and visually represent data.

CREPT-MCNP code for efficiency calibration of HPGe detectors with the representative point method.

PubMed

Saegusa, Jun

2008-01-01

The representative point method for the efficiency calibration of volume samples has been previously proposed. For smoothly implementing the method, a calculation code named CREPT-MCNP has been developed. The code estimates the position of a representative point which is intrinsic to each shape of volume sample. The self-absorption correction factors are also given to make correction on the efficiencies measured at the representative point with a standard point source. Features of the CREPT-MCNP code are presented.

Development and Application of NUMIT for Realistic Modeling of Deep-Dielectric Spacecraft Charging in the Space Environment

DTIC Science & Technology

2014-04-21

Dixon, a graduate student at the University of New Mexico who introduced us to MCNP . Using what we learned from Dixon, we were able to produce a...curves were produced with MCNP for incident electron energies from 10 to 100 keV in increments of 10 keV, see Figure 9. In this case, the same...the algorithm. Since MCNP does take backscatter into consideration, the comparisons on the vertical scales (energy or number of electrons deposited

Simplification of an MCNP model designed for dose rate estimation

NASA Astrophysics Data System (ADS)

Laptev, Alexander; Perry, Robert

2017-09-01

A study was made to investigate the methods of building a simplified MCNP model for radiological dose estimation. The research was done using an example of a complicated glovebox with extra shielding. The paper presents several different calculations for neutron and photon dose evaluations where glovebox elements were consecutively excluded from the MCNP model. The analysis indicated that to obtain a fast and reasonable estimation of dose, the model should be realistic in details that are close to the tally. Other details may be omitted.

National Dam Inspection Program. Jennings Pond Dam (NDI I.D. PA-0891 DER I.D. 066-012) Susquehanna River Basin, Little Mehoopany Creek, Wyoming County, Pennsylvania. Phase I Inspection Report,

DTIC Science & Technology

1981-03-19

Area 7.9 square miles(1) b. Discharge at Dam Site ( cfs ) Maximum known flood at dam site Unknown Outlet conduit at maximum pool Unknown Gated spillway...700 cfs , based on the available 2.4-foot freeboard relative to the low spot on the left abutment. b. Experience Data. As previously stated, Jennings...in Appendix D. The inflow hydrograph for one-half PMF was found to have a peak flow of 6835 cfs . Computer input and summary of computer output are

High Velocity Jet Noise Source Location and Reduction. Task 6. Supplement. Computer Programs: Engineering Correlation (M*S) Jet Noise Prediction Method and Unified Aeroacoustic Prediction Model (M*G*B) for Nozzles of Arbitary Shape.

DTIC Science & Technology

1979-03-01

LSPFIT 112 4.3.5 SLICE 112 4.3.6 CRD 113 4.3.7 OUTPUT 113 4.3.8 SHOCK 115 4.3.9 ATMOS 115 4.3.10 PNLC 115 4.4 Program Usage and Logic 116 4.5 Description...number MAIN, SLICE, OUTPUT F Intermediate variable LSPFIT FAC Intermediate variable PNLC FC Center frequency SLICE FIRSTU Flight velocity Ua MAIN, SLICE...Index CRD J211 Index CRD K Index, also wave number MAIN, SLICE, PNLC KN Surrounding boundary index MAIN KNCAS Case counter MAIN KNK Surrounding

Optimization of radiation shielding material aiming at compactness, lightweight, and low activation for a vehicle-mounted accelerator-driven D-T neutron source.

PubMed

Cai, Yao; Hu, Huasi; Lu, Shuangying; Jia, Qinggang

2018-05-01

To minimize the size and weight of a vehicle-mounted accelerator-driven D-T neutron source and protect workers from unnecessary irradiation after the equipment shutdown, a method to optimize radiation shielding material aiming at compactness, lightweight, and low activation for the fast neutrons was developed. The method employed genetic algorithm, combining MCNP and ORIGEN codes. A series of composite shielding material samples were obtained by the method step by step. The volume and weight needed to build a shield (assumed as a coaxial tapered cylinder) were adopted to compare the performance of the materials visually and conveniently. The results showed that the optimized materials have excellent performance in comparison with the conventional materials. The "MCNP6-ACT" method and the "rigorous two steps" (R2S) method were used to verify the activation grade of the shield irradiated by D-T neutrons. The types of radionuclide, the energy spectrum of corresponding decay gamma source, and the variation in decay gamma dose rate were also computed. Copyright © 2018 Elsevier Ltd. All rights reserved.

Numerical Tests for the Problem of U-Pu Fuel Burnup in Fuel Rod and Polycell Models Using the MCNP Code

NASA Astrophysics Data System (ADS)

Muratov, V. G.; Lopatkin, A. V.

An important aspect in the verification of the engineering techniques used in the safety analysis of MOX-fuelled reactors, is the preparation of test calculations to determine nuclide composition variations under irradiation and analysis of burnup problem errors resulting from various factors, such as, for instance, the effect of nuclear data uncertainties on nuclide concentration calculations. So far, no universally recognized tests have been devised. A calculation technique has been developed for solving the problem using the up-to-date calculation tools and the latest versions of nuclear libraries. Initially, in 1997, a code was drawn up in an effort under ISTC Project No. 116 to calculate the burnup in one VVER-1000 fuel rod, using the MCNP Code. Later on, the authors developed a computation technique which allows calculating fuel burnup in models of a fuel rod, or a fuel assembly, or the whole reactor. It became possible to apply it to fuel burnup in all types of nuclear reactors and subcritical blankets.

Middle School Teachers' Perceptions of Computer-Assisted Reading Intervention Programs

ERIC Educational Resources Information Center

Bippert, Kelli; Harmon, Janis

2017-01-01

Middle schools often turn to computer-assisted reading intervention programs to improve student reading. The questions guiding this study are (a) in what ways are computer-assisted reading intervention programs utilized, and (b) what are teachers' perceptions about these intervention programs? Nineteen secondary reading teachers were interviewed…

Characterization of the neutron irradiation system for use in the Low-Dose-Rate Irradiation Facility at Sandia National Laboratories.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Franco, Manuel

The objective of this work was to characterize the neutron irradiation system consisting of americium-241 beryllium (241AmBe) neutron sources placed in a polyethylene shielding for use at Sandia National Laboratories (SNL) Low Dose Rate Irradiation Facility (LDRIF). With a total activity of 0.3 TBq (9 Ci), the source consisted of three recycled 241AmBe sources of different activities that had been combined into a single source. The source in its polyethylene shielding will be used in neutron irradiation testing of components. The characterization of the source-shielding system was necessary to evaluate the radiation environment for future experiments. Characterization of the sourcemore » was also necessary because the documentation for the three component sources and their relative alignment within the Special Form Capsule (SFC) was inadequate. The system consisting of the source and shielding was modeled using Monte Carlo N-Particle transport code (MCNP). The model was validated by benchmarking it against measurements using multiple techniques. To characterize the radiation fields over the full spatial geometry of the irradiation system, it was necessary to use a number of instruments of varying sensitivities. First, the computed photon radiography assisted in determining orientation of the component sources. With the capsule properly oriented inside the shielding, the neutron spectra were measured using a variety of techniques. A N-probe Microspec and a neutron Bubble Dosimeter Spectrometer (BDS) set were used to characterize the neutron spectra/field in several locations. In the third technique, neutron foil activation was used to ascertain the neutron spectra. A high purity germanium (HPGe) detector was used to characterize the photon spectrum. The experimentally measured spectra and the MCNP results compared well. Once the MCNP model was validated to an adequate level of confidence, parametric analyses was performed on the model to optimize for potential experimental configurations and neutron spectra for component irradiation. The final product of this work is a MCNP model validated by measurements, an overall understanding of neutron irradiation system including photon/neutron transport and effective dose rates throughout the system, and possible experimental configurations for future irradiation of components.« less

Features of MCNP6 Relevant to Medical Radiation Physics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hughes, H. Grady III; Goorley, John T.

2012-08-29

MCNP (Monte Carlo N-Particle) is a general-purpose Monte Carlo code for simulating the transport of neutrons, photons, electrons, positrons, and more recently other fundamental particles and heavy ions. Over many years MCNP has found a wide range of applications in many different fields, including medical radiation physics. In this presentation we will describe and illustrate a number of significant recently-developed features in the current version of the code, MCNP6, having particular utility for medical physics. Among these are major extensions of the ability to simulate large, complex geometries, improvement in memory requirements and speed for large lattices, introduction of mesh-basedmore » isotopic reaction tallies, advances in radiography simulation, expanded variance-reduction capabilities, especially for pulse-height tallies, and a large number of enhancements in photon/electron transport.« less

Simple, rapid, and cost-effective modified Carba NP test for carbapenemase detection among Gram-negative bacteria

PubMed Central

Rudresh, Shoorashetty Manohara; Ravi, Giriyapur Siddappa; Sunitha, Lakshminarayanappa; Hajira, Sadiya Noor; Kalaiarasan, Ellappan; Harish, Belgode Narasimha

2017-01-01

PURPOSE: Detection of carbapenemases among Gram-negative bacteria (GNB) is important for both clinicians and infection control practitioners. The Clinical and Laboratory Standards Institute recommends Carba NP (CNP) as confirmatory test for carbapenemase production. The reagents required for CNP test are costly and hence the test cannot be performed on a routine basis. The present study evaluates modifications of CNP test for rapid detection of carbapenemases among GNB. MATERIALS AND METHODS: The GNB were screened for carbapenemase production using CNP, CarbAcineto NP (CANP), and modified CNP (mCNP) test. A multiplex polymerase chain reaction (PCR) was performed on all the carbapenem-resistant bacteria for carbapenemase genes. The results of three phenotypic tests were compared with PCR. RESULTS: A total of 765 gram negative bacteria were screened for carbapenem resistance. Carbapenem resistance was found in 144 GNB. The metallo-β-lactamases were most common carbapenemases followed by OXA-48-like enzymes. The CANP test was most sensitive (80.6%) for carbapenemases detection. The mCNP test was 62.1% sensitive for detection of carbapenemases. The mCNP, CNP, and CANP tests were equally sensitive (95%) for detection of NDM enzymes among Enterobacteriaceae. The mCNP test had poor sensitivity for detection of OXA-48-like enzymes. CONCLUSION: The mCNP test was rapid, cost-effective, and easily adoptable on routine basis. The early detection of carbapenemases using mCNP test will help in preventing the spread of multidrug-resistant organisms in the hospital settings. PMID:28966495

Simple, rapid, and cost-effective modified Carba NP test for carbapenemase detection among Gram-negative bacteria.

PubMed

Rudresh, Shoorashetty Manohara; Ravi, Giriyapur Siddappa; Sunitha, Lakshminarayanappa; Hajira, Sadiya Noor; Kalaiarasan, Ellappan; Harish, Belgode Narasimha

2017-01-01

Detection of carbapenemases among Gram-negative bacteria (GNB) is important for both clinicians and infection control practitioners. The Clinical and Laboratory Standards Institute recommends Carba NP (CNP) as confirmatory test for carbapenemase production. The reagents required for CNP test are costly and hence the test cannot be performed on a routine basis. The present study evaluates modifications of CNP test for rapid detection of carbapenemases among GNB. The GNB were screened for carbapenemase production using CNP, CarbAcineto NP (CANP), and modified CNP (mCNP) test. A multiplex polymerase chain reaction (PCR) was performed on all the carbapenem-resistant bacteria for carbapenemase genes. The results of three phenotypic tests were compared with PCR. A total of 765 gram negative bacteria were screened for carbapenem resistance. Carbapenem resistance was found in 144 GNB. The metallo-β-lactamases were most common carbapenemases followed by OXA-48-like enzymes. The CANP test was most sensitive (80.6%) for carbapenemases detection. The mCNP test was 62.1% sensitive for detection of carbapenemases. The mCNP, CNP, and CANP tests were equally sensitive (95%) for detection of NDM enzymes among Enterobacteriaceae. The mCNP test had poor sensitivity for detection of OXA-48-like enzymes. The mCNP test was rapid, cost-effective, and easily adoptable on routine basis. The early detection of carbapenemases using mCNP test will help in preventing the spread of multidrug-resistant organisms in the hospital settings.

Whole body counter calibration using Monte Carlo modeling with an array of phantom sizes based on national anthropometric reference data

NASA Astrophysics Data System (ADS)

Shypailo, R. J.; Ellis, K. J.

2011-05-01

During construction of the whole body counter (WBC) at the Children's Nutrition Research Center (CNRC), efficiency calibration was needed to translate acquired counts of 40K to actual grams of potassium for measurement of total body potassium (TBK) in a diverse subject population. The MCNP Monte Carlo n-particle simulation program was used to describe the WBC (54 detectors plus shielding), test individual detector counting response, and create a series of virtual anthropomorphic phantoms based on national reference anthropometric data. Each phantom included an outer layer of adipose tissue and an inner core of lean tissue. Phantoms were designed for both genders representing ages 3.5 to 18.5 years with body sizes from the 5th to the 95th percentile based on body weight. In addition, a spherical surface source surrounding the WBC was modeled in order to measure the effects of subject mass on room background interference. Individual detector measurements showed good agreement with the MCNP model. The background source model came close to agreement with empirical measurements, but showed a trend deviating from unity with increasing subject size. Results from the MCNP simulation of the CNRC WBC agreed well with empirical measurements using BOMAB phantoms. Individual detector efficiency corrections were used to improve the accuracy of the model. Nonlinear multiple regression efficiency calibration equations were derived for each gender. Room background correction is critical in improving the accuracy of the WBC calibration.

Development of the 3DHZETRN code for space radiation protection

NASA Astrophysics Data System (ADS)

Wilson, John; Badavi, Francis; Slaba, Tony; Reddell, Brandon; Bahadori, Amir; Singleterry, Robert

Space radiation protection requires computationally efficient shield assessment methods that have been verified and validated. The HZETRN code is the engineering design code used for low Earth orbit dosimetric analysis and astronaut record keeping with end-to-end validation to twenty percent in Space Shuttle and International Space Station operations. HZETRN treated diffusive leakage only at the distal surface limiting its application to systems with a large radius of curvature. A revision of HZETRN that included forward and backward diffusion allowed neutron leakage to be evaluated at both the near and distal surfaces. That revision provided a deterministic code of high computational efficiency that was in substantial agreement with Monte Carlo (MC) codes in flat plates (at least to the degree that MC codes agree among themselves). In the present paper, the 3DHZETRN formalism capable of evaluation in general geometry is described. Benchmarking will help quantify uncertainty with MC codes (Geant4, FLUKA, MCNP6, and PHITS) in simple shapes such as spheres within spherical shells and boxes. Connection of the 3DHZETRN to general geometry will be discussed.

Automated Instrumentation, Monitoring and Visualization of PVM Programs Using AIMS

NASA Technical Reports Server (NTRS)

Mehra, Pankaj; VanVoorst, Brian; Yan, Jerry; Tucker, Deanne (Technical Monitor)

1994-01-01

We present views and analysis of the execution of several PVM codes for Computational Fluid Dynamics on a network of Sparcstations, including (a) NAS Parallel benchmarks CG and MG (White, Alund and Sunderam 1993); (b) a multi-partitioning algorithm for NAS Parallel Benchmark SP (Wijngaart 1993); and (c) an overset grid flowsolver (Smith 1993). These views and analysis were obtained using our Automated Instrumentation and Monitoring System (AIMS) version 3.0, a toolkit for debugging the performance of PVM programs. We will describe the architecture, operation and application of AIMS. The AIMS toolkit contains (a) Xinstrument, which can automatically instrument various computational and communication constructs in message-passing parallel programs; (b) Monitor, a library of run-time trace-collection routines; (c) VK (Visual Kernel), an execution-animation tool with source-code clickback; and (d) Tally, a tool for statistical analysis of execution profiles. Currently, Xinstrument can handle C and Fortran77 programs using PVM 3.2.x; Monitor has been implemented and tested on Sun 4 systems running SunOS 4.1.2; and VK uses X11R5 and Motif 1.2. Data and views obtained using AIMS clearly illustrate several characteristic features of executing parallel programs on networked workstations: (a) the impact of long message latencies; (b) the impact of multiprogramming overheads and associated load imbalance; (c) cache and virtual-memory effects; and (4significant skews between workstation clocks. Interestingly, AIMS can compensate for constant skew (zero drift) by calibrating the skew between a parent and its spawned children. In addition, AIMS' skew-compensation algorithm can adjust timestamps in a way that eliminates physically impossible communications (e.g., messages going backwards in time). Our current efforts are directed toward creating new views to explain the observed performance of PVM programs. Some of the features planned for the near future include: (a) ConfigView, showing the physical topology of the virtual machine, inferred using specially formatted IP (Internet Protocol) packets; and (b) LoadView, synchronous animation of PVM-program execution and resource-utilization patterns.

SU-E-T-107: An Investigation Into Attenuation Effects On the Energy Spectrum for {sup 137}Cs Using Monte Carlo Techniques

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taneja, S; Bartol, L; Culberson, W

2015-06-15

Purpose: The calibration of radiation protection instrumentation including ionization chambers, scintillators, and Geiger Mueller (GM) counters used as survey meters are often done using {sup 137}Cs irradiators. During calibration, irradiators use a combination of attenuators with various thicknesses to modulate the beam to a known air-kerma rate. The variations in energy spectra as a result of these attenuators are not accounted for and may play a role in the energy-dependent response of survey meters. This study uses an experimentally validated irradiator geometry modeled in the MCNP5 transport code to characterize the effects of attenuation on the energy spectrum. Methods: Amore » Hopewell Designs G-10 {sup 137}Cs irradiator with lead attenuators of thicknesses of 0.635, 1.22, 2.22, and 4.32 cm, was used in this study. The irradiator geometry was modeled in MCNP5 and validated by comparing measured and simulated percent depth dose (PDD) and cross-field profiles. Variations in MCNP5 simulated spectra with increasing amounts of attenuation were characterized using the relative intensity of the 662 keV peak and the average energy. Results: Simulated and measured PDDs and profiles agreed within the associated uncertainty. The relative intensity of the 662 keV peak for simulated spectra normalized to the intensity of the unattenuated spectra ranged from 0.16% to 100% based on attenuation thickness. The average energy for simulated spectra for attenuators ranged from 582 keV with no attenuation to 653 keV with 5.54 cm of attenuation. Statistical uncertainty for MCNP5 simulations ranged from 0.11% to 3.69%. Conclusion: This study successfully used MCNP5 to validate a {sup 137}Cs irradiator geometry and characterize variations in energy spectra between different amounts of attenuation. Variations in the average energy of up to 12% were determined through simulations, and future work will aim to determine the effects of these differences on survey meter response and calibration.« less

Reducing statistical uncertainties in simulated organ doses of phantoms immersed in water

DOE PAGES

Hiller, Mauritius M.; Veinot, Kenneth G.; Easterly, Clay E.; ...

2016-08-13

In this study, methods are addressed to reduce the computational time to compute organ-dose rate coefficients using Monte Carlo techniques. Several variance reduction techniques are compared including the reciprocity method, importance sampling, weight windows and the use of the ADVANTG software package. For low-energy photons, the runtime was reduced by a factor of 10 5 when using the reciprocity method for kerma computation for immersion of a phantom in contaminated water. This is particularly significant since impractically long simulation times are required to achieve reasonable statistical uncertainties in organ dose for low-energy photons in this source medium and geometry. Althoughmore » the MCNP Monte Carlo code is used in this paper, the reciprocity technique can be used equally well with other Monte Carlo codes.« less

Background-Source Cosmic-Photon Elevation Scaling and Cosmic-Neutron/Photon Date Scaling in MCNP6

NASA Astrophysics Data System (ADS)

Tutt, J.; Anderson, C.; McKinney, G.

Cosmic neutron and photon fluxes are known to scale exponentially with elevation. Consequently, cosmic neutron elevation scaling was implemented for use with the background-source option shortly after its introduction into MCNP6, whereby the neutron flux weight factor was adjusted by the elevation scaling factor when the user-specified elevation differed from the selected background.dat grid-point elevation. At the same time, an elevation scaling factor was suggested for the cosmic photon flux, however, cosmic photon elevation scaling is complicated by the fact that the photon background consists of two components: cosmic and terrestrial. Previous versions of the background.dat file did not provide any way to separate these components. With Rel. 4 of this file in 2015, two new columns were added that provide the energy grid and differential cosmic photon flux separately from the total photon flux. Here we show that the cosmic photon flux component can now be scaled independently and combined with the terrestrial component to form the total photon flux at a user-specified elevation in MCNP6. Cosmic background fluxes also scale with the solar cycle due to solar modulation. This modulation has been shown to be nearly sinusoidal over time, with an inverse effect - increased modulation leads to a decrease in cosmic fluxes. This effect was initially included with the cosmic source option in MCNP6 and has now been extended for use with the background source option when: (1) the date is specified in the background.dat file, and (2) when the user specifies a date on the source definition card. A description of the cosmic-neutron/photon date scaling feature will be presented along with scaling results for past and future date extrapolations.

Direct Discrete Method for Neutronic Calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vosoughi, Naser; Akbar Salehi, Ali; Shahriari, Majid

The objective of this paper is to introduce a new direct method for neutronic calculations. This method which is named Direct Discrete Method, is simpler than the neutron Transport equation and also more compatible with physical meaning of problems. This method is based on physic of problem and with meshing of the desired geometry, writing the balance equation for each mesh intervals and with notice to the conjunction between these mesh intervals, produce the final discrete equations series without production of neutron transport differential equation and mandatory passing from differential equation bridge. We have produced neutron discrete equations for amore » cylindrical shape with two boundary conditions in one group energy. The correction of the results from this method are tested with MCNP-4B code execution. (authors)« less

Gamma-ray spectroscopy measurements and simulations for uranium mining

NASA Astrophysics Data System (ADS)

Marchais, T.; Pérot, B.; Carasco, C.; Allinei, P.-G.; Chaussonnet, P.; Ma, J.-L.; Toubon, H.

2018-01-01

AREVA Mines and the Nuclear Measurement Laboratory of CEA Cadarache are collaborating to improve the sensitivity and precision of uranium concentration evaluation by means of gamma measurements. This paper reports gamma-ray spectra, recorded with a high-purity coaxial germanium detector, on standard cement blocks with increasing uranium content, and the corresponding MCNP simulations. The detailed MCNP model of the detector and experimental setup has been validated by calculation vs. experiment comparisons. An optimization of the detector MCNP model is presented in this paper, as well as a comparison of different nuclear data libraries to explain missing or exceeding peaks in the simulation. Energy shifts observed between the fluorescence X-rays produced by MCNP and atomic data are also investigated. The qualified numerical model will be used in further studies to develop new gamma spectroscopy approaches aiming at reducing acquisition times, especially for ore samples with low uranium content.

Element analysis and calculation of the attenuation coefficients for gold, bronze and water matrixes using MCNP, WinXCom and experimental data

NASA Astrophysics Data System (ADS)

Esfandiari, M.; Shirmardi, S. P.; Medhat, M. E.

2014-06-01

In this study, element analysis and the mass attenuation coefficient for matrixes of gold, bronze and water with various impurities and the concentrations of heavy metals (Cu, Mn, Pb and Zn) are evaluated and calculated by the MCNP simulation code for photons emitted from Barium-133, Americium-241 and sources with energies between 1 and 100 keV. The MCNP data are compared with the experimental data and WinXCom code simulated results by Medhat. The results showed that the obtained results of bronze and gold matrix are in good agreement with the other methods for energies above 40 and 60 keV, respectively. However for water matrixes with various impurities, there is a good agreement between the three methods MCNP, WinXCom and the experimental one in low and high energies.

Multiprocessing MCNP on an IBM RS/6000 cluster

DOE Office of Scientific and Technical Information (OSTI.GOV)

McKinney, G.W.; West, J.T.

1993-01-01

The advent of high-performance computer systems has brought to maturity programming concepts like vectorization, multiprocessing, and multitasking. While there are many schools of thought as to the most significant factor in obtaining order-of-magnitude increases in performance, such speedup can only be achieved by integrating the computer system and application code. Vectorization leads to faster manipulation of arrays by overlapping instruction CPU cycles. Discrete ordinates codes, which require the solving of large matrices, have proved to be major benefactors of vectorization. Monte Carlo transport, on the other hand, typically contains numerous logic statements and requires extensive redevelopment to benefit from vectorization.more » Multiprocessing and multitasking provide additional CPU cycles via multiple processors. Such systems are generally designed with either common memory access (multitasking) or distributed memory access. In both cases, theoretical speedup, as a function of the number of processors (P) and the fraction of task time that multiprocesses (f), can be formulated using Amdahl's Law S ((f,P) = 1 f + f/P). However, for most applications this theoretical limit cannot be achieved, due to additional terms not included in Amdahl's Law. Monte Carlo transport is a natural candidate for multiprocessing, since the particle tracks are generally independent and the precision of the result increases as the square root of the number of particles tracked.« less

Positron Annihilation Gamma Ray Lineshape Studies of Defects in Solids.

DTIC Science & Technology

1980-06-24

of R. Waki in the development of anneal probably polygonized the zinc more completely the computer programs and for other experimental than did the...positron lifetime measurements. The assistance of R. Waki was greatly appreciated. References /1/ B.D. BOGGS and J.G. BYRNE, Metallurg.Trans. 4, 2153

Computer program for stress, stability, and vibration of complex branched shells of revolution: BOSOR 4

NASA Technical Reports Server (NTRS)

Bushnell, D.

1974-01-01

Code is easy to use yet is general with respect to: (a) type of analysis to be performed; (b) geometry of shell meridian; (c) type of wall construction; (d) type of boundary conditions, ring supports, and branching configuration; and (e) type of loading.

Processing TES Level-1B Data

NASA Technical Reports Server (NTRS)

DeBaca, Richard C.; Sarkissian, Edwin; Madatyan, Mariyetta; Shepard, Douglas; Gluck, Scott; Apolinski, Mark; McDuffie, James; Tremblay, Dennis

2006-01-01

TES L1B Subsystem is a computer program that performs several functions for the Tropospheric Emission Spectrometer (TES). The term "L1B" (an abbreviation of "level 1B"), refers to data, specific to the TES, on radiometric calibrated spectral radiances and their corresponding noise equivalent spectral radiances (NESRs), plus ancillary geolocation, quality, and engineering data. The functions performed by TES L1B Subsystem include shear analysis, monitoring of signal levels, detection of ice build-up, and phase correction and radiometric and spectral calibration of TES target data. Also, the program computes NESRs for target spectra, writes scientific TES level-1B data to hierarchical- data-format (HDF) files for public distribution, computes brightness temperatures, and quantifies interpixel signal variability for the purpose of first-order cloud and heterogeneous land screening by the level-2 software summarized in the immediately following article. This program uses an in-house-developed algorithm, called "NUSRT," to correct instrument line-shape factors.

Verification of Plutonium Content in PuBe Sources Using MCNP® 6.2.0 Beta with TENDL 2012 Libraries

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lockhart, Madeline Louise; McMath, Garrett Earl

Although the production of PuBe neutron sources has discontinued, hundreds of sources with unknown or inaccurately declared plutonium content are in existence around the world. Institutions have undertaken the task of assaying these sources, measuring, and calculating the isotopic composition, plutonium content, and neutron yield. The nominal plutonium content, based off the neutron yield per gram of pure 239Pu, has shown to be highly inaccurate. New methods of measuring the plutonium content allow a more accurate estimate of the true Pu content, but these measurements need verification. Using the TENDL 2012 nuclear data libraries, MCNP6 has the capability to simulatemore » the (α, n) interactions in a PuBe source. Theoretically, if the source is modeled according to the plutonium content, isotopic composition, and other source characteristics, the calculated neutron yield in MCNP can be compared to the experimental yield, offering an indication of the accuracy of the declared plutonium content. In this study, three sets of PuBe sources from various backgrounds were modeled in MCNP6 1.2 Beta, according to the source specifications dictated by the individuals who assayed the source. Verification of the source parameters with MCNP6 also serves as a means to test the alpha transport capabilities of MCNP6 1.2 Beta with TENDL 2012 alpha transport libraries. Finally, good agreement in the comparison would indicate the accuracy of the source parameters in addition to demonstrating MCNP's capabilities in simulating (α, n) interactions.« less

Verification of Plutonium Content in PuBe Sources Using MCNP® 6.2.0 Beta with TENDL 2012 Libraries

DOE PAGES

Lockhart, Madeline Louise; McMath, Garrett Earl

2017-10-26

Although the production of PuBe neutron sources has discontinued, hundreds of sources with unknown or inaccurately declared plutonium content are in existence around the world. Institutions have undertaken the task of assaying these sources, measuring, and calculating the isotopic composition, plutonium content, and neutron yield. The nominal plutonium content, based off the neutron yield per gram of pure 239Pu, has shown to be highly inaccurate. New methods of measuring the plutonium content allow a more accurate estimate of the true Pu content, but these measurements need verification. Using the TENDL 2012 nuclear data libraries, MCNP6 has the capability to simulatemore » the (α, n) interactions in a PuBe source. Theoretically, if the source is modeled according to the plutonium content, isotopic composition, and other source characteristics, the calculated neutron yield in MCNP can be compared to the experimental yield, offering an indication of the accuracy of the declared plutonium content. In this study, three sets of PuBe sources from various backgrounds were modeled in MCNP6 1.2 Beta, according to the source specifications dictated by the individuals who assayed the source. Verification of the source parameters with MCNP6 also serves as a means to test the alpha transport capabilities of MCNP6 1.2 Beta with TENDL 2012 alpha transport libraries. Finally, good agreement in the comparison would indicate the accuracy of the source parameters in addition to demonstrating MCNP's capabilities in simulating (α, n) interactions.« less

An accelerator-based Boron Neutron Capture Therapy (BNCT) facility based on the 7Li(p,n)7Be

NASA Astrophysics Data System (ADS)

Musacchio González, Elizabeth; Martín Hernández, Guido

2017-09-01

BNCT (Boron Neutron Capture Therapy) is a therapeutic modality used to irradiate tumors cells previously loaded with the stable isotope 10B, with thermal or epithermal neutrons. This technique is capable of delivering a high dose to the tumor cells while the healthy surrounding tissue receive a much lower dose depending on the 10B biodistribution. In this study, therapeutic gain and tumor dose per target power, as parameters to evaluate the treatment quality, were calculated. The common neutron-producing reaction 7Li(p,n)7Be for accelerator-based BNCT, having a reaction threshold of 1880.4 keV, was considered as the primary source of neutrons. Energies near the reaction threshold for deep-seated brain tumors were employed. These calculations were performed with the Monte Carlo N-Particle (MCNP) code. A simple but effective beam shaping assembly (BSA) was calculated producing a high therapeutic gain compared to previously proposed facilities with the same nuclear reaction.

A Patch to MCNP5 for Multiplication Inference: Description and User Guide

DOE Office of Scientific and Technical Information (OSTI.GOV)

Solomon, Jr., Clell J.

2014-05-05

A patch to MCNP5 has been written to allow generation of multiple neutrons from a spontaneous-fission event and generate list-mode output. This report documents the implementation and usage of this patch.

The design of a multisource americium-beryllium (Am-Be) neutron irradiation facility using MCNP for the neutronic performance calculation.

PubMed

Sogbadji, R B M; Abrefah, R G; Nyarko, B J B; Akaho, E H K; Odoi, H C; Attakorah-Birinkorang, S

2014-08-01

The americium-beryllium neutron irradiation facility at the National Nuclear Research Institute (NNRI), Ghana, was re-designed with four 20 Ci sources using Monte Carlo N-Particle (MCNP) code to investigate the maximum amount of flux that is produced by the combined sources. The results were compared with a single source Am-Be irradiation facility. The main objective was to enable us to harness the maximum amount of flux for the optimization of neutron activation analysis and to enable smaller sample sized samples to be irradiated. Using MCNP for the design construction and neutronic performance calculation, it was realized that the single-source Am-Be design produced a thermal neutron flux of (1.8±0.0007)×10(6) n/cm(2)s and the four-source Am-Be design produced a thermal neutron flux of (5.4±0.0007)×10(6) n/cm(2)s which is a factor of 3.5 fold increase compared to the single-source Am-Be design. The criticality effective, k(eff), of the single-source and the four-source Am-Be designs were found to be 0.00115±0.0008 and 0.00143±0.0008, respectively. Copyright © 2014 Elsevier Ltd. All rights reserved.

A Research Program in Computer Technology. 1986 Annual Technical Report

DTIC Science & Technology

1989-08-01

1986 (Annual Technical Report I July 1985 - June 1986 A Research Program in Computer Technology ISI/SR-87-178 U S C INFORMA-TION S C I EN C ES...Program in Computer Technology (Unclassified) 12. PERSONAL AUTHOR(S) 151 Research Staff 13a. TYPE OF REPORT 113b. TIME COVERED 14 DATE OF REPORT (Yeer...survivable networks 17. distributed processing, local networks, personal computers, workstation environment 18. computer acquisition, Strategic Computing 19

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pinilla, Maria Isabel

This report seeks to study and benchmark code predictions against experimental data; determine parameters to match MCNP-simulated detector response functions to experimental stilbene measurements; add stilbene processing capabilities to DRiFT; and improve NEUANCE detector array modeling and analysis using new MCNP6 and DRiFT features.

An Electron/Photon/Relaxation Data Library for MCNP6

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hughes, III, H. Grady

The capabilities of the MCNP6 Monte Carlo code in simulation of electron transport, photon transport, and atomic relaxation have recently been significantly expanded. The enhancements include not only the extension of existing data and methods to lower energies, but also the introduction of new categories of data and methods. Support of these new capabilities has required major additions to and redesign of the associated data tables. In this paper we present the first complete documentation of the contents and format of the new electron-photon-relaxation data library now available with the initial production release of MCNP6.

SABRINA - an interactive geometry modeler for MCNP

DOE Office of Scientific and Technical Information (OSTI.GOV)

West, J.T.; Murphy, J.

One of the most difficult tasks when analyzing a complex three-dimensional system with Monte Carlo is geometry model development. SABRINA attempts to make the modeling process more user-friendly and less of an obstacle. It accepts both combinatorial solid bodies and MCNP surfaces and produces MCNP cells. The model development process in SABRINA is highly interactive and gives the user immediate feedback on errors. Users can view their geometry from arbitrary perspectives while the model is under development and interactively find and correct modeling errors. An example of a SABRINA display is shown. It represents a complex three-dimensional shape.

Report on the FY17 Development of Computer Program for ASME Section III, Division 5, Subsection HB, Subpart B Rules

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swindeman, M. J.; Jetter, R. I.; Sham, T. -L.

One of the objectives of the high temperature design methodology activities is to develop and validate both improvements and the basic features of ASME Boiler and Pressure Vessel Code, Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors, Subsection HB, Subpart B (HBB). The overall scope of this task is to develop a computer program to aid assessment procedures of components under specified loading conditions in accordance with the elevated temperature design requirements for Division 5 Class A components. There are many features and alternative paths of varying complexity in HBB. The initial focus ofmore » this computer program is a basic path through the various options for a single reference material, 316H stainless steel. However, the computer program is being structured for eventual incorporation all of the features and permitted materials of HBB. This report will first provide a description of the overall computer program, particular challenges in developing numerical procedures for the assessment, and an overall approach to computer program development. This is followed by a more comprehensive appendix, which is the draft computer program manual for the program development. The strain limits rules have been implemented in the computer program. The evaluation of creep-fatigue damage will be implemented in future work scope.« less

The Effect of Birthrate Granularity on the Release- to- Birth Ratio for the AGR-1 In-core Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dawn Scates; John Walter

The AGR-1 Advanced Gas Reactor (AGR) tristructural-isotropic-particle fuel experiment underwent 13 irradiation intervals from December 2006 until November 2009 within the Idaho National Laboratory Advanced Test Reactor in support of the Next Generation Nuclear Power Plant program. During this multi-year experiment, release-to-birth rate ratios were computed at the end of each operating interval to provide information about fuel performance. Fission products released during irradiation were tracked daily by the Fission Product Monitoring System using 8-hour measurements. Birth rates calculated by MCNP with ORIGEN for as-run conditions were computed at the end of each irradiation interval. Each time step in MCNPmore » provided neutron flux, reaction rates and AGR-1 compact composition, which were used to determine birth rates using ORIGEN. The initial birth-rate data, consisting of four values for each irradiation interval at the beginning, end, and two intermediate times, were interpolated to obtain values for each 8-hour activity. The problem with this method is that any daily changes in heat rates or perturbations, such as shim control movement or core/lobe power fluctuations, would not be reflected in the interpolated data and a true picture of the system would not be presented. At the conclusion of the AGR-1 experiment, great efforts were put forth to compute daily birthrates, which were reprocessed with the 8-hour release activity. The results of this study are presented in this paper.« less

The effect of birthrate granularity on the release-to-birth ratio for the AGR-1 in-core experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. M. Scates; J. B. Walter; J. T. Maki

The AGR-1 Advanced Gas Reactor (AGR) tristructural-isotropic-particle fuel experiment underwent 13 irradiation intervals from December 2006 until November 2009 within the Idaho National Laboratory Advanced Test Reactor in support of the Next Generation Nuclear Power Plant program. During this multi-year experiment, release-to-birth rate ratios were computed at the end of each operating interval to provide information about fuel performance. Fission products released during irradiation were tracked daily by the Fission Product Monitoring System using 8-h measurements. Birth rate calculated by MCNP with ORIGEN for as-run conditions were computed at the end of each irradiation interval. Each time step in MCNPmore » provided neutron flux, reaction rates and AGR-1 compact composition, which were used to determine birth rate using ORIGEN. The initial birth-rate data, consisting of four values for each irradiation interval at the beginning, end, and two intermediate times, were interpolated to obtain values for each 8-h activity. The problem with this method is that any daily changes in heat rates or perturbations, such as shim control movement or core/lobe power fluctuations, would not be reflected in the interpolated data and a true picture of the system would not be presented. At the conclusion of the AGR-1 experiment, great efforts were put forth to compute daily birthrates, which were reprocessed with the 8-h release activity. The results of this study are presented in this paper.« less

48 CFR 227.7104 - Contracts under the Small Business Innovation Research (SBIR) Program.

Code of Federal Regulations, 2013 CFR

2013-10-01

... Data and Computer Software—Small Business Innovation Research (SBIR) Program, when technical data or computer software will be generated during performance of contracts under the SBIR program. (b) Under the clause at 252.227-7018, the Government obtains SBIR data rights in technical data and computer software...

48 CFR 227.7104 - Contracts under the Small Business Innovation Research (SBIR) Program.

Code of Federal Regulations, 2014 CFR

2014-10-01

... Data and Computer Software—Small Business Innovation Research (SBIR) Program, when technical data or computer software will be generated during performance of contracts under the SBIR program. (b) Under the clause at 252.227-7018, the Government obtains SBIR data rights in technical data and computer software...

Computer Program Development Specification for IDAMST Operational Flight Program Application, Software Type B5. Addendum 1.

DTIC Science & Technology

1976-07-30

Interface Requirements 4 3.1.1.1 Interface Block Diagram 4 3.1.1.2 Detailed Interface Definition 7 3.1.1.2.1 Subsystems 7 3.1.1.2.2 Controls & Displays 11 r...116 3.2.3.2 Navigation Brute Force 121 3.2.3.3 Cargo Brute Force 125 3.2.3.4 Sensor Brute Force 129 3.2.3.5 Controls /Displays Brute Force 135 3.2.3.6...STD-T553 Multiplex Data Bus, with the avionic subsystems, flight * control system, the controls /displays, engine sensors, and airframe sensors. 3.1

Low-Lying Electronic States of AlZn Calculated by MRCI+Q Method

NASA Astrophysics Data System (ADS)

Zhang, Shudong; Wang, Mingxu; Wang, Zifan; Hu, Kun; Dong, Jingping

2017-07-01

Some low-lying electronic states of AlZn have been studied by the ab initio calculation method of multireference configuration interaction (MRCI). The complete potential energy curves (PECs) of the three lowest doublet states (X2Π, A2Σ+, and B2Π) and the two lowest quartet states (a4Σ- and b4Π) are computed in the range of R = 0.1-0.9 nm and these states are correlated to three dissociation limits, X2Π and A2Σ+ to Zn(4s2,1S) + Al(3s23p1,2P), a4Σ- and b4Π to Zn(4s2,1S) + Al(3s13p2,4P), and B2Π to Zn(4s14p1,3P) + Al(3s23p1,2P). The calculated PECs indicate that the A2Σ+ state has a very shallow potential well and the other states show significant binding-state characteristics. The equilibrium internuclear distances Re, dissociation energies De, and term energies Te for the electronic excited states were obtained. All the possible vibrational levels, rotational constants, and spectral constants for the four bound states were computed by solving the radial Schrödinger equation of nuclear motion with the Level8.0 program provided by Le Roy.

48 CFR 27.404-4 - Contractor's release, publication, and use of data.

Code of Federal Regulations, 2013 CFR

2013-10-01

.... statutes. However, agencies may restrict the release or disclosure of computer software that is or is... software for purposes of established agency distribution programs, or where required to accomplish the purpose for which the software is acquired. (b) Except for the results of basic or applied research under...

Quantitative Measurements of the Effects of Variations in Panel Density and Distributions for Panel Method Computer Programs

DTIC Science & Technology

1980-01-01

AND ADDRESS 1 .PROGRAM ELEMENT. PROJECT, TASK AREA & WORK UNIT NUMBERS Aircraft and -.rew Systems Technology Di r. (C:ode 60 E61-.U o A0 Naval Air...0 . . . . . . . . . . . B- 1 B-If Input Statistics For Equations 5 and 6 .... ....... B- 2 B-Ill Computation of Coefficients of...trapezoidal panels and the formula for PAR can be derived for the case where equal spanwise and chordwise divisions are used: PAR (N/2M)/( 1 +X ( 2 = ( 2

Near-term hybrid vehicle program, phase 1. Appendix B: Design trade-off studies report. Volume 3: Computer program listings

NASA Technical Reports Server (NTRS)

1979-01-01

A description and listing is presented of two computer programs: Hybrid Vehicle Design Program (HYVELD) and Hybrid Vehicle Simulation Program (HYVEC). Both of the programs are modifications and extensions of similar programs developed as part of the Electric and Hybrid Vehicle System Research and Development Project.

MCNP6 model of the University of Washington clinical neutron therapy system (CNTS).

PubMed

Moffitt, Gregory B; Stewart, Robert D; Sandison, George A; Goorley, John T; Argento, David C; Jevremovic, Tatjana

2016-01-21

A MCNP6 dosimetry model is presented for the Clinical Neutron Therapy System (CNTS) at the University of Washington. In the CNTS, fast neutrons are generated by a 50.5 MeV proton beam incident on a 10.5 mm thick Be target. The production, scattering and absorption of neutrons, photons, and other particles are explicitly tracked throughout the key components of the CNTS, including the target, primary collimator, flattening filter, monitor unit ionization chamber, and multi-leaf collimator. Simulations of the open field tissue maximum ratio (TMR), percentage depth dose profiles, and lateral dose profiles in a 40 cm × 40 cm × 40 cm water phantom are in good agreement with ionization chamber measurements. For a nominal 10 × 10 field, the measured and calculated TMR values for depths of 1.5 cm, 5 cm, 10 cm, and 20 cm (compared to the dose at 1.7 cm) are within 0.22%, 2.23%, 4.30%, and 6.27%, respectively. For the three field sizes studied, 2.8 cm × 2.8 cm, 10.4 cm × 10.3 cm, and 28.8 cm × 28.8 cm, a gamma test comparing the measured and simulated percent depth dose curves have pass rates of 96.4%, 100.0%, and 78.6% (depth from 1.5 to 15 cm), respectively, using a 3% or 3 mm agreement criterion. At a representative depth of 10 cm, simulated lateral dose profiles have in-field (⩾ 10% of central axis dose) pass rates of 89.7% (2.8 cm × 2.8 cm), 89.6% (10.4 cm × 10.3 cm), and 100.0% (28.8 cm × 28.8 cm) using a 3% and 3 mm criterion. The MCNP6 model of the CNTS meets the minimum requirements for use as a quality assurance tool for treatment planning and provides useful insights and information to aid in the advancement of fast neutron therapy.

Using NJOY to Create MCNP ACE Files and Visualize Nuclear Data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kahler, Albert Comstock

We provide lecture materials that describe the input requirements to create various MCNP ACE files (Fast, Thermal, Dosimetry, Photo-nuclear and Photo-atomic) with the NJOY Nuclear Data Processing code system. Input instructions to visualize nuclear data with NJOY are also provided.

Adjoint-Based Sensitivity and Uncertainty Analysis for Density and Composition: A User’s Guide

DOE PAGES

Favorite, Jeffrey A.; Perko, Zoltan; Kiedrowski, Brian C.; ...

2017-03-01

The ability to perform sensitivity analyses using adjoint-based first-order sensitivity theory has existed for decades. This paper provides guidance on how adjoint sensitivity methods can be used to predict the effect of material density and composition uncertainties in critical experiments, including when these uncertain parameters are correlated or constrained. Two widely used Monte Carlo codes, MCNP6 (Ref. 2) and SCALE 6.2 (Ref. 3), are both capable of computing isotopic density sensitivities in continuous energy and angle. Additionally, Perkó et al. have shown how individual isotope density sensitivities, easily computed using adjoint methods, can be combined to compute constrained first-order sensitivitiesmore » that may be used in the uncertainty analysis. This paper provides details on how the codes are used to compute first-order sensitivities and how the sensitivities are used in an uncertainty analysis. Constrained first-order sensitivities are computed in a simple example problem.« less

Evaluation of the Impact Computer Program as a Linear Design Tool for Bird-Resistant Aircraft Transparencies.

DTIC Science & Technology

1980-03-01

Pressure on a Flat Plate, Arnold Engineering Development Center, Arnold Air Force Station, Tennessee 37389, AEDC-TR-79-14. 28. G. B. Thomas , Calculus and...Equation (6) was then 0.00177 sec. The average impact force from Equation (7) was 23,245 lb. The bird impact force-time history (28) G. B. Thomas ... Calculus and Analytic Geometry, Addison- Wesley, 1965. 60 Parallel to C Windshield N is unit vector B normal to windshieldNN panel at target point ... 4C

United States General Accounting Office Publications List.

DTIC Science & Technology

1981-06-30

114438 AFMD-81-25 nance of computer programs: Feb. 26, 1981 expensive and undermanaged 83 Secret Service has more corn- 114604 GGD-81-43 puter capacity...30, 1980 173 Examination of records of the 115066 AFMD-B1 -4 House of Representatives Fi- April 29, 1981 nance Office, fiscal 1980 174 Audit of the...199 The Sudden Infant Death 114727 HRD-81-25 Syndrome program helps Feb. 6, 1981 families but needs improve- ment 200 Circumstances of contract for

IViPP: A Tool for Visualization in Particle Physics

NASA Astrophysics Data System (ADS)

Tran, Hieu; Skiba, Elizabeth; Baldwin, Doug

2011-10-01

Experiments and simulations in physics generate a lot of data; visualization is helpful to prepare that data for analysis. IViPP (Interactive Visualizations in Particle Physics) is an interactive computer program that visualizes results of particle physics simulations or experiments. IViPP can handle data from different simulators, such as SRIM or MCNP. It can display relevant geometry and measured scalar data; it can do simple selection from the visualized data. In order to be an effective visualization tool, IViPP must have a software architecture that can flexibly adapt to new data sources and display styles. It must be able to display complicated geometry and measured data with a high dynamic range. We therefore organize it in a highly modular structure, we develop libraries to describe geometry algorithmically, use rendering algorithms running on the powerful GPU to display 3-D geometry at interactive rates, and we represent scalar values in a visual form of scientific notation that shows both mantissa and exponent. This work was supported in part by the US Department of Energy through the Laboratory for Laser Energetics (LLE), with special thanks to Craig Sangster at LLE.

MCNP6 unstructured mesh application to estimate the photoneutron distribution and induced activity inside a linac bunker

NASA Astrophysics Data System (ADS)

Juste, B.; Morató, S.; Miró, R.; Verdú, G.; Díez, S.

2017-08-01

Unwanted neutrons in radiation therapy treatments are typically generated by photonuclear reactions. High-energy beams emitted by medical Linear Accelerators (LinAcs) interact with high atomic number materials situated in the accelerator head and release neutrons. Since neutrons have a high relative biological effectiveness, even low neutron doses may imply significant exposure of patients. It is also important to study radioactivity induced by these photoneutrons when interacting with the different materials and components of the treatment head facility and the shielding room walls, since persons not present during irradiation (e.g. medical staff) may be exposed to them even when the accelerator is not operating. These problems are studied in this work in order to contribute to challenge the radiation protection in these treatment locations. The work has been performed by simulation using the latest state of the art of Monte-Carlo computer code MCNP6. To that, a detailed model of particles transport inside the bunker and treatment head has been carried out using a meshed geometry model. The LinAc studied is an Elekta Precise accelerator with a treatment photon energy of 15 MeV used at the Hospital Clinic Universitari de Valencia, Spain.

Argonne Bubble Experiment Thermal Model Development III

DOE Office of Scientific and Technical Information (OSTI.GOV)

Buechler, Cynthia Eileen

This report describes the continuation of the work reported in “Argonne Bubble Experiment Thermal Model Development” and “Argonne Bubble Experiment Thermal Model Development II”. The experiment was performed at Argonne National Laboratory (ANL) in 2014. A rastered 35 MeV electron beam deposited power in a solution of uranyl sulfate, generating heat and radiolytic gas bubbles. Irradiations were performed at beam power levels between 6 and 15 kW. Solution temperatures were measured by thermocouples, and gas bubble behavior was recorded. The previous report2 described the Monte-Carlo N-Particle (MCNP) calculations and Computational Fluid Dynamics (CFD) analysis performed on the as-built solution vesselmore » geometry. The CFD simulations in the current analysis were performed using Ansys Fluent, Ver. 17.2. The same power profiles determined from MCNP calculations in earlier work were used for the 12 and 15 kW simulations. The primary goal of the current work is to calculate the temperature profiles for the 12 and 15 kW cases using reasonable estimates for the gas generation rate, based on images of the bubbles recorded during the irradiations. Temperature profiles resulting from the CFD calculations are compared to experimental measurements.« less

Adjoint-Based Uncertainty Quantification with MCNP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Seifried, Jeffrey E.

2011-09-01

This work serves to quantify the instantaneous uncertainties in neutron transport simulations born from nuclear data and statistical counting uncertainties. Perturbation and adjoint theories are used to derive implicit sensitivity expressions. These expressions are transformed into forms that are convenient for construction with MCNP6, creating the ability to perform adjoint-based uncertainty quantification with MCNP6. These new tools are exercised on the depleted-uranium hybrid LIFE blanket, quantifying its sensitivities and uncertainties to important figures of merit. Overall, these uncertainty estimates are small (< 2%). Having quantified the sensitivities and uncertainties, physical understanding of the system is gained and some confidence inmore » the simulation is acquired.« less

LANL Summer 2016 Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mendoza, Paul Michael

The Monte Carlo N-Particle (MCNP) transport code developed at Los Alamos National Laboratory (LANL) utilizes nuclear cross-section data in a compact ENDF (ACE) format. The accuracy of MCNP calculations depends on the accuracy of nuclear ACE data tables, which depends on the accuracy of the original ENDF files. There are some noticeable differences in ENDF files from one generation to the next, even among the more common fissile materials. As the next generation of ENDF files is being prepared, several software tools were developed to simulate a large number of benchmarks in MCNP (over 1000), collect data from these simulations,more » and visually represent the results.« less

FT-IR, FT-Raman spectra, density functional computations of the vibrational spectra and molecular conformational analysis of 2,5-di-tert-butyl-hydroquinone

NASA Astrophysics Data System (ADS)

Subramanian, N.; Sundaraganesan, N.; Dereli, Ö.; Türkkan, E.

2011-12-01

The purpose of finding conformer among six different possible conformers of 2,5-di-tert-butyl-hydroquinone (DTBHQ), its equilibrium geometry and harmonic wavenumbers were calculated by the B3LYP/6-31G(d,p) method. The infrared and Raman spectra of DTBHQ were recorded in the region 400-4000 cm -1 and 50-3500 cm -1, respectively. In addition, the IR spectra in CCl 4 at various concentrations of DTBHQ are also recorded. The computed vibrational wavenumbers were compared with the IR and Raman experimental data. Computational calculations at B3LYP level with two different basis sets 6-31G(d,p) and 6-311++G(d,p) are also employed in the study of the possible conformer of DTBHQ. The complete assignments were performed on the basis of the potential energy distribution (PED) of the vibrational modes, calculated using VEDA 4 program. The general agreement between the observed and calculated frequencies was established.

On 3-D inelastic analysis methods for hot section components. Volume 1: Special finite element models

NASA Technical Reports Server (NTRS)

Nakazawa, S.

1987-01-01

This Annual Status Report presents the results of work performed during the third year of the 3-D Inelastic Analysis Methods for Hot Section Components program (NASA Contract NAS3-23697). The objective of the program is to produce a series of new computer codes that permit more accurate and efficient three-dimensional analysis of selected hot section components, i.e., combustor liners, turbine blades, and turbine vanes. The computer codes embody a progression of mathematical models and are streamlined to take advantage of geometrical features, loading conditions, and forms of material response that distinguish each group of selected components. This report is presented in two volumes. Volume 1 describes effort performed under Task 4B, Special Finite Element Special Function Models, while Volume 2 concentrates on Task 4C, Advanced Special Functions Models.

Modular space station, phase B extension. Information management advanced development. Volume 5: Software assembly

NASA Technical Reports Server (NTRS)

Gerber, C. R.

1972-01-01

The development of uniform computer program standards and conventions for the modular space station is discussed. The accomplishments analyzed are: (1) development of computer program specification hierarchy, (2) definition of computer program development plan, and (3) recommendations for utilization of all operating on-board space station related data processing facilities.

Computational Understanding: Analysis of Sentences and Context

DTIC Science & Technology

1974-05-01

Computer Science Department Stanford, California 9430b 10- PROGRAM ELEMENT. PROJECT. TASK AREA « WORK UNIT NUMBERS II. CONTROLLING OFFICE NAME...these is the need tor programs that can respond in useful ways to information expressed in a natural language. However a computational understanding...buying structure because "Mary" appears where it does. But the time for analysis was rarely over five seconds of computer time, when the Lisp program

INDIVIDUAL DOSIMETRY IN DISPOSAL REPOSITORY OF HEAT-GENERATING NUCLEAR WASTE.

PubMed

Pang, Bo; Saurí Suárez, Héctor; Becker, Frank

2016-09-01

Certain working scenarios in a disposal facility of heat-generating nuclear waste might lead to an enhanced level of radiation exposure for workers in such facilities. Hence, a realistic estimation of the personal dose during individual working scenarios is desired. In this study, the general-purpose Monte Carlo N-Particle code MCNP6 (Pelowitz, D. B. (ed). MCNP6 user manual LA-CP-13-00634, Rev. 0 (2013)) was applied to simulate a representative radiation field in a disposal facility. A tool to estimate the personal dose was then proposed by taking into account the influence of individual motion sequences during working scenarios. As basis for this approach, a movable whole-body phantom was developed to describe individual body gestures of the workers during motion sequences. In this study, the proposed method was applied to the German concept of geological disposal in rock salt. The feasibility of the proposed approach was demonstrated with an example of working scenario in an emplacement drift of a rock salt mine. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Neutronics Studies for the Nab Experiment

NASA Astrophysics Data System (ADS)

Scott, Elizabeth; Nab Collaboration

2017-09-01

The Nab experiment at the Spallation Neutron Source at ORNL aims to measure the neutron beta decay electron-neutrino correlation coefficient ``a'' and the Fierz interference term ``b'' with competitive precision. In Nab, the parameter ``a'' is extracted from the proton momentum and electron energy using an asymmetric magnetic spectrometer and two large-area highly pixelated Si detectors . To achieve 10-3 accuracy, there must be low background rates compared to our 1 kHz signal rates. The background is primarily reduced by using coincidence detection of the electron and photon from the decay. However, further reduction is still necessary. Neutron and gamma rates in the Si detectors can lead to false coincidences. The majority of this background radiation can be reduced by well designed collimation and shielding. The collimation design was done with McStas and the background shielding with MCNP6 (Monte Carlo N-Particle 6). Neutrons are absorbed by 6Li -loaded materials or borated polyethylene and gammas close to spectrometer with non magnetic materials such as lead and stainless steel. I will present the shielding design and MCNP6 results.

Reanalysis of tritium production in a sphere of /sup 6/LiD irradiated by 14-MeV neutrons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fawcett, L.R. Jr.

1985-08-01

Tritium production and activation of radiochemical detector foils in a sphere of /sup 6/LiD irradiated by a central source of 14-MeV neutrons has been reanalyzed. The /sup 6/LiD sphere consisted of 10 solid hemispherical nested shells with ampules of /sup 6/LiH, /sup 7/LiH, and activation foils located 2.2, 5, 7.7, 12.6, 20, and 30 cm from the center. The Los Alamos Monte Carlo Neutron Photon Transport Code (MCNP) was used to calculate neutron transport through the /sup 6/LiD, tritium production in the ampules, and foil activation. The MCNP input model was three-dimensional and employed ENDF/B-V cross sections for transport, tritiummore » production, and (where available) foil activation. The reanalyzed experimentally observed-to-calculated values of tritium production were 1.053 +- 2.1% in /sup 6/LiH and 0.999 +- 2.1% in /sup 7/LiH. The recalculated foil activation observed-to-calculated ratios were not generally improved over those reported in the original analysis.« less

Critical Low-Noise Technologies Being Developed for Engine Noise Reduction Systems Subproject

NASA Technical Reports Server (NTRS)

Grady, Joseph E.; Civinskas, Kestutis C.

2004-01-01

NASA's previous Advanced Subsonic Technology (AST) Noise Reduction Program delivered the initial technologies for meeting a 10-year goal of a 10-dB reduction in total aircraft system noise. Technology Readiness Levels achieved for the engine-noise-reduction technologies ranged from 4 (rig scale) to 6 (engine demonstration). The current Quiet Aircraft Technology (QAT) project is building on those AST accomplishments to achieve the additional noise reduction needed to meet the Aerospace Technology Enterprise's 10-year goal, again validated through a combination of laboratory rig and engine demonstration tests. In order to meet the Aerospace Technology Enterprise goal for future aircraft of a 50- reduction in the perceived noise level, reductions of 4 dB are needed in both fan and jet noise. The primary objectives of the Engine Noise Reduction Systems (ENRS) subproject are, therefore, to develop technologies to reduce both fan and jet noise by 4 dB, to demonstrate these technologies in engine tests, and to develop and experimentally validate Computational Aero Acoustics (CAA) computer codes that will improve our ability to predict engine noise.

Case Studies of Navigation Channel and Port Sedimentation

DTIC Science & Technology

2011-01-14

Engineering Supplement to Limited Reevaluation Report, Vol. II (1995) 1/14/2011 18 MCNP  Hypotheses Why is channel shoaling more than anticipated? (1...activities have modified shoaling patterns and magnitudes (trawling, dredging and disposal, subsidence/fluid withdrawal). MCNP Hypotheses Why is channel

PLNoise: a package for exact numerical simulation of power-law noises

NASA Astrophysics Data System (ADS)

Milotti, Edoardo

2006-08-01

Many simulations of stochastic processes require colored noises: here I describe a small program library that generates samples with a tunable power-law spectral density: the algorithm can be modified to generate more general colored noises, and is exact for all time steps, even when they are unevenly spaced (as may often happen in the case of astronomical data, see e.g. [N.R. Lomb, Astrophys. Space Sci. 39 (1976) 447]. The method is exact in the sense that it reproduces a process that is theoretically guaranteed to produce a range-limited power-law spectrum 1/f with -1<β⩽1. The algorithm has a well-behaved computational complexity, it produces a nearly perfect Gaussian noise, and its computational efficiency depends on the required degree of noise Gaussianity. Program summaryTitle of program: PLNoise Catalogue identifier:ADXV_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/ADXV_v1_0.html Program obtainable from: CPC Program Library, Queen's University of Belfast, N. Ireland Licensing provisions: none Programming language used: ANSI C Computer: Any computer with an ANSI C compiler: the package has been tested with gcc version 3.2.3 on Red Hat Linux 3.2.3-52 and gcc version 4.0.0 and 4.0.1 on Apple Mac OS X-10.4 Operating system: All operating systems capable of running an ANSI C compiler No. of lines in distributed program, including test data, etc.:6238 No. of bytes in distributed program, including test data, etc.:52 387 Distribution format:tar.gz RAM: The code of the test program is very compact (about 50 Kbytes), but the program works with list management and allocates memory dynamically; in a typical run (like the one discussed in Section 4 in the long write-up) with average list length 2ṡ10, the RAM taken by the list is 200 Kbytes. External routines: The package needs external routines to generate uniform and exponential deviates. The implementation described here uses the random number generation library ranlib freely available from Netlib [B.W. Brown, J. Lovato, K. Russell, ranlib, available from Netlib, http://www.netlib.org/random/index.html, select the C version ranlib.c], but it has also been successfully tested with the random number routines in Numerical Recipes [W.H. Press, S.A. Teulkolsky, W.T. Vetterling, B.P. Flannery, Numerical Recipes in C: The Art of Scientific Computing, second ed., Cambridge Univ. Press, Cambridge, 1992, pp. 274-290]. Notice that ranlib requires a pair of routines from the linear algebra package LINPACK, and that the distribution of ranlib includes the C source of these routines, in case LINPACK is not installed on the target machine. Nature of problem: Exact generation of different types of Gaussian colored noise. Solution method: Random superposition of relaxation processes [E. Milotti, Phys. Rev. E 72 (2005) 056701]. Unusual features: The algorithm is theoretically guaranteed to be exact, and unlike all other existing generators it can generate samples with uneven spacing. Additional comments: The program requires an initialization step; for some parameter sets this may become rather heavy. Running time: Running time varies widely with different input parameters, however in a test run like the one in Section 4 in this work, the generation routine took on average about 7 ms for each sample.

RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user`s manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.

This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARCmore » and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.« less

High Temperature Monotonic and Cyclic Deformation in a Directionally Solidified Nickel-Base Superalloy.

DTIC Science & Technology

1986-05-01

was conducted in air, using a SATEC Systems computer-controlled servohydraulic testing machine. This machine uses a minicomputer (Digital PDP 11/34...overall test program) was run. This test was performed using a feature of the SATEC machine called combinatorial feedback, which allowed a user-defined...Rn) l/T + (in Es /A)/n (4.3) Q can be calculated from 0*: b Q=n (4.4) Creep data for DS MAR-M246, containing no Hafnium, from Reference 99 was used to

The effect of tandem-ovoid titanium applicator on points A, B, bladder, and rectum doses in gynecological brachytherapy using 192Ir.

PubMed

Sadeghi, Mohammad Hosein; Sina, Sedigheh; Mehdizadeh, Amir; Faghihi, Reza; Moharramzadeh, Vahed; Meigooni, Ali Soleimani

2018-02-01

The dosimetry procedure by simple superposition accounts only for the self-shielding of the source and does not take into account the attenuation of photons by the applicators. The purpose of this investigation is an estimation of the effects of the tandem and ovoid applicator on dose distribution inside the phantom by MCNP5 Monte Carlo simulations. In this study, the superposition method is used for obtaining the dose distribution in the phantom without using the applicator for a typical gynecological brachytherapy (superposition-1). Then, the sources are simulated inside the tandem and ovoid applicator to identify the effect of applicator attenuation (superposition-2), and the dose at points A, B, bladder, and rectum were compared with the results of superposition. The exact dwell positions, times of the source, and positions of the dosimetry points were determined in images of a patient and treatment data of an adult woman patient from a cancer center. The MCNP5 Monte Carlo (MC) code was used for simulation of the phantoms, applicators, and the sources. The results of this study showed no significant differences between the results of superposition method and the MC simulations for different dosimetry points. The difference in all important dosimetry points was found to be less than 5%. According to the results, applicator attenuation has no significant effect on the calculated points dose, the superposition method, adding the dose of each source obtained by the MC simulation, can estimate the dose to points A, B, bladder, and rectum with good accuracy.

The MCNP-4C2 design of a two element photon/electron dosemeter that uses magnesium/copper/phosphorus doped lithium fluoride.

PubMed

Eakins, J S; Bartlett, D T; Hager, L G; Molinos-Solsona, C; Tanner, R J

2008-01-01

The Health Protection Agency is changing from using detectors made from 7LiF:Mg,Ti in its photon/electron personal dosemeters, to 7LiF:Mg,Cu,P. Specifically, the Harshaw TLD-700H card is to be adopted. As a consequence of this change, the dosemeter holder is also being modified not only to accommodate the shape of the new card, but also to optimize the photon and electron response characteristics of the device. This redesign process was achieved using MCNP-4C2 and the kerma approximation, electron range/energy tables with additional electron transport calculations, and experimental validation, with different potential filters compared; the optimum filter studied was a polytetrafluoroethylene disc of diameter 18 mm and thickness 4.3 mm. Calculated relative response characteristics at different angles of incidence and energies between 16 and 6174 keV are presented for this new dosemeter configuration and compared with measured type-test results. A new estimate for the energy-dependent relative light conversion efficiency appropriate to the 7LiF:Mg,Cu,P was also derived for determining the correct dosemeter response.

Users manual for the Chameleon parallel programming tools

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gropp, W.; Smith, B.

1993-06-01

Message passing is a common method for writing programs for distributed-memory parallel computers. Unfortunately, the lack of a standard for message passing has hampered the construction of portable and efficient parallel programs. In an attempt to remedy this problem, a number of groups have developed their own message-passing systems, each with its own strengths and weaknesses. Chameleon is a second-generation system of this type. Rather than replacing these existing systems, Chameleon is meant to supplement them by providing a uniform way to access many of these systems. Chameleon`s goals are to (a) be very lightweight (low over-head), (b) be highlymore » portable, and (c) help standardize program startup and the use of emerging message-passing operations such as collective operations on subsets of processors. Chameleon also provides a way to port programs written using PICL or Intel NX message passing to other systems, including collections of workstations. Chameleon is tracking the Message-Passing Interface (MPI) draft standard and will provide both an MPI implementation and an MPI transport layer. Chameleon provides support for heterogeneous computing by using p4 and PVM. Chameleon`s support for homogeneous computing includes the portable libraries p4, PICL, and PVM and vendor-specific implementation for Intel NX, IBM EUI (SP-1), and Thinking Machines CMMD (CM-5). Support for Ncube and PVM 3.x is also under development.« less

Gamma irradiator dose mapping simulation using the MCNP code and benchmarking with dosimetry.

PubMed

Sohrabpour, M; Hassanzadeh, M; Shahriari, M; Sharifzadeh, M

2002-10-01

The Monte Carlo transport code, MCNP, has been applied in simulating dose rate distribution in the IR-136 gamma irradiator system. Isodose curves, cumulative dose values, and system design data such as throughputs, over-dose-ratios, and efficiencies have been simulated as functions of product density. Simulated isodose curves, and cumulative dose values were compared with dosimetry values obtained using polymethyle-methacrylate, Fricke, ethanol-chlorobenzene, and potassium dichromate dosimeters. The produced system design data were also found to agree quite favorably with those of the system manufacturer's data. MCNP has thus been found to be an effective transport code for handling of various dose mapping excercises for gamma irradiators.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bily, T.

Thermoluminescent dosimeters represent very useful tool for gamma fields parameters measurements at nuclear research reactors, especially at zero power ones. {sup 7}LiF:Mg,Ti and {sup 7}LiF:Mg,Cu,P type TL dosimeters enable determination of only gamma component in mixed neutron - gamma field. At VR-1 reactor operated within the Faculty of Nuclear Sciences and Physical Engineering at the Czech Technical University in Prague the integral characteristics of gamma rays field were investigated, especially its spatial distribution and time behaviour, i.e. the non-saturated delayed gamma ray emission influence. Measured spatial distributions were compared with monte carlo code MCNP5 calculations. Although MCNP cannot generate delayedmore » gamma rays from fission, the relative gamma dose rate distribution is within {+-} 15% with measured values. The experiments were carried out with core configuration C1 consisting of LEU fuel IRT-4M (19.7 %). (author)« less

Gas Core Reactor Numerical Simulation Using a Coupled MHD-MCNP Model

NASA Technical Reports Server (NTRS)

Kazeminezhad, F.; Anghaie, S.

2008-01-01

Analysis is provided in this report of using two head-on magnetohydrodynamic (MHD) shocks to achieve supercritical nuclear fission in an axially elongated cylinder filled with UF4 gas as an energy source for deep space missions. The motivation for each aspect of the design is explained and supported by theory and numerical simulations. A subsequent report will provide detail on relevant experimental work to validate the concept. Here the focus is on the theory of and simulations for the proposed gas core reactor conceptual design from the onset of shock generations to the supercritical state achieved when the shocks collide. The MHD model is coupled to a standard nuclear code (MCNP) to observe the neutron flux and fission power attributed to the supercritical state brought about by the shock collisions. Throughout the modeling, realistic parameters are used for the initial ambient gaseous state and currents to ensure a resulting supercritical state upon shock collisions.

MCNP modelling of vaginal and uterine applicators used in intracavitary brachytherapy and comparison with radiochromic film measurements

NASA Astrophysics Data System (ADS)

Ceccolini, E.; Gerardy, I.; Ródenas, J.; van Dycke, M.; Gallardo, S.; Mostacci, D.

Brachytherapy is an advanced cancer treatment that is minimally invasive, minimising radiation exposure to the surrounding healthy tissues. Microselectron© Nucletron devices with 192Ir source can be used for gynaecological brachytherapy, in patients with vaginal or uterine cancer. Measurements of isodose curves have been performed in a PMMA phantom and compared with Monte Carlo calculations and TPS (Plato software of Nucletron BPS 14.2) evaluation. The isodose measurements have been performed with radiochromic films (Gafchromic EBT©). The dose matrix has been obtained after digitalisation and use of a dose calibration curve obtained with a 6 MV photon beam provided by a medical linear accelerator. A comparison between the calculated and the measured matrix has been performed. The calculated dose matrix is obtained with a simulation using the MCNP5 Monte Carlo code (F4MESH tally).

Determination of efficiency of an aged HPGe detector for gaseous sources by self absorption correction and point source methods

NASA Astrophysics Data System (ADS)

Sarangapani, R.; Jose, M. T.; Srinivasan, T. K.; Venkatraman, B.

2017-07-01

Methods for the determination of efficiency of an aged high purity germanium (HPGe) detector for gaseous sources have been presented in the paper. X-ray radiography of the detector has been performed to get detector dimensions for computational purposes. The dead layer thickness of HPGe detector has been ascertained from experiments and Monte Carlo computations. Experimental work with standard point and liquid sources in several cylindrical geometries has been undertaken for obtaining energy dependant efficiency. Monte Carlo simulations have been performed for computing efficiencies for point, liquid and gaseous sources. Self absorption correction factors have been obtained using mathematical equations for volume sources and MCNP simulations. Self-absorption correction and point source methods have been used to estimate the efficiency for gaseous sources. The efficiencies determined from the present work have been used to estimate activity of cover gas sample of a fast reactor.

Measurement of 89Y(n,2n) spectral averaged cross section in LR-0 special core reactor spectrum

NASA Astrophysics Data System (ADS)

Košťál, Michal; Losa, Evžen; Baroň, Petr; Šolc, Jaroslav; Švadlenková, Marie; Koleška, Michal; Mareček, Martin; Uhlíř, Jan

2017-12-01

The present paper describes reaction rate measurement of 89Y(n,2n)88Y in a well-defined reactor spectrum of a special core assembled in the LR-0 reactor and compares this value with results of simulation. The reaction rate is derived from the measurement of activity of 88Y using gamma-ray spectrometry of irradiated Y2O3 sample. The resulting cross section value averaged in spectrum is 43.9 ± 1.5 μb, averaged in the 235U spectrum is 0.172 ± 0.006 mb. This cross-section is important as it is used as high energy neutron monitor and is therefore included in the International Reactor Dosimetry and Fusion File. Calculations of reaction rates were performed with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND-2010, CENDL-3.1 and IRDFF nuclear data libraries. The agreement with uranium description by CIELO library is very good, while in ENDF/B-VII.0 description of uranium, underprediction about 10% in average can be observed.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hanan, N. A.; Matos, J. E.

At The request of the Czech Technical University in Prague, ANL has performed independent verification calculations using the MCNP Monte Carlo code for three core configurations of the VR-1 reactor: a current core configuration B1 with HEU (36%) IRT-3M fuel assemblies and planned core configurations C1 and C2 with LEU (19.7%) IRT-4M fuel assemblies. Details of these configurations were provided to ANL by CTU. For core configuration B1, criticality calculations were performed for two sets of control rod positions provided to ANL by CTU. For core configurations C1 and C2, criticality calculations were done for cases with all control rodsmore » at the top positions, all control rods at the bottom positions, and two critical states of the reactor for different control rod positions. In addition, sensitivity studies for variation of the {sup 235}U mass in each fuel assembly and variation of the fuel meat and cladding thicknesses in each of the fuel tubes were done for the C1 core configuration. Finally the reactivity worth of the individual control rods was calculated for the B1, C1, and C2 core configurations.« less

Quantitative basis for component factors of gas flow proportional counting efficiencies

NASA Astrophysics Data System (ADS)

Nichols, Michael C.

This dissertation investigates the counting efficiency calibration of a gas flow proportional counter with beta-particle emitters in order to (1) determine by measurements and simulation the values of the component factors of beta-particle counting efficiency for a proportional counter, (2) compare the simulation results and measured counting efficiencies, and (3) determine the uncertainty of the simulation and measurements. Monte Carlo simulation results by the MCNP5 code were compared with measured counting efficiencies as a function of sample thickness for 14C, 89Sr, 90Sr, and 90Y. The Monte Carlo model simulated strontium carbonate with areal thicknesses from 0.1 to 35 mg cm-2. The samples were precipitated as strontium carbonate with areal thicknesses from 3 to 33 mg cm-2 , mounted on membrane filters, and counted on a low background gas flow proportional counter. The estimated fractional standard deviation was 2--4% (except 6% for 14C) for efficiency measurements of the radionuclides. The Monte Carlo simulations have uncertainties estimated to be 5 to 6 percent for carbon-14 and 2.4 percent for strontium-89, strontium-90, and yttrium-90. The curves of simulated counting efficiency vs. sample areal thickness agreed within 3% of the curves of best fit drawn through the 25--49 measured points for each of the four radionuclides. Contributions from this research include development of uncertainty budgets for the analytical processes; evaluation of alternative methods for determining chemical yield critical to the measurement process; correcting a bias found in the MCNP normalization of beta spectra histogram; clarifying the interpretation of the commonly used ICRU beta-particle spectra for use by MCNP; and evaluation of instrument parameters as applied to the simulation model to obtain estimates of the counting efficiency from simulated pulse height tallies.

Computational-Experimental Processing of Boride/Carbide Composites by Reactive Infusion of Hf Alloy Melts into B4C

DTIC Science & Technology

2015-09-16

AFRL-AFOSR-VA-TR-2015-0314 Computational -Experimental Processing of Boride /Carbide Composites by Reactive Infusion of Hf Alloy Melts into B4C...Computational -Experimental Processing of Boride /Carbide Composites by Reactive Infusion of Hf Alloy Melts into B4C 5a.  CONTRACT NUMBER 5b.  GRANT...with a packed bed of B4C to form boride - carbide precipitates. Although the ultimate goal of the research endeavor is to enhance significantly the

MCNP5 evaluation of photoneutron production from the Alexandria University 15 MV Elekta Precise medical LINAC.

PubMed

Abou-Taleb, W M; Hassan, M H; El Mallah, E A; Kotb, S M

2018-05-01

Photoneutron production, and the dose equivalent, in the head assembly of the 15 MV Elekta Precise medical linac; operating in the faculty of Medicine at Alexandria University were estimated with the MCNP5 code. Photoneutron spectra were calculated in air and inside a water phantom to different depths as a function of the radiation field sizes. The maximum neutron fluence is 3.346×10 -9 n/cm 2 -e for a 30×30 cm 2 field size to 2-4 cm-depth in the phantom. The dose equivalent due to fast neutron increases as the field size increases, being a maximum of 0.912 ± 0.05 mSv/Gy at depth between 2 and 4 cm in the water phantom for 40×40 cm 2 field size. Photoneutron fluence and dose equivalent are larger to 100 cm from the isocenter than to 35 cm from the treatment room wall. Copyright © 2018 Elsevier Ltd. All rights reserved.

Multiprocessing MCNP on an IBN RS/6000 cluster

DOE Office of Scientific and Technical Information (OSTI.GOV)

McKinney, G.W.; West, J.T.

1993-01-01

The advent of high-performance computer systems has brought to maturity programming concepts like vectorization, multiprocessing, and multitasking. While there are many schools of thought as to the most significant factor in obtaining order-of-magnitude increases in performance, such speedup can only be achieved by integrating the computer system and application code. Vectorization leads to faster manipulation of arrays by overlapping instruction CPU cycles. Discrete ordinates codes, which require the solving of large matrices, have proved to be major benefactors of vectorization. Monte Carlo transport, on the other hand, typically contains numerous logic statements and requires extensive redevelopment to benefit from vectorization.more » Multiprocessing and multitasking provide additional CPU cycles via multiple processors. Such systems are generally designed with either common memory access (multitasking) or distributed memory access. In both cases, theoretical speedup, as a function of the number of processors P and the fraction f of task time that multiprocesses, can be formulated using Amdahl's law: S(f, P) =1/(1-f+f/P). However, for most applications, this theoretical limit cannot be achieved because of additional terms (e.g., multitasking overhead, memory overlap, etc.) that are not included in Amdahl's law. Monte Carlo transport is a natural candidate for multiprocessing because the particle tracks are generally independent, and the precision of the result increases as the square Foot of the number of particles tracked.« less

Multiprocessing MCNP on an IBM RS/6000 cluster

DOE Office of Scientific and Technical Information (OSTI.GOV)

McKinney, G.W.; West, J.T.

1993-03-01

The advent of high-performance computer systems has brought to maturity programming concepts like vectorization, multiprocessing, and multitasking. While there are many schools of thought as to the most significant factor in obtaining order-of-magnitude increases in performance, such speedup can only be achieved by integrating the computer system and application code. Vectorization leads to faster manipulation of arrays by overlapping instruction CPU cycles. Discrete ordinates codes, which require the solving of large matrices, have proved to be major benefactors of vectorization. Monte Carlo transport, on the other hand, typically contains numerous logic statements and requires extensive redevelopment to benefit from vectorization.more » Multiprocessing and multitasking provide additional CPU cycles via multiple processors. Such systems are generally designed with either common memory access (multitasking) or distributed memory access. In both cases, theoretical speedup, as a function of the number of processors (P) and the fraction of task time that multiprocesses (f), can be formulated using Amdahl`s Law S ((f,P) = 1 f + f/P). However, for most applications this theoretical limit cannot be achieved, due to additional terms not included in Amdahl`s Law. Monte Carlo transport is a natural candidate for multiprocessing, since the particle tracks are generally independent and the precision of the result increases as the square root of the number of particles tracked.« less

Computer programming for nucleic acid studies. II. Total chemical shifts calculation of all protons of double-stranded helices.

PubMed

Giessner-Prettre, C; Ribas Prado, F; Pullman, B; Kan, L; Kast, J R; Ts'o, P O

1981-01-01

A FORTRAN computer program called SHIFTS is described. Through SHIFTS, one can calculate the NMR chemical shifts of the proton resonances of single and double-stranded nucleic acids of known sequences and of predetermined conformations. The program can handle RNA and DNA for an arbitrary sequence of a set of 4 out of the 6 base types A,U,G,C,I and T. Data files for the geometrical parameters are available for A-, A'-, B-, D- and S-conformations. The positions of all the atoms are calculated using a modified version of the SEQ program [1]. Then, based on this defined geometry three chemical shift effects exerted by the atoms of the neighboring nucleotides on the protons of each monomeric unit are calculated separately: the ring current shielding effect: the local atomic magnetic susceptibility effect (including both diamagnetic and paramagnetic terms); and the polarization or electric field effect. Results of the program are compared with experimental results for a gamma (ApApGpCpUpU) 2 helical duplex and with calculated results on this same helix based on model building of A'-form and B-form and on graphical procedure for evaluating the ring current effects.

Effect of particle size and percentages of Boron carbide on the thermal neutron radiation shielding properties of HDPE/B4C composite: Experimental and simulation studies

NASA Astrophysics Data System (ADS)

Soltani, Zahra; Beigzadeh, Amirmohammad; Ziaie, Farhood; Asadi, Eskandar

2016-10-01

In this paper the effects of particle size and weight percentage of the reinforcement phase on the absorption ability of thermal neutron by HDPE/B4C composites were investigated by means of Monte-Carlo simulation method using MCNP code and experimental studies. The composite samples were prepared using the HDPE filled with different weight percentages of Boron carbide powder in the form of micro and nano particles. Micro and nano composite were prepared under the similar mixing and moulding processes. The samples were subjected to thermal neutron radiation. Neutron shielding efficiency in terms of the neutron transmission fractions of the composite samples were investigated and compared with simulation results. According to the simulation results, the particle size of the radiation shielding material has an important role on the shielding efficiency. By decreasing the particle size of shielding material in each weight percentages of the reinforcement phase, better radiation shielding properties were obtained. It seems that, decreasing the particle size and homogeneous distribution of nano forms of B4C particles, cause to increase the collision probability between the incident thermal neutron and the shielding material which consequently improve the radiation shielding properties. So, this result, propose the feasibility of nano composite as shielding material to have a high performance shielding characteristic, low weight and low thick shielding along with economical benefit.

Radioactive ion beams produced by neutron-induced fission at ISOLDE

NASA Astrophysics Data System (ADS)

Catherall, R.; Lettry, J.; Gilardoni, S.; Köster, U.; Isolde Collaboration

2003-05-01

The production rates of neutron-rich fission products for the next-generation radioactive beam facility EURISOL [EU-RTD Project EURISOL (HPRI-CT-1999-50001)] are mainly limited by the maximum amount of power deposited by protons in the target. An alternative approach is to use neutron beams to induce fission in actinide targets. This has the advantage of reducing: the energy deposited by the proton beam in the target; contamination from neutron-deficient isobars that would be produced by spallation; and mechanical stress on the target. At ISOLDE CERN [E. Kugler, Hyperfine Interact. 129 (2000) 23], tests have been made on standard ISOLDE actinide targets using fast-neutron bunches produced by bombarding thick, high- Z metal converters with 1 and 1.4 GeV proton pulses. This paper reviews the first applications of converters used at ISOLDE. It highlights the different geometries and the techniques used to compare fission yields produced by the proton beam directly on the target with neutron-induced fission. Results from the six targets already tested, namely UC 2/graphite and ThO 2 targets with tungsten and tantalum converters, are presented. To gain further knowledge for the design of a dedicated target as required by the TARGISOL project [EU-RTD Project TARGISOL (HPRI-CT-2001-50033)], the results are compared to simulations, using the MARS [N.V. Mokhov, S.I. Striganov, A. Van Ginneken, S.G. Mashnik, A.J. Sierk, J. Ranft, MARS code developments, in: 4th Workshop on Simulating Accelerator Radiation Environments, SARE-4, Knoxville, USA, 14-15.9.1998, FERMILAB-PUB-98-379, nucl-th/9812038; N.V. Mokhov, The Mars Code System User's Guide, Fermilab-FN-628, 1995; N.V. Mokhov, MARS Code Developments, Benchmarking and Applications, Fermilab-Conf-00-066, 2000; O.E. Krivosheev, N.V. Mokhov, A New MARS and its Applications, Fermilab-Conf-98/43, 1998] code interfaced with MCNP [J.S. Hendrics, MCNP4C LANL Memo X-5; JSH-2000-3; J.F. Briemesteir (Ed.), MCNP - A General Montecarlo N-Particle Transport Code, Version 4C, LA-13709-M] libraries, of the neutron flux from the converters interacting with the actinide targets.

Radioactive ion beams produced by neutron-induced fission at ISOLDE

NASA Astrophysics Data System (ADS)

Isolde Collaboration; Catherall, R.; Lettry, J.; Gilardoni, S.; Köster, U.

2003-05-01

The production rates of neutron-rich fission products for the next-generation radioactive beam facility EURISOL [EU-RTD Project EURISOL (HPRI-CT-1999-50001)] are mainly limited by the maximum amount of power deposited by protons in the target. An alternative approach is to use neutron beams to induce fission in actinide targets. This has the advantage of reducing: the energy deposited by the proton beam in the target; contamination from neutron-deficient isobars that would be produced by spallation; and mechanical stress on the target. At ISOLDE CERN [E. Kugler, Hyperfine Interact. 129 (2000) 23], tests have been made on standard ISOLDE actinide targets using fast-neutron bunches produced by bombarding thick, high-/Z metal converters with 1 and 1.4 GeV proton pulses. This paper reviews the first applications of converters used at ISOLDE. It highlights the different geometries and the techniques used to compare fission yields produced by the proton beam directly on the target with neutron-induced fission. Results from the six targets already tested, namely UC2/graphite and ThO2 targets with tungsten and tantalum converters, are presented. To gain further knowledge for the design of a dedicated target as required by the TARGISOL project [EU-RTD Project TARGISOL (HPRI-CT-2001-50033)], the results are compared to simulations, using the MARS [N.V. Mokhov, S.I. Striganov, A. Van Ginneken, S.G. Mashnik, A.J. Sierk, J. Ranft, MARS code developments, in: 4th Workshop on Simulating Accelerator Radiation Environments, SARE-4, Knoxville, USA, 14-15.9.1998, FERMILAB-PUB-98-379, nucl-th/9812038; N.V. Mokhov, The Mars Code System User's Guide, Fermilab-FN-628, 1995; N.V. Mokhov, MARS Code Developments, Benchmarking and Applications, Fermilab-Conf-00-066, 2000; O.E. Krivosheev, N.V. Mokhov, A New MARS and its Applications, Fermilab-Conf-98/43, 1998] code interfaced with MCNP [J.S. Hendrics, MCNP4C LANL Memo X-5; JSH-2000-3; J.F. Briemesteir (Ed.), MCNP - A General Montecarlo N-Particle Transport Code, Version 4C, LA-13709-M] libraries, of the neutron flux from the converters interacting with the actinide targets.

22 CFR 1101.4 - Reports on new systems of records; computer matching programs.

Code of Federal Regulations, 2012 CFR

2012-04-01

... 22 Foreign Relations 2 2012-04-01 2009-04-01 true Reports on new systems of records; computer matching programs. 1101.4 Section 1101.4 Foreign Relations INTERNATIONAL BOUNDARY AND WATER COMMISSION... records; computer matching programs. (a) Before establishing any new systems of records, or making any...

22 CFR 1101.4 - Reports on new systems of records; computer matching programs.

Code of Federal Regulations, 2014 CFR

2014-04-01

... 22 Foreign Relations 2 2014-04-01 2014-04-01 false Reports on new systems of records; computer matching programs. 1101.4 Section 1101.4 Foreign Relations INTERNATIONAL BOUNDARY AND WATER COMMISSION... records; computer matching programs. (a) Before establishing any new systems of records, or making any...

22 CFR 1101.4 - Reports on new systems of records; computer matching programs.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 22 Foreign Relations 2 2013-04-01 2009-04-01 true Reports on new systems of records; computer matching programs. 1101.4 Section 1101.4 Foreign Relations INTERNATIONAL BOUNDARY AND WATER COMMISSION... records; computer matching programs. (a) Before establishing any new systems of records, or making any...

22 CFR 1101.4 - Reports on new systems of records; computer matching programs.

Code of Federal Regulations, 2011 CFR

2011-04-01

... 22 Foreign Relations 2 2011-04-01 2009-04-01 true Reports on new systems of records; computer matching programs. 1101.4 Section 1101.4 Foreign Relations INTERNATIONAL BOUNDARY AND WATER COMMISSION... records; computer matching programs. (a) Before establishing any new systems of records, or making any...

Source terms, shielding calculations and soil activation for a medical cyclotron.

PubMed

Konheiser, J; Naumann, B; Ferrari, A; Brachem, C; Müller, S E

2016-12-01

Calculations of the shielding and estimates of soil activation for a medical cyclotron are presented in this work. Based on the neutron source term from the 18 O(p,n) 18 F reaction produced by a 28 MeV proton beam, neutron and gamma dose rates outside the building were estimated with the Monte Carlo code MCNP6 (Goorley et al 2012 Nucl. Technol. 180 298-315). The neutron source term was calculated with the MCNP6 code and FLUKA (Ferrari et al 2005 INFN/TC_05/11, SLAC-R-773) code as well as with supplied data by the manufacturer. MCNP and FLUKA calculations yielded comparable results, while the neutron yield obtained using the manufacturer-supplied information is about a factor of 5 smaller. The difference is attributed to the missing channels in the manufacturer-supplied neutron source terms which considers only the 18 O(p,n) 18 F reaction, whereas the MCNP and FLUKA calculations include additional neutron reaction channels. Soil activation was performed using the FLUKA code. The estimated dose rate based on MCNP6 calculations in the public area is about 0.035 µSv h -1 and thus significantly below the reference value of 0.5 µSv h -1 (2011 Strahlenschutzverordnung, 9 Auflage vom 01.11.2011, Bundesanzeiger Verlag). After 5 years of continuous beam operation and a subsequent decay time of 30 d, the activity concentration of the soil is about 0.34 Bq g -1 .

Comparison of Three Methods of Calculation, Experimental and Monte Carlo Simulation in Investigation of Organ Doses (Thyroid, Sternum, Cervical Vertebra) in Radioiodine Therapy

PubMed Central

Shahbazi-Gahrouei, Daryoush; Ayat, Saba

2012-01-01

Radioiodine therapy is an effective method for treating thyroid cancer carcinoma, but it has some affects on normal tissues, hence dosimetry of vital organs is important to weigh the risks and benefits of this method. The aim of this study is to measure the absorbed doses of important organs by Monte Carlo N Particle (MCNP) simulation and comparing the results of different methods of dosimetry by performing a t-paired test. To calculate the absorbed dose of thyroid, sternum, and cervical vertebra using the MCNP code, *F8 tally was used. Organs were simulated by using a neck phantom and Medical Internal Radiation Dosimetry (MIRD) method. Finally, the results of MCNP, MIRD, and Thermoluminescent dosimeter (TLD) measurements were compared by SPSS software. The absorbed dose obtained by Monte Carlo simulations for 100, 150, and 175 mCi administered 131I was found to be 388.0, 427.9, and 444.8 cGy for thyroid, 208.7, 230.1, and 239.3 cGy for sternum and 272.1, 299.9, and 312.1 cGy for cervical vertebra. The results of paired t-test were 0.24 for comparing TLD dosimetry and MIRD calculation, 0.80 for MCNP simulation and MIRD, and 0.19 for TLD and MCNP. The results showed no significant differences among three methods of Monte Carlo simulations, MIRD calculation and direct experimental dosimetry using TLD. PMID:23717806

Evolution of a standard microprocessor-based space computer

NASA Technical Reports Server (NTRS)

Fernandez, M.

1980-01-01

An existing in inventory computer hardware/software package (B-1 RFS/ECM) was repackaged and applied to multiple missile/space programs. Concurrent with the application efforts, low risk modifications were made to the computer from program to program to take advantage of newer, advanced technology and to meet increasingly more demanding requirements (computational and memory capabilities, longer life, and fault tolerant autonomy). It is concluded that microprocessors hold promise in a number of critical areas for future space computer applications. However, the benefits of the DoD VHSIC Program are required and the old proliferation problem must be revised.

Rotorcraft Flight Simulation Computer Program C81 with Datamap Interface, Volume 2. Programmer’s Manual

DTIC Science & Technology

1981-10-01

39 :CIRK M p VPJoJ 29 4 31 j ii 33bAqRKI H4NLSP Ř A 8. 86 08 I N 1*Li I0 4t 102 103 to4 BATV Lo AoT 142 SN 6 kNkM TV UQIIRTV I 16 " COMT kL A1I N t 3...TENNIS RACKET MOMENT EXPRESS ION OANDOIT AUl THETA DOT FROM CYCLIC OANOOI T AERCO DRAG RISE COEFFICIENT (DEFAULT - 1.9/RAD) CC AGUST (13) GUST

Brachytherapy dosimetry of 125I and 103Pd sources using an updated cross section library for the MCNP Monte Carlo transport code.

PubMed

Bohm, Tim D; DeLuca, Paul M; DeWerd, Larry A

2003-04-01

Permanent implantation of low energy (20-40 keV) photon emitting radioactive seeds to treat prostate cancer is an important treatment option for patients. In order to produce accurate implant brachytherapy treatment plans, the dosimetry of a single source must be well characterized. Monte Carlo based transport calculations can be used for source characterization, but must have up to date cross section libraries to produce accurate dosimetry results. This work benchmarks the MCNP code and its photon cross section library for low energy photon brachytherapy applications. In particular, we calculate the emitted photon spectrum, air kerma, depth dose in water, and radial dose function for both 125I and 103Pd based seeds and compare to other published results. Our results show that MCNP's cross section library differs from recent data primarily in the photoelectric cross section for low energies and low atomic number materials. In water, differences as large as 10% in the photoelectric cross section and 6% in the total cross section occur at 125I and 103Pd photon energies. This leads to differences in the dose rate constant of 3% and 5%, and differences as large as 18% and 20% in the radial dose function for the 125I and 103Pd based seeds, respectively. Using a partially updated photon library, calculations of the dose rate constant and radial dose function agree with other published results. Further, the use of the updated photon library allows us to verify air kerma and depth dose in water calculations performed using MCNP's perturbation feature to simulate updated cross sections. We conclude that in order to most effectively use MCNP for low energy photon brachytherapy applications, we must update its cross section library. Following this update, the MCNP code system will be a very effective tool for low energy photon brachytherapy dosimetry applications.

Ubiquitous occurrence of chlorinated byproducts of bisphenol A and nonylphenol in bleached food contacting papers and their implications for human exposure.

PubMed

Zhou, Yuyin; Chen, Mo; Zhao, Fanrong; Mu, Di; Zhang, Zhaobin; Hu, Jianying

2015-06-16

The occurrence of bisphenol A (BPA), nonylphenol (NP), and their six chlorinated byproducts were investigated in 74 food contacting papers (FCPs) from China, the U.S.A., Japan, and Europe using a sensitive dansylation LC-MS/MS method. BPA (

Numerical nonlinear inelastic analysis of stiffened shells of revolution. Volume 4: Satellite-1P program for STARS-2P digital computer program

NASA Technical Reports Server (NTRS)

Svalbonas, V.; Ogilvie, P.

1975-01-01

A special data debugging package called SAT-1P created for the STARS-2P computer program is described. The program was written exclusively in FORTRAN 4 for the IBM 370-165 computer, and then converted to the UNIVAC 1108.

FY2012 summary of tasks completed on PROTEUS-thermal work.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, C.H.; Smith, M.A.

2012-06-06

PROTEUS is a suite of the neutronics codes, both old and new, that can be used within the SHARP codes being developed under the NEAMS program. Discussion here is focused on updates and verification and validation activities of the SHARP neutronics code, DeCART, for application to thermal reactor analysis. As part of the development of SHARP tools, the different versions of the DeCART code created for PWR, BWR, and VHTR analysis were integrated. Verification and validation tests for the integrated version were started, and the generation of cross section libraries based on the subgroup method was revisited for the targetedmore » reactor types. The DeCART code has been reorganized in preparation for an efficient integration of the different versions for PWR, BWR, and VHTR analysis. In DeCART, the old-fashioned common blocks and header files have been replaced by advanced memory structures. However, the changing of variable names was minimized in order to limit problems with the code integration. Since the remaining stability problems of DeCART were mostly caused by the CMFD methodology and modules, significant work was performed to determine whether they could be replaced by more stable methods and routines. The cross section library is a key element to obtain accurate solutions. Thus, the procedure for generating cross section libraries was revisited to provide libraries tailored for the targeted reactor types. To improve accuracy in the cross section library, an attempt was made to replace the CENTRM code by the MCNP Monte Carlo code as a tool obtaining reference resonance integrals. The use of the Monte Carlo code allows us to minimize problems or approximations that CENTRM introduces since the accuracy of the subgroup data is limited by that of the reference solutions. The use of MCNP requires an additional set of libraries without resonance cross sections so that reference calculations can be performed for a unit cell in which only one isotope of interest includes resonance cross sections, among the isotopes in the composition. The OECD MHTGR-350 benchmark core was simulated using DeCART as initial focus of the verification/validation efforts. Among the benchmark problems, Exercise 1 of Phase 1 is a steady-state benchmark case for the neutronics calculation for which block-wise cross sections were provided in 26 energy groups. This type of problem was designed for a homogenized geometry solver like DIF3D rather than the high-fidelity code DeCART. Instead of the homogenized block cross sections given in the benchmark, the VHTR-specific 238-group ENDF/B-VII.0 library of DeCART was directly used for preliminary calculations. Initial results showed that the multiplication factors of a fuel pin and a fuel block with or without a control rod hole were off by 6, -362, and -183 pcm Dk from comparable MCNP solutions, respectively. The 2-D and 3-D one-third core calculations were also conducted for the all-rods-out (ARO) and all-rods-in (ARI) configurations, producing reasonable results. Figure 1 illustrates the intermediate (1.5 eV - 17 keV) and thermal (below 1.5 eV) group flux distributions. As seen from VHTR cores with annular fuels, the intermediate group fluxes are relatively high in the fuel region, but the thermal group fluxes are higher in the inner and outer graphite reflector regions than in the fuel region. To support the current project, a new three-year I-NERI collaboration involving ANL and KAERI was started in November 2011, focused on performing in-depth verification and validation of high-fidelity multi-physics simulation codes for LWR and VHTR. The work scope includes generating improved cross section libraries for the targeted reactor types, developing benchmark models for verification and validation of the neutronics code with or without thermo-fluid feedback, and performing detailed comparisons of predicted reactor parameters against both Monte Carlo solutions and experimental measurements. The following list summarizes the work conducted so far for PROTEUS-Thermal Tasks: Unification of different versions of DeCART was initiated, and at the same time code modernization was conducted to make code unification efficient; (2) Regeneration of cross section libraries was attempted for the targeted reactor types, and the procedure for generating cross section libraries was updated by replacing CENTRM with MCNP for reference resonance integrals; (3) The MHTGR-350 benchmark core was simulated using DeCART with VHTR-specific 238-group ENDF/B-VII.0 library, and MCNP calculations were performed for comparison; and (4) Benchmark problems for PWR and BWR analysis were prepared for the DeCART verification/validation effort. In the coming months, the work listed above will be completed. Cross section libraries will be generated with optimized group structures for specific reactor types.« less

Human Adaptation to the Computer.

DTIC Science & Technology

1986-09-01

advance in man’s civilization [Ref. 3:p. 4]. This advance has caused dramatic changes in man’s way of life. It has caused several emotions in man that...their program run or com- pleting their work project can create within them a loss of human feeling and emotion . There have been a lot of jokes made...aggression 44 3. violence 4. antisocial activity 5. grievances 6. work slowdowns/missed deadlines 7. strikes B. absenteeism 9. tardiness 10. turnover

Passive TCP Reconstruction and Forensic Analysis with tcpflow

DTIC Science & Technology

2013-09-01

Detection Evaluation [17]. 6 Future Work The original tcpflow was developed at a time when computers had relatively small memories and, as a result...4. TITLE AND SUBTITLE 5a. CONTRACT NUMBER 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 5d. PROJECT NUMBER 5e. TASK NUMBER 5f. WORK UNIT NUMBER 6...1 2 Related Work . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3 Requirements

Modification and benchmarking of MCNP for low-energy tungsten spectra.

PubMed

Mercier, J R; Kopp, D T; McDavid, W D; Dove, S B; Lancaster, J L; Tucker, D M

2000-12-01

The MCNP Monte Carlo radiation transport code was modified for diagnostic medical physics applications. In particular, the modified code was thoroughly benchmarked for the production of polychromatic tungsten x-ray spectra in the 30-150 kV range. Validating the modified code for coupled electron-photon transport with benchmark spectra was supplemented with independent electron-only and photon-only transport benchmarks. Major revisions to the code included the proper treatment of characteristic K x-ray production and scoring, new impact ionization cross sections, and new bremsstrahlung cross sections. Minor revisions included updated photon cross sections, electron-electron bremsstrahlung production, and K x-ray yield. The modified MCNP code is benchmarked to electron backscatter factors, x-ray spectra production, and primary and scatter photon transport.

Monte Carol-Based Dosimetry of Beta-Emitters for Intravascular Brachytherapy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Choi, C.K.

2002-06-25

Monte Carlo simulations for radiation dosimetry and the experimental verifications of the simulations have been developed for the treatment geometry of intravascular brachytherapy, a form of radionuclide therapy for occluded coronary disease (restenosis). Monte Carlo code, MCNP4C, has been used to calculate the radiation dose from the encapsulated array of B-emitting seeds (Sr/Y-source train). Solid water phantoms have been fabricated to measure the dose on the radiochromic films that were exposed to the beta source train for both linear and curved coronary vessel geometries. While the dose difference for the 5-degree curved vessel at the prescription point of f+2.0 mmmore » is within the 10% guideline set by the AAPM, however, the difference increased dramatically to 16.85% for the 10-degree case which requires additional adjustment for the acceptable dosimetry planning. The experimental dose measurements agree well with the simulation results« less

Reduced Variance using ADVANTG in Monte Carlo Calculations of Dose Coefficients to Stylized Phantoms

NASA Astrophysics Data System (ADS)

Hiller, Mauritius; Bellamy, Michael; Eckerman, Keith; Hertel, Nolan

2017-09-01

The estimation of dose coefficients of external radiation sources to the organs in phantoms becomes increasingly difficult for lower photon source energies. This study focus on the estimation of photon emitters around the phantom. The computer time needed to calculate a result within a certain precision can be lowered by several orders of magnitude using ADVANTG compared to a standard run. Using ADVANTG which employs the DENOVO adjoint calculation package enables the user to create a fully populated set of weight windows and source biasing instructions for an MCNP calculation.

CPsuperH2.0: An improved computational tool for Higgs phenomenology in the MSSM with explicit CP violation

NASA Astrophysics Data System (ADS)

Lee, J. S.; Carena, M.; Ellis, J.; Pilaftsis, A.; Wagner, C. E. M.

2009-02-01

We describe the Fortran code CPsuperH2.0, which contains several improvements and extensions of its predecessor CPsuperH. It implements improved calculations of the Higgs-boson pole masses, notably a full treatment of the 4×4 neutral Higgs propagator matrix including the Goldstone boson and a more complete treatment of threshold effects in self-energies and Yukawa couplings, improved treatments of two-body Higgs decays, some important three-body decays, and two-loop Higgs-mediated contributions to electric dipole moments. CPsuperH2.0 also implements an integrated treatment of several B-meson observables, including the branching ratios of B→μμ, B→ττ, B→τν, B→Xγ and the latter's CP-violating asymmetry A, and the supersymmetric contributions to the Bs,d0-B¯s,d0 mass differences. These additions make CPsuperH2.0 an attractive integrated tool for analyzing supersymmetric CP and flavour physics as well as searches for new physics at high-energy colliders such as the Tevatron, LHC and linear colliders. Program summaryProgram title: CPsuperH2.0 Catalogue identifier: ADSR_v2_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/ADSR_v2_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 13 290 No. of bytes in distributed program, including test data, etc.: 89 540 Distribution format: tar.gz Programming language: Fortran 77 Computer: PC running under Linux and computers in Unix environment Operating system: Linux RAM: 32 Mbytes Classification: 11.1 Catalogue identifier of the previous version: ADSR_v1_0 Journal reference of the previous version: CPC 156 (2004) 283 Does the new version supersede the previous version?: Yes Nature of problem: The calculations of mass spectrum, decay widths and branching ratios of the neutral and charged Higgs bosons in the Minimal Supersymmetric Standard Model with explicit CP violation have been improved. The program is based on recent renormalization-group-improved diagrammatic calculations that include dominant higher-order logarithmic and threshold corrections, b-quark Yukawa-coupling resummation effects and improved treatment of Higgs-boson pole-mass shifts. The couplings of the Higgs bosons to the Standard Model gauge bosons and fermions, to their supersymmetric partners and all the trilinear and quartic Higgs-boson self-couplings are also calculated. The new implementations include a full treatment of the 4×4(2×2) neutral (charged) Higgs propagator matrix together with the center-of-mass dependent Higgs-boson couplings to gluons and photons, two-loop Higgs-mediated contributions to electric dipole moments, and an integrated treatment of several B-meson observables. Solution method: One-dimensional numerical integration for several Higgs-decay modes, iterative treatment of the threshold corrections and Higgs-boson pole masses, and the numerical diagonalization of the neutralino mass matrix. Reasons for new version: Mainly to provide a coherent numerical framework which calculates consistently observables for both low- and high-energy experiments. Summary of revisions: Improved treatment of Higgs-boson masses and propagators. Improved treatment of Higgs-boson couplings and decays. Higgs-mediated two-loop electric dipole moments. B-meson observables. Running time: Less than 0.1 seconds. The program may be obtained from http://www.hep.man.ac.uk/u/jslee/CPsuperH.html.

SU-F-T-111: Investigation of the Attila Deterministic Solver as a Supplement to Monte Carlo for Calculating Out-Of-Field Radiotherapy Dose

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mille, M; Lee, C; Failla, G

Purpose: To use the Attila deterministic solver as a supplement to Monte Carlo for calculating out-of-field organ dose in support of epidemiological studies looking at the risks of second cancers. Supplemental dosimetry tools are needed to speed up dose calculations for studies involving large-scale patient cohorts. Methods: Attila is a multi-group discrete ordinates code which can solve the 3D photon-electron coupled linear Boltzmann radiation transport equation on a finite-element mesh. Dose is computed by multiplying the calculated particle flux in each mesh element by a medium-specific energy deposition cross-section. The out-of-field dosimetry capability of Attila is investigated by comparing averagemore » organ dose to that which is calculated by Monte Carlo simulation. The test scenario consists of a 6 MV external beam treatment of a female patient with a tumor in the left breast. The patient is simulated by a whole-body adult reference female computational phantom. Monte Carlo simulations were performed using MCNP6 and XVMC. Attila can export a tetrahedral mesh for MCNP6, allowing for a direct comparison between the two codes. The Attila and Monte Carlo methods were also compared in terms of calculation speed and complexity of simulation setup. A key perquisite for this work was the modeling of a Varian Clinac 2100 linear accelerator. Results: The solid mesh of the torso part of the adult female phantom for the Attila calculation was prepared using the CAD software SpaceClaim. Preliminary calculations suggest that Attila is a user-friendly software which shows great promise for our intended application. Computational performance is related to the number of tetrahedral elements included in the Attila calculation. Conclusion: Attila is being explored as a supplement to the conventional Monte Carlo radiation transport approach for performing retrospective patient dosimetry. The goal is for the dosimetry to be sufficiently accurate for use in retrospective epidemiological investigations.« less

Corps Helicopter Attack Planning System (CHAPS). Positional Handbook. Appendix A. Messages. Appendix B. Statespace Construction Sample Session

DTIC Science & Technology

1989-10-01

REVIEW MENU PROGRAM (S) CHAPS PURPOSE AND OVERVIEV The Do Review menu allows the user to select which missions to perform detailed analysis on and...input files must be resident on the computer you are running SUPR on. Any interface or file transfer programs must be successfully executed prior to... COMPUTER PROGRAM WAS DEVELOPED BY SYSTEMS CONTROL TECHNOLOGY FOR THE DEPUTY CHIEF OF STAFF/OPERATIONS,HQ USAFE. THE USE OF THE COMPUTER PROGRAM IS

2012 SARA Students Technical Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Briccetti, Angelo; Lorei, Nathan; Yonkings, David

The Service Academy Research Associates (SARA) program provides an opportunity for Midshipmen and Cadets from US Service Academies to participate in research at Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL), and Sandia National Laboratory for several weeks during the summer as part of their summer training assignments. During the summer of 2012, three Midshipmen were assigned to work with the XCP Division at LANL for approximately 5-6 weeks. As one of the nation's top national security science laboratories, LANL stretches across 36 square miles, has over 2,100 facilities, and employs over 9,000 individuals including a significant numbermore » of students and postdocs. LANL's mission is to 'apply science and technology to: ensure the safety, security, and reliability of the US nuclear deterrent, reduce global threats, and solve other emerging national security challenges.' While LANL officially operates under the US Department of Energy (DoE), fulfilling this mission requires mutual cooperation with the US Department of Defense (DoD) as well. LANL's high concentration of knowledge and experience provides interns a chance to perform research in many disciplines, and its connection with the DoD in both operation and personnel gives SARA students insight to career possibilities both during and after military service. SARA students have plenty of opportunity to enjoy hiking, camping, the Los Alamos YMCA, and many other outdoor activities in New Mexico while staying at the Buffalo Thunder Resort, located 20 miles east of the lab. XCP Division is the Computational Physics division of LANL's Weapons Department. Working with XCP Division requires individuals to be Q cleared by the DoE. This means it is significantly more convenient for SARA students to be assigned to XCP Division than their civilian counterparts as the DoD CNWDI clearance held by SARA students is easily transferred to the lab prior to the students arriving at the start of the summer. SARA students working with XCP Division were given a comprehensive introduction into nuclear engineering and physics, nuclear weapons, and radiation transport and detection via texts and lectures at various classification levels. Students also attended tours of several prominent facilities at LANL including TA-41 Ice House, TA-55 PF-4 plutonium facility, the Nicholas C. Metropolis Center for Modeling and Simulation, also known as the Secure Computing Center (SCC), and the Dual-Axis Radiological Hydro Test (DARHT) facility; in addition, SARA students accompanied by LANL staff traveled to Minot AFB in North Dakota for tours of the 5th Bomb Wing and 91st Missile Wing facilities. Students participated in a week long class on the Monte Carlo N Particle (MCNP) code to supplement their understanding of radiation transport simulations. SARA students were then tasked with using this knowledge to model radiation detectors and use MCNP to compare their models to experimental data and previously accepted models.« less

3D element imaging using NSECT for the detection of renal cancer: a simulation study in MCNP.

PubMed

Viana, R S; Agasthya, G A; Yoriyaz, H; Kapadia, A J

2013-09-07

This work describes a simulation study investigating the application of neutron stimulated emission computed tomography (NSECT) for noninvasive 3D imaging of renal cancer in vivo. Using MCNP5 simulations, we describe a method of diagnosing renal cancer in the body by mapping the 3D distribution of elements present in tumors using the NSECT technique. A human phantom containing the kidneys and other major organs was modeled in MCNP5. The element composition of each organ was based on values reported in literature. The two kidneys were modeled to contain elements reported in renal cell carcinoma (RCC) and healthy kidney tissue. Simulated NSECT scans were executed to determine the 3D element distribution of the phantom body. Elements specific to RCC and healthy kidney tissue were then analyzed to identify the locations of the diseased and healthy kidneys and generate tomographic images of the tumor. The extent of the RCC lesion inside the kidney was determined using 3D volume rendering. A similar procedure was used to generate images of each individual organ in the body. Six isotopes were studied in this work - (32)S, (12)C, (23)Na, (14)N, (31)P and (39)K. The results demonstrated that through a single NSECT scan performed in vivo, it is possible to identify the location of the kidneys and other organs within the body, determine the extent of the tumor within the organ, and to quantify the differences between cancer and healthy tissue-related isotopes with p ≤ 0.05. All of the images demonstrated appropriate concentration changes between the organs, with some discrepancy observed in (31)P, (39)K and (23)Na. The discrepancies were likely due to the low concentration of the elements in the tissue that were below the current detection sensitivity of the NSECT technique.

Treating electron transport in MCNP{sup trademark}

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hughes, H.G.

1996-12-31

The transport of electrons and other charged particles is fundamentally different from that of neutrons and photons. A neutron, in aluminum slowing down from 0.5 MeV to 0.0625 MeV will have about 30 collisions; a photon will have fewer than ten. An electron with the same energy loss will undergo 10{sup 5} individual interactions. This great increase in computational complexity makes a single- collision Monte Carlo approach to electron transport unfeasible for many situations of practical interest. Considerable theoretical work has been done to develop a variety of analytic and semi-analytic multiple-scattering theories for the transport of charged particles. Themore » theories used in the algorithms in MCNP are the Goudsmit-Saunderson theory for angular deflections, the Landau an theory of energy-loss fluctuations, and the Blunck-Leisegang enhancements of the Landau theory. In order to follow an electron through a significant energy loss, it is necessary to break the electron`s path into many steps. These steps are chosen to be long enough to encompass many collisions (so that multiple-scattering theories are valid) but short enough that the mean energy loss in any one step is small (for the approximations in the multiple-scattering theories). The energy loss and angular deflection of the electron during each step can then be sampled from probability distributions based on the appropriate multiple- scattering theories. This subsumption of the effects of many individual collisions into single steps that are sampled probabilistically constitutes the ``condensed history`` Monte Carlo method. This method is exemplified in the ETRAN series of electron/photon transport codes. The ETRAN codes are also the basis for the Integrated TIGER Series, a system of general-purpose, application-oriented electron/photon transport codes. The electron physics in MCNP is similar to that of the Integrated TIGER Series.« less

BCVEGPY2.0: An upgraded version of the generator BCVEGPY with the addition of hadroproduction of the P-wave B states

NASA Astrophysics Data System (ADS)

Chang, Chao-Hsi; Wang, Jian-Xiong; Wu, Xing-Gang

2006-02-01

The generator BCVEGPY is upgraded by improving some of its features and by adding the hadroproduction of the P-wave excited B states (denoted by BcJ,L=1∗ or by hB_c and χB_c). In order to make the generator more efficient, we manipulate the amplitude as compact as possible with special effort. The correctness of the program is tested by various checks. We denote it as BCVEGPY2.0. As for the added part of the P-wave production, only the dominant gluon-gluon fusion mechanism ( gg→BcJ,L=1∗+c¯+b) is taken into account. Moreover, in the program, not only the ability to compute the contributions from the color-singlet components ( to the P-wave production but also the ability to compute the contributions from the color-octet components ( are available. With BCVEGPY2.0 the contributions from the two 'color components' to the production of each of the P-wave states may be computed separately by an option, furthermore, besides individually the event samples of the S-wave and P-wave ( cb¯)-heavy-quarkonium in various correct (realistic) mixtures can be generated by relevant options too. Program summaryTitle of program: BCVEGPY Version: 2.0 (December, 2004) Catalogue identifier: ADWQ Program summary URL:http://cpc.cs.qub.ac.uk/summaries/ADWQ Program obtained from: CPC Program Library, Queen's University of Belfast, N. Ireland Reference to original program: ADTJ (BCVEGPY1.0) Reference in CPC: Comput. Phys. Comm. 159 (2004) 192 Does the new version supersede the old program: yes Computer: Any computer with FORTRAN 77 or 90 compiler. The program has been tested on HP-SC45 Sigma-X parallel computer, Linux PCs and Windows PCs with Visual Fortran Operating systems: UNIX, Linux and Windows Programming language used: FORTRAN 77/90 Memory required to execute with typical data: About 2.0 MB No. of lines in distributed program, including test data, etc.: 124 297 No. of bytes in distributed program, including test data, etc.: 1 137 177 Distribution format: tar.g2 Nature of physical problem: Hadronic production of B meson itself and its excited states. Method of solution: The code with option can generate weighted and unweighted events. For jet hadronization, an interface to PYTHIA is provided. Reason for the new version: There are two reasons. One is to provide additional codes for the hadronic production of P-wave excited B states: the four via color-singlet P-wave state directly and the two via color-octet S-wave state accordingly. The other one is to decompose the color-flow factor for the amplitude by an approximate way, that is adopted in PYTHIA. Summary of Revisions: (1) The integration efficiency over the momentum fractions of the initial partons x and x are improved; (2) The amplitudes for the hadronic production of the color-singlet components corresponding to the four P-wave states, BcJ,L=1∗ or P1 and P3 ( J=0,1,2), are included; (3) The amplitudes for P-wave production via the two color-octet components |((S1)g> and |((S3)g> are included; (4) For comparison, the S-wave ( S1 and S3) hadronic production via the light quark-antiquark annihilation mechanism is also included; (5) For convenience, 24 data files to record the information of the generated events in one run are added; (6) An additional file, parameter.for, is added to set the initial values of the parameters; (7) Two new parameters 'IMIX' (IMIX = 0 or 1) and 'IMIXTYPE' (IMIXTYPE = 1, = 2 or = 3) are added to meet the needs of generating the events for simulating 'mixing' or 'separate' event samples for various B and its excited states correctly; (8) One switch, 'IVEGGRADE', is added to determine whether to use the existed importance sampling function to generate a more precise importance sampling function or not; (9) Two parameters, 'IOUTPDF' and 'IPDFNUM', are added to determine which type of PDFs to use; (10) The color-flow decomposition for the amplitudes is rewritten by an approximate way, that is adopted in PYTHIA. Restrictions on the complexity of the problem: The hadronic production of (cb¯)-quarkonium in S-wave and P-wave states via the mechanism of gluon-gluon fusion are given by the 'complete calculation' approach of the leading order QCD. The contributions from the other mechanisms for P-wave production which are small comparatively are not included. Typical running time: Generally speaking, it depends on which option is used to drive PYTHIA when generating the B events. Typically, for the hadronic production of the S-wave (cb¯)-quarkonium, if the PYTHIA parameter IDWTUP = 1, then it takes about 20 hours on a 1.8 GHz Intel P4-processor machine to generate 1000 events; however if IDWTUP = 3, to generate 10 6 events, it takes only about 40 minutes. For the hadronic production of the P-wave (cb¯)-quarkonium, the necessary time will be almost two times longer than the S-wave quarkonium production.

Neutronic Calculation Analysis for CN HCCB TBM-Set

NASA Astrophysics Data System (ADS)

Cao, Qixiang; Zhao, Fengchao; Zhao, Zhou; Wu, Xinghua; Li, Zaixin; Wang, Xiaoyu; Feng, Kaiming

2015-07-01

Using the Monte Carlo transport code MCNP, neutronic calculation analysis for China helium cooled ceramic breeder test blanket module (CN HCCB TBM) and the associated shield block (together called TBM-set) has been carried out based on the latest design of HCCB TBM-set and C-lite model. Key nuclear responses of HCCB TBM-set, such as the neutron flux, tritium production rate, nuclear heating and radiation damage, have been obtained and discussed. These nuclear performance data can be used as the basic input data for other analyses of HCCB TBM-set, such as thermal-hydraulics, thermal-mechanics and safety analysis. supported by the Major State Basic Research Development Program of China (973 Program) (No. 2013GB108000)

The effect of tandem-ovoid titanium applicator on points A, B, bladder, and rectum doses in gynecological brachytherapy using 192Ir

PubMed Central

Sadeghi, Mohammad Hosein; Mehdizadeh, Amir; Faghihi, Reza; Moharramzadeh, Vahed; Meigooni, Ali Soleimani

2018-01-01

Purpose The dosimetry procedure by simple superposition accounts only for the self-shielding of the source and does not take into account the attenuation of photons by the applicators. The purpose of this investigation is an estimation of the effects of the tandem and ovoid applicator on dose distribution inside the phantom by MCNP5 Monte Carlo simulations. Material and methods In this study, the superposition method is used for obtaining the dose distribution in the phantom without using the applicator for a typical gynecological brachytherapy (superposition-1). Then, the sources are simulated inside the tandem and ovoid applicator to identify the effect of applicator attenuation (superposition-2), and the dose at points A, B, bladder, and rectum were compared with the results of superposition. The exact dwell positions, times of the source, and positions of the dosimetry points were determined in images of a patient and treatment data of an adult woman patient from a cancer center. The MCNP5 Monte Carlo (MC) code was used for simulation of the phantoms, applicators, and the sources. Results The results of this study showed no significant differences between the results of superposition method and the MC simulations for different dosimetry points. The difference in all important dosimetry points was found to be less than 5%. Conclusions According to the results, applicator attenuation has no significant effect on the calculated points dose, the superposition method, adding the dose of each source obtained by the MC simulation, can estimate the dose to points A, B, bladder, and rectum with good accuracy. PMID:29619061

A Review of Australian Investigations on Aeronautical Fatigue during the Period April 1979 to March 1981.

DTIC Science & Technology

1981-03-01

RD73 9. COST CODE: b. Sponsoring Agency: 27003 SUPPLY 50/2 10. IMPRINT: 11. COMPUTER PROGRAM(S) Aeronautical Research (Title(s) and language(s...laminates. 9/24 An advanced iso -parametric element is also being Jeveloped specifically for the analysis of disbonds and internal flaws in composite...FAILURE - STATION 119 iso I f FIG. 9.3 NOMAD STRLFCI URAl I AlT 10(L TESI FIG. 9.4 FAILED NOMAD STRUT UPPER END FITTING FIG. 9.5 FRACTURE FACES OF FAILED

Doppler Temperature Coefficient Calculations Using Adjoint-Weighted Tallies and Continuous Energy Cross Sections in MCNP6

NASA Astrophysics Data System (ADS)

Gonzales, Matthew Alejandro

The calculation of the thermal neutron Doppler temperature reactivity feedback co-efficient, a key parameter in the design and safe operation of advanced reactors, using first order perturbation theory in continuous energy Monte Carlo codes is challenging as the continuous energy adjoint flux is not readily available. Traditional approaches of obtaining the adjoint flux attempt to invert the random walk process as well as require data corresponding to all temperatures and their respective temperature derivatives within the system in order to accurately calculate the Doppler temperature feedback. A new method has been developed using adjoint-weighted tallies and On-The-Fly (OTF) generated continuous energy cross sections within the Monte Carlo N-Particle (MCNP6) transport code. The adjoint-weighted tallies are generated during the continuous energy k-eigenvalue Monte Carlo calculation. The weighting is based upon the iterated fission probability interpretation of the adjoint flux, which is the steady state population in a critical nuclear reactor caused by a neutron introduced at that point in phase space. The adjoint-weighted tallies are produced in a forward calculation and do not require an inversion of the random walk. The OTF cross section database uses a high order functional expansion between points on a user-defined energy-temperature mesh in which the coefficients with respect to a polynomial fitting in temperature are stored. The coefficients of the fits are generated before run- time and called upon during the simulation to produce cross sections at any given energy and temperature. The polynomial form of the OTF cross sections allows the possibility of obtaining temperature derivatives of the cross sections on-the-fly. The use of Monte Carlo sampling of adjoint-weighted tallies and the capability of computing derivatives of continuous energy cross sections with respect to temperature are used to calculate the Doppler temperature coefficient in a research version of MCNP6. Temperature feedback results from the cross sections themselves, changes in the probability density functions, as well as changes in the density of the materials. The focus of this work is specific to the Doppler temperature feedback which result from Doppler broadening of cross sections as well as changes in the probability density function within the scattering kernel. This method is compared against published results using Mosteller's numerical benchmark to show accurate evaluations of the Doppler temperature coefficient, fuel assembly calculations, and a benchmark solution based on the heavy gas model for free-gas elastic scattering. An infinite medium benchmark for neutron free gas elastic scattering for large scattering ratios and constant absorption cross section has been developed using the heavy gas model. An exact closed form solution for the neutron energy spectrum is obtained in terms of the confluent hypergeometric function and compared against spectra for the free gas scattering model in MCNP6. Results show a quick increase in convergence of the analytic energy spectrum to the MCNP6 code with increasing target size, showing absolute relative differences of less than 5% for neutrons scattering with carbon. The analytic solution has been generalized to accommodate piecewise constant in energy absorption cross section to produce temperature feedback. Results reinforce the constraints in which heavy gas theory may be applied resulting in a significant target size to accommodate increasing cross section structure. The energy dependent piecewise constant cross section heavy gas model was used to produce a benchmark calculation of the Doppler temperature coefficient to show accurate calculations when using the adjoint-weighted method. Results show the Doppler temperature coefficient using adjoint weighting and cross section derivatives accurately obtains the correct solution within statistics as well as reduce computer runtimes by a factor of 50.

Using Machine Learning to Predict MCNP Bias

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grechanuk, Pavel Aleksandrovi

For many real-world applications in radiation transport where simulations are compared to experimental measurements, like in nuclear criticality safety, the bias (simulated - experimental k eff) in the calculation is an extremely important quantity used for code validation. The objective of this project is to accurately predict the bias of MCNP6 [1] criticality calculations using machine learning (ML) algorithms, with the intention of creating a tool that can complement the current nuclear criticality safety methods. In the latest release of MCNP6, the Whisper tool is available for criticality safety analysts and includes a large catalogue of experimental benchmarks, sensitivity profiles,more » and nuclear data covariance matrices. This data, coming from 1100+ benchmark cases, is used in this study of ML algorithms for criticality safety bias predictions.« less

Design tradeoff studies and sensitivity analysis, appendices B1 - B4. [hybrid electric vehicles

NASA Technical Reports Server (NTRS)

1979-01-01

Documentation is presented for a program which separately computes fuel and energy consumption for the two modes of operation of a hybrid electric vehicle. The distribution of daily travel is specified as input data as well as the weights which the component driving cycles are given in each of the composite cycles. The possibility of weight reduction through the substitution of various materials is considered as well as the market potential for hybrid vehicles. Data relating to battery compartment weight distribution and vehicle handling analysis is tabulated.

The National Shipbuilding Research Program. Proceedings of the REAPS Technical Symposium (4th) held in New Orleans, Louisiana on 21-22 June, 1977

DTIC Science & Technology

1977-06-01

meeting. The Agenda in Appendix A lists topics and speakers, while Appendix B identifies Symposium attendees. TABLE OF CONTENTS Preface...225 253 271 I I T R E S E A R C H I N S T I T U T E 111 TABLE OF CONTENTS (Continued) Page USERS EXPERIENCE WITH REAPS SIMPLIFIED ALKON B.J. Breen...the development of computer graphics systems for ship design and lofting applications, software for marine transportation system economic analysis , and

A computer program for two-particle generalized coefficients of fractional parentage

NASA Astrophysics Data System (ADS)

Deveikis, A.; Juodagalvis, A.

2008-10-01

We present a FORTRAN90 program GCFP for the calculation of the generalized coefficients of fractional parentage (generalized CFPs or GCFP). The approach is based on the observation that the multi-shell CFPs can be expressed in terms of single-shell CFPs, while the latter can be readily calculated employing a simple enumeration scheme of antisymmetric A-particle states and an efficient method of construction of the idempotent matrix eigenvectors. The program provides fast calculation of GCFPs for a given particle number and produces results possessing numerical uncertainties below the desired tolerance. A single j-shell is defined by four quantum numbers, (e,l,j,t). A supplemental C++ program parGCFP allows calculation to be done in batches and/or in parallel. Program summaryProgram title:GCFP, parGCFP Catalogue identifier: AEBI_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEBI_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 17 199 No. of bytes in distributed program, including test data, etc.: 88 658 Distribution format: tar.gz Programming language: FORTRAN 77/90 ( GCFP), C++ ( parGCFP) Computer: Any computer with suitable compilers. The program GCFP requires a FORTRAN 77/90 compiler. The auxiliary program parGCFP requires GNU-C++ compatible compiler, while its parallel version additionally requires MPI-1 standard libraries Operating system: Linux (Ubuntu, Scientific) (all programs), also checked on Windows XP ( GCFP, serial version of parGCFP) RAM: The memory demand depends on the computation and output mode. If this mode is not 4, the program GCFP demands the following amounts of memory on a computer with Linux operating system. It requires around 2 MB of RAM for the A=12 system at E⩽2. Computation of the A=50 particle system requires around 60 MB of RAM at E=0 and ˜70 MB at E=2 (note, however, that the calculation of this system will take a very long time). If the computation and output mode is set to 4, the memory demands by GCFP are significantly larger. Calculation of GCFPs of A=12 system at E=1 requires 145 MB. The program parGCFP requires additional 2.5 and 4.5 MB of memory for the serial and parallel version, respectively. Classification: 17.18 Nature of problem: The program GCFP generates a list of two-particle coefficients of fractional parentage for several j-shells with isospin. Solution method: The method is based on the observation that multishell coefficients of fractional parentage can be expressed in terms of single-shell CFPs [1]. The latter are calculated using the algorithm [2,3] for a spectral decomposition of an antisymmetrization operator matrix Y. The coefficients of fractional parentage are those eigenvectors of the antisymmetrization operator matrix Y that correspond to unit eigenvalues. A computer code for these coefficients is available [4]. The program GCFP offers computation of two-particle multishell coefficients of fractional parentage. The program parGCFP allows a batch calculation using one input file. Sets of GCFPs are independent and can be calculated in parallel. Restrictions:A<86 when E=0 (due to the memory constraints); small numbers of particles allow significantly higher excitations, though the shell with j⩾11/2 cannot get full (it is the implementation constraint). Unusual features: Using the program GCFP it is possible to determine allowed particle configurations without the GCFP computation. The GCFPs can be calculated either for all particle configurations at once or for a specified particle configuration. The values of GCFPs can be printed out with a complete specification in either one file or with the parent and daughter configurations printed in separate files. The latter output mode requires additional time and RAM memory. It is possible to restrict the ( J,T) values of the considered particle configurations. (Here J is the total angular momentum and T is the total isospin of the system.) The program parGCFP produces several result files the number of which equals to the number of particle configurations. To work correctly, the program GCFP needs to be compiled to read parameters from the standard input (the default setting). Running time: It depends on the size of the problem. The minimum time is required, if the computation and output mode ( CompMode) is not 4, but the resulting file is larger. A system with A=12 particles at E=0 (all 9411 GCFPs) took around 1 sec on a Pentium4 2.8 GHz processor with 1 MB L2 cache. The program required about 14 min to calculate all 1.3×10 GCFPs of E=1. The time for all 5.5×10 GCFPs of E=2 was about 53 hours. For this number of particles, the calculation time of both E=0 and E=1 with CompMode = 1 and 4 is nearly the same, when no other processes are running. The case of E=2 could not be calculated with CompMode = 4, because the RAM memory was insufficient. In general, the latter CompMode requires a longer computation time, although the resulting files are smaller in size. The program parGCFP puts virtually no time overhead. Its parallel version speeds-up the calculation. However, the results need to be collected from several files created for each configuration. References: [1] J. Levinsonas, Works of Lithuanian SSR Academy of Sciences 4 (1957) 17. [2] A. Deveikis, A. Bončkus, R. Kalinauskas, Lithuanian Phys. J. 41 (2001) 3. [3] A. Deveikis, R.K. Kalinauskas, B.R. Barrett, Ann. Phys. 296 (2002) 287. [4] A. Deveikis, Comput. Phys. Comm. 173 (2005) 186. (CPC Catalogue ID. ADWI_v1_0)

Undergraduate Student Task Group Approach to Complex Problem Solving Employing Computer Programming.

ERIC Educational Resources Information Center

Brooks, LeRoy D.

A project formulated a computer simulation game for use as an instructional device to improve financial decision making. The author constructed a hypothetical firm, specifying its environment, variables, and a maximization problem. Students, assisted by a professor and computer consultants and having access to B5500 and B6700 facilities, held 16…

Computer program for determining rotational line intensity factors for diatomic molecules

NASA Technical Reports Server (NTRS)

Whiting, E. E.

1973-01-01

A FORTRAN IV computer program, that provides a new research tool for determining reliable rotational line intensity factors (also known as Honl-London factors), for most electric and magnetic dipole allowed diatomic transitions, is described in detail. This users manual includes instructions for preparing the input data, a program listing, detailed flow charts, and three sample cases. The program is applicable to spin-allowed dipole transitions with either or both states intermediate between Hund's case (a) and Hund's case (b) coupling and to spin-forbidden dipole transitions with either or both states intermediate between Hund's case (c) and Hund's case (b) coupling.

Spectroscopic (FT-IR, FT-Raman) and quantum mechanical studies of 3t-pentyl-2r,6c-diphenylpiperidin-4-one thiosemicarbazone

NASA Astrophysics Data System (ADS)

Savithiri, S.; Arockia doss, M.; Rajarajan, G.; Thanikachalam, V.; Bharanidharan, S.; Saleem, H.

2015-02-01

In this study, the molecular structure and vibrational spectra of 3t-pentyl2r,6c-diphenylpiperidin-4-one thiosemicarbazone (PDPOTSC) were studied. The ground-state molecular geometry was ascertained by using the density functional theory (DFT)/B3LYP method using 6-31++G(d,p) as a basis set. The vibrational (FT-IR and FT-Raman) spectra of PDPOTSC were computed using DFT/B3LYP and HF methods with 6-31++G(d,p) basis set. The fundamental vibrations were assigned on the basis of the total energy distribution (TED ⩾ 10%) of the vibrational modes, calculated with scaled quantum mechanics (SQM) methods PQS program. The electrical dipole moment (μ) and first hyperpolarizability (βo) values have been computed using DFT/B3LYP and HF methods. The calculated result (βo) shows that the title molecule might have nonlinear optical (NLO) behavior. Atomic charges of C, N, S and molecular electrostatic potential (MEP) were calculated using B3LYP/6-31G++(d,p). The HOMO-LUMO energies were calculated and natural bonding orbital (NBO) analysis has also been carried out.

An Organization Design for the PERSPAY Consolidated Data Center.

DTIC Science & Technology

1983-06-01

GENERAL 25 B. PROGRAM DEVELOPMENT 30 C. COMPUTER OPERATIONS 33 D. TECHNICAL SUPPORT BRANCH 36 E. MANAGEMENT OF THE D.P. ORGANIZATION...support, and 4) management , including planning, administration, and control ( Brandon , Norton Gaydasch, Frank). The following list shows the... Management policy on just what the ADP installation is expected to do is an information requirement that must be satisfied in order for the operation to

Short-term effects of a randomized computer-based out-of-school smoking prevention trial aimed at elementary schoolchildren.

PubMed

Ausems, Marlein; Mesters, Ilse; van Breukelen, Gerard; De Vries, Hein

2002-06-01

Smoking prevention programs usually run during school hours. In our study, an out-of-school program was developed consisting of a computer-tailored intervention aimed at the age group before school transition (11- to 12-year-old elementary schoolchildren). The aim of this study is to evaluate the additional effect of out-of-school smoking prevention. One hundred fifty-six participating schools were randomly allocated to one of four research conditions: (a) the in-school condition, an existing seven-lesson program; (b) the out-of-school condition, three computer-tailored letters sent to the students' homes; (c) the in-school and out-of-school condition, a combined approach; (d) the control condition. Pretest and 6 months follow-up data on smoking initiation and continuation, and data on psychosocial variables were collected from 3,349 students. Control and out-of-school conditions differed regarding posttest smoking initiation (18.1 and 10.4%) and regarding posttest smoking continuation (23.5 and 13.1%). Multilevel logistic regression analyses showed positive effects regarding the out-of-school program. Significant effects were not found regarding the in-school program, nor did the combined approach show stronger effects than the single-method approaches. The findings of this study suggest that smoking prevention trials for elementary schoolchildren can be effective when using out-of-school computer-tailored interventions. Copyright 2002 Elsevier Science (USA).

Update of aircraft profile data for the Integrated Noise Model computer program, vol. 3 : appendix B aircraft performance coefficients

DOT National Transportation Integrated Search

1992-03-01

This report provides aircraft takeoff and landing profiles, : aircraft aerodynamic performance coefficients and engine : performance coefficients for the aircraft data base : (Database 9) in the Integrated Noise Model (INM) computer : program. Flight...

WE-DE-207B-11: Implementation of Size-Specific 3D Beam Modulation Filters On a Dedicated Breast CT Platform Using Breast Immobilization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hernandez, A; Boone, J

Purpose: To implement a 3D beam modulation filter (3D-BMF) in dedicated breast CT (bCT) and develop a method for conforming the patient’s breast to a pre-defined shape, optimizing the effects of the filter. This work expands on previous work reporting the methodology for designing a 3D-BMF that can spare unnecessary dose and improve signal equalization at the detector by preferentially filtering the beam in the thinner anterior and peripheral breast regions. Methods: Effective diameter profiles were measured for 219 segmented bCT images, grouped into volume quintiles, and averaged within each group to represent the range of breast sizes found clinically.more » These profiles were then used to generate five size-specific computational phantoms and fabricate five size-specific UHMW phantoms. Each computational phantom was utilized for designing a size-specific 3D-BMF using previously reported methods. Glandular dose values and projection images were simulated in MCNP6 with and without the 3DBMF using the system specifications of our prototype bCT scanner “Doheny”. Lastly, thermoplastic was molded around each of the five phantom sizes and used to produce a series of breast immobilizers for use in conforming the patient’s breast during bCT acquisition. Results: After incorporating the 3D-BMF, MC simulations estimated an 80% average reduction in the detector dynamic range requirements across all phantom sizes. The glandular dose was reduced on average 57% after normalizing by the number of quanta reaching the detector under the thickest region of the breast. Conclusion: A series of bCT-derived breast phantoms were used to design size-specific 3D-BMFs and breast immobilizers that can be used on the bCT platform to conform the patient’s breast and therefore optimally exploit the benefits of the 3D-BMF. Current efforts are focused on fabricating several prototype 3D-BMFs and performing phantom scans on Doheny for MC simulation validation and image quality analysis. Research reported in this paper was supported in part by the National Cancer Institute of the National Institutes of Health under award R01CA181081. The content is solely the responsibility of the authors and does not necessarily represent the official views of the National Institue of Health.« less

Considerations of MCNP Monte Carlo code to be used as a radiotherapy treatment planning tool.

PubMed

Juste, B; Miro, R; Gallardo, S; Verdu, G; Santos, A

2005-01-01

The present work has simulated the photon and electron transport in a Theratron 780® (MDS Nordion)⁶⁰Co radiotherapy unit, using the Monte Carlo transport code, MCNP (Monte Carlo N-Particle). This project explains mainly the different methodologies carried out to speedup calculations in order to apply this code efficiently in radiotherapy treatment planning.

Inclusion of Scatter in HADES: Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aufderheide, M B

Covert nuclear attack is one of the foremost threats facing the United States and is a primary focus of the War on Terror. The Domestic Nuclear Detection Office (DNDO), within the Department of Homeland Security (DHS), is chartered to develop, and improve domestic systems to detect and interdict smuggling for the illicit use of a nuclear explosive device, fissile material or radiologica1 material. The CAARS (Cargo Advanced Automated Radiography System) program is a major part of the DHS effort to enhance US security by harnessing cutting-edge technologies to detect radiological and nuclear threats at points of entry to the Unitedmore » States. DNDO has selected vendors to develop complete radiographic systems. It is crucial that the initial design and testing concepts for the systems be validated and compared prior to the substantial efforts to build and deploy prototypes and subsequent large-scale production. An important aspect of these systems is the scatter which interferes with imaging. Monte Carlo codes, such as MCNP (X-5 Monte Carlo Team, 2005 Revision) allow scatter to be calculatied, but these calculations are very time consuming. It would be useful to have a fast scatter estimation algorithm in a fast ray tracing code. We have been extending the HADES ray-tracing radiographic simulation code to model vendor systems in a flexible and quick fashion and to use this tool to study a variety of questions involving system performance and the comparative value of surrogates. To enable this work, HADES has been linked to the BRL-CAD library (BRL-CAD Open Source Project, 2010), in order to enable the inclusion of complex CAD geometries in simulations, scanner geometries have been implemented in HADES, and the novel detector responses have been included in HADES. A major extension of HADES which has been required by this effort is the inclusion of scatter in these radiographic simulations. Ray tracing codes generally do not easily allow the inclusion of scatter, because these codes define a source and a grid of detector pixels and only compute the attenuation along rays between these points. Scatter is an extremely complex set of processes which can involve rays which change directions many times between the source and detector. Scatter from outside the field of view of the imaging system, as well as within the field of view, can have an important role in image formation. In this report, we will describe how we implemented a treatment of scatter in HADES. We begin with a discussion of how we define scatter in Section 2, followed by a description of how single Compton scatter is now included in HADES in Section 3. In Section 4 we report a set of verification tests against MCNP and tests of how the technique scales with image size, number of scatters allowed and number of processors used in the calculations. In Section 5, we describe how we plan to extend this approach to other forms of scatter and conclude in Section 6. It should be emphasized that the purpose of this report is to show that a form of scatter has been implemented in HADES and has been verified against MCNP. Validation, the process of comparing simulation and experiment, is a future task.« less

New version of PLNoise: a package for exact numerical simulation of power-law noises

NASA Astrophysics Data System (ADS)

Milotti, Edoardo

2007-08-01

In a recent paper I have introduced a package for the exact simulation of power-law noises and other colored noises [E. Milotti, Comput. Phys. Comm. 175 (2006) 212]: in particular, the algorithm generates 1/f noises with 0<α⩽2. Here I extend the algorithm to generate 1/f noises with 2<α⩽4 (black noises). The method is exact in the sense that it produces a sampled process with a theoretically guaranteed range-limited power-law spectrum for any arbitrary sequence of sampling intervals, i.e. the sampling times may be unevenly spaced. Program summaryTitle of program: PLNoise Catalogue identifier:ADXV_v2_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/ADXV_v2_0.html Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html Program obtainable from: CPC Program Library, Queen's University of Belfast, N. Ireland Programming language used: ANSI C Computer: Any computer with an ANSI C compiler: the package has been tested with gcc version 3.2.3 on Red Hat Linux 3.2.3-52 and gcc version 4.0.0 and 4.0.1 on Apple Mac OS X-10.4 Operating system: All operating systems capable of running an ANSI C compiler RAM: The code of the test program is very compact (about 60 Kbytes), but the program works with list management and allocates memory dynamically; in a typical run with average list length 2ṡ10, the RAM taken by the list is 200 Kbytes External routines: The package needs external routines to generate uniform and exponential deviates. The implementation described here uses the random number generation library ranlib freely available from Netlib [B.W. Brown, J. Lovato, K. Russell: ranlib, available from Netlib, http://www.netlib.org/random/index.html, select the C version ranlib.c], but it has also been successfully tested with the random number routines in Numerical Recipes [W.H. Press, S.A. Teulkolsky, W.T. Vetterling, B.P. Flannery, Numerical Recipes in C: The Art of Scientific Computing, second ed., Cambridge Univ. Press., Cambridge, 1992, pp. 274-290]. Notice that ranlib requires a pair of routines from the linear algebra package LINPACK, and that the distribution of ranlib includes the C source of these routines, in case LINPACK is not installed on the target machine. No. of lines in distributed program, including test data, etc.:2975 No. of bytes in distributed program, including test data, etc.:194 588 Distribution format:tar.gz Catalogue identifier of previous version: ADXV_v1_0 Journal reference of previous version: Comput. Phys. Comm. 175 (2006) 212 Does the new version supersede the previous version?: Yes Nature of problem: Exact generation of different types of colored noise. Solution method: Random superposition of relaxation processes [E. Milotti, Phys. Rev. E 72 (2005) 056701], possibly followed by an integration step to produce noise with spectral index >2. Reasons for the new version: Extension to 1/f noises with spectral index 2<α⩽4: the new version generates both noises with spectral with spectral index 0<α⩽2 and with 2<α⩽4. Summary of revisions: Although the overall structure remains the same, one routine has been added and several changes have been made throughout the code to include the new integration step. Unusual features: The algorithm is theoretically guaranteed to be exact, and unlike all other existing generators it can generate samples with uneven spacing. Additional comments: The program requires an initialization step; for some parameter sets this may become rather heavy. Running time: Running time varies widely with different input parameters, however in a test run like the one in Section 3 in the long write-up, the generation routine took on average about 75 μs for each sample.

Estimation of coolant void reactivity for CANDU-NG lattice using DRAGON and validation using MCNP5 and TRIPOLI-4.3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Karthikeyan, R.; Tellier, R. L.; Hebert, A.

2006-07-01

The Coolant Void Reactivity (CVR) is an important safety parameter that needs to be estimated at the design stage of a nuclear reactor. It helps to have an a priori knowledge of the behavior of the system during a transient initiated by the loss of coolant. In the present paper, we have attempted to estimate the CVR for a CANDU New Generation (CANDU-NG) lattice, as proposed at an early stage of the Advanced CANDU Reactor (ACR) development. We have attempted to estimate the CVR with development version of the code DRAGON, using the method of characteristics. DRAGON has several advancedmore » self-shielding models incorporated in it, each of them compatible with the method of characteristics. This study will bring to focus the performance of these self-shielding models, especially when there is voiding of such a tight lattice. We have also performed assembly calculations in 2 x 2 pattern for the CANDU-NG fuel, with special emphasis on checkerboard voiding. The results obtained have been validated against Monte Carlo codes MCNP5 and TRIPOLI-4.3. (authors)« less

Performance of the MTR core with MOX fuel using the MCNP4C2 code.

PubMed

Shaaban, Ismail; Albarhoum, Mohamad

2016-08-01

The MCNP4C2 code was used to simulate the MTR-22 MW research reactor and perform the neutronic analysis for a new fuel namely: a MOX (U3O8&PuO2) fuel dispersed in an Al matrix for One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) in its core. Its new characteristics were compared to its original characteristics based on the U3O8-Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U3O8-Al fuel at nominal power were used to validate the calculated values and were found acceptable. The achieved results seem to confirm that the use of MOX fuel in the MTR-22 MW will not degrade the safe operational conditions of the reactor. In addition, the use of MOX fuel in the MTR-22 MW core leads to reduce the uranium fuel enrichment with (235)U and the amount of loaded (235)U in the core by about 34.84% and 15.21% for the ONT and TNTs cases, respectively. Copyright © 2016 Elsevier Ltd. All rights reserved.

Monte Carlo simulations of medical imaging modalities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Estes, G.P.

Because continuous-energy Monte Carlo radiation transport calculations can be nearly exact simulations of physical reality (within data limitations, geometric approximations, transport algorithms, etc.), it follows that one should be able to closely approximate the results of many experiments from first-principles computations. This line of reasoning has led to various MCNP studies that involve simulations of medical imaging modalities and other visualization methods such as radiography, Anger camera, computerized tomography (CT) scans, and SABRINA particle track visualization. It is the intent of this paper to summarize some of these imaging simulations in the hope of stimulating further work, especially as computermore » power increases. Improved interpretation and prediction of medical images should ultimately lead to enhanced medical treatments. It is also reasonable to assume that such computations could be used to design new or more effective imaging instruments.« less

Procedures for Computing Site Seismicity

DTIC Science & Technology

1994-02-01

Fourth World Conference on Earthquake Engineering, Santiago, Chile , 1969. Schnabel, P.B., J. Lysmer, and H.B. Seed (1972). SHAKE, a computer program for...This fault system is composed of the Elsinore and Whittier fault zones, Agua Caliente fault, and Earthquake Valley fault. Five recent earthquakes of

Radiation shielding quality assurance

NASA Astrophysics Data System (ADS)

Um, Dallsun

For the radiation shielding quality assurance, the validity and reliability of the neutron transport code MCNP, which is now one of the most widely used radiation shielding analysis codes, were checked with lot of benchmark experiments. And also as a practical example, follows were performed in this thesis. One integral neutron transport experiment to measure the effect of neutron streaming in iron and void was performed with Dog-Legged Void Assembly in Knolls Atomic Power Laboratory in 1991. Neutron flux was measured six different places with the methane detectors and a BF-3 detector. The main purpose of the measurements was to provide benchmark against which various neutron transport calculation tools could be compared. Those data were used in verification of Monte Carlo Neutron & Photon Transport Code, MCNP, with the modeling for that. Experimental results and calculation results were compared in both ways, as the total integrated value of neutron fluxes along neutron energy range from 10 KeV to 2 MeV and as the neutron spectrum along with neutron energy range. Both results are well matched with the statistical error +/-20%. MCNP results were also compared with those of TORT, a three dimensional discrete ordinates code which was developed by Oak Ridge National Laboratory. MCNP results are superior to the TORT results at all detector places except one. This means that MCNP is proved as a very powerful tool for the analysis of neutron transport through iron & air and further it could be used as a powerful tool for the radiation shielding analysis. For one application of the analysis of variance (ANOVA) to neutron and gamma transport problems, uncertainties for the calculated values of critical K were evaluated as in the ANOVA on statistical data.

Development of low level 226Ra analysis for live fish using gamma-ray spectrometry

NASA Astrophysics Data System (ADS)

Chandani, Z.; Prestwich, W. V.; Byun, S. H.

2017-06-01

A low level 226Ra analysis method for live fish was developed using a 4π NaI(Tl) gamma-ray spectrometer. In order to find out the best algorithm for accomplishing the lowest detection limit, the gamma-ray spectrum from a 226Ra point was collected and nine different methods were attempted for spectral analysis. The lowest detection limit of 0.99 Bq for an hour counting occurred when the spectrum was integrated in the energy region of 50-2520 keV. To extend 226Ra analysis to live fish, a Monte Carlo simulation model with a cylindrical fish in a water container was built using the MCNP code. From simulation results, the spatial distribution of the efficiency and the efficiency correction factor for the live fish model were determined. The MCNP model will be able to be conveniently modified when a different fish or container geometry is employed as fish grow up in real experiments.

Multi-canister overpack project -- verification and validation, MCNP 4A

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goldmann, L.H.

This supporting document contains the software verification and validation (V and V) package used for Phase 2 design of the Spent Nuclear Fuel Multi-Canister Overpack. V and V packages for both ANSYS and MCNP are included. Description of Verification Run(s): This software requires that it be compiled specifically for the machine it is to be used on. Therefore to facilitate ease in the verification process the software automatically runs 25 sample problems to ensure proper installation and compilation. Once the runs are completed the software checks for verification by performing a file comparison on the new output file and themore » old output file. Any differences between any of the files will cause a verification error. Due to the manner in which the verification is completed a verification error does not necessarily indicate a problem. This indicates that a closer look at the output files is needed to determine the cause of the error.« less

Using the MCNP Taylor series perturbation feature (efficiently) for shielding problems

NASA Astrophysics Data System (ADS)

Favorite, Jeffrey

2017-09-01

The Taylor series or differential operator perturbation method, implemented in MCNP and invoked using the PERT card, can be used for efficient parameter studies in shielding problems. This paper shows how only two PERT cards are needed to generate an entire parameter study, including statistical uncertainty estimates (an additional three PERT cards can be used to give exact statistical uncertainties). One realistic example problem involves a detailed helium-3 neutron detector model and its efficiency as a function of the density of its high-density polyethylene moderator. The MCNP differential operator perturbation capability is extremely accurate for this problem. A second problem involves the density of the polyethylene reflector of the BeRP ball and is an example of first-order sensitivity analysis using the PERT capability. A third problem is an analytic verification of the PERT capability.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hannan, N. A.; Matos, J. E.; Stillman, J. A.

At the request of the Czech Technical University (CTU) in Prague, ANL has performed independent verification calculations using the MCNP Monte Carlo code for three core configurations of the VR-1 reactor: a current core configuration B1 with HEU (36%) IRT-3M fuel assemblies and planned core configurations C1 and C2 with LEU (19.7%) IRT-4M fuel assemblies. Details of these configurations were provided to ANL by CTU. For core configuration B1, criticality calculations were performed for two sets of control rod positions provided to ANL by CTU. Fore core configurations C1 and C2, criticality calculations were done for cases with all controlmore » rods at the top positions, all control rods at the bottom positions, and two critical states of the reactor for different control rod positions. In addition, sensitivity studies for variation of the {sup 235}U mass in each fuel assembly and variation of the fuel meat and cladding thicknesses in each of the fuel tubes were doe for the C1 core configuration. The reactivity worth of the individual control rods was calculated for the B1, C1, and C2 core configurations. Finally, the reactivity feedback coefficients, the prompt neutron lifetime, and the total effective delay neutron fraction were calculated for each of the three cores.« less

46 CFR 62.25-25 - Programable systems and devices.

Code of Federal Regulations, 2013 CFR

2013-10-01

...-25 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING VITAL SYSTEM... range of the equipment. (b) Operating programs for microprocessor-based or computer-based vital control... power resumption. (c) If a microprocessor-based or computer-based system serves both vital and non-vital...

46 CFR 62.25-25 - Programable systems and devices.

Code of Federal Regulations, 2010 CFR

2010-10-01

...-25 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING VITAL SYSTEM... range of the equipment. (b) Operating programs for microprocessor-based or computer-based vital control... power resumption. (c) If a microprocessor-based or computer-based system serves both vital and non-vital...

46 CFR 62.25-25 - Programable systems and devices.

Code of Federal Regulations, 2012 CFR

2012-10-01

...-25 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING VITAL SYSTEM... range of the equipment. (b) Operating programs for microprocessor-based or computer-based vital control... power resumption. (c) If a microprocessor-based or computer-based system serves both vital and non-vital...

46 CFR 62.25-25 - Programable systems and devices.

Code of Federal Regulations, 2011 CFR

2011-10-01

...-25 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING VITAL SYSTEM... range of the equipment. (b) Operating programs for microprocessor-based or computer-based vital control... power resumption. (c) If a microprocessor-based or computer-based system serves both vital and non-vital...

Information Security: Federal Guidance Needed to Address Control Issues With Implementing Cloud Computing

DTIC Science & Technology

2010-05-01

Figure 2: Cloud Computing Deployment Models 13 Figure 3: NIST Essential Characteristics 14 Figure 4: NASA Nebula Container 37...Access Computing Environment (RACE) program, the National Aeronautics and Space Administration’s (NASA) Nebula program, and the Department of...computing programs: the DOD’s RACE program; NASA’s Nebula program; and Department of Transportation’s CARS program, including lessons learned related

Modeling gamma radiation dose in dwellings due to building materials.

PubMed

de Jong, Peter; van Dijk, Willem

2008-01-01

A model is presented that calculates the absorbed dose rate in air of gamma radiation emitted by building materials in a rectangular body construction. The basis for these calculations is formed by a fixed set of specific absorbed dose rates (the dose rate per Bq kg(-1) 238U, 232Th, and 40K), as determined for a standard geometry with the dimensions 4 x 5 x 2.8 m3. Using the computer codes Marmer and MicroShield, correction factors are assessed that quantify the influence of several room and material related parameters on the specific absorbed dose rates. The investigated parameters are the position in the construction; the thickness, density, and dimensions of the construction parts; the contribution from the outer leave; the presence of doors and windows; the attenuation by internal partition walls; the contribution from building materials present in adjacent rooms; and the effect of non-equilibrium due to 222Rn exhalation. To verify the precision, the proposed method is applied to three Dutch reference dwellings, i.e., a row house, a coupled house, and a gallery apartment. The averaged difference with MCNP calculations is found to be 4%.

Prompt Radiation Protection Factors

DTIC Science & Technology

2018-02-01

dimensional Monte-Carlo radiation transport code MCNP (Monte Carlo N-Particle) and the evaluation of the protection factors (ratio of dose in the open to...radiation was performed using the three dimensional Monte- Carlo radiation transport code MCNP (Monte Carlo N-Particle) and the evaluation of the protection...by detonation of a nuclear device have placed renewed emphasis on evaluation of the consequences in case of such an event. The Defense Threat

An investigation of MCNP6.1 beryllium oxide S(α, β) cross sections

DOE PAGES

Sartor, Raymond F.; Glazener, Natasha N.

2016-03-08

In MCNP6.1, materials are constructed by identifying the constituent isotopes (or elements in a few cases) individually. This list selects the corresponding microscopic cross sections calculated from the free-gas model to create the material macroscopic cross sections. Furthermore, the free-gas model and the corresponding material macroscopic cross sections assume that the interactions of atoms do not affect the nuclear cross sections.

Characterization of Filters Loaded With Reactor Strontium Carbonate - 13203

DOE Office of Scientific and Technical Information (OSTI.GOV)

Josephson, Walter S.; Steen, Franciska H.

A collection of three highly radioactive filters containing reactor strontium carbonate were being prepared for disposal. All three filters were approximately characterized at the time of manufacture by gravimetric methods. The first filter had been partially emptied, and the quantity of residual activity was uncertain. Dose rate to activity modeling using the Monte-Carlo N Particle (MCNP) code was selected to confirm the gravimetric characterization of the full filters, and to fully characterize the partially emptied filter. Although dose rate to activity modeling using MCNP is a common technique, it is not often used for Bremsstrahlung-dominant materials such as reactor strontium.more » As a result, different MCNP modeling options were compared to determine the optimum approach. This comparison indicated that the accuracy of the results were heavily dependent on the MCNP modeling details and the location of the dose rate measurement point. The optimum model utilized a photon spectrum generated by the Oak Ridge Isotope Generation and Depletion (ORIGEN) code and dose rates measured at 30 cm. Results from the optimum model agreed with the gravimetric estimates within 15%. It was demonstrated that dose rate to activity modeling can be successful for Bremsstrahlung-dominant radioactive materials. However, the degree of success is heavily dependent on the choice of modeling techniques. (authors)« less

Evaluation of the Pool Critical Assembly Benchmark with Explicitly-Modeled Geometry using MCNP6

DOE PAGES

Kulesza, Joel A.; Martz, Roger Lee

2017-03-01

Despite being one of the most widely used benchmarks for qualifying light water reactor (LWR) radiation transport methods and data, no benchmark calculation of the Oak Ridge National Laboratory (ORNL) Pool Critical Assembly (PCA) pressure vessel wall benchmark facility (PVWBF) using MCNP6 with explicitly modeled core geometry exists. As such, this paper provides results for such an analysis. First, a criticality calculation is used to construct the fixed source term. Next, ADVANTG-generated variance reduction parameters are used within the final MCNP6 fixed source calculations. These calculations provide unadjusted dosimetry results using three sets of dosimetry reaction cross sections of varyingmore » ages (those packaged with MCNP6, from the IRDF-2002 multi-group library, and from the ACE-formatted IRDFF v1.05 library). These results are then compared to two different sets of measured reaction rates. The comparison agrees in an overall sense within 2% and on a specific reaction- and dosimetry location-basis within 5%. Except for the neptunium dosimetry, the individual foil raw calculation-to-experiment comparisons usually agree within 10% but is typically greater than unity. Finally, in the course of developing these calculations, geometry that has previously not been completely specified is provided herein for the convenience of future analysts.« less

Scattered Dose Calculations and Measurements in a Life-Like Mouse Phantom

PubMed Central

Welch, David; Turner, Leah; Speiser, Michael; Randers-Pehrson, Gerhard; Brenner, David J.

2017-01-01

Anatomically accurate phantoms are useful tools for radiation dosimetry studies. In this work, we demonstrate the construction of a new generation of life-like mouse phantoms in which the methods have been generalized to be applicable to the fabrication of any small animal. The mouse phantoms, with built-in density inhomogeneity, exhibit different scattering behavior dependent on where the radiation is delivered. Computer models of the mouse phantoms and a small animal irradiation platform were devised in Monte Carlo N-Particle code (MCNP). A baseline test replicating the irradiation system in a computational model shows minimal differences from experimental results from 50 Gy down to 0.1 Gy. We observe excellent agreement between scattered dose measurements and simulation results from X-ray irradiations focused at either the lung or the abdomen within our phantoms. This study demonstrates the utility of our mouse phantoms as measurement tools with the goal of using our phantoms to verify complex computational models. PMID:28140787

FORTRAN manpower account program

NASA Technical Reports Server (NTRS)

Strand, J. N.

1972-01-01

Computer program for determining manpower costs for full time, part time, and contractor personnel is discussed. Twelve different tables resulting from computer output are described. Program is written in FORTRAN 4 for IBM 360/65 computer.

Principles for Integrating Mars Analog Science, Operations, and Technology Research

NASA Technical Reports Server (NTRS)

Clancey, William J.

2003-01-01

During the Apollo program, the scientific community and NASA used terrestrial analog sites for understanding planetary features and for training astronauts to be scientists. Human factors studies (Harrison, Clearwater, & McKay 1991; Stuster 1996) have focused on the effects of isolation in extreme environments. More recently, with the advent of wireless computing, we have prototyped advanced EVA technologies for navigation, scheduling, and science data logging (Clancey 2002b; Clancey et al., in press). Combining these interests in a single expedition enables tremendous synergy and authenticity, as pioneered by Pascal Lee's Haughton-Mars Project (Lee 2001; Clancey 2000a) and the Mars Society s research stations on a crater rim on Devon Island in the High Canadian Arctic (Clancey 2000b; 2001b) and the Morrison Formation of southeast Utah (Clancey 2002a). Based on this experience, the following principles are proposed for conducting an integrated science, operations, and technology research program at analog sites: 1) Authentic work; 2) PI-based projects; 3) Unencumbered baseline studies; 4) Closed simulations; and 5) Observation and documentation. Following these principles, we have been integrating field science, operations research, and technology development at analog sites on Devon Island and in Utah over the past five years. Analytic methods include work practice simulation (Clancey 2002c; Sierhuis et a]., 2000a;b), by which the interaction of human behavior, facilities, geography, tools, and procedures are formalized in computer models. These models are then converted into the runtime EVA system we call mobile agents (Clancey 2002b; Clancey et al., in press). Furthermore, we have found that the Apollo Lunar Surface Journal (Jones, 1999) provides a vast repository or understanding astronaut and CapCom interactions, serving as a baseline for Mars operations and quickly highlighting opportunities for computer automation (Clancey, in press).

Optimized Waterspace Management and Scheduling Using Mixed-Integer Linear Programming

DTIC Science & Technology

2016-01-01

Complete [30]. Proposition 4.1 satisfies the first criterion. For the second criterion, we will use the Traveling Salesman Problem (TSP), which has been...A branch and cut algorithm for the symmetric generalized traveling salesman problem , Operations Research 45 (1997) 378–394. [33] J. Silberholz, B...Golden, The generalized traveling salesman problem : A new genetic algorithm ap- proach, Extended Horizons: Advances in Computing, Optimization, and

Fatigue strength reduction model: RANDOM3 and RANDOM4 user manual, appendix 2

NASA Technical Reports Server (NTRS)

Boyce, Lola; Lovelace, Thomas B.

1989-01-01

The FORTRAN programs RANDOM3 and RANDOM4 are documented. They are based on fatigue strength reduction, using a probabilistic constitutive model. They predict the random lifetime of an engine component to reach a given fatigue strength. Included in this user manual are details regarding the theoretical backgrounds of RANDOM3 and RANDOM4. Appendix A gives information on the physical quantities, their symbols, FORTRAN names, and both SI and U.S. Customary units. Appendix B and C include photocopies of the actual computer printout corresponding to the sample problems. Appendices D and E detail the IMSL, Version 10(1), subroutines and functions called by RANDOM3 and RANDOM4 and SAS/GRAPH(2) programs that can be used to plot both the probability density functions (p.d.f.) and the cumulative distribution functions (c.d.f.).

A suspended boron foil multi-wire proportional counter neutron detector

NASA Astrophysics Data System (ADS)

Nelson, Kyle A.; Edwards, Nathaniel S.; Hinson, Niklas J.; Wayant, Clayton D.; McGregor, Douglas S.

2014-12-01

Three natural boron foils, approximately 1.0 cm in diameter and 1.0 μm thick, were obtained from The Lebow Company and suspended in a multi-wire proportional counter. Suspending the B foils allowed the alpha particle and Li ion reaction products to escape simultaneously, one on each side of the foil, and be measured concurrently in the gas volume. The thermal neutron response pulse-height spectrum was obtained and two obvious peaks appear from the 94% and 6% branches of the 10B(n,α)7Li neutron reaction. Scanning electron microscope images were collected to obtain the exact B foil thicknesses and MCNP6 simulations were completed for those same B thicknesses. Pulse-height spectra obtained from the simulations were compared to experimental data and matched well. The theoretical intrinsic thermal-neutron detection efficiency for enriched 10B foils was calculated and is presented. Additionally, the intrinsic thermal neutron detection efficiency of the three natural B foils was calculated to be 3.2±0.2%.

Computer Science 205. Interim Guide, 1983.

ERIC Educational Resources Information Center

Manitoba Dept. of Education, Winnipeg.

This guide to a 4-unit, required high school computer science course emphasizes problem solving and computer programming and is designed for use with a variety of hardware configurations and programming languages. An overview covers the program rationale, goals and objectives, program design and description, program implementation, time allotment,…

DYGABCD: A program for calculating linear A, B, C, and D matrices from a nonlinear dynamic engine simulation

NASA Technical Reports Server (NTRS)

Geyser, L. C.

1978-01-01

A digital computer program, DYGABCD, was developed that generates linearized, dynamic models of simulated turbofan and turbojet engines. DYGABCD is based on an earlier computer program, DYNGEN, that is capable of calculating simulated nonlinear steady-state and transient performance of one- and two-spool turbojet engines or two- and three-spool turbofan engines. Most control design techniques require linear system descriptions. For multiple-input/multiple-output systems such as turbine engines, state space matrix descriptions of the system are often desirable. DYGABCD computes the state space matrices commonly referred to as the A, B, C, and D matrices required for a linear system description. The report discusses the analytical approach and provides a users manual, FORTRAN listings, and a sample case.

Computer programs for thermodynamic and transport properties of hydrogen

NASA Technical Reports Server (NTRS)

Hall, W. J.; Mc Carty, R. D.; Roder, H. M.

1968-01-01

Computer program subroutines provide the thermodynamic and transport properties of hydrogen in tabular form. The programs provide 18 combinations of input and output variables. This program is written in FORTRAN 4 for use on the IBM 7044 or CDC 3600 computers.

Counterfactuals cannot count: a rejoinder to David Chalmers.

PubMed

Bishop, Mark

2002-12-01

The initial argument presented herein is not significantly original--it is a simple reflection upon a notion of computation originally developed by Putnam (Putnam 1988; see also Searle, 1990) and criticised by Chalmers et al. (Chalmers, 1994; 1996a, b; see also the special issue, What is Computation?, in Minds and Machines, 4:4, November 1994). In what follows, instead of seeking to justify Putnam's conclusion that every open system implements every Finite State Automaton (FSA) and hence that psychological states of the brain cannot be functional states of a computer, I will establish the weaker result that, over a finite time window every open system implements the trace of FSA Q, as it executes program (P) on input (I). If correct the resulting bold philosophical claim is that phenomenal states--such as feelings and visual experiences--can never be understood or explained functionally. Copyright 2002 Elsevier Science (USA)

Performance Study of Monte Carlo Codes on Xeon Phi Coprocessors — Testing MCNP 6.1 and Profiling ARCHER Geometry Module on the FS7ONNi Problem

NASA Astrophysics Data System (ADS)

Liu, Tianyu; Wolfe, Noah; Lin, Hui; Zieb, Kris; Ji, Wei; Caracappa, Peter; Carothers, Christopher; Xu, X. George

2017-09-01

This paper contains two parts revolving around Monte Carlo transport simulation on Intel Many Integrated Core coprocessors (MIC, also known as Xeon Phi). (1) MCNP 6.1 was recompiled into multithreading (OpenMP) and multiprocessing (MPI) forms respectively without modification to the source code. The new codes were tested on a 60-core 5110P MIC. The test case was FS7ONNi, a radiation shielding problem used in MCNP's verification and validation suite. It was observed that both codes became slower on the MIC than on a 6-core X5650 CPU, by a factor of 4 for the MPI code and, abnormally, 20 for the OpenMP code, and both exhibited limited capability of strong scaling. (2) We have recently added a Constructive Solid Geometry (CSG) module to our ARCHER code to provide better support for geometry modelling in radiation shielding simulation. The functions of this module are frequently called in the particle random walk process. To identify the performance bottleneck we developed a CSG proxy application and profiled the code using the geometry data from FS7ONNi. The profiling data showed that the code was primarily memory latency bound on the MIC. This study suggests that despite low initial porting e_ort, Monte Carlo codes do not naturally lend themselves to the MIC platform — just like to the GPUs, and that the memory latency problem needs to be addressed in order to achieve decent performance gain.

Full-Scale Model of Subionospheric VLF Signal Propagation Based on First-Principles Charged Particle Transport Calculations

NASA Astrophysics Data System (ADS)

Kouznetsov, A.; Cully, C. M.; Knudsen, D. J.

2016-12-01

Changes in D-Region ionization caused by energetic particle precipitation are monitored by the Array for Broadband Observations of VLF/ELF Emissions (ABOVE) - a network of receivers deployed across Western Canada. The observed amplitudes and phases of subionospheric-propagating VLF signals from distant artificial transmitters depend sensitively on the free electron population created by precipitation of energetic charged particles. Those include both primary (electrons, protons and heavier ions) and secondary (cascades of ionized particles and electromagnetic radiation) components. We have designed and implemented a full-scale model to predict the received VLF signals based on first-principle charged particle transport calculations coupled to the Long Wavelength Propagation Capability (LWPC) software. Calculations of ionization rates and free electron densities are based on MCNP-6 (a general-purpose Monte Carlo N- Particle) software taking advantage of its capability of coupled neutron/photon/electron transport and novel library of cross-sections for low-energetic electron and photon interactions with matter. Cosmic ray calculations of background ionization are based on source spectra obtained both from PAMELA direct Cosmic Rays spectra measurements and based on the recently-implemented MCNP 6 galactic cosmic-ray source, scaled using our (Calgary) neutron monitor measurement results. Conversion from calculated fluxes (MCNP F4 tallies) to ionization rates for low-energy electrons are based on the total ionization cross-sections for oxygen and nitrogen molecules from the National Institute of Standard and Technology. We use our model to explore the complexity of the physical processes affecting VLF propagation.

Computational and spectral studies of 6-phenylazo-3-(p-tolyl)-2H-chromen-2-one.

PubMed

Manimekalai, A; Vijayalakshmi, N

2015-02-05

6-Phenylazo-3-(p-tolyl)-2H-chromen-2-one 4 was prepared and characterized by IR, (1)H, and (13)C NMR spectral studies. The optimized structure of the chromen-2-one 4 was investigated by the Gaussian 03 B3LYP density functional method calculations at 6-31G(d,p) basis set. The gauge-independent atomic orbital (GIAO) (13)C and (1)H chemical shift calculations for the synthesized chromen-2-one in CDCl3 were also made by the same method. The computed IR frequencies of the chromen-2-one and the corresponding vibrational assignments were analyzed by means of potential energy distribution (PED%) calculation using vibrational energy distribution analysis (VEDA) program. The first order hyperpolarizability (βtot), polarizability (α) and dipole moment (μ) were calculated using 6-311G(d,p) basis set and the nonlinear optical (NLO) properties are also addressed theoretically. Stability of the chromen-2-one 4 molecule has been analyzed by calculating the intramolecular charge transfer using natural bond order (NBO) analysis. The molecular electrostatic potentials, HOMO-LUMO energy gap and geometrical parameters were also computed. Topological properties of the electronic charge density in chromen-2-one 4 were analyzed employing the Bader's Atoms in Molecule (AIM) theory which indicated the presence of intramolecular hydrogen bond in the molecule. Copyright © 2014 Elsevier B.V. All rights reserved.

Panasonic HR-1800 Hand-Held Computer Solutions to Composite Materials Formulas.

DTIC Science & Technology

1983-09-01

TR-83-4093 PROGRAM LISTING (Cont’d) I, : E-6 :PE*u3 D RINT 3 =U4---4*U3 gtr -BINf X𔃾)=U5-C4§U3 GS fQ :T SEXT *lJ 3 54A@ ;OSUB 6)i6’S2*U2J ,’-54 PLA... SEXT * 3’E6 ":W=X(ue) E:=E1*A(1i:E2:- * * :628800)E=6. ~.-- (33B4*A~ )2A QPmI 5 39 a@ G$=" T r4EH 310 1APUT p~~4 ~2’ -;C SUB 3800 ŗUCK!:4G (T’...u (?00

MuSim, a Graphical User Interface for Multiple Simulation Programs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Roberts, Thomas; Cummings, Mary Anne; Johnson, Rolland

2016-06-01

MuSim is a new user-friendly program designed to interface to many different particle simulation codes, regardless of their data formats or geometry descriptions. It presents the user with a compelling graphical user interface that includes a flexible 3-D view of the simulated world plus powerful editing and drag-and-drop capabilities. All aspects of the design can be parametrized so that parameter scans and optimizations are easy. It is simple to create plots and display events in the 3-D viewer (with a slider to vary the transparency of solids), allowing for an effortless comparison of different simulation codes. Simulation codes: G4beamline, MAD-X,more » and MCNP; more coming. Many accelerator design tools and beam optics codes were written long ago, with primitive user interfaces by today's standards. MuSim is specifically designed to make it easy to interface to such codes, providing a common user experience for all, and permitting the construction and exploration of models with very little overhead. For today's technology-driven students, graphical interfaces meet their expectations far better than text-based tools, and education in accelerator physics is one of our primary goals.« less

Relationship of In-Cylinder Gaseous and Particulate Concentration to Radiative Heat Transfer in Direct Injection-Type Diesel Combustion.

DTIC Science & Technology

1986-04-15

spectral reflectivities, a,.,, are included in the expression for the radiosity of surface 1, Bk1, as (I I l cos eI I(0)) 4 B cos e e 2 + B 2 cos 92 i...OAAG29-83-K-0042 (Scientific Program Officer, Dr. David M. Mann). NOMENCLATURE A band absorptance a constant determining species distribution 8 radiosity ...Fig. 1), the radiosity of the surface 1, 81, i.e., to compute the thermal and optical is expressed as properties of the species and to implement

Core–Shell Nanoparticle-Based Peptide Therapeutics and Combined Hyperthermia for Enhanced Cancer Cell Apoptosis

PubMed Central

2015-01-01

Mitochondria-targeting peptides have garnered immense interest as potential chemotherapeutics in recent years. However, there is a clear need to develop strategies to overcome the critical limitations of peptides, such as poor solubility and the lack of target specificity, which impede their clinical applications. To this end, we report magnetic core–shell nanoparticle (MCNP)-mediated delivery of a mitochondria-targeting pro-apoptotic amphipathic tail-anchoring peptide (ATAP) to malignant brain and metastatic breast cancer cells. Conjugation of ATAP to the MCNPs significantly enhanced the chemotherapeutic efficacy of ATAP, while the presence of targeting ligands afforded selective delivery to cancer cells. Induction of MCNP-mediated hyperthermia further potentiated the efficacy of ATAP. In summary, a combination of MCNP-mediated ATAP delivery and subsequent hyperthermia resulted in an enhanced effect on mitochondrial dysfunction, thus resulting in increased cancer cell apoptosis. PMID:25133971

MCNP simulation to optimise in-pile and shielding parts of the Portuguese SANS instrument.

PubMed

Gonçalves, I F; Salgado, J; Falcão, A; Margaça, F M A; Carvalho, F G

2005-01-01

A Small Angle Neutron Scattering instrument is being installed at one end of the tangential beam tube of the Portuguese Research Reactor. The instrument is fed using a neutron scatterer positioned in the middle of the beam tube. The scatterer consists of circulating H2O contained in a hollow disc of Al. The in-pile shielding components and the shielding installed around the neutron selector have been the object of an MCNP simulation study. The quantities calculated were the neutron and gamma-ray fluxes in different positions, the energy deposited in the material by the neutron and gamma-ray fields, the material activation resulting from the neutron field and radiation doses at the exit wall of the shutter and around the shielding. The MCNP results are presented and compared with results of an analytical approach and with experimental data collected after installation.

A D-D/D-T fusion reaction based neutron generator system for liver tumor BNCT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koivunoro, H.; Lou, T.P.; Leung, K. N.

2003-04-02

Boron-neutron capture therapy (BNCT) is an experimental radiation treatment modality used for highly malignant tumor treatments. Prior to irradiation with low energetic neutrons, a 10B compound is located selectively in the tumor cells. The effect of the treatment is based on the high LET radiation released in the {sup 10}B(n,{alpha}){sup 7}Li reaction with thermal neutrons. BNCT has been used experimentally for brain tumor and melanoma treatments. Lately applications of other severe tumor type treatments have been introduced. Results have shown that liver tumors can also be treated by BNCT. At Lawrence Berkeley National Laboratory, various compact neutron generators based onmore » D-D or D-T fusion reactions are being developed. The earlier theoretical studies of the D-D or D-T fusion reaction based neutron generators have shown that the optimal moderator and reflector configuration for brain tumor BNCT can be created. In this work, the applicability of 2.5 MeV neutrons for liver tumor BNCT application was studied. The optimal neutron energy for external liver treatments is not known. Neutron beams of different energies (1eV < E < 100 keV) were simulated and the dose distribution in the liver was calculated with the MCNP simulation code. In order to obtain the optimal neutron energy spectrum with the D-D neutrons, various moderator designs were performed using MCNP simulations. In this article the neutron spectrum and the optimized beam shaping assembly for liver tumor treatments is presented.« less

Deflagration-to-Detonation Transition in Heteorogeneous Solids: A Bibliography.

DTIC Science & Technology

1980-11-01

and Rockets, Vol. 9. No. 6, 1972, pp. 415-419. 1.4 Francois, D., and L. Joly; " La Rupture des Metaux; Ecole d’ete de la Colle sur Loup ," Masson et Cie...Computer Program for Multifield Fluid Flows," Los Alamos Scientific Laboratory, LA -5680, 1974. 5.3, 9 Nnderssen, K. E. B.; "Pressure Drop in Ideal...5.6, 6, 9 Forest, C. A.: "Burning and Detonation," Los Alamos Scientific Laboratory, LA -7245, July 1978. 2, 3, 4 Fox, J.; "Flow Regimes in

Computer Programs for Producing Single-Event Aircraft Noise Data for Specific Engine Power and Meteorological Conditions for Use with USAF (United States Air Force) Community Noise Model (NOISEMAP).

DTIC Science & Technology

1983-04-01

containing the atmospheric ab- sorption coefficients in dB per 1000 for standard day temperature and relative humidity. ATN8(24) Atmospheric...t L% Ias0% ar.. r- N0%aO CD -F PFa -N CaN.F F-S I-.6 0CP-. NF O3rIl DN X C CP.4 (a.,MJ-Nft In0 0N M66 %L 0% P- .FK..4 N t-4 DC2 a OAa*A e Sn U fU 4 0

Insight into the theoretical and experimental studies of 1-phenyl-3-methyl-4-benzoyl-5-pyrazolone N(4)-methyl-N(4)- phenylthiosemicarbazone - A potential NLO material

NASA Astrophysics Data System (ADS)

Sangeetha, K. G.; Aravindakshan, K. K.; Safna Hussan, K. P.

2017-12-01

The synthesis, geometrical parameters, spectroscopic studies, optimised molecular structure, vibrational analysis, Mullikan population analysis, MEP, NBO, frontier molecular orbitals and NLO effects of 1-phenyl-3-methyl-4-benzoyl-5-pyrazolone N-(4)-methyl-N-(4)-phenylthiosemicarbazone, C25H23N5OS (L1) have been communicated in this paper. A combined experimental and theoretical approach was used to explore the structure and properties of the compound. For computational studies, Gaussian 09 program was used. Starting geometry of molecule was taken from X-ray refinement data and has been optimized by using DFT (B3LYP) method with the 6-31+G (d, p) basis sets. NBO analysis gave insight into the strongly delocalized structure, responsible for the nonlinearity and hence the stability of the molecule. Frontier molecular orbitals have been defined to forecast the global reactivity descriptors of L1. The computed first-order hyperpolarizability (β) of the compound is 2 times higher than that of urea and this account for its nonlinear optical property. Simultaneously, a molecular docking study of the compound was performed using GLIDE Program. For this, three biological enzymes, histone deacetylase, ribonucleotide reductase and DNA methyl transferase, were selected as receptor molecules.

Fast computation of close-coupling exchange integrals using polynomials in a tree representation

NASA Astrophysics Data System (ADS)

Wallerberger, Markus; Igenbergs, Katharina; Schweinzer, Josef; Aumayr, Friedrich

2011-03-01

The semi-classical atomic-orbital close-coupling method is a well-known approach for the calculation of cross sections in ion-atom collisions. It strongly relies on the fast and stable computation of exchange integrals. We present an upgrade to earlier implementations of the Fourier-transform method. For this purpose, we implement an extensive library for symbolic storage of polynomials, relying on sophisticated tree structures to allow fast manipulation and numerically stable evaluation. Using this library, we considerably speed up creation and computation of exchange integrals. This enables us to compute cross sections for more complex collision systems. Program summaryProgram title: TXINT Catalogue identifier: AEHS_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEHS_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 12 332 No. of bytes in distributed program, including test data, etc.: 157 086 Distribution format: tar.gz Programming language: Fortran 95 Computer: All with a Fortran 95 compiler Operating system: All with a Fortran 95 compiler RAM: Depends heavily on input, usually less than 100 MiB Classification: 16.10 Nature of problem: Analytical calculation of one- and two-center exchange matrix elements for the close-coupling method in the impact parameter model. Solution method: Similar to the code of Hansen and Dubois [1], we use the Fourier-transform method suggested by Shakeshaft [2] to compute the integrals. However, we heavily speed up the calculation using a library for symbolic manipulation of polynomials. Restrictions: We restrict ourselves to a defined collision system in the impact parameter model. Unusual features: A library for symbolic manipulation of polynomials, where polynomials are stored in a space-saving left-child right-sibling binary tree. This provides stable numerical evaluation and fast mutation while maintaining full compatibility with the original code. Additional comments: This program makes heavy use of the new features provided by the Fortran 90 standard, most prominently pointers, derived types and allocatable structures and a small portion of Fortran 95. Only newer compilers support these features. Following compilers support all features needed by the program. GNU Fortran Compiler "gfortran" from version 4.3.0 GNU Fortran 95 Compiler "g95" from version 4.2.0 Intel Fortran Compiler "ifort" from version 11.0

The Berry Informational Technology (B.I.T.S.) Student Work Program: An Effective Environment for Collaborative Learning, Leadership, Technological Training, and Certification.

ERIC Educational Resources Information Center

Cornelius, Amy; Macaluso, Paul

The Berry Informational Technology (B.I.T.S.) program at Berry College (Georgia) is an apprenticeship opportunity associated with student work. The program gives students the opportunity to seek technological training in areas, such as building computer systems, trouble-shooting, networking, Web development, and user and technical support. In…

Monte Carlo dose calculations of beta-emitting sources for intravascular brachytherapy: a comparison between EGS4, EGSnrc, and MCNP.

PubMed

Wang, R; Li, X A

2001-02-01

The dose parameters for the beta-particle emitting 90Sr/90Y source for intravascular brachytherapy (IVBT) have been calculated by different investigators. At a distant distance from the source, noticeable differences are seen in these parameters calculated using different Monte Carlo codes. The purpose of this work is to quantify as well as to understand these differences. We have compared a series of calculations using an EGS4, an EGSnrc, and the MCNP Monte Carlo codes. Data calculated and compared include the depth dose curve for a broad parallel beam of electrons, and radial dose distributions for point electron sources (monoenergetic or polyenergetic) and for a real 90Sr/90Y source. For the 90Sr/90Y source, the doses at the reference position (2 mm radial distance) calculated by the three code agree within 2%. However, the differences between the dose calculated by the three codes can be over 20% in the radial distance range interested in IVBT. The difference increases with radial distance from source, and reaches 30% at the tail of dose curve. These differences may be partially attributed to the different multiple scattering theories and Monte Carlo models for electron transport adopted in these three codes. Doses calculated by the EGSnrc code are more accurate than those by the EGS4. The two calculations agree within 5% for radial distance <6 mm.

Computer Program for Steady Transonic Flow over Thin Airfoils by Finite Elements

DTIC Science & Technology

1975-10-01

COMPUTER PROGRAM FOR STEADY JJ TRANSONIC FLOW OVER THIN AIRFOILS BY g FINITE ELEMENTS • *q^^ r ̂ c HUNTSVILLE RESEARCH & ENGINEERING CENTER...jglMMi B Jun’ INC ORGANIMTION NAME ANO ADDRESS Lö^kfteed Missiles & Space Company, Inc. Huntsville Research & Engineering Center,^ Huntsville, Alab...This report was prepared by personnel in the Computational Mechamcs Section of the Lockheed Missiles fc Space Company, Inc.. Huntsville Research

Synthesis, characterization and vibrational spectra analysis of ethyl (2Z)-2-(2-amino-4-oxo-1,3-oxazol-5(4H)-ylidene)-3-oxo-3-phenylpropanoate.

PubMed

Kıbrız, Ibrahim Evren; Sert, Yusuf; Saçmacı, Mustafa; Sahin, Ertan; Yıldırım, Ismail; Ucun, Fatih

2013-10-01

In the present study, the experimental and theoretical vibrational spectra of ethyl (2Z)-2-(2-amino-4-oxo-1,3-oxazol-5(4H)-ylidene)-3-oxo-3-phenylpropanoate (AOX) were investigated. The experimental FT-IR (400-4000 cm(-1)) and Laser-Raman spectra (100-4000 cm(-1)) of the molecule in the solid phase were recorded. Theoretical vibrational frequencies and geometric parameters (bond lengths, bond angles and torsion angles) were calculated using ab initio Hartree Fock (HF), Density Functional Theory (B3LYP and B3PW91) methods with 6-311++G(d,p) basis set by Gaussian 03 program, for the first time. The computed values of frequencies are scaled using a suitable scale factor to yield good coherence with the observed values. The assignments of the vibrational frequencies were performed by potential energy distribution (PED) analysis by using VEDA 4 program. The theoretical optimized geometric parameters and vibrational frequencies were compared with the corresponding experimental X-ray diffraction data, and they were seen to be in a good agreement with each other. The hydrogen bonding geometry of the molecule was also simulated to evaluate the effect of intermolecular hydrogen bonding on the vibrational frequencies. Also, the highest occupied molecular orbital (HOMO) and lowest unoccupied molecular orbital (LUMO) energies were found. Published by Elsevier B.V.

Los Alamos radiation transport code system on desktop computing platforms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Briesmeister, J.F.; Brinkley, F.W.; Clark, B.A.

The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. These codes were originally developed many years ago and have undergone continual improvement. With a large initial effort and continued vigilance, the codes are easily portable from one type of hardware to another. The performance of scientific work-stations (SWS) has evolved to the point that such platforms can be used routinely to perform sophisticated radiation transport calculations. As the personal computer (PC) performance approaches that of the SWS, the hardware options for desk-top radiation transport calculations expands considerably. Themore » current status of the radiation transport codes within the LARTCS is described: MCNP, SABRINA, LAHET, ONEDANT, TWODANT, TWOHEX, and ONELD. Specifically, the authors discuss hardware systems on which the codes run and present code performance comparisons for various machines.« less

PLOT3D/AMES, GENERIC UNIX VERSION USING DISSPLA (WITH TURB3D)

NASA Technical Reports Server (NTRS)

Buning, P.

1994-01-01

PLOT3D is an interactive graphics program designed to help scientists visualize computational fluid dynamics (CFD) grids and solutions. Today, supercomputers and CFD algorithms can provide scientists with simulations of such highly complex phenomena that obtaining an understanding of the simulations has become a major problem. Tools which help the scientist visualize the simulations can be of tremendous aid. PLOT3D/AMES offers more functions and features, and has been adapted for more types of computers than any other CFD graphics program. Version 3.6b+ is supported for five computers and graphic libraries. Using PLOT3D, CFD physicists can view their computational models from any angle, observing the physics of problems and the quality of solutions. As an aid in designing aircraft, for example, PLOT3D's interactive computer graphics can show vortices, temperature, reverse flow, pressure, and dozens of other characteristics of air flow during flight. As critical areas become obvious, they can easily be studied more closely using a finer grid. PLOT3D is part of a computational fluid dynamics software cycle. First, a program such as 3DGRAPE (ARC-12620) helps the scientist generate computational grids to model an object and its surrounding space. Once the grids have been designed and parameters such as the angle of attack, Mach number, and Reynolds number have been specified, a "flow-solver" program such as INS3D (ARC-11794 or COS-10019) solves the system of equations governing fluid flow, usually on a supercomputer. Grids sometimes have as many as two million points, and the "flow-solver" produces a solution file which contains density, x- y- and z-momentum, and stagnation energy for each grid point. With such a solution file and a grid file containing up to 50 grids as input, PLOT3D can calculate and graphically display any one of 74 functions, including shock waves, surface pressure, velocity vectors, and particle traces. PLOT3D's 74 functions are organized into five groups: 1) Grid Functions for grids, grid-checking, etc.; 2) Scalar Functions for contour or carpet plots of density, pressure, temperature, Mach number, vorticity magnitude, helicity, etc.; 3) Vector Functions for vector plots of velocity, vorticity, momentum, and density gradient, etc.; 4) Particle Trace Functions for rake-like plots of particle flow or vortex lines; and 5) Shock locations based on pressure gradient. TURB3D is a modification of PLOT3D which is used for viewing CFD simulations of incompressible turbulent flow. Input flow data consists of pressure, velocity and vorticity. Typical quantities to plot include local fluctuations in flow quantities and turbulent production terms, plotted in physical or wall units. PLOT3D/TURB3D includes both TURB3D and PLOT3D because the operation of TURB3D is identical to PLOT3D, and there is no additional sample data or printed documentation for TURB3D. Graphical capabilities of PLOT3D version 3.6b+ vary among the implementations available through COSMIC. Customers are encouraged to purchase and carefully review the PLOT3D manual before ordering the program for a specific computer and graphics library. There is only one manual for use with all implementations of PLOT3D, and although this manual generally assumes that the Silicon Graphics Iris implementation is being used, informative comments concerning other implementations appear throughout the text. With all implementations, the visual representation of the object and flow field created by PLOT3D consists of points, lines, and polygons. Points can be represented with dots or symbols, color can be used to denote data values, and perspective is used to show depth. Differences among implementations impact the program's ability to use graphical features that are based on 3D polygons, the user's ability to manipulate the graphical displays, and the user's ability to obtain alternate forms of output. The UNIX/DISSPLA implementation of PLOT3D supports 2-D polygons as well as 2-D and 3-D lines, but does not support graphics features requiring 3-D polygons (shading and hidden line removal, for example). Views can be manipulated using keyboard commands. This version of PLOT3D is potentially able to produce files for a variety of output devices; however, site-specific capabilities will vary depending on the device drivers supplied with the user's DISSPLA library. The version 3.6b+ UNIX/DISSPLA implementations of PLOT3D (ARC-12788) and PLOT3D/TURB3D (ARC-12778) were developed for use on computers running UNIX SYSTEM 5 with BSD 4.3 extensions. The standard distribution media for each ofthese programs is a 9track, 6250 bpi magnetic tape in TAR format. Customers purchasing one implementation version of PLOT3D or PLOT3D/TURB3D will be given a $200 discount on each additional implementation version ordered at the same time. Version 3.6b+ of PLOT3D and PLOT3D/TURB3D are also supported for the following computers and graphics libraries: (1) generic UNIX Supercomputer and IRIS, suitable for CRAY 2/UNICOS, CONVEX, Alliant with remote IRIS 2xxx/3xxx or IRIS 4D (ARC-12779, ARC-12784); (2) Silicon Graphics IRIS 2xxx/3xxx or IRIS 4D (ARC-12783, ARC-12782); (3) VAX computers running VMS Version 5.0 and DISSPLA Version 11.0 (ARC-12777, ARC-12781); and (4) Apollo computers running UNIX and GMR3D Version 2.0 (ARC-12789, ARC-12785 which have no capabilities to put text on plots). Silicon Graphics Iris, IRIS 4D, and IRIS 2xxx/3xxx are trademarks of Silicon Graphics Incorporated. VAX and VMS are trademarks of Digital Electronics Corporation. DISSPLA is a trademark of Computer Associates. CRAY 2 and UNICOS are trademarks of CRAY Research, Incorporated. CONVEX is a trademark of Convex Computer Corporation. Alliant is a trademark of Alliant. Apollo and GMR3D are trademarks of Hewlett-Packard, Incorporated. System 5 is a trademark of Bell Labs, Incorporated. BSD4.3 is a trademark of the University of California at Berkeley. UNIX is a registered trademark of AT&T.

PLOT3D/AMES, GENERIC UNIX VERSION USING DISSPLA (WITHOUT TURB3D)

NASA Technical Reports Server (NTRS)

Buning, P.

1994-01-01

PLOT3D is an interactive graphics program designed to help scientists visualize computational fluid dynamics (CFD) grids and solutions. Today, supercomputers and CFD algorithms can provide scientists with simulations of such highly complex phenomena that obtaining an understanding of the simulations has become a major problem. Tools which help the scientist visualize the simulations can be of tremendous aid. PLOT3D/AMES offers more functions and features, and has been adapted for more types of computers than any other CFD graphics program. Version 3.6b+ is supported for five computers and graphic libraries. Using PLOT3D, CFD physicists can view their computational models from any angle, observing the physics of problems and the quality of solutions. As an aid in designing aircraft, for example, PLOT3D's interactive computer graphics can show vortices, temperature, reverse flow, pressure, and dozens of other characteristics of air flow during flight. As critical areas become obvious, they can easily be studied more closely using a finer grid. PLOT3D is part of a computational fluid dynamics software cycle. First, a program such as 3DGRAPE (ARC-12620) helps the scientist generate computational grids to model an object and its surrounding space. Once the grids have been designed and parameters such as the angle of attack, Mach number, and Reynolds number have been specified, a "flow-solver" program such as INS3D (ARC-11794 or COS-10019) solves the system of equations governing fluid flow, usually on a supercomputer. Grids sometimes have as many as two million points, and the "flow-solver" produces a solution file which contains density, x- y- and z-momentum, and stagnation energy for each grid point. With such a solution file and a grid file containing up to 50 grids as input, PLOT3D can calculate and graphically display any one of 74 functions, including shock waves, surface pressure, velocity vectors, and particle traces. PLOT3D's 74 functions are organized into five groups: 1) Grid Functions for grids, grid-checking, etc.; 2) Scalar Functions for contour or carpet plots of density, pressure, temperature, Mach number, vorticity magnitude, helicity, etc.; 3) Vector Functions for vector plots of velocity, vorticity, momentum, and density gradient, etc.; 4) Particle Trace Functions for rake-like plots of particle flow or vortex lines; and 5) Shock locations based on pressure gradient. TURB3D is a modification of PLOT3D which is used for viewing CFD simulations of incompressible turbulent flow. Input flow data consists of pressure, velocity and vorticity. Typical quantities to plot include local fluctuations in flow quantities and turbulent production terms, plotted in physical or wall units. PLOT3D/TURB3D includes both TURB3D and PLOT3D because the operation of TURB3D is identical to PLOT3D, and there is no additional sample data or printed documentation for TURB3D. Graphical capabilities of PLOT3D version 3.6b+ vary among the implementations available through COSMIC. Customers are encouraged to purchase and carefully review the PLOT3D manual before ordering the program for a specific computer and graphics library. There is only one manual for use with all implementations of PLOT3D, and although this manual generally assumes that the Silicon Graphics Iris implementation is being used, informative comments concerning other implementations appear throughout the text. With all implementations, the visual representation of the object and flow field created by PLOT3D consists of points, lines, and polygons. Points can be represented with dots or symbols, color can be used to denote data values, and perspective is used to show depth. Differences among implementations impact the program's ability to use graphical features that are based on 3D polygons, the user's ability to manipulate the graphical displays, and the user's ability to obtain alternate forms of output. The UNIX/DISSPLA implementation of PLOT3D supports 2-D polygons as well as 2-D and 3-D lines, but does not support graphics features requiring 3-D polygons (shading and hidden line removal, for example). Views can be manipulated using keyboard commands. This version of PLOT3D is potentially able to produce files for a variety of output devices; however, site-specific capabilities will vary depending on the device drivers supplied with the user's DISSPLA library. The version 3.6b+ UNIX/DISSPLA implementations of PLOT3D (ARC-12788) and PLOT3D/TURB3D (ARC-12778) were developed for use on computers running UNIX SYSTEM 5 with BSD 4.3 extensions. The standard distribution media for each ofthese programs is a 9track, 6250 bpi magnetic tape in TAR format. Customers purchasing one implementation version of PLOT3D or PLOT3D/TURB3D will be given a $200 discount on each additional implementation version ordered at the same time. Version 3.6b+ of PLOT3D and PLOT3D/TURB3D are also supported for the following computers and graphics libraries: (1) generic UNIX Supercomputer and IRIS, suitable for CRAY 2/UNICOS, CONVEX, Alliant with remote IRIS 2xxx/3xxx or IRIS 4D (ARC-12779, ARC-12784); (2) Silicon Graphics IRIS 2xxx/3xxx or IRIS 4D (ARC-12783, ARC-12782); (3) VAX computers running VMS Version 5.0 and DISSPLA Version 11.0 (ARC-12777, ARC-12781); and (4) Apollo computers running UNIX and GMR3D Version 2.0 (ARC-12789, ARC-12785 which have no capabilities to put text on plots). Silicon Graphics Iris, IRIS 4D, and IRIS 2xxx/3xxx are trademarks of Silicon Graphics Incorporated. VAX and VMS are trademarks of Digital Electronics Corporation. DISSPLA is a trademark of Computer Associates. CRAY 2 and UNICOS are trademarks of CRAY Research, Incorporated. CONVEX is a trademark of Convex Computer Corporation. Alliant is a trademark of Alliant. Apollo and GMR3D are trademarks of Hewlett-Packard, Incorporated. System 5 is a trademark of Bell Labs, Incorporated. BSD4.3 is a trademark of the University of California at Berkeley. UNIX is a registered trademark of AT&T.

36 CFR § 404.7 - Fees to be charged-general.

Code of Federal Regulations, 2013 CFR

2013-07-01

... operating statutory-based fee schedule programs (see definition in § 404.6(b)), such as the NTIS, ABMC...(s) making the search. (b) Computer searches for records. ABMC will charge at the actual direct cost.... For copies prepared by computer, such as tapes or printouts, ABMC shall charge the actual cost...

5 CFR 1303.40 - Fees to be charged-general.

Code of Federal Regulations, 2013 CFR

2013-01-01

... distribution by agencies operating statutory-based fee schedule programs (see definition in Sections 1303.30(b... percent) of the employee(s) making the search. (b) Computer searches for records. OMB will charge at the... page. For copies prepared by computer, such as tapes or printouts, OMB shall charge the actual cost...

5 CFR 1303.40 - Fees to be charged-general.

Code of Federal Regulations, 2014 CFR

2014-01-01

... distribution by agencies operating statutory-based fee schedule programs (see definition in Sections 1303.30(b... percent) of the employee(s) making the search. (b) Computer searches for records. OMB will charge at the... page. For copies prepared by computer, such as tapes or printouts, OMB shall charge the actual cost...

5 CFR 1303.40 - Fees to be charged-general.

Code of Federal Regulations, 2012 CFR

2012-01-01

... distribution by agencies operating statutory-based fee schedule programs (see definition in Sections 1303.30(b... percent) of the employee(s) making the search. (b) Computer searches for records. OMB will charge at the... page. For copies prepared by computer, such as tapes or printouts, OMB shall charge the actual cost...

36 CFR 404.7 - Fees to be charged-general.

Code of Federal Regulations, 2014 CFR

2014-07-01

... operating statutory-based fee schedule programs (see definition in § 404.6(b)), such as the NTIS, ABMC...(s) making the search. (b) Computer searches for records. ABMC will charge at the actual direct cost.... For copies prepared by computer, such as tapes or printouts, ABMC shall charge the actual cost...

36 CFR 404.7 - Fees to be charged-general.

Code of Federal Regulations, 2012 CFR

2012-07-01

... operating statutory-based fee schedule programs (see definition in § 404.6(b)), such as the NTIS, ABMC...(s) making the search. (b) Computer searches for records. ABMC will charge at the actual direct cost.... For copies prepared by computer, such as tapes or printouts, ABMC shall charge the actual cost...

Measurement of leakage neutron spectra from graphite cylinders irradiated with D-T neutrons for validation of evaluated nuclear data.

PubMed

Luo, F; Han, R; Chen, Z; Nie, Y; Shi, F; Zhang, S; Lin, W; Ren, P; Tian, G; Sun, Q; Gou, B; Ruan, X; Ren, J; Ye, M

2016-10-01

A benchmark experiment for validation of graphite data evaluated from nuclear data libraries was conducted for 14MeV neutrons irradiated on graphite cylinder samples. The experiments were performed using the benchmark experimental facility at the China Institute of Atomic Energy (CIAE). The leakage neutron spectra from the surface of graphite (Φ13cm×20cm) at 60° and 120° and graphite (Φ13cm×2cm) at 60° were measured by the time-of-flight (TOF) method. The obtained results were compared with the measurements made by the Monte Carlo neutron transport code MCNP-4C with the ENDF/B-VII.1, CENDL-3.1 and JENDL-4.0 libraries. The results obtained from a 20cm-thick sample revealed that the calculation results with CENDL-3.1 and JENDL-4.0 libraries showed good agreements with the experiments conducted in the whole energy region. However, a large discrepancy of approximately 40% was observed below the 3MeV energy region with the ENDF/B-VII.1 library. For the 2cm-thick sample, the calculated results obtained from the abovementioned three libraries could not reproduce the experimental data in the energy range of 5-7MeV. The graphite data in CENDL-3.1 were verified for the first time and were proved to be reliable. Copyright © 2016 Elsevier Ltd. All rights reserved.

Space shuttle environmental and thermal control life support system computer program

NASA Technical Reports Server (NTRS)

1972-01-01

A computer program for the design and operation of the space shuttle environmental and thermal control life support system is presented. The subjects discussed are: (1) basic optimization program, (2) off design performance, (3) radiator/evaporator expendable usage, (4) component weights, and (5) computer program operating procedures.

Arbitrator Evaluation and Selection: A Policy Capturing Approach.

DTIC Science & Technology

1980-09-01

b. M -years S (In-house). 4. Often it is not possible to attach equivalent dollar values to research, although the results of the research may, in...149 M . WILLIAMS-MODIFIED MODEL FULL SEVEN CUE REGRESSION MODEL ... ............ 152 N. COMPUTER PROGRAM TO TEST FOR POLICY DIFFERENCES...idams Management Rights .188* .320 .098 .112 2.939 2.693 C3ontract Language * Past Practice M * ** Fairness ** .195 ** .095 2.982 Effect on the Worker R2

An Interactive Computer Program for the Preliminary Design and Analysis of Marine Reduction Gears.

DTIC Science & Technology

1982-03-01

4 E-4 c 043 04 4 E- PQ = 24 0 0 E- m 2054 4a =44 Env)2 d0 a b43 09 0 4 I6.4 44 = w. 0k2. . U)3c m 00 I= =C * 4 0 . O)U ad i 0 0~P In3 U= 4 40:M = 4 m...pa *Q4 2 c4~ .~ co N ra a9.* o H 00a = = :b4 H~ H U (3 C) U- of *4= t. E- H)40 UUO *n04 043 q N. W E-4 E-4 1.4-0 31 H 000 oi-4. H4 - M0 2I 2 04 E- HU...aO o %-c ’. %D in0 o f 3W -+V E4pa a noM pa pa pCa 0 pa H14 . pa H 1.4 pa ~00 MtH x 4 -4 U 4E4E4 0 E-4 H E ~ CaM N4 -40 0 UE- E.UE- u 0 c E4 0 9 E-4

Assessment of two-temperature kinetic model for dissociating and weakly-ionizing nitrogen

NASA Technical Reports Server (NTRS)

Park, C.

1986-01-01

The validity of the author's two-temperature, chemical/kinetic model which the author has recently improved is assessed by comparing the calculated results with the existing experimental data for nitrogen in the dissociating and weakly ionizing regime produced behind a normal shock wave. The computer program Shock Tube Radiation Program (STRAP) based on the two-temperature model is used in calculating the flow properties behind the shock wave and the Nonequilibrium Air Radiation (NEQAIR) program, in determining the radiative characteristics of the flow. Both programs were developed earlier. Comparison is made between the calculated and the existing shock tube data on (1) spectra in the equilibrium region, (2) rotational temperature of the N2(+) B state, (3) vibrational temperature of the N2(+) B state, (4) electronic excitation temperature of the N2 B state, (5) the shape of time-variation of radiation intensities, (6) the times to reach the peak in radiation intensity and equilibrium, and (7) the ratio of nonequilibrium to equilibrium radiative heat fluxes. Good agreement is seen between the experimental data and the present calculation except for the vibrational temperature. A possible reason for the discrepancy is given.

A study of low-cost reliable actuators for light aircraft. Part B: Appendices

NASA Technical Reports Server (NTRS)

Eijsink, H.; Rice, M.

1978-01-01

Computer programs written in FORTRAN are given for time response calculations on pneumatic and linear hydraulic actuators. The programs are self-explanatory with comment statements. Program output is also included.

10 CFR 605.5 - The Office of Energy Research Financial Assistance Program.

Code of Federal Regulations, 2012 CFR

2012-01-01

...) Scientific Computing Staff (7) Superconducting Super Collider (8) University and Science Education Programs... appendix A of this part. (b) The Program areas are: (1) Basic Energy Sciences (2) Field Operations...

10 CFR 605.5 - The Office of Energy Research Financial Assistance Program.

Code of Federal Regulations, 2013 CFR

2013-01-01

...) Scientific Computing Staff (7) Superconducting Super Collider (8) University and Science Education Programs... appendix A of this part. (b) The Program areas are: (1) Basic Energy Sciences (2) Field Operations...

10 CFR 605.5 - The Office of Energy Research Financial Assistance Program.

Code of Federal Regulations, 2011 CFR

2011-01-01

...) Scientific Computing Staff (7) Superconducting Super Collider (8) University and Science Education Programs... appendix A of this part. (b) The Program areas are: (1) Basic Energy Sciences (2) Field Operations...

10 CFR 605.5 - The Office of Energy Research Financial Assistance Program.

Code of Federal Regulations, 2014 CFR

2014-01-01

...) Scientific Computing Staff (7) Superconducting Super Collider (8) University and Science Education Programs... appendix A of this part. (b) The Program areas are: (1) Basic Energy Sciences (2) Field Operations...

General Chemistry Collection for Students, 6th Edition

NASA Astrophysics Data System (ADS)

2002-05-01

System requirements are given in Tables 2a and b. Some programs have additional special requirements. Please see the individual program abstracts at JCE Online or the documentation included on the CD-ROM for more specific information.