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

Energy Technology Data Exchange (ETDEWEB)

Ay, M R [Department of Physics and Nuclear Sciences, AmirKabir University of Technology, Tehran (Iran, Islamic Republic of); Shahriari, M [Department of Nuclear Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of); Sarkar, S [Department of Medical Physics, Tehran University of Medical Science, Tehran (Iran, Islamic Republic of); Adib, M [TPP Co., GE Medical Systems, Iran Authorized Distributor, Tehran (Iran, Islamic Republic of); Zaidi, H [Division of Nuclear Medicine, Geneva University Hospital, 1211 Geneva (Switzerland)

2004-11-07

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.

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

Science.gov (United States)

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.

GB - a preliminary linking code between MCNP4C and Origen2.1 - DEN/UFMG version

International Nuclear Information System (INIS)

Campolina, Daniel; Pereira, Claubia; Veloso, Maria Auxiliadora Fortini; Cavatoni, Andre

2009-01-01

Nowadays it is possible to perform burnup simulation in a detailed 3D geometry and a continuous energy description by the Monte Carlo method. This paper describes an initial project to create and verify a connection code to link Origen2.1 (Oak Ridge National Laboratory) and MCNP4C (Los Alamos National Laboratory). Essentially the code includes point depletion capability to the MCNP code. The incorporation of point depletion capability is explicit and can be summarized by three steps: 1-Monte Carlo determines reaction rates, 2-the reaction rates are used to determine microscopic cross sections for depletion equations, 3-solution of depletion equations (given by Origen2.1) determines number densities for next MCNP step. To evaluate the initial version of the program, we focused on comparing the results with one of the major Monte Carlo burnup codes: MCNPX version 2.6.0. The input files for all codes share the same MCNP geometry, nuclear data library and core thermal power. While simulating 75 time steps at 800 kw of a Heat Pipe Power System model, we have found that the codes generate very similar results. The neutron flux and criticality value of the core agree, especially in the begin of burnup when the influence of fission products are not very considerable. The small difference encountered was probably caused by the difference in the number of isotopes considered in the transport models (89 MCNPX x 25 GB (author)

MCNP(trademark) Version 5

International Nuclear Information System (INIS)

Cox, Lawrence J.; Barrett, Richard F.; Booth, Thomas Edward; Briesmeister, Judith F.; Brown, Forrest B.; Bull, Jeffrey S.; Giesler, Gregg Carl; Goorley, John T.; Mosteller, Russell D.; Forster, R. Arthur; Post, Susan E.; Prael, Richard E.; Selcow, Elizabeth Carol; Sood, Avneet

2002-01-01

The Monte Carlo transport workhorse, MCNP, is undergoing a massive renovation at Los Alamos National Laboratory (LANL) in support of the Eolus Project of the Advanced Simulation and Computing (ASCI) Program. MCNP Version 5 (V5) (expected to be released to RSICC in Spring, 2002) will consist of a major restructuring from FORTRAN-77 (with extensions) to ANSI-standard FORTRAN-90 with support for all of the features available in the present release (MCNP-4C2/4C3). To most users, the look-and-feel of MCNP will not change much except for the improvements (improved graphics, easier installation, better online documentation). For example, even with the major format change, full support for incremental patching will still be provided. In addition to the language and style updates, MCNP V5 will have various new user features. These include improved photon physics, neutral particle radiography, enhancements and additions to variance reduction methods, new source options, and improved parallelism support (PVM, MPI, OpenMP).

KENO2MCNP, Version 5L, Conversion of Input Data between KENOV.a and MCNP File Formats

International Nuclear Information System (INIS)

2008-01-01

1 - Description of program or function: The KENO2MCNP program was written to convert KENO V.a input files to MCNP Format. This program currently only works with KENO Va geometries and will not work with geometries that contain more than a single array. A C++ graphical user interface was created that was linked to Fortran routines from KENO V.a that read the material library and Fortran routines from the MCNP Visual Editor that generate the MCNP input file. Either SCALE 5.0 or SCALE 5.1 cross section files will work with this release. 2 - Methods: The C++ binary executable reads the KENO V.a input file, the KENO V.a material library and SCALE data libraries. When an input file is read in, the input is stored in memory. The converter goes through and loads different sections of the input file into memory including parameters, composition, geometry information, array information and starting information. Many of the KENO V.a materials represent compositions that must be read from the KENO V.a material library. KENO2MCNP includes the KENO V.a FORTRAN routines used to read this material file for creating the MCNP materials. Once the file has been read in, the user must select 'Convert' to convert the file from KENO V.a to MCNP. This will generate the MCNP input file along with an output window that lists the KENO V.a composition information for the materials contained in the KENO V.a input file. The program can be run interactively by clicking on the executable or in batch mode from the command prompt. 3 - Restrictions on the complexity of the problem: Not all KENO V.a input files are supported. Only one array is allowed in the input file. Some of the more complex material descriptions also may not be converted

Gamma ray shielding study of barium-bismuth-borosilicate glasses as transparent shielding materials using MCNP-4C code, XCOM program, and available experimental data

Energy Technology Data Exchange (ETDEWEB)

Bagheri, Reza; Yousefinia, Hassan [Nuclear Fuel Cycle Research School (NFCRS), Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran, Tehran (Iran, Islamic Republic of); Moghaddam, Alireza Khorrami [Radiology Department, Paramedical Faculty, Mazandaran University of Medical Sciences, Sari (Iran, Islamic Republic of)

2017-02-15

In this work, linear and mass attenuation coefficients, effective atomic number and electron density, mean free paths, and half value layer and 10th value layer values of barium-bismuth-borosilicate glasses were obtained for 662 keV, 1,173 keV, and 1,332 keV gamma ray energies using MCNP-4C code and XCOM program. Then obtained data were compared with available experimental data. The MCNP-4C code and XCOM program results were in good agreement with the experimental data. Barium-bismuth-borosilicate glasses have good gamma ray shielding properties from the shielding point of view.

Depleted Reactor Analysis With MCNP-4B

International Nuclear Information System (INIS)

Caner, M.; Silverman, L.; Bettan, M.

2004-01-01

Monte Carlo neutronics calculations are mostly done for fresh reactor cores. There is today an ongoing activity in the development of Monte Carlo plus burnup code systems made possible by the fast gains in computer processor speeds. In this work we investigate the use of MCNP-4B for the calculation of a depleted core of the Soreq reactor (IRR-1). The number densities as function of burnup were taken from the WIMS-D/4 cell code calculations. This particular code coupling has been implemented before. The Monte Carlo code MCNP-4B calculates the coupled transport of neutrons and photons for complicated geometries. We have done neutronics calculations of the IRR-1 core with the WIMS and CITATION codes in the past Also, we have developed an MCNP model of the IRR-1 standard fuel for a criticality safety calculation of a spent fuel storage pool

Calculation of the power distribution in the fuel rods of the low power research reactor using the MCNP4C code

International Nuclear Information System (INIS)

Dawahra, S.; Khattab, K.

2011-01-01

Highlights: → The MCNP4C code was used to calculate the power distribution in 3-D geometry in the MNSR reactor. → The maximum power of the individual rod was found in the fuel ring number 2 and was found to be 105 W. → The minimum power was found in the fuel ring number 9 and was 79.9 W. → The total power in the total fuel rods was 30.9 kW. - Abstract: The Monte Carlo method, using the MCNP4C code, was used in this paper to calculate the power distribution in 3-D geometry in the fuel rods of the Syrian Miniature Neutron Source Reactor (MNSR). To normalize the MCNP4C result to the steady state nominal thermal power, the appropriate scaling factor was defined to calculate the power distribution precisely. The maximum power of the individual rod was found in the fuel ring number 2 and was found to be 105 W. The minimum power was found in the fuel ring number 9 and was 79.9 W. The total power in the total fuel rods was 30.9 kW. This result agrees very well with nominal power reported in the reactor safety analysis report which equals 30 kW. Finally, the peak power factors, which are defined as the ratios between the maximum to the average and the maximum to the minimum powers were calculated to be 1.18 and 1.31 respectively.

MatMCNP: A Code for Producing Material Cards for MCNP

Energy Technology Data Exchange (ETDEWEB)

DePriest, Kendall Russell [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Saavedra, Karen C. [American Structurepoint, Inc., Indianapolis, IN (United States)

2014-09-01

A code for generating MCNP material cards (MatMCNP) has been written and verified for naturally occurring, stable isotopes. The program allows for material specification as either atomic or weight percent (fractions). MatMCNP also permits the specification of enriched lithium, boron, and/or uranium. In addition to producing the material cards for MCNP, the code calculates the atomic (or number) density in atoms/barn-cm as well as the multiplier that should be used to convert neutron and gamma fluences into dose in the material specified.

Gamma Ray Shielding Study of Barium–Bismuth–Borosilicate Glasses as Transparent Shielding Materials using MCNP-4C Code, XCOM Program, and Available Experimental Data

Directory of Open Access Journals (Sweden)

Reza Bagheri

2017-02-01

Full Text Available In this work, linear and mass attenuation coefficients, effective atomic number and electron density, mean free paths, and half value layer and 10th value layer values of barium–bismuth–borosilicate glasses were obtained for 662 keV, 1,173 keV, and 1,332 keV gamma ray energies using MCNP-4C code and XCOM program. Then obtained data were compared with available experimental data. The MCNP-4C code and XCOM program results were in good agreement with the experimental data. Barium–bismuth–borosilicate glasses have good gamma ray shielding properties from the shielding point of view.

Definition of neutron lifespan and neutron lifetime in MCNP4B

International Nuclear Information System (INIS)

Busch, R.D.; Spriggs, G.D.; Hendricks, J.S.

1997-01-01

MCNP4B was released in early 1997. In this new version, several major changes were made to the underlying theory used to estimate the non-adjoint-weighted removal, fission, capture, and escape prompt-neutron lifetimes. These four lifetimes are now being calculated in accordance to the neutron-balance theory described by Spriggs et al. in which the non-adjoint-weighted lifetime for a particular type of reaction (i.e., fission, capture, escape, removal, etc.) is defined as the total neutron population in the system divided by that reaction rate

Improvement of Monte Carlo code A3MCNP for large-scale shielding problems

International Nuclear Information System (INIS)

Miyake, Y.; Ohmura, M.; Hasegawa, T.; Ueki, K.; Sato, O.; Haghighat, A.; Sjoden, G.E.

2004-01-01

A 3 MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, that automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic 'importance' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A 3 MCNP uses the 3-D Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A 3 MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A 3 MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A 3 MCNP (referred to as A 3 MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A 3 MCNPV for a concrete cask streaming problem and a PWR dosimetry problem. (author)

MOCUP: MCNP-ORIGEN2 coupled utility program

International Nuclear Information System (INIS)

Moore, R.L.; Schnitzler, B.G.; Wemple, C.A.

1995-01-01

MOCUP is a system of external processors that allow for a limited treatment of the temporal composition of the user-selected MCNP cells in a time-dependent flux environment. The ORIGEN2 code computes the time-dependent compositions of these individually selected MCNP cells. All data communication between the two codes is accomplished through the MCNP and ORIGEN2 input/output files, the MOCUP Processor Output files, and two user supplied tables. MOCUP is either command line or interactively driven. The interactive interface is based on the portable XII window environment and the Motif tool kit. MOCUP was constructed so that no modifications to either MCNP or ORIGEN2 were necessary. Section 4 of the writeup contains the input instructions needed to set up the MOCUP run. MOCUP is extremely useful for analysts who perform isotope production, material transformation, and depletion and isotope analyses on complex, non-lattice geometries, and uniform and non-uniform lattices

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

Science.gov (United States)

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.

Analysis of parallel computing performance of the code MCNP

International Nuclear Information System (INIS)

Wang Lei; Wang Kan; Yu Ganglin

2006-01-01

Parallel computing can reduce the running time of the code MCNP effectively. With the MPI message transmitting software, MCNP5 can achieve its parallel computing on PC cluster with Windows operating system. Parallel computing performance of MCNP is influenced by factors such as the type, the complexity level and the parameter configuration of the computing problem. This paper analyzes the parallel computing performance of MCNP regarding with these factors and gives measures to improve the MCNP parallel computing performance. (authors)

Importance sampling techniques and treatment of electron transport in MCNP 4A

International Nuclear Information System (INIS)

Ueki, K.

1994-01-01

The continuous energy Monte Carlo code MCNP was developed by the Radiation Transport Group at Los Alamos National Laboratory and the MCNP 4A version is available, now. The MCNP 4A is able to do the coupled neutron-secondary gamma-ray-electron-bremsstrahlung calculation. The calculated results, such as energy spectra, tally fluctuation chart, and geometrical input data can be displayed by using a work station. The document of the MCNP 4A code has no description on the subroutines, except few ones of 'SOURCE', 'TALLYX'. However, when we want to improve the MCNP Monte Carlo sampling techniques to get more accuracy or efficiency results for some problems, some subroutines are required or needed to revised. Three subroutines have been revised and built in the MCNP 4A code. (author)

Comparisons between MCNP, EGS4 and experiment for clinical electron beams

International Nuclear Information System (INIS)

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

1999-01-01

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. (author)

Analysis of the variation of the attenuation curve in function of the radiation field size for k Vp X-ray beams using the MCNP-5C code

Energy Technology Data Exchange (ETDEWEB)

Fernandes, Marco A.R., E-mail: marco@cetea.com.b, E-mail: marfernandes@fmb.unesp.b [Universidade Estadual Paulista Julio de Mesquita Filho (FMB/UNESP), Botucatu, SP (Brazil). Fac. de Medicina; Ribeiro, Victor A.B. [Universidade Estadual Paulista Julio de Mesquita Filho (IBB/UNESP), Botucatu, SP (Brazil). Inst. de Biociencias; Viana, Rodrigo S.S.; Coelho, Talita S. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

The paper illustrates the use of the Monte Carlo method, MCNP-5C code, to analyze the attenuation curve behavior of the 50 kVp radiation beam from superficial radiotherapy equipment as Dermopan2 model. The simulations seek to verify the MCNP-5C code performance to study the variation of the attenuation curve - percentage depth dose (PDD) curve - in function of the radiation field dimension used at radiotherapy of skin tumors with 50 kVp X-ray beams. The PDD curve was calculated for six different radiation field sizes with circular geometry of 1.0, 2.0, 3.0, 4.0, 5.0 and 6.0 cm in diameter. The radiation source was modeled considering a tungsten target with inclination 30 deg, focal point of 6.5 mm in diameter and energy beam of 50 kVp; the X-ray spectrum was calculated with the MCNP-5C code adopting total filtration (beryllium window of 1 mm and aluminum additional filter of 1 mm). The PDD showed decreasing behavior with the attenuation depth similar what is presented on the literature. There was not significant variation at the PDD values for the radiation field between 1.0 and 4.0 cm in diameter. The differences increased for fields of 5.0 and 6.0 cm and at attenuation depth higher than 1.0 cm. When it is compared the PDD values for fields of 3.0 and 6.0 cm in diameter, it verifies the greater difference (12.6 %) at depth of 5.7 cm, proving the scattered radiation effect. The MCNP-5C code showed as an appropriate procedure to analyze the attenuation curves of the superficial radiotherapy beams. (author)

New developments enhancing MCNP for criticality safety

International Nuclear Information System (INIS)

Hendricks, J.S.; McKinney, G.W.; Forster, R.A.

1993-01-01

Since the early 80's MCNP has had three estimates of k eff : collision, absorption, and track length. MCNP has also had collision and absorption estimators of removal lifetime. These are calculated for every cycle and are averaged over the cycles as simple averages and covariance weighted averages. Correlation coefficients between estimators are also calculated. These criticality estimators are all in addition to the extensive summary information and tally edits used in shielding and other problems. A number of significant new developments have been made to enhance the MCNP Monte Carlo radiation transport code for criticality safety applications. These are available in the newly released MCNP4A version of the code

S values at voxels level for 188Re and 90Y calculated with the MCNP-4C code

International Nuclear Information System (INIS)

Coca, M.A.; Torres, L.A.; Cornejo, N.; Martin, G.

2008-01-01

Full text: MIRD formalism at voxel level has been suggested as an optional methodology to perform internal radiation dosimetry calculation during internal radiation therapy in Nuclear Medicine. Voxel S values for Y 90 , 131 I, 32 P, 99m Tc and 89 Sr have been published to different sizes. Currently, 188 Re has been proposed as a promising radionuclide for therapy due to its physical features and availability from generators. The main objective of this work was to estimate the voxel S values for 188 Re at cubical geometry using the MCNP-4C code for the simulations of radiation transport and energy deposition. Mean absorbed dose to target voxels per radioactive decay in a source voxel were estimated and reported for 188 Re and Y 90 . A comparison of voxel S values computed with the MCNP code and the data reported in MIRD Pamphlet 17 for 90 Y was performed in order to evaluate our results. (author)

Performance of the improved version of Monte Carlo Code A3MCNP for cask shielding design

International Nuclear Information System (INIS)

Hasegawa, T.; Ueki, K.; Sato, O.; Sjoden, G.E.; Miyake, Y.; Ohmura, M.; Haghighat, A.

2004-01-01

A 3 MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, that automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic ''importance'' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A 3 MCNP uses the 3-D Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A3MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A 3 MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A 3 MCNP (referred to as A 3 MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A 3 MCNPV for cask neutron and gamma-ray shielding problem

Calculation of the power distribution in the fuel rods of the low power research reactor using the MCNP4C code

International Nuclear Information System (INIS)

Dawahra, S.; Khattab, K.

2012-01-01

The Monte Carlo method, using the MCNP4C code, was used in this paper to calculate the power distribution in 3-D geometry in the fuel rods of the Syrian Miniature Neutron Source Reactor (MNSR). To normalize the MCNP4C result to the steady state nominal thermal power, the appropriate scaling factor was defined to calculate the power distribution precisely. The maximum power of the individual rod was found in the fuel ring number 2 and was found to be 105 W. The minimum power was found in the fuel ring number 9 and was 79.9 W. The total power in the total fuel rods was 30.9 k W. This result agrees very well with nominal power reported in the reactor safety analysis report which equals 30 k W. Finally, the peak power factors, which are defined as the ratios between the maximum to the average and the maximum to the minimum powers were calculated to be 1.18 and 1.31 respectively. (author)

DXRaySMCS. First user friendly interface developed for prediction of diagnostic radiology X-ray spectra produced by Monte Carlo (MCNP-4C) simulation in Iran

International Nuclear Information System (INIS)

Bahreyni Toossi, M.T.; Zare, H.; Moradi Faradanbe, H.

2008-01-01

An accurate knowledge of the output energy spectra of an x-ray tube is essential in many areas of radiological studies. It forms the basis of almost all image quality simulations and enable system designers to predict patient dose more accurately. Many radiological physics problems that can be solved by Monte Carlo simulation methods require an x-ray spectra as input data. Computer simulation of x-ray spectra is one of the most important tools for investigation of patient dose and image quality in diagnostic radiology systems. In this work the general purpose Monte Carlo N-particle radiation transport computer code (MCNP-4C) was used for the simulation of x-ray spectra in diagnostic radiology, Electron's path in the target was followed until it's energy was reduced to 10 keV. A user friendly interface named 'Diagnostic X-ray Spectra by Monte Carlo Simulation (DXRaySMCS)' was developed to facilitate the application of MCNP-4C code for diagnostic radiology spectrum prediction. The program provides a user friendly interface for modifying the MCNP input file, launching the MCNP program to simulate electron and photon transport and processing the MCNP output file to yield a summary of the results (Relative Photon Number per Energy Bin). In this article the development and characteristics of DXRaySMCS are outlined. As part of the validation process, out put spectra for 46 diagnostic radiology system settings produced by DXRaySMCS were compared with the corresponding IPEM78. Generally, there is a good agreement between the two sets of spectra. No statistically significant differences have been observed between IPEM78 reported spectra and the simulated spectra generated in this study. (author)

A comparison of MCNP4C electron transport with ITS 3.0 and experiment at incident energies between 100 keV and 20 MeV: influence of voxel size, substeps and energy indexing algorithm

International Nuclear Information System (INIS)

Schaart, Dennis R.; Jansen, Jan Th.M.; Zoetelief, Johannes; Leege, Piet F.A. de

2002-01-01

The condensed-history electron transport algorithms in the Monte Carlo code MCNP4C are derived from ITS 3.0, which is a well-validated code for coupled electron-photon simulations. This, combined with its user-friendliness and versatility, makes MCNP4C a promising code for medical physics applications. Such applications, however, require a high degree of accuracy. In this work, MCNP4C electron depth-dose distributions in water are compared with published ITS 3.0 results. The influences of voxel size, substeps and choice of electron energy indexing algorithm are investigated at incident energies between 100 keV and 20 MeV. Furthermore, previously published dose measurements for seven beta emitters are simulated. Since MCNP4C does not allow tally segmentation with the *F8 energy deposition tally, even a homogeneous phantom must be subdivided in cells to calculate the distribution of dose. The repeated interruption of the electron tracks at the cell boundaries significantly affects the electron transport. An electron track length estimator of absorbed dose is described which allows tally segmentation. In combination with the ITS electron energy indexing algorithm, this estimator appears to reproduce ITS 3.0 and experimental results well. If, however, cell boundaries are used instead of segments, or if the MCNP indexing algorithm is applied, the agreement is considerably worse. (author)

Validation of MCNP4A for repository scattered radiation analysis

International Nuclear Information System (INIS)

Haas, M.N.; Su, S.

1998-02-01

Comparison is made between experimentally determined albedo (scattered) radiation and MCNP4A predictions in order to provide independent validation for repository shielding analysis. Both neutron and gamma scattered radiation fields from concrete ducts are compared in this paper. Satisfactory agreement is found between actual and calculated results with conservative values calculated by the MCNP4A code for all conditions

The application of weight windows to 'Global' Monte Carlo problems

International Nuclear Information System (INIS)

Becker, T. L.; Larsen, E. W.

2009-01-01

This paper describes two basic types of global deep-penetration (shielding) problems-the global flux problem and the global response problem. For each of these, two methods for generating weight windows are presented. The first approach, developed by the authors of this paper and referred to generally as the Global Weight Window, constructs a weight window that distributes Monte Carlo particles according to a user-specified distribution. The second approach, developed at Oak Ridge National Laboratory and referred to as FW-CADIS, constructs a weight window based on intuitively extending the concept of the source-detector problem to global problems. The numerical results confirm that the theory used to describe the Monte Carlo particle distribution for a given weight window is valid and that the figure of merit is strongly correlated to the Monte Carlo particle distribution. Furthermore, they illustrate that, while both methods are capable of obtaining the correct solution, the Global Weight Window distributes particles much more uniformly than FW-CADIS. As a result, the figure of merit is higher for the Global Weight Window. (authors)

MCNP trademark directions

International Nuclear Information System (INIS)

Hendricks, J.S.

1994-01-01

The MCNP code development program is a relatively large and rapidly changing project in the small and highly-specialized field of radiation transport, specifically radiation protection and shielding. A number of major new MCNP initiatives are described in the subsequent papers in this session. The focus of this paper is the important new developments not described elsewhere and a number of recent developments that have been available since MCNP4A but have gone unnoticed. In particular, we report for the first time a new MCNP quality assurance initiative providing 97% test coverage, a new MCNP feature enabling plotting of nuclear data, and the other new features developed so far for MCNP4B. Finally, an attempt is made to articulate how all these fit together into the overall MCNP development program

MCNP speed advances for boron neutron capture therapy

International Nuclear Information System (INIS)

Goorley, J.T.; McKinney, G.; Adams, K.; Estes, G.

1998-04-01

The Boron Neutron Capture Therapy (BNCT) treatment planning process of the Beth Israel Deaconess Medical Center-M.I.T team relies on MCNP to determine dose rates in the subject's head for various beam orientations. In this time consuming computational process, four or five potential beams are investigated. Of these, one or two final beams are selected and thoroughly evaluated. Recent advances greatly decreased the time needed to do these MCNP calculations. Two modifications to the new MCNP4B source code, lattice tally and tracking enhancements, reduced the wall-clock run times of a typical one million source neutrons run to one hour twenty five minutes on a 200 MHz Pentium Pro computer running Linux and using the GNU FORTRAN compiler. Previously these jobs used a special version of MCNP4AB created by Everett Redmond, which completed in two hours two minutes. In addition to this 30% speedup, the MCNP4B version was adapted for use with Parallel Virtual Machine (PVM) on personal computers running the Linux operating system. MCNP, using PVM, can be run on multiple computers simultaneously, offering a factor of speedup roughly the same as the number of computers used. With two 200 MHz Pentium Pro machines, the run time was reduced to forty five minutes, a 1.9 factor of improvement over the single Linux computer. While the time of a single run was greatly reduced, the advantages associated with PVM derive from using computational power not already used. Four possible beams, currently requiring four separate runs, could be run faster when each is individually run on a single machine under Windows NT, rather than using Linux and PVM to run one after another with each multiprocessed across four computers. It would be advantageous, however, to use PVM to distribute the final two beam orientations over four computers

Lecture Notes on Criticality Safety Validation Using MCNP & Whisper

Energy Technology Data Exchange (ETDEWEB)

Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-03-11

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 usage are discussed.

Parallelization of MCNP4 code by using simple FORTRAN algorithms

International Nuclear Information System (INIS)

Yazid, P.I.; Takano, Makoto; Masukawa, Fumihiro; Naito, Yoshitaka.

1993-12-01

Simple FORTRAN algorithms, that rely only on open, close, read and write statements, together with disk files and some UNIX commands have been applied to parallelization of MCNP4. The code, named MCNPNFS, maintains almost all capabilities of MCNP4 in solving shielding problems. It is able to perform parallel computing on a set of any UNIX workstations connected by a network, regardless of the heterogeneity in hardware system, provided that all processors produce a binary file in the same format. Further, it is confirmed that MCNPNFS can be executed also on Monte-4 vector-parallel computer. MCNPNFS has been tested intensively by executing 5 photon-neutron benchmark problems, a spent fuel cask problem and 17 sample problems included in the original code package of MCNP4. Three different workstations, connected by a network, have been used to execute MCNPNFS in parallel. By measuring CPU time, the parallel efficiency is determined to be 58% to 99% and 86% in average. On Monte-4, MCNPNFS has been executed using 4 processors concurrently and has achieved the parallel efficiency of 79% in average. (author)

Dosimetry analysis of distributions radials dose profiles of {sup 90}Sr + {sup 90}Y beta therapy applicators using the MCNP-4C code and radio chromium films; Analise dosimetrica de perfis de distribuicoes radias de doses relativas de um aplicador de betaterapia de {sup 90}Sr + {sup 90}Y utilizando o codigo MCNP-4C e filmes radiocromicos

Energy Technology Data Exchange (ETDEWEB)

Coelho, Talita S.; Yoriyaz, Helio [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Fernandes, Marco A.R., E-mail: tasallesc@gmail.co [UNESP, Botucatu, SP (Brazil). Faculdade de Medicina. Servico de Radioterapia; Louzada, Mario J.Q. [UNESP, Aracatuba, SP (Brazil). Curso de Medicina Veterinaria

2011-07-01

Although they are no longer manufactured, the applicators of {sup 90}Sr + {sup 90}Y acquired in the decades of 1990 are still in use, by having half-life of 28.5 years. These applicators have calibration certificate given by their manufacturers, where few have been re calibrated. Thus it becomes necessary to accomplish thorough dosimetry of these applicators. This paper presents a dosimetric analysis distribution radial dose profiles for emitted by an {sup 90}Sr + {sup 90}Y beta therapy applicator, using the MCNP-4C code to simulate the distribution radial dose profiles and radio chromium films to get them experimentally . The results with the simulated values were compared with the results of experimental measurements, where both curves show similar behavior, which may validate the use of MCNP-4C and radio chromium films for this type of dosimetry. (author)

Dosimetry analysis of distribution radial dose profiles of {sup 90}Sr + {sup 90}Y beta therapy applicators using the MCNP-4C code and radio chromium films; Analise dosimetrica de perfis de distribuicoes radiais de doses relativas de um aplicador de betaterapia de {sup 90}Sr + {sup 90}Y utilizando o codigo MCNP-4C e filmes radiocromicos

Energy Technology Data Exchange (ETDEWEB)

Coelho, T.S.; Yoriyaz, H. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Fernandes, M.A.R. [Universidade Estadual Paulista Julio de Mesquita Filho (UNESP), Botucatu, SP (Brazil). Fac. de Medicina. Servico de Radioterapia; Louzada, M.J.Q. [Universidade Estadual Paulista Julio de Mesquita Filho (UNESP), Aracatuba, SP (Brazil). Curso de Medicina Veterinaria

2010-07-01

Although they are no longer manufactured, the applicators of {sup 90}Sr +{sup 90}Y acquired in the decades of 1990 are still in use, by having half-life of 28.5 years. These applicators have calibration certificate given by their manufacturers, where few have been recalibrated. Thus it becomes necessary to accomplish thorough dosimetry of these applicators. This paper presents a dosimetric analysis distribution radial dose profiles for emitted by an {sup 90}Sr+{sup 90}Y beta therapy applicator, using the MCNP-4C code to simulate the distribution radial dose profiles and radiochromium films to get them experimentally . The results with the simulated values were compared with the results of experimental measurements, where both curves show similar behavior, which may validate the use of MCNP-4C and radiochromium films for this type of dosimetry. (author)

Performance of the improved version of Monte Carlo code A 3MCNP for large-scale shielding problems

International Nuclear Information System (INIS)

Omura, M.; Miyake, Y.; Hasegawa, T.; Ueki, K.; Sato, O.; Haghighat, A.; Sjoden, G. E.

2005-01-01

A 3MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, which automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic 'importance' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A 3MCNP uses the three-dimensional (3-D) Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A 3MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A 3MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A 3MCNP (referred to as A 3MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A 3MCNPV for a concrete cask neutron and gamma-ray shielding problem, and a PWR dosimetry problem. (authors)

A new MCNP trademark test set

International Nuclear Information System (INIS)

Brockhoff, R.C.; Hendricks, J.S.

1994-09-01

The MCNP test set is used to test the MCNP code after installation on various computer platforms. For MCNP4 and MCNP4A this test set included 25 test problems designed to test as many features of the MCNP code as possible. A new and better test set has been devised to increase coverage of the code from 85% to 97% with 28 problems. The new test set is as fast as and shorter than the MCNP4A test set. The authors describe the methodology for devising the new test set, the features that were not covered in the MCNP4A test set, and the changes in the MCNP4A test set that have been made for MCNP4B and its developmental versions. Finally, new bugs uncovered by the new test set and a compilation of all known MCNP4A bugs are presented

Nuclear densimeter of soil simulated in MCNP-4C code

International Nuclear Information System (INIS)

Braga, Mario R.M.S.S.; Penna, Rodrigo; Vasconcelos, Danilo C.; Pereira, Claubia; Guerra, Bruno T.; Silva, Clemente J.G.C.

2009-01-01

The Monte Carlo code (MCNPX) was used to simulate a nuclear densimeter for measuring soil density. An Americium source (E = 60 keV) and a NaI (Tl) detector were placed on soil surface. Results from MCNP shown that scattered photon fluxes may be used to determining soil density. Linear regressions between scattered photons fluxes and soil density were calculated and shown correlation coefficients near unity. (author)

LEU-fueled SLOWPOKE-2 modelling with MCNP4A

International Nuclear Information System (INIS)

Pierre, J.R.M.; Bonin, H.W.J.

1996-01-01

Following the commissioning of the Low Enrichment Uranium (LEU) Fueled SLOWPOKE-2 research reactor at Royal Military College,excess reactivity measurements were conducted over a range of temperature and power. Given the advance in computer technology, the use of Monte Carlo N-Particle Transport Code System MCNP 4A appeared possible for the simulation of the LEU-fueled SLOWPOKE-2 reactor core, and this work demonstrates that this is indeed the case. MCNP 4A is a full three dimensional program allowing the user to enter a large amount of complexity. The limit on the geometry complexity is the computing time required to achieve a reasonable standard deviation. To this point several models of the SLOWPOKE-2 have been developed giving some insight on the sensitivity of the code. MCNP4A can use various cross section libraries. The aim of this work is to calculate accurately the reactivity of the core and reproduce The temperature trend of the reactivity. The model preserved as much as possible the details of the core and facility in order to allow further study in the flux mapping

Elaborate SMART MCNP Modelling Using ANSYS and Its Applications

Science.gov (United States)

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.

Comparison of MCNP4C and experimental results on neutron and gamma ray shielding effects for materials

Energy Technology Data Exchange (ETDEWEB)

Cha, Kyoon Ho; Lee, Eun Ki [KEPRI, Taejon (Korea, Republic of)

2004-07-01

MCNP code is a general-purpose Monte Carlo radiation transport code that can numerically simulate neutron, photon, and electron transport. Increasing the speed of computing machine is making numerical transport simulation more attractive and has led to the widespread use of such code. This code can be used for general radiation shielding and criticality accident alarm system related dose calculations, so that the version 4C2 of this code was used to evaluate the shielding effect against neutron and gamma ray experiments. The Ueki experiments were used for neutron shielding effects for materials, and the Kansas State University (KSU) photon skyshine experiments of 1977 were tested for gamma ray shielding effects.

Shielding calculations for neutron calibration bunker using Monte Carlo code MCNP-4C

International Nuclear Information System (INIS)

Suman, H.; Kharita, M. H.; Yousef, S.

2008-02-01

In this work, the dose arising from an Am-Be source of 10 8 neutron/sec strength located inside the newly constructed neutron calibration bunker in the National Radiation Metrology Laboratories, was calculated using MCNP-4C code. It was found that the shielding of the neutron calibration bunker is sufficient. As the calculated dose is not expected to exceed in inhabited areas 0.183 μSv/hr, which is 10 times smaller than the regulatory dose constraints. Hence, it can be concluded that the calibration bunker can house - from the external exposure point of view - an Am-Be neutron source of 10 9 neutron/sec strength. It turned out that the neutron dose from the source is few times greater than the photon dose. The sky shine was found to contribute significantly to the total dose. This contribution was estimated to be 60% of the neutron dose and 10% of the photon dose. The systematic uncertainties due to various factors have been assessed and was found to be between 4 and 10% due to concrete density variations; 15% due to the dose estimation method; 4 -10% due to weather variations (temperature and moisture). The calculated dose was highly sensitive to the changes in source spectra. The uncertainty due to the use of two different neutron spectra is about 70%.(author)

MCNP5 CRITICALITY VALIDATION AND BIAS FOR INTERMEDIATE ENRICHED URANIUM SYSTEMS

International Nuclear Information System (INIS)

Finfrock, S.H.

2009-01-01

The purpose of this analysis is to validate the Monte Carlo N-Particle 5 (MCNP5) code Version 1.40 (LA-UR-03-1987, 2005) and its cross-section database for k-code calculations of intermediate enriched uranium systems on INTEL(reg s ign) processor based PC's running any version of the WINDOWS operating system. Configurations with intermediate enriched uranium were modeled with the moderator range of 39 (le) H/Fissile (le) 1438. See Table 2-1 for brief descriptions of selected cases and Table 3-1 for the range of applicability for this validation. A total of 167 input cases were evaluated including bare and reflected systems in a single body or arrays. The 167 cases were taken directly from the previous (Version 4C [Lan 2005]) validation database. Section 2.0 list data used to calculate k-effective (k eff ) for the 167 experimental criticality benchmark cases using the MCNP5 code v1.40 and its cross section database. Appendix B lists the MCNP cross-section database entries validated for use in evaluating the intermediate enriched uranium systems for criticality safety. The dimensions and atom densities for the intermediate enriched uranium experiments were taken from NEA/NSC/DOC(95)03, September 2005, which will be referred to as the benchmark handbook throughout the report. For these input values, the experimental benchmark k eff is approximately 1.0. The MCNP validation computer runs ran to an accuracy of approximately ± 0.001. For the cases where the reported benchmark k eff was not equal to 1.0000 the MCNP calculational results were normalized. The difference between the MCNP validation computer runs and the experimentally measured k eff is the MCNP5 v1.40 bias. The USLSTATS code (ORNL 1998) was utilized to perform the statistical analysis and generate an acceptable maximum k eff limit for calculations of the intermediate enriched uranium type systems.

SABRINA, Geometry Plot Program for MCNP

International Nuclear Information System (INIS)

SEIDL, Marcus

2003-01-01

1 - Description of program or function: SABRINA is an interactive, three-dimensional, geometry-modeling code system, primarily for use with CCC-200/MCNP. SABRINA's capabilities include creation, visualization, and verification of three-dimensional geometries specified by either surface- or body-base combinatorial geometry; display of particle tracks are calculated by MCNP; and volume fraction generation. 2 - Method of solution: Rendering is performed by ray tracing or an edge and intersection algorithm. Volume fraction calculations are made by ray tracing. 3 - Restrictions on the complexity of the problem: A graphics display with X Window capability is required

Weight window/importance generator for Monte Carlo streaming problems

International Nuclear Information System (INIS)

Booth, T.E.

1983-01-01

A Monte Carlo method for solving highly angle dependent streaming problems is described. The method uses a DXTRAN-like angle biasing scheme, a space-angle weight window to reduce weight fluctuations introduced by the angle biasing, and a space-angle importance generator to set parameters for the space-angle weight window. Particle leakage through a doubly-bent duct is calculated to demonstrate the method's use

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

International Nuclear Information System (INIS)

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

2004-01-01

The expanding clinical use of low-energy photon emitting 125 I and 103 Pd 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

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

Science.gov (United States)

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.

MCNP application for the 21 century

International Nuclear Information System (INIS)

McKinney, G.W.

2000-01-01

The Los Alamos National Laboratory (LANL) Monte Carlo N-Particle radiation transport code, MCNP, has become an international standard for a wide spectrum of neutron, photon, and electron radiation transport applications. The latest version of the code, MCNP 4C, was released to the Radiation Safety Information Computational Center (RSICC) in February 2000. This paper describes the code development philosophy, new features and capabilities, applicability to various problems, and future directions

Absorbed body dose simulation in Thyroid cancer therapy using MCNP4Cand ITScodes and comparison to experimental results

International Nuclear Information System (INIS)

Hadad, K.; Gorji, Y.

2004-01-01

Two standard particle transport codes of MCNP4C and integrated tiger series were used to estimate the total body dose in a thyroid cancer therapy study, with I-131 as the radionuclide source. Human body was modeled by water and soft tissue ellipsoids. Phantoms' dimensions were selected according to Brow nell recommendation. Absorbed fractions were calculated by both codes for different phantoms and for gammas with 0.364 MeV energy, which has the highest fraction in I-131 emitting gammas. Results were compared to the data published by Brow nell et.al.. Figure 1 shows the results of MCNP4C and Integrated Tiger Series with results published by Brow nell et. al.

Development of visual platform of MCNP4B

International Nuclear Information System (INIS)

Fan Jiajin; Wang Yi; Cheng Jianping

2002-01-01

For convenience of using MCNP, the authors successfully developed a new code named McnpClient. With friend man-machine interface, the users can create input files very easily. If any error occurs during running process, McnpClient will give detailed fatal error or bad trouble messages. When the running is done, all the data can be obtained and in the mean time the curves associated with the data can be displayed

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

Science.gov (United States)

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.

Monte Carlo simulation applied to radiosurgery narrow beams using MCNP-4C

International Nuclear Information System (INIS)

Chaves, A.; Lopes, M.C.; Oliveira, C.

2001-01-01

Dose measurements for the narrow photon beams used in radiosurgery are complicated by the lack of electron equilibrium which is a requirement namely for ionometric methods. To overcome this difficulty the use of different dosimetric supports is strongly recommended in order to appreciate the influence of each type of detector. Monte Carlo simulation is another kind of tool to assess the details of the energy deposition phenomena in such narrow photon beams. In this study output factors and depth dose calculated by the Monte Carlo MCNP-4C code are presented and compared with experimental data measured with a diode, a Markus chamber, a 0.125 cc thimble chamber and a Pinpoint chamber. Simulated energy spectra for narrow beams are also presented in order to compare them with the reference 10 cm x 10 cm beam field size and thus discuss the different contributions of the absorbed energy in water, in each case. A detailed analysis on the photon energy spectra showed a slight decrease on the photon mean energy that can be explained by the increased scattering inside the additional collimators. Calculated and measured depth doses curves are in good agreement for most of the collimators. For the two smallest collimators some differences have been pointed and explained according to the characteristics of the detectors (author)

Calibration of a foot borne spectrometry system using the MCNP 4C code

International Nuclear Information System (INIS)

Nylen, T.; Agren, G.

2004-01-01

The increased interest for the cycling of radioactive Caesium in natural ecosystems has gained need for rapid and reliable methods to investigate the deposition density in natural soils. One commonly used method, soil sampling, is a good method that correctly used gives information of both the horizontal and vertical distribution of the desired nuclide. The main disadvantage is that the method is time consuming regarding sampling, preparation and measurements. An alternative method is the use of semiconductors or scintillation detectors in the field i.e. in cars, airplanes, or helicopters. Theses methods are rapid and integrate over large areas which gives a more reliable mean value provided that the operator has some basic knowledge about the depth distribution of the radio nuclides and bulk density in the soil. To be effective the systems are often connected to a GPS to give the exact coordinate for each measurement. In a situation where the area of interest is too large to cover by soil samples and measurements by airplane not will give a spatial resolution good enough, one feasible method is to use a foot borne gamma spectrometry system. The advantage of a foot borne system is that the operator can cover a quite large area within a few hours and that the method can detect small anomalies in the deposition field which may be difficult to discover with soil samples. This abstract describes the calibration of a foot borne gamma-spectrometry system carried in a back-pack and consisting of a NaI-detector, a GPS and a system for logging activity and position. The detector system and surroundings has been modeled in the Monte Carlo code MCNP 4C (Figure 1). The Monte Carlo method gives the possibility to study the influence of complex geometries that are difficult to create for a practical calibration using real activity. The results of the MCNP calibration model, has been compared to foot borne gamma-spectrometry field measurements in a Cs-137 deposition area. A

Calculation of the store house worker dose in a lost wax foundry using MCNP-4C.

Science.gov (United States)

Alegría, Natalia; Legarda, Fernando; Herranz, Margarita; Idoeta, Raquel

2005-01-01

Lost wax casting is an industrial process which permits the transmutation into metal of models made in wax. The wax model is covered with a silicaceous shell of the required thickness and once this shell is built the set is heated and wax melted. Liquid metal is then cast into the shell replacing the wax. When the metal is cool, the shell is broken away in order to recover the metallic piece. In this process zircon sands are used for the preparation of the silicaceous shell. These sands have varying concentrations of natural radionuclides: 238U, 232Th and 235U together with their progenics. The zircon sand is distributed in bags of 50 kg, and 30 bags are on a pallet, weighing 1,500 kg. The pallets with the bags have dimensions 80 cm x 120 cm x 80 cm, and constitute the radiation source in this case. The only pathway of exposure to workers in the store house is external radiation. In this case there is no dust because the bags are closed and covered by plastic, the store house has a good ventilation rate and so radon accumulation is not possible. The workers do not touch with their hands the bags and consequently skin contamination will not take place. In this study all situations of external irradiation to the workers have been considered; transportation of the pallets from vehicle to store house, lifting the pallets to the shelf, resting of the stock on the shelf, getting down the pallets, and carrying the pallets to production area. Using MCNP-4C exposure situations have been simulated, considering that the source has a homogeneous composition, the minimum stock in the store house is constituted by 7 pallets, and the several distances between pallets and workers when they are at work. The photons flux obtained by MCNP-4C is multiplied by the conversion factor of Flux to Kerma for air by conversion factor to Effective Dose by Kerma unit, and by the number of emitted photons. Those conversion factors are obtained of ICRP 74 table 1 and table 17 respectively. This

Calculation of the store house worker dose in a lost wax foundry using MCNP-4C

International Nuclear Information System (INIS)

Alegria, N.; Legarda, F.; Herranz, M.; Idoeta, R.

2005-01-01

Lost wax casting is an industrial process which permits the transmutation into metal of models made in wax. The wax model is covered with a siliceous shell of the required thickness and once this shell is built the set is heated and wax melted. Liquid metal is then cast into the shell replacing the wax. When the metal is cool, the shell is broken away in order to recover the metallic piece. In this process zircon sands are used for the preparation of the siliceous shell. These sands have varying concentrations of natural radionuclides: 238 U, 232 Th and 235 U together with their progenics. The zircon sand is distributed in bags of 50 kg, and 30 bags are on a pallet, weighing 1,500 kg. The pallets with the bags have dimensions 80 cm x 120 cm x 80 cm, and constitute the radiation source in this case. The only pathway of exposure to workers in the store house is external radiation. In this case there is no dust because the bags are closed and covered by plastic, the store house has a good ventilation rate and so radon accumulation is not possible. The workers do not touch with their hands the bags and consequently skin contamination will not take place. In this study all situations of external irradiation to the workers have been considered; transportation of the pallets from vehicle to store house, lifting the pallets to the shelf, resting of the stock on the shelf, getting down the pallets, and carrying the pallets to production area. Using MCNP-4C exposure situations have been simulated, considering that the source has a homogeneous composition, the minimum stock in the store house is constituted by 7 pallets, and the several distances between pallets and workers when they are at work. The photons flux obtained by MCNP-4C is multiplied by the conversion factor of Flux to Kerma for air by conversion factor to Effective Dose by Kerma unit, and by the number of emitted photons. Those conversion factors are obtained of ICRP 74 table 1 and table 17 respectively. This is

Installation of Monte Carlo neutron and photon transport code system MCNP4

International Nuclear Information System (INIS)

Takano, Makoto; Sasaki, Mikio; Kaneko, Toshiyuki; Yamazaki, Takao.

1993-03-01

The continuous energy Monte Carlo code MCNP-4 including its graphic functions has been installed on the Sun-4 sparc-2 work station with minor corrections. In order to validate the installed MCNP-4 code, 25 sample problems have been executed on the work station and these results have been compared with the original ones. And, the most of the graphic functions have been demonstrated by using 3 sample problems. Further, additional 14 nuclides have been included to the continuous cross section library edited from JENDL-3. (author)

Electron/Photon Verification Calculations Using MCNP4B

Energy Technology Data Exchange (ETDEWEB)

D. P. Gierga; K. J. Adams

1999-04-01

MCNP4BW was released in February 1997 with significant enhancements to electron/photon transport methods. These enhancements have been verified against a wide range of published electron/photon experiments, spanning high energy bremsstrahlung production to electron transmission and reflection. The impact of several MCNP tally options and physics parameters was explored in detail. The agreement between experiment and simulation was usually within two standard deviations of the experimental and calculational errors. Furthermore, sub-step artifacts for bremsstrahlung production were shown to be mitigated. A detailed suite of electron depth dose calculations in water is also presented. Areas for future code development have also been explored and include the dependence of cell and detector tallies on different bremsstrahlung angular models and alternative variance reduction splitting schemes for bremsstrahlung production.

Comparison of CdZnTe neutron detector models using MCNP6 and Geant4

Science.gov (United States)

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.

Windowed multipole sensitivity to target accuracy of the optimization procedure

International Nuclear Information System (INIS)

Josey, Colin; Forget, Benoit; Smith, Kord

2015-01-01

This paper compares the accuracy of the windowed multipole direct Doppler broadening method to that of the ENDF-B/VII.1 libraries that come with MCNP6. Various windowed multipole libraries were generated with different maximum allowed relative errors. Then, the libraries were compared to the MCNP6 data via resonance integral and through single assembly Monte Carlo analysis. Since the windowed multipole uses resonance parameters, resonance integrals are only affected by the number of resonances included in the library and not by the order of the background fitting function. The relative performance of each library with varying maximum allowed error was evaluated. It was found that setting a maximum target relative error of 0.1% in the library provided highly accurate data that closely matches the MCNP6 data for all temperatures of interest, while still having suitable computational performance. Additionally, a library with a maximum relative error of 1% also provided reasonable accuracy on eigenvalue and reaction rates with a noticeable improvement on performance, but with a few statistically significant differences with the MCNP6 data. (author)

Evaluation of Tehran research reactor (TRR) control rod worth using MCNP4C computer code

International Nuclear Information System (INIS)

Hosseini, Mohammad; Vosoughi, Naser; Hosseini, Seyed Abolfazl

2010-01-01

The main objective of reactor control system is to provide a safe reactor starting up, operation and shutting down. Calculation or measurement of precise values of control rod worth is of great importance in Tehran Research Reactor (TRR), considering the fact that they are the only controlling tools in the reactor. In present paper, simulation of TRR in First Operation Cycle (FOC) and in cold and clean core for the calculation of total and integral worth of control nods is reported. MCNP4C computer code has been used for all simulation process. Two method have been used for control rods worth calculation in this paper, namely the direct approach and perturbation method. It is shown that while the direct approach is appropriate for worth calculation of both the shim and the regulating control rods, the perturbation method is just suitable for tiny reactivity changes, i.e. for small initial part of regulating rods. Results of simulation are compared with the reported data in Safety Analysis Report (SAR) of Tehran research reactor and showed satisfactory agreement. (author)

MCNP(TM) Release 6.1.1 beta: Creating and Testing the Code Distribution

Energy Technology Data Exchange (ETDEWEB)

Cox, Lawrence J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Casswell, Laura [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2014-06-12

This report documents the preparations for and testing of the production release of MCNP6™1.1 beta through RSICC at ORNL. It addresses tests on supported operating systems (Linux, MacOSX, Windows) with the supported compilers (Intel, Portland Group and gfortran). Verification and Validation test results are documented elsewhere. This report does not address in detail the overall packaging of the distribution. Specifically, it does not address the nuclear and atomic data collection, the other included software packages (MCNP5, MCNPX and MCNP6) and the collection of reference documents.

TET_2MCNP: A conversion program to implement tetrahearal-mesh models in MCNP

International Nuclear Information System (INIS)

Han, Min Cheol; Yeom, Yeon Soo; Nguyen, Thng Tat; Choi, Chan Soo; Lee, Hyun Su; Kim, Chan Hyeong

2016-01-01

Tetrahedral-mesh geometries can be used in the MCNP code, but the MCNP code accepts only the geometry in the Abaqus input file format; hence, the existing tetrahedral-mesh models first need to be converted to the Abacus input file format to be used in the MCNP code. In the present study, we developed a simple but useful computer program, TET_2MCNP, for converting TetGen-generated tetrahedral-mesh models to the Abacus input file format. TET_2MCNP is written in C++ and contains two components: one for converting a TetGen output file to the Abacus input file and the other for the reverse conversion process. The TET_2MCP program also produces an MCNP input file. Further, the program provides some MCNP-specific functions: the maximum number of elements (i.e., tetrahedrons) per part can be limited, and the material density of each element can be transferred to the MCNP input file. To test the developed program, two tetrahedral-mesh models were generated using TetGen and converted to the Abaqus input file format using TET_2MCNP. Subsequently, the converted files were used in the MCNP code to calculate the object- and organ-averaged absorbed dose in the sphere and phantom, respectively. The results show that the converted models provide, within statistical uncertainties, identical dose values to those obtained using the PHITS code, which uses the original tetrahedral-mesh models produced by the TetGen program. The results show that the developed program can successfully convert TetGen tetrahedral-mesh models to Abacus input files. In the present study, we have developed a computer program, TET_2MCNP, which can be used to convert TetGen-generated tetrahedral-mesh models to the Abaqus input file format for use in the MCNP code. We believe this program will be used by many MCNP users for implementing complex tetrahedral-mesh models, including computational human phantoms, in the MCNP code

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

Science.gov (United States)

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.

MCNP output data analysis with ROOT (MODAR)

Science.gov (United States)

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

TET{sub 2}MCNP: A conversion program to implement tetrahearal-mesh models in MCNP

Energy Technology Data Exchange (ETDEWEB)

Han, Min Cheol; Yeom, Yeon Soo; Nguyen, Thng Tat; Choi, Chan Soo; Lee, Hyun Su; Kim, Chan Hyeong [Dept. of Nuclear Engineering, Hanyang University, Seoul (Korea, Republic of)

2016-12-15

Tetrahedral-mesh geometries can be used in the MCNP code, but the MCNP code accepts only the geometry in the Abaqus input file format; hence, the existing tetrahedral-mesh models first need to be converted to the Abacus input file format to be used in the MCNP code. In the present study, we developed a simple but useful computer program, TET{sub 2}MCNP, for converting TetGen-generated tetrahedral-mesh models to the Abacus input file format. TET{sub 2}MCNP is written in C++ and contains two components: one for converting a TetGen output file to the Abacus input file and the other for the reverse conversion process. The TET{sub 2}MCP program also produces an MCNP input file. Further, the program provides some MCNP-specific functions: the maximum number of elements (i.e., tetrahedrons) per part can be limited, and the material density of each element can be transferred to the MCNP input file. To test the developed program, two tetrahedral-mesh models were generated using TetGen and converted to the Abaqus input file format using TET{sub 2}MCNP. Subsequently, the converted files were used in the MCNP code to calculate the object- and organ-averaged absorbed dose in the sphere and phantom, respectively. The results show that the converted models provide, within statistical uncertainties, identical dose values to those obtained using the PHITS code, which uses the original tetrahedral-mesh models produced by the TetGen program. The results show that the developed program can successfully convert TetGen tetrahedral-mesh models to Abacus input files. In the present study, we have developed a computer program, TET{sub 2}MCNP, which can be used to convert TetGen-generated tetrahedral-mesh models to the Abaqus input file format for use in the MCNP code. We believe this program will be used by many MCNP users for implementing complex tetrahedral-mesh models, including computational human phantoms, in the MCNP code.

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

Science.gov (United States)

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.

MCNP Perturbation Capability for Monte Carlo Criticality Calculations

International Nuclear Information System (INIS)

Hendricks, J.S.; Carter, L.L.; McKinney, G.W.

1999-01-01

The differential operator perturbation capability in MCNP4B has been extended to automatically calculate perturbation estimates for the track length estimate of k eff in MCNP4B. The additional corrections required in certain cases for MCNP4B are no longer needed. Calculating the effect of small design changes on the criticality of nuclear systems with MCNP is now straightforward

Computational analysis of Bangladesh 3 MW TRIGA research reactor using MCNP4C, JENDL-3.3 and ENDF/B-Vl data libraries

International Nuclear Information System (INIS)

Huda, M.Q.

2006-01-01

The three-dimensional continuous energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Validation of the JENDL-3.3 and ENDF/BVI continuous energy cross-section data for MCNP4C was performed against some well-known benchmark lattices. For TRIGA analysis, data from JENDL-3.3 and ENDF/B-VI in combination with the JENDL-3.2 and ENDF/B-V data files (for nat Zr, nat Mo, nat Cr, nat Fe, nat Ni, nat Si, and nat Mg) at 300 K evaluations were used. Full S(α, β) scattering functions from ENDF/B-V for Zr in ZrH, H in ZrH and water molecule, and for graphite were used in both cases. The validation of the model was performed against the criticality and reactivity benchmark experiments of the TRIGA reactor. There is ∼20.0% decrease of thermal neutron flux occurs when the thermal library is removed during the calculation. Effect of erbium isotope that is present in the TRIGA fuel was also studied. In addition to the effective multiplication values, the well-known integral parameters: δ 28 , δ 25 , ρ 25 , and C * were calculated and compared for both JENDL3.3 and ENDF/B-VI libraries and were found to be in very good agreement. Results are also reported for most of the analyses performed by JENDL-3.2 and ENDF/B-V data libraries

MCNP and OMEGA criticality calculations

International Nuclear Information System (INIS)

Seifert, E.

1998-04-01

The reliability of OMEGA criticality calculations is shown by a comparison with calculations by the validated and widely used Monte Carlo code MCNP. The criticality of 16 assemblies with uranium as fissionable is calculated with the codes MCNP (Version 4A, ENDF/B-V cross sections), MCNP (Version 4B, ENDF/B-VI cross sections), and OMEGA. Identical calculation models are used for the three codes. The results are compared mutually and with the experimental criticality of the assemblies. (orig.)

How to Build MCNP 6.2

Energy Technology Data Exchange (ETDEWEB)

Bull, Jeffrey S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-11-13

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.

RBMK fuel channel blockage analysis by MCNP5, DRAGON and RELAP5-3D codes

International Nuclear Information System (INIS)

Parisi, C.; D'Auria, F.

2007-01-01

The aim of this work was to perform precise criticality analyses by Monte-Carlo code MCNP5 for a Fuel Channel (FC) flow blockage accident, considering as calculation domain a single FC and a 3x3 lattice of RBMK cells. Boundary conditions for MCNP5 input were derived by a previous transient calculation by state-of-the-art codes HELIOS/RELAP5-3D. In a preliminary phase, suitable MCNP5 models of a single cell and of a small lattice of RBMK cells were set-up; criticality analyses were performed at reference conditions for 2.0% and 2.4% enriched fuel. These analyses were compared with results obtained by University of Pisa (UNIPI) using deterministic transport code DRAGON and with results obtained by NIKIET Institute using MCNP4C. Then, the changes of the main physical parameters (e.g. fuel and water/steam temperature, water density, graphite temperature) at different time intervals of the FC blockage transient were evaluated by a RELAP5-3D calculation. This information was used to set up further MCNP5 inputs. Criticality analyses were performed for different systems (single channel and lattice) at those transient' states, obtaining global criticality versus transient time. Finally the weight of each parameter's change (fuel overheating and channel voiding) on global criticality was assessed. The results showed that reactivity of a blocked FC is always negative; nevertheless, when considering the effect of neighboring channels, the global reactivity trend reverts, becoming slightly positive or not changing at all, depending in inverse relation to the fuel enrichment. (author)

MCNP-DSP, Monte Carlo Neutron-Particle Transport Code with Digital Signal Processing

International Nuclear Information System (INIS)

2002-01-01

1 - Description of program or function: MCNP-DSP is recommended only for experienced MCNP users working with subcritical measurements. It is a modification of the Los Alamos National Laboratory's Monte Carlo code MCNP4a that is used to simulate a variety of subcritical measurements. The DSP version was developed to simulate frequency analysis measurements, correlation (Rossi-) measurements, pulsed neutron measurements, Feynman variance measurements, and multiplicity measurements. CCC-700/MCNP4C is recommended for general purpose calculations. 2 - Methods:MCNP-DSP performs calculations very similarly to MCNP and uses the same generalized geometry capabilities of MCNP. MCNP-DSP can only be used with the continuous-energy cross-section data. A variety of source and detector options are available. However, unlike standard MCNP, the source and detector options are limited to those described in the manual because these options are specified in the MCNP-DSP extra data file. MCNP-DSP is used to obtain the time-dependent response of detectors that are modeled in the simulation geometry. The detectors represent actual detectors used in measurements. These time-dependent detector responses are used to compute a variety of quantities such as frequency analysis signatures, correlation signatures, multiplicity signatures, etc., between detectors or sources and detectors. Energy ranges are 0-60 MeV for neutrons (data generally only available up to 20 MeV) and 1 keV - 1 GeV for photons and electrons. 3 - Restrictions on the complexity of the problem: None noted

EchoSeed Model 6733 Iodine-125 brachytherapy source: Improved dosimetric characterization using the MCNP5 Monte Carlo code

Energy Technology Data Exchange (ETDEWEB)

Mosleh-Shirazi, M. A.; Hadad, K.; Faghihi, R.; Baradaran-Ghahfarokhi, M.; Naghshnezhad, Z.; Meigooni, A. S. [Center for Research in Medical Physics and Biomedical Engineering and Physics Unit, Radiotherapy Department, Shiraz University of Medical Sciences, Shiraz 71936-13311 (Iran, Islamic Republic of); Radiation Research Center and Medical Radiation Department, School of Engineering, Shiraz University, Shiraz 71936-13311 (Iran, Islamic Republic of); Comprehensive Cancer Center of Nevada, Las Vegas, Nevada 89169 (United States)

2012-08-15

This study primarily aimed to obtain the dosimetric characteristics of the Model 6733 {sup 125}I seed (EchoSeed) with improved precision and accuracy using a more up-to-date Monte-Carlo code and data (MCNP5) compared to previously published results, including an uncertainty analysis. Its secondary aim was to compare the results obtained using the MCNP5, MCNP4c2, and PTRAN codes for simulation of this low-energy photon-emitting source. The EchoSeed geometry and chemical compositions together with a published {sup 125}I spectrum were used to perform dosimetric characterization of this source as per the updated AAPM TG-43 protocol. These simulations were performed in liquid water material in order to obtain the clinically applicable dosimetric parameters for this source model. Dose rate constants in liquid water, derived from MCNP4c2 and MCNP5 simulations, were found to be 0.993 cGyh{sup -1} U{sup -1} ({+-}1.73%) and 0.965 cGyh{sup -1} U{sup -1} ({+-}1.68%), respectively. Overall, the MCNP5 derived radial dose and 2D anisotropy functions results were generally closer to the measured data (within {+-}4%) than MCNP4c and the published data for PTRAN code (Version 7.43), while the opposite was seen for dose rate constant. The generally improved MCNP5 Monte Carlo simulation may be attributed to a more recent and accurate cross-section library. However, some of the data points in the results obtained from the above-mentioned Monte Carlo codes showed no statistically significant differences. Derived dosimetric characteristics in liquid water are provided for clinical applications of this source model.

Determination of finite-difference weights using scaled binomial windows

KAUST Repository

Chu, Chunlei; Stoffa, Paul L.

2012-01-01

The finite-difference method evaluates a derivative through a weighted summation of function values from neighboring grid nodes. Conventional finite-difference weights can be calculated either from Taylor series expansions or by Lagrange interpolation polynomials. The finite-difference method can be interpreted as a truncated convolutional counterpart of the pseudospectral method in the space domain. For this reason, we also can derive finite-difference operators by truncating the convolution series of the pseudospectral method. Various truncation windows can be employed for this purpose and they result in finite-difference operators with different dispersion properties. We found that there exists two families of scaled binomial windows that can be used to derive conventional finite-difference operators analytically. With a minor change, these scaled binomial windows can also be used to derive optimized finite-difference operators with enhanced dispersion properties. © 2012 Society of Exploration Geophysicists.

Determination of finite-difference weights using scaled binomial windows

KAUST Repository

Chu, Chunlei

2012-05-01

The finite-difference method evaluates a derivative through a weighted summation of function values from neighboring grid nodes. Conventional finite-difference weights can be calculated either from Taylor series expansions or by Lagrange interpolation polynomials. The finite-difference method can be interpreted as a truncated convolutional counterpart of the pseudospectral method in the space domain. For this reason, we also can derive finite-difference operators by truncating the convolution series of the pseudospectral method. Various truncation windows can be employed for this purpose and they result in finite-difference operators with different dispersion properties. We found that there exists two families of scaled binomial windows that can be used to derive conventional finite-difference operators analytically. With a minor change, these scaled binomial windows can also be used to derive optimized finite-difference operators with enhanced dispersion properties. © 2012 Society of Exploration Geophysicists.

A photoneutron production option for MCNP4A

International Nuclear Information System (INIS)

Gallmeier, F.X.

1996-01-01

A photoneutron production option was implemented in the MCNP4A code, mainly to supply a tool for reactor shielding calculations in beryllium and heavy water environments of complicated three dimensional geometries. Subroutines were developed to calculate the probability of the photoneutron production at the photon collision sites and the energy and flight direction of the created photoneutrons with the help of user supplied data. These subroutines are accessed through subroutine colidp which processes the photon collisions

MCNP6 Status

International Nuclear Information System (INIS)

Goorley, John T.

2012-01-01

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, 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.

Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system

International Nuclear Information System (INIS)

Iga, Kiminori; Takada, Hiroshi; Nagao, Tadashi.

1998-01-01

In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B 4 C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)

Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system

Energy Technology Data Exchange (ETDEWEB)

Iga, Kiminori [Kyushu Univ., Fukuoka (Japan); Takada, Hiroshi; Nagao, Tadashi

1998-01-01

In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B{sub 4}C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)

Criticality Calculations with MCNP6 - Practical Lectures

Energy Technology Data Exchange (ETDEWEB)

Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3)

2016-11-29

These slides are used to teach MCNP (Monte Carlo N-Particle) usage to nuclear criticality safety analysts. The following are the lecture topics: course information, introduction, MCNP basics, criticality calculations, advanced geometry, tallies, adjoint-weighted tallies and sensitivities, physics and nuclear data, parameter studies, NCS validation I, NCS validation II, NCS validation III, case study 1 - solution tanks, case study 2 - fuel vault, case study 3 - B&W core, case study 4 - simple TRIGA, case study 5 - fissile mat. vault, criticality accident alarm systems. After completion of this course, you should be able to: Develop an input model for MCNP; Describe how cross section data impact Monte Carlo and deterministic codes; Describe the importance of validation of computer codes and how it is accomplished; Describe the methodology supporting Monte Carlo codes and deterministic codes; Describe pitfalls of Monte Carlo calculations; Discuss the strengths and weaknesses of Monte Carlo and Discrete Ordinants codes; The diffusion theory model is not strictly valid for treating fissile systems in which neutron absorption, voids, and/or material boundaries are present. In the context of these limitations, identify a fissile system for which a diffusion theory solution would be adequate.

Criticality Calculations with MCNP6 - Practical Lectures

International Nuclear Information System (INIS)

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

2016-01-01

These slides are used to teach MCNP (Monte Carlo N-Particle) usage to nuclear criticality safety analysts. The following are the lecture topics: course information, introduction, MCNP basics, criticality calculations, advanced geometry, tallies, adjoint-weighted tallies and sensitivities, physics and nuclear data, parameter studies, NCS validation I, NCS validation II, NCS validation III, case study 1 - solution tanks, case study 2 - fuel vault, case study 3 - B&W core, case study 4 - simple TRIGA, case study 5 - fissile mat. vault, criticality accident alarm systems. After completion of this course, you should be able to: Develop an input model for MCNP; Describe how cross section data impact Monte Carlo and deterministic codes; Describe the importance of validation of computer codes and how it is accomplished; Describe the methodology supporting Monte Carlo codes and deterministic codes; Describe pitfalls of Monte Carlo calculations; Discuss the strengths and weaknesses of Monte Carlo and Discrete Ordinants codes; The diffusion theory model is not strictly valid for treating fissile systems in which neutron absorption, voids, and/or material boundaries are present. In the context of these limitations, identify a fissile system for which a diffusion theory solution would be adequate.

Windows 8 app projects XAML and C#

CERN Document Server

Vermeir, Nico

2013-01-01

Become a leading Windows 8 app developer by using Windows 8 App Projects - XAML and C# Edition to learn techniques, tools, and ideas to create successful, 5-star apps. Windows 8 App Projects - XAML and C# Edition shows you the nuts and bolts of the Windows 8 development ecosystem. Then, through a series of example driven chapters, you'll discover how to leverage the platform's unique features. With each project, you'll be one step closer to building full-featured, responsive, and well designed apps that feel like they're a part of the operating system. Windows 8 App Projects - XAML and C# Edit

Benchmarking ENDF/B-VII.1, JENDL-4.0 and JEFF-3.1.1 with MCNP6

International Nuclear Information System (INIS)

Marck, Steven C. van der

2012-01-01

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), to 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 6 Li, 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

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

Science.gov (United States)

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.

General introduction to MCNP

International Nuclear Information System (INIS)

Naito, Yoshitaka

2001-01-01

To assist succeeding reports which will be presented in this research meeting, following items on the computer code MCNP developed in USA are presented: (1) history of development of MCNP, (2) meaning of the development, (3) progress of study on Monte Carlo codes in the nuclear code committee and (4) expectation to Monte Carlo codes. (author)

C-Window Peaks on CE-HPLC are Extremely Rare in Northern India, and Only Infrequently Represent HbC.

Science.gov (United States)

Dass, Jasmita; Mittal, Suchi; Saraf, Amrita; Kotwal, Jyoti

2018-01-01

Hemoglobin C (HbC, HBB:c.19G > A) is a structural variant that has been reported rarely from India. This was a retrospective review of all high performance liquid chromatography (HPLCs) submitted over a 14 year period to a tertiary care center in North India with an aim of finding hemoglobins that elute in the C-window. Of the 32,364 HPLCs screened, 6 cases showed peaks in the C-window. Of these 6 cases, only two cases contained hemoglobin C. These was one case each of HbC/β thalassemia and compound heterozygosity for HbC and HbD. There were 4 cases which showed very similar red cell indices and chromatograms with multiple peaks eluting in D-window, C-window and an additional peak with a retention time of 4.74 min. These four cases were compound heterozygous for an α chain variant HbQ-India and a β-chain variant HbD.

Status of electron transport in MCNP trademark

International Nuclear Information System (INIS)

Hughes, H.G.

1997-01-01

The latest version of MCNP, the Los Alamos Monte Carlo transport code, has now been officially released. MCNP4B has been sent to the Radiation Safety Information Computational Center (RSICC), in Oak Ridge, Tennessee, which is responsible for the further distribution of the code within the US. International distribution of MCNP is done by the Nuclear Energy Agency (ECD/NEA), in Paris, France. Readers with access to the World-Wide-Web should consult the MCNP distribution site http://www-xdiv.lanl.gov/XTM/mcnp/about.html for specific information about contacting RSICC and OECD/NEA. A variety of new features are available in MCNP4B. Among these are differential operator perturbations, cross-section plotting capabilities, enhanced diagnostics for transport in repeated structures and lattices, improved efficiency in distributed-memory multiprocessing, corrected particle lifetime and lifespan estimators, and expanded software quality assurance procedures and testing, including testing of the multigroup Boltzmann-Fokker-Planck capability. New and improved cross section sets in the form of ENDF/B-VI evaluations have also been recently released and can be used in MCNP4B. Perhaps most significant for the interests of this special session, the electron transport algorithm has been improved, especially in the collisional energy-loss straggling and the angular-deflection treatments. In this paper, the author concentrates on a fairly complete documentation of the current status of the electron transport methods in MCNP

Biasing secondary particle interaction physics and production in MCNP6

International Nuclear Information System (INIS)

Fensin, M.L.; James, M.R.

2016-01-01

Highlights: • Biasing secondary production and interactions of charged particles in the tabular energy regime. • Examining lower weight window bounds for rare events when using Russian roulette. • The new biasing strategy can speedup calculations by a factor of 1 million or more. - Abstract: Though MCNP6 will transport elementary charged particles and light ions to low energies (i.e. less than 20 MeV), MCNP6 has historically relied on model physics with suggested minimum energies of ∼20 to 200 MeV. Use of library data for the low energy regime was developed for MCNP6 1.1.Beta to read and use light ion libraries. Thick target yields of neutron production for alphas on fluoride result in 1 production event per roughly million sampled alphas depending on the energy of the alpha (for other isotopes the yield can be even rarer). Calculation times to achieve statistically significant and converged thick target yields are quite laborious, needing over one hundred processor hours. The MUCEND code possess a biasing technique for improving the sampling of secondary particle production by forcing a nuclear interaction to occur per each alpha transported. We present here a different biasing strategy for secondary particle production from charged particles. During each substep, as the charged particle slows down, we bias both a nuclear collision event to occur at each substep and the production of secondary particles at the collision event, while still continuing to progress the charged particle until reaching a region of zero importance or an energy/time cutoff. This biasing strategy is capable of speeding up calculations by a factor of a million or more as compared to the unbiased calculation. Further presented here are both proof that the biasing strategy is capable of producing the same results as the unbiased calculation and the limitations to consider in order to achieve accurate results of secondary particle production. Though this strategy was developed for MCNP

Doppler Temperature Coefficient Calculations Using Adjoint-Weighted Tallies and Continuous Energy Cross Sections in MCNP6

Science.gov (United States)

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

MCNP-DSP users manual

International Nuclear Information System (INIS)

Valentine, T.E.

1997-01-01

The Monte Carlo code MCNP-DSP was developed from the Los Alamos MCNP4a code to calculate the time and frequency response statistics obtained from the 252 Cf-source-driven frequency analysis measurements. This code can be used to validate calculational methods and cross section data sets from subcritical experiments. This code provides a more general model for interpretation and planning of experiments for nuclear criticality safety, nuclear safeguards, and nuclear weapons identification and replaces the use of point kinetics models for interpreting the measurements. The use of MCNP-DSP extends the usefulness of this measurement method to systems with much lower neutron multiplication factors

Neutrons Flux Distributions of the Pu-Be Source and its Simulation by the MCNP-4B Code

Science.gov (United States)

Faghihi, F.; Mehdizadeh, S.; Hadad, K.

Neutron Fluence rate of a low intense Pu-Be source is measured by Neutron Activation Analysis (NAA) of 197Au foils. Also, the neutron fluence rate distribution versus energy is calculated using the MCNP-4B code based on ENDF/B-V library. Theoretical simulation as well as our experimental performance are a new experience for Iranians to make reliability with the code for further researches. In our theoretical investigation, an isotropic Pu-Be source with cylindrical volume distribution is simulated and relative neutron fluence rate versus energy is calculated using MCNP-4B code. Variation of the fast and also thermal neutrons fluence rate, which are measured by NAA method and MCNP code, are compared.

MCNP4c JEFF-3.1 Based Libraries. Eccolib-Jeff-3.1 libraries

International Nuclear Information System (INIS)

Sublet, J.Ch.

2006-01-01

Continuous-energy and multi-temperatures MCNP Ace types libraries, derived from the Joint European Fusion-Fission JEFF-3.1 evaluations, have been generated using the NJOY-99.111 processing code system. They include the continuous-energy neutron JEFF-3.1/General Purpose, JEFF-3.1/Activation-Dosimetry and thermal S(α,β) JEFF-3.1/Thermal libraries and data tables. The processing steps and features are explained together with the Quality Assurance processes and records linked to the generation of such multipurpose libraries. (author)

Evaluation of dose equivalent to the people accompanying patients in diagnostic radiology using MCNP4C Monte Carlo code

International Nuclear Information System (INIS)

Mehdizadeh, S.; Faghihi, R.; Sina, S.; Zehtabian, M.

2007-01-01

Complete text of publication follows. Objective: X rays used in diagnostic radiology contribute a major share to population doses from man-made sources of radiation. In some branches of radiology, it is necessary that another person stay in the imaging room and immobilize the patient to carry out radiological operation. ICRP 70 recommends that this should be done by parents or accompanying nursing or ancillary personnel and not in any case by radiation workers. Methods: Dose measurements were made previously using standard methods employing LiF TLD-100 dosimeters. A TLD card was installed on the main trunk of the body of the accompanying people where the maximum dose was probable. In this research the general purpose Monte Carlo N-particle radiation transport computer code (MCNP4C) is used to calculate the equivalent dose to the people accompanying patients exposed to radiation scattered from the patient (Without protective clothing). To do the simulations, all components of the geometry are placed within an air-filled box. Two homogeneous water phantoms are used to simulate the patient and the accompanying person. The accompanying person leans against the table at one side of the patient. Finally in case of source specification, only the focus of the X-ray tube is modelled, i.e. as a standard MCNP point source emitting a cone of photons. Photon stopping material is used as a collimator model to reduce the circular cross section of the cone to a rectangle. The X-ray spectra to be used in the MCNP simulations are generated with spectrum generator software, taking the X-ray voltage and all filtration applied in the clinic as input parameters. These calculations are done for different patient sizes and for different radiological operations. Results: In case of TL dosimetry, for a group of 100 examinations, the dose equivalents ranged from 0.01 μsv to 0.13 msv with the average of 0.05 msv. The results are seen to be in close agreement with Monte Carlo simulations

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

Science.gov (United States)

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.

Verification of MCNP6.2 for Nuclear Criticality Safety Applications

Energy Technology Data Exchange (ETDEWEB)

Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

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 suites 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.

Mock Target Window OTR and IR Design and Testing

Energy Technology Data Exchange (ETDEWEB)

Wass, Alexander Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-09-19

In order to fully verify temperature measurements made on the target window using infrared (IR) optical non-contact methods, actual comparative measurements are made with a real beam distribution as the heat source using Argonne National Laboratory’s (ANL) 35 MeV electron accelerator. Using Monte Carlo N-Particle (MCNP) simulations and thermal Finite Element Analysis (FEA), a cooled mock target window with thermocouple implants is designed to be used in such a test to achieve window temperatures up to 700°C. An uncoated and blackcoated mock window is designed to enhance the IR temperature measurements and verify optical transmitted radiation (OTR) imagery. This allows us to fully verify and characterize our temperature accuracy with our current IR camera method and any future method we may wish to explore using actual production conditions. This test also provides us with valuable conclusions/concerns regarding the calibration method we developed using our IR test stand at TA-53 in MPF-14.

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

Science.gov (United States)

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.

PWR Facility Dose Modeling Using MCNP5 and the CADIS/ADVANTG Variance-Reduction Methodology

Energy Technology Data Exchange (ETDEWEB)

Blakeman, Edward D [ORNL; Peplow, Douglas E. [ORNL; Wagner, John C [ORNL; Murphy, Brian D [ORNL; Mueller, Don [ORNL

2007-09-01

The feasibility of modeling a pressurized-water-reactor (PWR) facility and calculating dose rates at all locations within the containment and adjoining structures using MCNP5 with mesh tallies is presented. Calculations of dose rates resulting from neutron and photon sources from the reactor (operating and shut down for various periods) and the spent fuel pool, as well as for the photon source from the primary coolant loop, were all of interest. Identification of the PWR facility, development of the MCNP-based model and automation of the run process, calculation of the various sources, and development of methods for visually examining mesh tally files and extracting dose rates were all a significant part of the project. Advanced variance reduction, which was required because of the size of the model and the large amount of shielding, was performed via the CADIS/ADVANTG approach. This methodology uses an automatically generated three-dimensional discrete ordinates model to calculate adjoint fluxes from which MCNP weight windows and source bias parameters are generated. Investigative calculations were performed using a simple block model and a simplified full-scale model of the PWR containment, in which the adjoint source was placed in various regions. In general, it was shown that placement of the adjoint source on the periphery of the model provided adequate results for regions reasonably close to the source (e.g., within the containment structure for the reactor source). A modification to the CADIS/ADVANTG methodology was also studied in which a global adjoint source is weighted by the reciprocal of the dose response calculated by an earlier forward discrete ordinates calculation. This method showed improved results over those using the standard CADIS/ADVANTG approach, and its further investigation is recommended for future efforts.

PWR Facility Dose Modeling Using MCNP5 and the CADIS/ADVANTG Variance-Reduction Methodology

International Nuclear Information System (INIS)

Blakeman, Edward D.; Peplow, Douglas E.; Wagner, John C.; Murphy, Brian D.; Mueller, Don

2007-01-01

The feasibility of modeling a pressurized-water-reactor (PWR) facility and calculating dose rates at all locations within the containment and adjoining structures using MCNP5 with mesh tallies is presented. Calculations of dose rates resulting from neutron and photon sources from the reactor (operating and shut down for various periods) and the spent fuel pool, as well as for the photon source from the primary coolant loop, were all of interest. Identification of the PWR facility, development of the MCNP-based model and automation of the run process, calculation of the various sources, and development of methods for visually examining mesh tally files and extracting dose rates were all a significant part of the project. Advanced variance reduction, which was required because of the size of the model and the large amount of shielding, was performed via the CADIS/ADVANTG approach. This methodology uses an automatically generated three-dimensional discrete ordinates model to calculate adjoint fluxes from which MCNP weight windows and source bias parameters are generated. Investigative calculations were performed using a simple block model and a simplified full-scale model of the PWR containment, in which the adjoint source was placed in various regions. In general, it was shown that placement of the adjoint source on the periphery of the model provided adequate results for regions reasonably close to the source (e.g., within the containment structure for the reactor source). A modification to the CADIS/ADVANTG methodology was also studied in which a global adjoint source is weighted by the reciprocal of the dose response calculated by an earlier forward discrete ordinates calculation. This method showed improved results over those using the standard CADIS/ADVANTG approach, and its further investigation is recommended for future efforts

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

Science.gov (United States)

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

Evaluation of the OSCAR-4/MCNP calculation methodology for radioisotope production in the SAFARI-1 reactor

International Nuclear Information System (INIS)

Karriem, Z.; Zamonsky, O.M.

2014-01-01

The South African Nuclear Energy Corporation SOC Ltd (Necsa) is a state owned nuclear facility which owns and operates SAFARI-1, a 20 MW material testing reactor. SAFARI-1 is a multi-purpose reactor and is used for the production of radioisotopes through in-core sample irradiation. The Radiation and Reactor Theory (RRT) Section of Necsa supports SAFARI-1 operations with nuclear engineering analyses which include core-reload design, core-follow and radiation transport analyses. The primary computer codes that are used for the analyses are the OSCAR-4 nodal diffusion core simulator and the Monte Carlo transport code MCNP. RRT has developed a calculation methodology based on OSCAR-4 and MCNP to simulate the diverse in-core irradiation conditions in SAFARI-1, for the purpose of radioisotope production. In this paper we present the OSCAR-4/MCNP calculation methodology and the software tools that were developed for rapid and reliable construction of MCNP analysis models. The paper will present the application and accuracy of the methodology for the production of yttrium-90 ( 90 Y) and will include comparisons between calculation results and experimental measurements. The paper will also present sensitivity analyses that were performed to determine the effects of control rod bank position, representation of core depletion state and sample loading configuration, on the calculated 90 Y sample activity. (author)

Features of MCNP6

International Nuclear Information System (INIS)

Goorley, T.; James, M.; Booth, T.; Brown, F.; Bull, J.; Cox, L.J.; Durkee, J.; Elson, J.; Fensin, M.; Forster, R.A.; Hendricks, J.; Hughes, H.G.; Johns, R.; Kiedrowski, B.; Martz, R.; Mashnik, S.; McKinney, G.; Pelowitz, D.; Prael, R.; Sweezy, J.

2016-01-01

Highlights: • MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, but it is much more than the sum of these two computer codes. • MCNP6 is the result of six years of effort by the MCNP5 and MCNPX code development teams. • These groups of people, residing in Los Alamos National Laboratory’s X Computational Physics Division, Monte Carlo Codes Group (XCP-3) and Nuclear Engineering and Nonproliferation Division, Radiation Transport Modeling Team (NEN-5) respectively, have combined their code development efforts to produce the next evolution of MCNP. • While maintenance and major bug fixes will continue for MCNP5 1.60 and MCNPX 2.7.0 for upcoming years, new code development capabilities only will be developed and released in MCNP6. • In fact, the initial release of MCNP6 contains numerous new features not previously found in either code. • These new features are summarized in this document. • Packaged with MCNP6 is also the new production release of the ENDF/B-VII.1 nuclear data files usable by MCNP. • The high quality of the overall merged code, usefulness of these new features, along with the desire in the user community to start using the merged code, have led us to make the first MCNP6 production release: MCNP6 version 1. • High confidence in the MCNP6 code is based on its performance with the verification and validation test suites, comparisons to its predecessor codes, our automated nightly software debugger tests, the underlying high quality nuclear and atomic databases, and significant testing by many beta testers. - Abstract: MCNP6 can be described as the merger of MCNP5 and MCNPX capabilities, but it is much more than the sum of these two computer codes. MCNP6 is the result of six years of effort by the MCNP5 and MCNPX code development teams. These groups of people, residing in Los Alamos National Laboratory’s X Computational Physics Division, Monte Carlo Codes Group (XCP-3) and Nuclear Engineering and

Performance of the MTR core with MOX fuel using the MCNP4C2 code

International Nuclear Information System (INIS)

Shaaban, Ismail; Albarhoum, Mohamad

2016-01-01

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 (U 3 O 8 &PuO 2 ) 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 U 3 O 8 -Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U 3 O 8 -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. - Highlights: • Re-cycling of the ETRR-2 reactor by MOX fuel. • Increase the number of the neutronic traps from one neutronic trap to three neutronic trap. • Calculation of the criticality safety and neutronic parameters of the ETRR-2 reactor for the U 3 O 8 -Al original fuel and the MOX fuel.

Monte Carlo MCNP-4B-based absorbed dose distribution estimates for patient-specific dosimetry.

Science.gov (United States)

Yoriyaz, H; Stabin, M G; dos Santos, A

2001-04-01

This study was intended to verify the capability of the Monte Carlo MCNP-4B code to evaluate spatial dose distribution based on information gathered from CT or SPECT. A new three-dimensional (3D) dose calculation approach for internal emitter use in radioimmunotherapy (RIT) was developed using the Monte Carlo MCNP-4B code as the photon and electron transport engine. It was shown that the MCNP-4B computer code can be used with voxel-based anatomic and physiologic data to provide 3D dose distributions. This study showed that the MCNP-4B code can be used to develop a treatment planning system that will provide such information in a time manner, if dose reporting is suitably optimized. If each organ is divided into small regions where the average energy deposition is calculated with a typical volume of 0.4 cm(3), regional dose distributions can be provided with reasonable central processing unit times (on the order of 12-24 h on a 200-MHz personal computer or modest workstation). Further efforts to provide semiautomated region identification (segmentation) and improvement of marrow dose calculations are needed to supply a complete system for RIT. It is envisioned that all such efforts will continue to develop and that internal dose calculations may soon be brought to a similar level of accuracy, detail, and robustness as is commonly expected in external dose treatment planning. For this study we developed a code with a user-friendly interface that works on several nuclear medicine imaging platforms and provides timely patient-specific dose information to the physician and medical physicist. Future therapy with internal emitters should use a 3D dose calculation approach, which represents a significant advance over dose information provided by the standard geometric phantoms used for more than 20 y (which permit reporting of only average organ doses for certain standardized individuals)

Characteristics of multiprocessing MCNP5 on small personal computer clusters

International Nuclear Information System (INIS)

Robinson, S M; Mc Conn, R J Jr; Pagh, R T; Schweppe, J E; Siciliano, E R

2006-01-01

The feasibility and efficiency of performing MCNP5 calculations with a small, heterogeneous computing cluster built from Microsoft ( R) Windows TM personal computers (PC) are explored. The performance increases that may be expected with such clusters are estimated for cases that typify general radiation-shielding calculations. Our results show that the speed increase from additional slave PCs is nearly linear up to 10 processors. Guidance is given as to the specific advantages of changing various parameters present in the system. Implementing load balancing, and reducing the overhead from the MCNP rendezvous mechanism add to heterogeneous cluster efficiency. Hyper-threading technology and matching the total number of slave processes to the total number of logical processors also yield modest speed increases in the range below 7 processors. Because of the ease of acquisition of heterogeneous desktop computers, and the peak in efficiency at the level of a few physical processors, a strong case is made for the use of small clusters as a tool for producing MCNP5 calculations rapidly, and detailed instructions for constructing such clusters are provided

MCNP neutron benchmarks

International Nuclear Information System (INIS)

Hendricks, J.S.; Whalen, D.J.; Cardon, D.A.; Uhle, J.L.

1991-01-01

Over 50 neutron benchmark calculations have recently been completed as part of an ongoing program to validate the MCNP Monte Carlo radiation transport code. The new and significant aspects of this work are as follows: These calculations are the first attempt at a validation program for MCNP and the first official benchmarking of version 4 of the code. We believe the chosen set of benchmarks is a comprehensive set that may be useful for benchmarking other radiation transport codes and data libraries. These calculations provide insight into how well neutron transport calculations can be expected to model a wide variety of problems

Criticality Analysis Of TCA Critical Lattices With MNCP-4C Monte Carlo Calculation

International Nuclear Information System (INIS)

Zuhair

2002-01-01

The use of uranium-plutonium mixed oxide (MOX) fuel in electric generation light water reactor (PWR, BWR) is being planned in Japan. Therefore, the accuracy evaluations of neutronic analysis code for MOX cores have been employed by many scientists and reactor physicists. Benchmark evaluations for TCA was done using various calculation methods. The Monte Carlo become the most reliable method to predict criticality of various reactor types. In this analysis, the MCNP-4C code was chosen because various superiorities the code has. All in all, the MCNP-4C calculation for TCA core with 38 MOX critical lattice configurations gave the results with high accuracy. The JENDL-3.2 library showed significantly closer results to the ENDF/B-V. The k eff values calculated with the ENDF/B-VI library gave underestimated results. The ENDF/B-V library gave the best estimation. It can be concluded that MCNP-4C calculation, especially with ENDF/B-V and JENDL-3.2 libraries, for MOX fuel utilized NPP design in reactor core is the best choice

Modelling of HTR (High Temperature Reactor Pebble-Bed 10 MW to Determine Criticality as A Variations of Enrichment and Radius of the Fuel (Kernel With the Monte Carlo Code MCNP4C

Directory of Open Access Journals (Sweden)

Hammam Oktajianto

2014-12-01

Full Text Available Gas-cooled nuclear reactor is a Generation IV reactor which has been receiving significant attention due to many desired characteristics such as inherent safety, modularity, relatively low cost, short construction period, and easy financing. High temperature reactor (HTR pebble-bed as one of type of gas-cooled reactor concept is getting attention. In HTR pebble-bed design, radius and enrichment of the fuel kernel are the key parameter that can be chosen freely to determine the desired value of criticality. This paper models HTR pebble-bed 10 MW and determines an effective of enrichment and radius of the fuel (Kernel to get criticality value of reactor. The TRISO particle coated fuel particle which was modelled explicitly and distributed in the fuelled region of the fuel pebbles using a Simple-Cubic (SC lattice. The pebble-bed balls and moderator balls distributed in the core zone using a Body-Centred Cubic lattice with assumption of a fresh fuel by the fuel enrichment was 7-17% at 1% range and the size of the fuel radius was 175-300 µm at 25 µm ranges. The geometrical model of the full reactor is obtained by using lattice and universe facilities provided by MCNP4C. The details of model are discussed with necessary simplifications. Criticality calculations were conducted by Monte Carlo transport code MCNP4C and continuous energy nuclear data library ENDF/B-VI. From calculation results can be concluded that an effective of enrichment and radius of fuel (Kernel to achieve a critical condition was the enrichment of 15-17% at a radius of 200 µm, the enrichment of 13-17% at a radius of 225 µm, the enrichments of 12-15% at radius of 250 µm, the enrichments of 11-14% at a radius of 275 µm and the enrichment of 10-13% at a radius of 300 µm, so that the effective of enrichments and radii of fuel (Kernel can be considered in the HTR 10 MW. Keywords—MCNP4C, HTR, enrichment, radius, criticality 

Right-to-left-shunt detected by c-TCD using the orbital window in comparison with temporal bone windows.

Science.gov (United States)

Kobayashi, Kazuto; Kimura, Kazumi; Iguchi, Yasuyuki; Sakai, Kenichirou; Aoki, Junya; Iwanaga, Takeshi; Shibazaki, Kensaku

2012-01-01

There have been some reports on right-to-left shunt as a cause of cryptogenic stroke. Although contrast transcranial Doppler (c-TCD) can detect RLS, an insufficient temporal window has occasionally restricted its applicability. Thus, we compared the rates of detecting RLS among temporal windows for the middle cerebral arteries (MCAs) and the orbital window for the internal carotid artery (ICA) on c-TCD. We used c-TCD to detect RLS in patients with suspected ischemic stroke. We enrolled patients who had both sufficient bilateral temporal windows for MCAs and a right orbital window for ICA and performed c-TCD using all three windows simultaneously. We enrolled 106 consecutive patients and identified microembolic signals (MES) in 30 (28%) of them. Among these 30 patients, 15 had MES from all 3 windows. When these 30 patients were defined as being positive for RLS, the rates of detection were 67%, 73%, and 80% from the right temporal, left temporal, and right orbital windows, respectively (P= .795). The right orbital window as well as the temporal window for c-TCD could detect RLS. Insonation from the orbital window should be useful for patients who lack temporal windows. Copyright © 2010 by the American Society of Neuroimaging.

Calculation of ex-core detector weighting functions for a sodium-cooled tru burner mockup using MCNP5

International Nuclear Information System (INIS)

Pham Nhu Viet Ha; Min Jae Lee; Sunghwan Yun; Sang Ji Kim

2015-01-01

Power regulation systems of fast reactors are based on the signals of excore detectors. The excore detector weighting functions, which establish correspondence between the core power distribution and detector signal, are very useful for detector response analyses, e.g., in rod drop experiments. This paper presents the calculation of the weighting functions for a TRU burner mockup of the Korean Prototype Generation-IV Sodium-cooled Fast Reactor (named BFS-76-1A) using the MCNP5 multi-group adjoint capability. For generation of the weighting functions, all fuel assemblies were considered and each of them was divided into ten horizontal layers. Then the weighting functions for individual fuel assembly horizontal layers, the assembly weighting functions, and the shape annealing functions at RCP (Reactor Critical Point) and at conditions under which a control rod group was fully inserted into the core while other control rods at RCP were determined and evaluated. The results indicate that the weighting functions can be considered relatively insensitive to the control rods position during the rod drop experiments and therefore those weighting values at RCP can be applied to the dynamic rod worth simulation for the BFS-76-1A. (author)

Use of the MCNP Monte Carlo code for characterization of a pencil-type ionization chamber; Uso do código de Monte Carlo MCNP para caracterização de uma câmara de ionização tipo lápis

Energy Technology Data Exchange (ETDEWEB)

Mendonça, Dalila Souza Costa; Santos, William S.; Perini, Ana Paula, E-mail: anapaula.perini@ufu.br [Universidade Federal de Uberlândia (INFIS/UFU), MG (Brazil). Instituto de Física; Neves, Lucio Pereira; Caldas, Linda V. E. [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Belinato, Walmir [Instituto Federal de Educação, Ciência e Tecnologia da Bahia (IFBA), Vitória da Conquista, BA (Brazil)

2017-07-01

Ionization chambers are widely used in diagnostic radiology dosimetry. In this work, a special pencil-type ionization chamber, with different dimensions, configuration and materials in relation to commercial ones, was studied computationally. For this, the MCNP-4C Monte Carlo code and different radiation spectra were used to determine the influence of its components on its response. It was possible to observe that the highest influence was for the PVC wall. (author)

MCNP trademark Monte Carlo: A precis of MCNP

International Nuclear Information System (INIS)

Adams, K.J.

1996-01-01

MCNP trademark is a general purpose three-dimensional time-dependent neutron, photon, and electron transport code. It is highly portable and user-oriented, and backed by stringent software quality assurance practices and extensive experimental benchmarks. The cross section database is based upon the best evaluations available. MCNP incorporates state-of-the-art analog and adaptive Monte Carlo techniques. The code is documented in a 600 page manual which is augmented by numerous Los Alamos technical reports which detail various aspects of the code. MCNP represents over a megahour of development and refinement over the past 50 years and an ongoing commitment to excellence

Absorbed fractions in a voxel-based phantom calculated with the MCNP-4B code.

Science.gov (United States)

Yoriyaz, H; dos Santos, A; Stabin, M G; Cabezas, R

2000-07-01

A new approach for calculating internal dose estimates was developed through the use of a more realistic computational model of the human body. The present technique shows the capability to build a patient-specific phantom with tomography data (a voxel-based phantom) for the simulation of radiation transport and energy deposition using Monte Carlo methods such as in the MCNP-4B code. MCNP-4B absorbed fractions for photons in the mathematical phantom of Snyder et al. agreed well with reference values. Results obtained through radiation transport simulation in the voxel-based phantom, in general, agreed well with reference values. Considerable discrepancies, however, were found in some cases due to two major causes: differences in the organ masses between the phantoms and the occurrence of organ overlap in the voxel-based phantom, which is not considered in the mathematical phantom.

Simulation of Photon energy Spectra Using MISC, SOURCES, MCNP and GADRAS

International Nuclear Information System (INIS)

Tucker, Lucas P.; Shores, Erik F.; Myers, Steven C.; Felsher, Paul D.; Garner, Scott E.; Solomon, Clell J. Jr.

2012-01-01

The detector response functions included in the Gamma Detector Response and Analysis Software (GADRAS) are a valuable resource for simulating radioactive source emission spectra. Application of these response functions to the results of three-dimensional transport calculations is a useful modeling capability. Using a 26.2 kg shell of depleted uranium (DU) as a simple test problem, this work illustrates a method for manipulating current tally results from MCNP into the GAM file format necessary for a practical link to GADRAS detector response functions. MISC (MCNP Intrinsic Source Constructor) and SOURCES 4C were used to develop photon and neutron source terms for subsequent MCNP transport, and the resultant spectrum is shown to be in good agreement with that from GADRAS. A 1 kg DU sphere was also modeled with the method described here and showed similarly encouraging results.

Simulation of Photon energy Spectra Using MISC, SOURCES, MCNP and GADRAS

Energy Technology Data Exchange (ETDEWEB)

Tucker, Lucas P. [Los Alamos National Laboratory; Shores, Erik F. [Los Alamos National Laboratory; Myers, Steven C. [Los Alamos National Laboratory; Felsher, Paul D. [Los Alamos National Laboratory; Garner, Scott E. [Los Alamos National Laboratory; Solomon, Clell J. Jr. [Los Alamos National Laboratory

2012-08-14

The detector response functions included in the Gamma Detector Response and Analysis Software (GADRAS) are a valuable resource for simulating radioactive source emission spectra. Application of these response functions to the results of three-dimensional transport calculations is a useful modeling capability. Using a 26.2 kg shell of depleted uranium (DU) as a simple test problem, this work illustrates a method for manipulating current tally results from MCNP into the GAM file format necessary for a practical link to GADRAS detector response functions. MISC (MCNP Intrinsic Source Constructor) and SOURCES 4C were used to develop photon and neutron source terms for subsequent MCNP transport, and the resultant spectrum is shown to be in good agreement with that from GADRAS. A 1 kg DU sphere was also modeled with the method described here and showed similarly encouraging results.

Neutronic analysis for core conversion (HEU–LEU of the low power research reactor using the MCNP4C code

Directory of Open Access Journals (Sweden)

Aldawahra Saadou

2015-06-01

Full Text Available Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The proposed LEU core contained the same number of fuel pins as the HEU core. All other structure materials and dimensions of HEU and LEU cores were the same except the increase in the radius of control rod material from 0.195 to 0.205 cm and keeping the outer diameter of the control rod unchanged in the LEU core. The effective multiplication factor (keff, excess reactivity (ρex, control rod worth (CRW, shutdown margin (SDM, safety reactivity factor (SRF, delayed neutron fraction (βeff and the neutron fluxes in the irradiation tubes for the existing and the potential LEU fuel were investigated. The results showed that the safety parameters and the neutron fluxes in the irradiation tubes of the LEU fuels were in good agreements with the HEU results. Therefore, the LEU fuel was validated to be a suitable choice for fuel conversion of the MNSR in the future.

The MCNP-4C2 design of a two element photon/electron dosemeter that uses magnesium/copper/phosphorus doped lithium fluoride.

Science.gov (United States)

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.

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

Science.gov (United States)

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.

New calculations for critical assemblies using MCNP4B

International Nuclear Information System (INIS)

Adams, A.A.; Frankle, S.C.; Little, R.C.

1997-07-01

A suite of 41 criticality benchmarks has been modeled using MCNP trademark (version 4B). Most of the assembly specifications were obtained from the Cross Section Evaluation Working Group (CSEWG) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) compendiums of experimental benchmarks. A few assembly specifications were obtained from experimental papers. The suite contains thermal and fast assemblies, bare and reflected assemblies, and emphasizes 233 U, 235 U, 238 U, and 239 Pu. The values of k eff for each assembly in the suite were calculated using MCNP libraries derived primarily from release 2 of ENDF/B-V and release 2 of ENDF/B-VI. The results show that the new ENDF/B-VI.2 evaluations for H, O, N, B, 235 U, 238 U, and 239 Pu can have a significant impact on the values of k eff . In addition to the integral quantity k eff , several additional experimental measurements were performed and documented. These experimental measurements include central fission and reaction-rate ratios for various isotopes, and neutron leakage and flux spectra. They provide more detailed information about the accuracy of the nuclear data than can k eff . Comparison calculations were performed using both ENDF/B-V.2 and ENDF/B-VI.2-based data libraries. The purpose of this paper is to compare the results of these additional calculations with experimental data, and to use these results to assess the quality of the nuclear data

A methodology for evaluating weighting functions using MCNP and its application to PWR ex-core analyses

International Nuclear Information System (INIS)

Pecchia, Marco; Vasiliev, Alexander; Ferroukhi, Hakim; Pautz, Andreas

2017-01-01

Highlights: • Evaluation of neutron source importance for a given tally. • Assessment of ex-core detector response plus its uncertainty. • Direct use of neutron track evaluated by a Monte Carlo neutron transport code. - Abstract: The ex-core neutron detectors are commonly used to control reactor power in light water reactors. Therefore, it is relevant to understand the importance of a neutron source to the ex-core detectors response. In mathematical terms, this information is conveniently represented by the so called weighting functions. A new methodology based on the MCNP code for evaluating the weighting functions starting from the neutron history database is presented in this work. A simultaneous evaluation of the weighting functions in a user-given Cartesian coverage mesh is the main advantage of the method. The capability to generate weighting functions simultaneously in both spatial and energy ranges is the innovative part of this work. Then, an interpolation tool complements the methodology, allowing the generation of weighting functions up to the pin-by-pin fuel segment, where a direct evaluation is not possible due to low statistical precision. A comparison to reference results provides a verification of the methodology. Finally, an application to investigate the role of ex-core detectors spatial location and core burnup for a Swiss nuclear power plant is provided.

Development of interface between MCNP-FISPACT-MCNP (IPR-MFM) based on rigorous two step method

International Nuclear Information System (INIS)

Shaw, A.K.; Swami, H.L.; Danani, C.

2015-01-01

In this work we present the development of interface tool between MCNP-FISPACT-MCNP (MFM) based on Rigorous Two Step method for the shutdown dose rate (SDDR) calculation. The MFM links MCNP radiation transport and the FISPACT inventory code through a suitable coupling scheme. MFM coupling scheme has three steps. In first step it picks neutron spectrum and total flux from MCNP output file to use as input parameter for FISPACT. It prepares the FISPACT input files by using irradiation history, neutron flux and neutron spectrum and then execute the FISPACT input file in the second step. Third step of MFM coupling scheme extracts the decay gammas from the FISPACT output file and prepares MCNP input file for decay gamma transport followed by execution of MCNP input file and estimation of SDDR. Here detailing of MFM methodology and flow scheme has been described. The programming language PYTHON has been chosen for this development of the coupling scheme. A complete loop of MCNP-FISPACT-MCNP has been developed to handle the simplified geometrical problems. For validation of MFM interface a manual cross-check has been performed which shows good agreements. The MFM interface also has been validated with exiting MCNP-D1S method for a simple geometry with 14 MeV cylindrical neutron source. (author)

MCNP variance reduction overview

International Nuclear Information System (INIS)

Hendricks, J.S.; Booth, T.E.

1985-01-01

The MCNP code is rich in variance reduction features. Standard variance reduction methods found in most Monte Carlo codes are available as well as a number of methods unique to MCNP. We discuss the variance reduction features presently in MCNP as well as new ones under study for possible inclusion in future versions of the code

MCNP load balancing and fault tolerance with PVM

International Nuclear Information System (INIS)

McKinney, G.W.

1995-01-01

Version 4A of the Monte Carlo neutron, photon, and electron transport code MCNP, developed by LANL (Los Alamos National Laboratory), supports distributed-memory multiprocessing through the software package PVM (Parallel Virtual Machine, version 3.1.4). Using PVM for interprocessor communication, MCNP can simultaneously execute a single problem on a cluster of UNIX-based workstations. This capability provided system efficiencies that exceeded 80% on dedicated workstation clusters, however, on heterogeneous or multiuser systems, the performance was limited by the slowest processor (i.e., equal work was assigned to each processor). The next public release of MCNP will provide multiprocessing enhancements that include load balancing and fault tolerance which are shown to dramatically increase multiuser system efficiency and reliability

Using MCNP-4C code for design of the thermal neutron beam for neutron radiography at the MNSR

International Nuclear Information System (INIS)

Shaaban, I.

2009-11-01

Studies were carried out for determination of the parameters of a thermal neutron beam at the MNSR reactor (MNSR-30 kW) for neutron radiography in the vertical beam port by using the MCNP-4C (Monte Carlo Neutron - Photon transport). Thermal, epithermal and fast neutron energy ranges were selected as 10 keV respectively. To produce a good neutron beam in terms of intensity and quality, several materials Lead (Pb), Bismuth (Bi), Borated polyethelyene and Alumina Oxide (Al 2 O 3 ) were used as neutron and photon filters. Based on the current design, the L/D of the facility ranges between 125, 110 and 90. The thermal neutron flux at the beam exit is 1.436x10 5 n/cm2 .s ,1.843x10 5 n/cm2 .s and 2.845x10 5 n/cm2 .s respectively, middots with a Cd-ratio of ∼ 2.829, 2.766, 3.191 for the L/D = 125, 110, 90 respectively. The estimated values for gamma doses are 6.705x10 -2 Rem/h and 1.275x10 -1 Rem/h and 2.678x10 -1 Rem/ h with bismuth. The divergent angle of the collimator is 1.348 degree - 2.021 degree. Such neutron beams, if built into the Syrian MNSR reactor, could support the application of NRG in Syria. (author)

Analysis of inconsistent source sampling in monte carlo weight-window variance reduction methods

Directory of Open Access Journals (Sweden)

David P. Griesheimer

2017-09-01

Full Text Available The application of Monte Carlo (MC to large-scale fixed-source problems has recently become possible with new hybrid methods that automate generation of parameters for variance reduction techniques. Two common variance reduction techniques, weight windows and source biasing, have been automated and popularized by the consistent adjoint-driven importance sampling (CADIS method. This method uses the adjoint solution from an inexpensive deterministic calculation to define a consistent set of weight windows and source particles for a subsequent MC calculation. One of the motivations for source consistency is to avoid the splitting or rouletting of particles at birth, which requires computational resources. However, it is not always possible or desirable to implement such consistency, which results in inconsistent source biasing. This paper develops an original framework that mathematically expresses the coupling of the weight window and source biasing techniques, allowing the authors to explore the impact of inconsistent source sampling on the variance of MC results. A numerical experiment supports this new framework and suggests that certain classes of problems may be relatively insensitive to inconsistent source sampling schemes with moderate levels of splitting and rouletting.

MCNP capabilities at the dawn of the 21st century: Neutron-gamma applications

International Nuclear Information System (INIS)

Selcow, E.C.; McKinney, G.W.

2000-01-01

The Los Alamos National Laboratory Monte Carlo N-Particle radiation transport code, MCNP, has become an international standard for a wide spectrum of neutron-gamma radiation transport applications. These include nuclear criticality safety, radiation shielding, nuclear safeguards, nuclear well-logging, fission and fusion reactor design, accelerator target design, detector design and analysis, health physics, medical radiation therapy and imaging, radiography, decontamination and decommissioning, and waste storage and disposal. The latest version of the code, MCNP4C, was released to the Radiation Safety Information Computational Center (RSICC) in February 2000.This paper described the new features and capabilities of the code, and discusses the specific applicability to neutron-gamma problems. We will also discuss the future directions for MCNP code development, including rewriting the code in Fortran 90

MOCUP, MCNP/ORIGEN Coupling Utility Programs

International Nuclear Information System (INIS)

SEIDL, Marcus

2003-01-01

1 - Description of program or function: MOCUP is a series of utility and data manipulation programs to solve time and space-dependent coupled neutronics/isotopics problems. 2 - Methods: The neutronics calculation is performed by the Los Alamos National Laboratory code system, version 4a or later (CCC-200 or CCC-660),and the depletion and isotopics calculation is performed by CCC-371/ORIGEN2.1 developed at Oak Ridge National Laboratory. MCNP and ORIGEN2.1 are NOT included in this package. MOCUP consists of three utility programs (mcnpPRO, origenPRO, compPRO) to, respectively, search the MCNP output and tally files for relevant cell and tally parameters, prepare ORIGEN2.1 input files and execute the ORIGEN2.1 runs, and search ORIGEN2.1 punch files for relevant isotope concentrations and produce new MCNP input files. A graphical user interface is provided for execution convenience. 3 - Restrictions on the complexity of the problem: At present, no mechanism exists for automatic serial execution of the program modules. The user must interface with the GUI to run each of the modules

Particle Track Visualization using the MCNP Visual Editor

International Nuclear Information System (INIS)

Schwarz, Randolph A.; Carter, Lee; Brown, Wendi A.

2001-01-01

The Monte Carlo N-Particle (MCNP) visual editor1,2,3 is used throughout the world for displaying and creating complex MCNP geometries. The visual editor combines the Los Alamos MCNP Fortran code with a C front end to provide a visual interface. A big advantage of this approach is that the particle transport routines for MCNP are available to the visual front end. The latest release of the visual editor by Pacific Northwest National Laboratory enables the user to plot transport data points on top of a two-dimensional geometry plot. The user can plot source points, collisions points, surface crossings, and tally contributions. This capability can be used to show where particle collisions are occurring, verify the effectiveness of the particle biasing, or show which collisions contribute to a tally. For a KCODE (criticality source) calculation, the visual editor can be used to plot the source points for specific cycles

Fixed versus dynamic co-occurrence windows in TextRank term weights for information retrieval

DEFF Research Database (Denmark)

Lu, Wei; Cheng, Qikai; Lioma, Christina

2012-01-01

iteratively is a score for each vertex, i.e. a term weight, that can be used for information retrieval (IR) just like conventional term frequency based term weights. So far, when computing TextRank term weights over co-occurrence graphs, the window of term co-occurrence is always fixed. This work departs from...

MCNP: Photon benchmark problems

International Nuclear Information System (INIS)

Whalen, D.J.; Hollowell, D.E.; Hendricks, J.S.

1991-09-01

The recent widespread, markedly increased use of radiation transport codes has produced greater user and institutional demand for assurance that such codes give correct results. Responding to these pressing requirements for code validation, the general purpose Monte Carlo transport code MCNP has been tested on six different photon problem families. MCNP was used to simulate these six sets numerically. Results for each were compared to the set's analytical or experimental data. MCNP successfully predicted the analytical or experimental results of all six families within the statistical uncertainty inherent in the Monte Carlo method. From this we conclude that MCNP can accurately model a broad spectrum of photon transport problems. 8 refs., 30 figs., 5 tabs

Thermal lattice benchmarks for testing basic evaluated data files, developed with MCNP4B

International Nuclear Information System (INIS)

Maucec, M.; Glumac, B.

1996-01-01

The development of unit cell and full reactor core models of DIMPLE S01A and TRX-1 and TRX-2 benchmark experiments, using Monte Carlo computer code MCNP4B is presented. Nuclear data from ENDF/B-V and VI version of cross-section library were used in the calculations. In addition, a comparison to results obtained with the similar models and cross-section data from the EJ2-MCNPlib library (which is based upon the JEF-2.2 evaluation) developed in IRC Petten, Netherlands is presented. The results of the criticality calculation with ENDF/B-VI data library, and a comparison to results obtained using JEF-2.2 evaluation, confirm the MCNP4B full core model of a DIMPLE reactor as a good benchmark for testing basic evaluated data files. On the other hand, the criticality calculations results obtained using the TRX full core models show less agreement with experiment. It is obvious that without additional data about the TRX geometry, our TRX models are not suitable as Monte Carlo benchmarks. (author)

Utility of re-windowing for MR T2-weighted images in differentiating between benign tumors and cysts

International Nuclear Information System (INIS)

Yamamoto, A.; Nishikawa, K.; Otonari-Yamamoto, M.; Sano, T.

2009-01-01

Both benign tumors and cysts in the oral and maxillofacial region show clear borders and homogeneously high signal intensity on magnetic resonance (MR T2-weighted images, making differentiation difficult without contrast enhancement. Windowing for brightness and contrast adjustment may be helpful in interpreting relative signal intensities on MR images. This study was performed to determine whether re-windowing against targeted lesions on T2-weighted images was a useful procedure that would enhance differentiation without invasive contrast enhancement. Twenty-six lesions (13 benign tumors, 13 cysts) that showed clear borders and homogeneously high signal intensity on T2-weighted images were examined. The windowing parameters of axial images were readjusted to emphasize contrast only inside the lesions using automatic density adjustment. Re-windowed images were reviewed by three experienced oral radiologists and categorized based on the internal homogeneity of the lesion into four grades: 0, heterogeneous; 1, slightly heterogeneous; 2, slightly homogeneous; 3, homogeneous. Re-windowing was then evaluated for its usefulness in differentiating between benign tumors and cysts. For cysts, the rates of homogeneous (grades 3 and 2) and heterogeneous intensity (grades 1 and 0) were 66.7 (26/39) and 33.3% (13/39), respectively. For benign tumors, these rates were 33.3 (13/39) and 66.7% (26/39), respectively. Cysts showed a higher rate of homogeneous intensity, while the opposite was true for benign tumors. A significant difference in distribution was observed between cysts and benign tumors (P 2 test). Re-windowing for T2-weighted images is helpful in differentiating between benign tumors and cysts with clear borders and homogeneously high signal intensity on T2-weighted images. (author)

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

Science.gov (United States)

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.

Investigation of the Effects of Tissue Inhomogeneities on the Dosimetric Parameters of a Cs-137 Brachytherapy Source using the MCNP4C Code

Directory of Open Access Journals (Sweden)

Mehdi Zehtabian

2010-09-01

Full Text Available Introduction: Brachytherapy is the use of small encapsulated radioactive sources in close vicinity of tumors. Various methods are used to obtain the dose distribution around brachytherapy sources. TG-43 is a dosimetry protocol proposed by the AAPM for determining dose distributions around brachytherapy sources. The goal of this study is to update this protocol for presence of bone and air inhomogenities.  Material and Methods: To update the dose rate constant parameter of the TG-43 formalism, the MCNP4C simulations were performed in phantoms composed of water-bone and water-air combinations. The values of dose at different distances from the source in both homogeneous and inhomogeneous phantoms were estimated in spherical tally cells of 0.5 mm radius using the F6 tally. Results: The percentages of dose reductions in presence of air and bone inhomogenities for the Cs-137 source were found to be 4% and 10%, respectively. Therefore, the updated dose rate constant (Λ will also decrease by the same percentages.   Discussion and Conclusion: It can be easily concluded that such dose variations are more noticeable when using lower energy sources such as Pd-103 or I-125.

A parallelization study of the general purpose Monte Carlo code MCNP4 on a distributed memory highly parallel computer

International Nuclear Information System (INIS)

Yamazaki, Takao; Fujisaki, Masahide; Okuda, Motoi; Takano, Makoto; Masukawa, Fumihiro; Naito, Yoshitaka

1993-01-01

The general purpose Monte Carlo code MCNP4 has been implemented on the Fujitsu AP1000 distributed memory highly parallel computer. Parallelization techniques developed and studied are reported. A shielding analysis function of the MCNP4 code is parallelized in this study. A technique to map a history to each processor dynamically and to map control process to a certain processor was applied. The efficiency of parallelized code is up to 80% for a typical practical problem with 512 processors. These results demonstrate the advantages of a highly parallel computer to the conventional computers in the field of shielding analysis by Monte Carlo method. (orig.)

MCNP Version 6.2 Release Notes

Energy Technology Data Exchange (ETDEWEB)

Werner, Christopher John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bull, Jeffrey S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Solomon, C. J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McKinney, Gregg Walter [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dixon, David A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martz, Roger Lee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hughes, Henry G. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Cox, Lawrence James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Zukaitis, Anthony J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Armstrong, J. C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Forster, Robert Arthur [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Casswell, Laura [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2018-02-05

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 guide 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).

Sweeping-window arc therapy: an implementation of rotational IMRT with automatic beam-weight calculation

International Nuclear Information System (INIS)

Cameron, C

2005-01-01

Sweeping-window arc therapy (SWAT) is a variation of intensity-modulated radiation therapy (IMRT) with direct aperture optimization (DAO) that is initialized with a leaf sequence of sweeping windows that move back and forth periodically across the target as the gantry rotates. This initial sequence induces modulation in the dose and is assumed to be near enough to a minimum to allow successful optimization, done with simulated annealing, without requiring excessive leaf speeds. Optimal beam weights are calculated analytically, with easy extension to allow for variable beam weights. In this paper SWAT is tested on a phantom model and clinical prostate case. For the phantom, constant and variable beam weights are used. Although further work (in particular, improving the dose model) is required, the results show SWAT to be a feasible approach to generating deliverable dynamic arc treatments that are optimized

Sweeping-window arc therapy: an implementation of rotational IMRT with automatic beam-weight calculation

Energy Technology Data Exchange (ETDEWEB)

Cameron, C [Division of Radiation Physics, Department of Radiation Oncology, Stanford Cancer Center, 875 Blake Wilbur Drive, Rm G-233, Stanford, CA 94305-5847 (United States)

2005-09-21

Sweeping-window arc therapy (SWAT) is a variation of intensity-modulated radiation therapy (IMRT) with direct aperture optimization (DAO) that is initialized with a leaf sequence of sweeping windows that move back and forth periodically across the target as the gantry rotates. This initial sequence induces modulation in the dose and is assumed to be near enough to a minimum to allow successful optimization, done with simulated annealing, without requiring excessive leaf speeds. Optimal beam weights are calculated analytically, with easy extension to allow for variable beam weights. In this paper SWAT is tested on a phantom model and clinical prostate case. For the phantom, constant and variable beam weights are used. Although further work (in particular, improving the dose model) is required, the results show SWAT to be a feasible approach to generating deliverable dynamic arc treatments that are optimized.

A new effective Monte Carlo Midway coupling method in MCNP applied to a well logging problem

Energy Technology Data Exchange (ETDEWEB)

Serov, I.V.; John, T.M.; Hoogenboom, J.E

1998-12-01

The background of the Midway forward-adjoint coupling method including the black absorber technique for efficient Monte Carlo determination of radiation detector responses is described. The method is implemented in the general purpose MCNP Monte Carlo code. The utilization of the method is fairly straightforward and does not require any substantial extra expertise. The method was applied to a standard neutron well logging porosity tool problem. The results exhibit reliability and high efficiency of the Midway method. For the studied problem the efficiency gain is considerably higher than for a normal forward calculation, which is already strongly optimized by weight-windows. No additional effort is required to adjust the Midway model if the position of the detector or the porosity of the formation is changed. Additionally, the Midway method can be used with other variance reduction techniques if extra gain in efficiency is desired.

Preliminary evaluation of pin power distribution for fuel assemblies of SMART by MCNP

International Nuclear Information System (INIS)

Kim, Kyo Youn

1998-08-01

Monte Carlo transport code MCNP can describe an object sophisticately by use of three-dimensional modelling and can adopt a continuous energy cross-section library. Therefore MCNP has been widely utilized in the field of radiation physics to estimate fluxes and dose rates for nuclear facilities and to review results from conventional methods such a as discrete ordinates method and point kernel method. The Monte Carlo method has recently been introduced to estimated the neutron multiplication factor and pin power distribution in the fuel assembly of a reactor core. The operating thermal power of SMART core is 330 MWt and there are 57 fuel assemblies in the core. In this study it was assumed that the core has 4 types of fuel assemblies. In this study, MCNP4a was used to perform to estimate criticality and normalized pin power distribution in a fuel assembly of SMART core. The results from MCNP4a calculations are able to be used review those from nuclear design/analysis code. It is very complicated to pick up interested data from MCNP output list and to normalize pin power distribution in a fuel assembly because MCNP is not only a nuclear design/analysis code. In this study a program FAPIN was developed to generated a generate a normalized pin power distribution from the MCNP output list. (author). 11 refs

The new MCNP6 depletion capability

International Nuclear Information System (INIS)

Fensin, M. L.; James, M. R.; Hendricks, J. S.; Goorley, J. T.

2012-01-01

The first MCNP based in-line Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. Both the MCNP5 and MCNPX codes have historically provided a successful combinatorial geometry based, continuous energy, Monte Carlo radiation transport solution for advanced reactor modeling and simulation. However, due to separate development pathways, useful simulation capabilities were dispersed between both codes and not unified in a single technology. MCNP6, the next evolution in the MCNP suite of codes, now combines the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. We describe here the new capabilities of the MCNP6 depletion code dating from the official RSICC release MCNPX 2.6.0, reported previously, to the now current state of MCNP6. NEA/OECD benchmark results are also reported. The MCNP6 depletion capability enhancements beyond MCNPX 2.6.0 reported here include: (1) new performance enhancing parallel architecture that implements both shared and distributed memory constructs; (2) enhanced memory management that maximizes calculation fidelity; and (3) improved burnup physics for better nuclide prediction. MCNP6 depletion enables complete, relatively easy-to-use depletion calculations in a single Monte Carlo code. The enhancements described here help provide a powerful capability as well as dictate a path forward for future development to improve the usefulness of the technology. (authors)

The New MCNP6 Depletion Capability

International Nuclear Information System (INIS)

Fensin, Michael Lorne; James, Michael R.; Hendricks, John S.; Goorley, John T.

2012-01-01

The first MCNP based inline Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. Both the MCNP5 and MCNPX codes have historically provided a successful combinatorial geometry based, continuous energy, Monte Carlo radiation transport solution for advanced reactor modeling and simulation. However, due to separate development pathways, useful simulation capabilities were dispersed between both codes and not unified in a single technology. MCNP6, the next evolution in the MCNP suite of codes, now combines the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. We describe here the new capabilities of the MCNP6 depletion code dating from the official RSICC release MCNPX 2.6.0, reported previously, to the now current state of MCNP6. NEA/OECD benchmark results are also reported. The MCNP6 depletion capability enhancements beyond MCNPX 2.6.0 reported here include: (1) new performance enhancing parallel architecture that implements both shared and distributed memory constructs; (2) enhanced memory management that maximizes calculation fidelity; and (3) improved burnup physics for better nuclide prediction. MCNP6 depletion enables complete, relatively easy-to-use depletion calculations in a single Monte Carlo code. The enhancements described here help provide a powerful capability as well as dictate a path forward for future development to improve the usefulness of the technology.

Impact of MCNP unresolved resonance probability-table treatment on uranium and plutonium benchmarks

International Nuclear Information System (INIS)

Mosteller, R.D.; Little, R.C.

1998-01-01

Versions of MCNP up through and including 4B have not accurately modeled neutron self-shielding effects in the unresolved resonance energy region. Recently, a probability-table treatment has been incorporated into a developmental version of MCNP. This paper presents MCNP results for a variety of uranium and plutonium critical benchmarks, calculated with and without the probability-table treatment

Impact of patient weight on tumor visibility based on human-shaped phantom simulation study in PET imaging system

International Nuclear Information System (INIS)

Musarudin, M.; Saripan, M.I.; Mashohor, S.; Saad, W.H.M.; Nordin, A.J.; Hashim, S.

2015-01-01

Energy window technique has been implemented in all positron emission tomography (PET) imaging protocol, with the aim to remove the unwanted low energy photons. Current practices in our institution however are performed by using default energy threshold level regardless of the weight of the patient. Phantom size, which represents the size of the patient's body, is the factor that determined the level of scatter fraction during PET imaging. Thus, the motivation of this study is to determine the optimum energy threshold level for different sizes of human-shaped phantom, to represent underweight, normal, overweight and obese patients. In this study, the scanner was modeled by using Monte Carlo code, version MCNP5. Five different sizes of elliptical-cylinder shaped of human-sized phantoms with diameter ranged from 15 to 30 cm were modeled. The tumor was modeled by a cylindrical line source filled with 1.02 MeV positron emitters at the center of the phantom. Various energy window widths, in the ranged of 10–50% were implemented to the data. In conclusion, the phantom mass volume did influence the scatter fraction within the volume. Bigger phantom caused more scattering events and thus led to coincidence counts lost. We evaluated the impact of phantom sizes on the sensitivity and visibility of the simulated models. Implementation of wider energy window improved the sensitivity of the system and retained the coincidence photons lost. Visibility of the tumor improved as an appropriate energy window implemented for the different sizes of phantom. - Highlights: • Optimizing the energy window improved the sensitivity of the PET system. • Improving the visibility of the tumors using the optimized energy window. • Recommendations on the optimized energy windows for different body sizes. • Using simulated phantom using MCNP to determine various body sizes

Potential MCNP enhancements for NCT

International Nuclear Information System (INIS)

Estes, G.P.; Taylor, W.M.

1992-01-01

MCNP a Monte Carlo radiation transport code, is currently widely used in the medical community for a variety of purposes including treatment planning, diagnostics, beam design, tomographic studies, and radiation protection. This is particularly true in the Neutron Capture Therapy (NCT) community. The current widespread medical use of MCNP after its general public distribution in about 1980 attests to the code's general versatility and usefulness, particularly since its development to date has not been influenced by medical applications. This paper discusses enhancements to MCNP that could be implemented at Los Alamos for the benefit of the NCT community. These enhancements generally fall into two categories, namely those that have already been developed to some extent but are not yet publicly available, and those that seem both needed based on our current understanding of NCT goals, and achievable based on our working knowledge of the MCNP code. MCNP is a general, coupled neutron/photon/electron Monte Carlo code developed and maintained by the Radiation Transport Group at Los Alamos. It has been used extensively for radiation shielding studies, reactor analysis, detector design, physics experiment interpretation, oil and gas well logging, radiation protection studies, accelerator design, etc. over the years. MCNP is a three-dimensional geometry, continuous energy physics code capable of modeling complex geometries, specifying material regions such as organs by the intersections of analytical surfaces

SU-E-T-521: Investigation of the Uncertainties Involved in Secondary Neutron/gamma Production in Geant4/MCNP6 Monte Carlo Codes for Proton Therapy Application

International Nuclear Information System (INIS)

Mirzakhanian, L; Enger, S; Giusti, V

2015-01-01

Purpose: A major concern in proton therapy is the production of secondary neutrons causing secondary cancers, especially in young adults and children. Most utilized Monte Carlo codes in proton therapy are Geant4 and MCNP. However, the default versions of Geant4 and MCNP6 do not have suitable cross sections or physical models to properly handle secondary particle production in proton energy ranges used for therapy. In this study, default versions of Geant4 and MCNP6 were modified to better handle production of secondaries by adding the TENDL-2012 cross-section library. Methods: In-water proton depth-dose was measured at the “The Svedberg Laboratory” in Uppsala (Sweden). The proton beam was mono-energetic with mean energy of 178.25±0.2 MeV. The measurement set-up was simulated by Geant4 version 10.00 (default and modified version) and MCNP6. Proton depth-dose, primary and secondary particle fluence and neutron equivalent dose were calculated. In case of Geant4, the secondary particle fluence was filtered by all the physics processes to identify the main process responsible for the difference between the default and modified version. Results: The proton depth-dose curves and primary proton fluence show a good agreement between both Geant4 versions and MCNP6. With respect to the modified version, default Geant4 underestimates the production of secondary neutrons while overestimates that of gammas. The “ProtonInElastic” process was identified as the main responsible process for the difference between the two versions. MCNP6 shows higher neutron production and lower gamma production than both Geant4 versions. Conclusion: Despite the good agreement on the proton depth dose curve and primary proton fluence, there is a significant discrepancy on secondary neutron production between MCNP6 and both versions of Geant4. Further studies are thus in order to find the possible cause of this discrepancy or more accurate cross-sections/models to handle the nuclear

Modeling the effect in of criticality from changes in key parameters for small High Temperature Nuclear Reactor (U-BatteryTM) using MCNP4C

International Nuclear Information System (INIS)

Pauzi, A M

2013-01-01

The neutron transport code, Monte Carlo N-Particle (MCNP) which was wellkown as the gold standard in predicting nuclear reaction was used to model the small nuclear reactor core called U -battery TM, which was develop by the University of Manchester and Delft Institute of Technology. The paper introduces on the concept of modeling the small reactor core, a high temperature reactor (HTR) type with small coated TRISO fuel particle in graphite matrix using the MCNPv4C software. The criticality of the core were calculated using the software and analysed by changing key parameters such coolant type, fuel type and enrichment levels, cladding materials, and control rod type. The criticality results from the simulation were validated using the SCALE 5.1 software by [1] M Ding and J L Kloosterman, 2010. The data produced from these analyses would be used as part of the process of proposing initial core layout and a provisional list of materials for newly design reactor core. In the future, the criticality study would be continued with different core configurations and geometries.

Methodology for converting CT medical images to MCNP input using the Scan2MCNP system

International Nuclear Information System (INIS)

Boia, L.S.; Silva, A.X.; Cardoso, S.C.; Castro, R.C.

2009-01-01

This paper develops a methodology for the application software Scan2MCNP, which converts medical images DICOM (Digital Imaging and Communications in Medicine) for MCNP input file. The Scan2MCNP handles, processes and executes the medical images generated by CT equipment, allowing the user to perform the selection and parameterization of the study area in question (tissues and organs). The details of these worked in medical imaging software, therefore, will be converted to equity to the process of language analysis of MCNP radiation transport, through the generation of a code input file. With this file, it is possible to simulate any situation/problem of the type and level of radiation to the proposed treatment chosen by the medical staff responsible for the patient. Within a computational process oriented, the Scan2MCNP can contribute along with other software that has been used recently in the area of medical physics, to improve the levels of quality and precision of radiotherapy treatments. In this work, medical images DICOM of the Anthropomorphic Rando Phantom were used in the process of analysis and development of computer software Scan2MCNP. However, it emphasized that the software is successful in certain situations, depending upon a number of auxiliary procedures and software that can help in the solution of certain problems in the natural radiation treatment or express agility by the team of medical physics. (author)

MCNP Progress & Performance Improvements

Energy Technology Data Exchange (ETDEWEB)

Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bull, Jeffrey S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-04-14

Twenty-eight slides give information about the work of the US DOE/NNSA Nuclear Criticality Safety Program on MCNP6 under the following headings: MCNP6.1.1 Release, with ENDF/B-VII.1; Verification/Validation; User Support & Training; Performance Improvements; and Work in Progress. Whisper methodology will be incorporated into the code, and run speed should be increased.

Professional WCF 4 Windows Communication Foundation with NET 4

CERN Document Server

Cibraro, Pablo; Cozzolino, Fabio

2010-01-01

A guide to architecting, designing, and building distributed applications with Windows Communication Foundation. Windows Communication Foundation is the .NET technology that is used to build service-oriented applications, exchange messages in various communication scenarios, and run workflows. This guide enables developers to create state-of-the-art applications using this technology. Written by a team of Microsoft MVPs and WCF experts, this book explains how the pieces of WCF 4.0 build on each other to provide a comprehensive framework to support distributed enterprise applications. Experienc

MS Windows domēna darbstaciju migrācija no MS Windows XP uz Windows Vista.

OpenAIRE

Tetere, Agate

2009-01-01

Kvalifikācijas darbā izpētīju darbstaciju migrācijas no Windows XP uz Windows Vista plusus un mīnusus. Darba gaitā tika veikti sekojoši uzdevumi: 1.Veikta Windows XP un Windows Vista darbstaciju instalācija, iestatījumu konfigurēšana un tika pārbaudīta sistēmas darbība 2.Veikta Windows Server 2003 un Windows Server 2008 instalācija, iestatījumu konfigurēšana un tika pārbaudīta sistēmas darbība 3.Izstrādāts migrācijas modelis 4.Veikta migrācijas optimizēšana 5.Veikta datu migrāc...

Simulación con el código MCNP del reactor nuclear RP-10 en su configuración #14, BOC

OpenAIRE

Lázaro, Gerardo; Parreño, Fernando

2001-01-01

Se presenta los resultados de exceso de reactividad del núcleo del reactor RP-10 en su configuración 14. Este exceso de reactividad ha sido calculado con MCNP4B con un modelo que describe en detalle las características de los elementos combustibles normales y de control, así como de cada elemento que constituye la configuración de trabajo #14. Este modelo fue previamente utilizado en el reactor RP-0 y ha sido aplicado en la configuración de arranque para el cálculo del exceso de reactividad y...

Impact of MCNP Unresolved Resonance Probability-Table Treatment on Uranium and Plutonium Benchmarks

International Nuclear Information System (INIS)

Mosteller, R.D.; Little, R.C.

1999-01-01

A probability-table treatment recently has been incorporated into an intermediate version of the MCNP Monte Carlo code named MCNP4XS. This paper presents MCNP4XS results for a variety of uranium and plutonium criticality benchmarks, calculated with and without the probability-table treatment. It is shown that the probability-table treatment can produce small but significant reactivity changes for plutonium and 233 U systems with intermediate spectra. More importantly, it can produce substantial reactivity increases for systems with large amounts of 238 U and intermediate spectra

Dose mapping simulation using the MCNP code for the Syrian gamma irradiation facility and benchmarking

International Nuclear Information System (INIS)

Khattab, K.; Boush, M.; Alkassiri, H.

2013-01-01

Highlights: • The MCNP4C was used to calculate the gamma ray dose rate spatial distribution in for the SGIF. • Measurement of the gamma ray dose rate spatial distribution using the Chlorobenzene dosimeter was conducted as well. • Good agreements were noticed between the calculated and measured results. • The maximum relative differences were less than 7%, 4% and 4% in the x, y and z directions respectively. - Abstract: A three dimensional model for the Syrian gamma irradiation facility (SGIF) is developed in this paper to calculate the gamma ray dose rate spatial distribution in the irradiation room at the 60 Co source board using the MCNP-4C code. Measurement of the gamma ray dose rate spatial distribution using the Chlorobenzene dosimeter is conducted as well to compare the calculated and measured results. Good agreements are noticed between the calculated and measured results with maximum relative differences less than 7%, 4% and 4% in the x, y and z directions respectively. This agreement indicates that the established model is an accurate representation of the SGIF and can be used in the future to make the calculation design for a new irradiation facility

Suitability study of MCNP Monte Carlo program for use in medical physics

International Nuclear Information System (INIS)

Jeraj, R.

1998-01-01

MCNP is widely used Monte Carlo program in reactor and nuclear physics. However, an option of simulating electrons was added into the code a few years ago. With this extension MCNP became a code, potentially applicable for applications in medical physics. In 1997, a new version of the code, named MCNP4B was released, which contains several improvements in electron transport modeling. To test suitability of the code, several important issues were considered and examined. Default sampling in MCNP electron transport was found to be inappropriate, because it gives wrong depth dose curves for electron energies of interest in radiotherapy (Me V range). The problem can be solved if ITS-style energy sampling is used instead. One of the most difficult problems in electron transport is simulation of electron backscattering, which MCNP predicts well for all, low and high Z materials. One of the potential drawbacks, if somebody wanted to use MCNP for dosimetry on real patient geometries is that MCNP lattice calculation (e.g. when calculating dose distributions) becomes very slow for large number of scoring voxels. However, if just one scoring voxel is used, the number of geometry voxels only slightly affects the speed. In the study it was found that MCNP could be reliability used for many applications in medical physics. However, the established limitations should be taken into account when MCNP is used for a particular application.(author)

Benchmark analysis of MCNP trademark ENDF/B-VI iron

International Nuclear Information System (INIS)

Court, J.D.; Hendricks, J.S.

1994-12-01

The MCNP ENDF/B-VI iron cross-section data was subjected to four benchmark studies as part of the Hiroshima/Nagasaki dose re-evaluation for the National Academy of Science and the Defense Nuclear Agency. The four benchmark studies were: (1) the iron sphere benchmarks from the Lawrence Livermore Pulsed Spheres; (2) the Oak Ridge National Laboratory Fusion Reactor Shielding Benchmark; (3) a 76-cm diameter iron sphere benchmark done at the University of Illinois; (4) the Oak Ridge National Laboratory Benchmark for Neutron Transport through Iron. MCNP4A was used to model each benchmark and computational results from the ENDF/B-VI iron evaluations were compared to ENDF/B-IV, ENDF/B-V, the MCNP Recommended Data Set (which includes Los Alamos National Laboratory Group T-2 evaluations), and experimental data. The results show that the ENDF/B-VI iron evaluations are as good as, or better than, previous data sets

MCNP5 development, verification, and performance

International Nuclear Information System (INIS)

Forrest B, Brown

2003-01-01

MCNP is a well-known and widely used Monte Carlo code for neutron, photon, and electron transport simulations. During the past 18 months, MCNP was completely reworked to provide MCNP5, a modernized version with many new features, including plotting enhancements, photon Doppler broadening, radiography image tallies, enhancements to source definitions, improved variance reduction, improved random number generator, tallies on a superimposed mesh, and edits of criticality safety parameters. Significant improvements in software engineering and adherence to standards have been made. Over 100 verification problems have been used to ensure that MCNP5 produces the same results as before and that all capabilities have been preserved. Testing on large parallel systems shows excellent parallel scaling. (author)

MCNP5 development, verification, and performance

Energy Technology Data Exchange (ETDEWEB)

Forrest B, Brown [Los Alamos National Laboratory (United States)

2003-07-01

MCNP is a well-known and widely used Monte Carlo code for neutron, photon, and electron transport simulations. During the past 18 months, MCNP was completely reworked to provide MCNP5, a modernized version with many new features, including plotting enhancements, photon Doppler broadening, radiography image tallies, enhancements to source definitions, improved variance reduction, improved random number generator, tallies on a superimposed mesh, and edits of criticality safety parameters. Significant improvements in software engineering and adherence to standards have been made. Over 100 verification problems have been used to ensure that MCNP5 produces the same results as before and that all capabilities have been preserved. Testing on large parallel systems shows excellent parallel scaling. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Romero–Barrientos, Jaime, E-mail: jaromero@ing.uchile.cl [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, Las Condes, Santiago (Chile); Universidad de Chile, DFI, Facultad de Ciencias Físicas Y Matemáticas, Avenida Blanco Encalada 2008, Santiago (Chile); Molina, F. [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, Las Condes, Santiago (Chile); Aguilera, Pablo [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, Las Condes, Santiago (Chile); Universidad de Chile, Depto. de Física, Facultad de Ciencias, Las Palmeras 3425, Ñuñoa, Santiago (Chile); Arellano, H. F. [Universidad de Chile, DFI, Facultad de Ciencias Físicas Y Matemáticas, Avenida Blanco Encalada 2008, Santiago (Chile)

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.

Comparison calculations of WWER-1000 fuel assemblies by using the MCNP 4.2 a KASSETA codes

International Nuclear Information System (INIS)

Trgina, M.

1993-12-01

The power multiplication and distribution factors are compared for various geometries and material configurations of WWER-1000 fuel assemblies. The calculations were performed in 2 ways: (i) using nuclear data, employing older and current data collections, and (ii) using the author's own model based on the KASSETA code. The comparison code MCNP 4.2 is described, intended for computerized simulation of the transport of neutrons, photons and electrons. This code uses its own cross section library. The methodology is outlined and a specification of the Monte Carlo method employed is given. The use of the refined data library gave rise to appreciable deviations of the multiplication factors in all variants. The use of the older data library led to identical criticality results for the variant with water holes. For inserted absorbers the discrepancies in criticality and in power distribution data are appreciable. The marked disagreement between the results of application of the MCNP 4.2 and KASSETA codes for the variants with inserted control elements is indicative of inappropriateness of the approximation procedure in the latter code. (J.B.). 2 tabs., 11 figs., 11 refs

MCNP and MATXS cross section libraries based on JENDL-3.3

International Nuclear Information System (INIS)

Kosako, Kazuaki; Konno, Chikara; Fukahori, Tokio; Shibata, Keiichi

2003-01-01

The continuous energy cross section library for the Monte Carlo transport code MCNP-4C, FSXLIB-J33, has been generated from the latest version of JENDL-3.3. The multigroup cross section library with the MATXS format, MATXS-J33, has been generated also from JENDL-3.3. Both libraries contain all nuclides in JENDL-3.3 and are processed at 300 K by the nuclear data processing system NJOY99. (author)

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

Science.gov (United States)

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 (MCNP results depend significantly on the electron energy-indexing method.

30 CFR 18.66 - Tests of windows and lenses.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Tests of windows and lenses. 18.66 Section 18... Tests § 18.66 Tests of windows and lenses. (a) Impact tests. A 4-pound cylindrical weight with a 1-inch-diameter hemispherical striking surface shall be dropped (free fall) to strike the window or lens in its...

Reactor physics verification of the MCNP6 unstructured mesh capability

International Nuclear Information System (INIS)

Burke, T. P.; Kiedrowski, B. C.; Martz, R. L.; Martin, W. R.

2013-01-01

The Monte Carlo software package MCNP6 has the ability to transport particles on unstructured meshes generated from the Computed-Aided Engineering software Abaqus. Verification is performed using benchmarks with features relevant to reactor physics - Big Ten and the C5G7 computational benchmark. Various meshing strategies are tested and results are compared to reference solutions. Computational performance results are also given. The conclusions show MCNP6 is capable of producing accurate calculations for reactor physics geometries and the computational requirements for small lattice benchmarks are reasonable on modern computing platforms. (authors)

Reactor physics verification of the MCNP6 unstructured mesh capability

Energy Technology Data Exchange (ETDEWEB)

Burke, T. P. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Boulevard, Ann Arbor, MI 48109 (United States); Kiedrowski, B. C.; Martz, R. L. [X-Computational Physics Division, Monte Carlo Codes Group, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Martin, W. R. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Boulevard, Ann Arbor, MI 48109 (United States)

2013-07-01

The Monte Carlo software package MCNP6 has the ability to transport particles on unstructured meshes generated from the Computed-Aided Engineering software Abaqus. Verification is performed using benchmarks with features relevant to reactor physics - Big Ten and the C5G7 computational benchmark. Various meshing strategies are tested and results are compared to reference solutions. Computational performance results are also given. The conclusions show MCNP6 is capable of producing accurate calculations for reactor physics geometries and the computational requirements for small lattice benchmarks are reasonable on modern computing platforms. (authors)

Verification of the AZNHEX code v.1.4 with MCNP6 for different reference cases; Verificacion del codigo AZNHEX v.1.4 con MCNP6 para diferentes casos de referencia

Energy Technology Data Exchange (ETDEWEB)

Galicia A, J.; Francois L, J. L.; Bastida O, G. E. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, Circuito Exterior s/n, 04510 Ciudad de Mexico (Mexico); Del Valle G, E., E-mail: jgaliciaa87@gmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, 07738 Ciudad de Mexico (Mexico)

2017-09-15

The codes that make up the AZTLAN platform (AZTHECA, AZTRAN, AZKIND and AZNHEX) are currently in the testing phase simulating a variety of nuclear reactor assemblies and cores to compare and validate the results obtained for a particular case, with codes globally used in the nuclear area such as CASMO, Serpent and MCNP. The objective of this work is to continue improving the future versions of the codes of the AZTLAN platform so that accurate and reliable results can be obtained for the user. To test the current version of the AZNHEX code, 3 cases were taken into account, the first being the simulation of a VVER-440 reactor assembly; for the second case, the assembly of a fast reactor cooled with helium was simulated and for the third case it was decided to take up the case of the core of a fast reactor cooled with sodium, this because the previous versions of AZNHEX did not show adequate results and, in addition, they presented a considerable amount of limitations. The comparison and validation of the results (neutron multiplication factor, radial power, radial flow, axial power) for these three cases were made using the code MCNP6. The results obtained show that this version of AZNHEX produces values of the neutron multiplication factor and the neutron and power flow distributions very close to those of MCNP6. (Author)

Comparisons of the MCNP criticality benchmark suite with ENDF/B-VI.8, JENDL-3.3, and JEFF-3.0

International Nuclear Information System (INIS)

Kim, Do Heon; Gil, Choong-Sup; Kim, Jung-Do; Chang, Jonghwa

2003-01-01

A comparative study has been performed with the latest evaluated nuclear data libraries ENDF/B-VI.8, JENDL-3.3, and JEFF-3.0. The study has been conducted through the benchmark calculations for 91 criticality problems with the libraries processed for MCNP4C. The calculation results have been compared with those of the ENDF60 library. The self-shielding effects of the unresolved-resonance (UR) probability tables have also been estimated for each library. The χ 2 differences between the MCNP results and experimental data were calculated for the libraries. (author)

MCNP4c JEFF-3.1 Based Libraries. Eccolib-Jeff-3.1 libraries; Les bibliotheques Eccolib-Jeff-3.1

Energy Technology Data Exchange (ETDEWEB)

Sublet, J.Ch

2006-07-01

Continuous-energy and multi-temperatures MCNP Ace types libraries, derived from the Joint European Fusion-Fission JEFF-3.1 evaluations, have been generated using the NJOY-99.111 processing code system. They include the continuous-energy neutron JEFF-3.1/General Purpose, JEFF-3.1/Activation-Dosimetry and thermal S({alpha},{beta}) JEFF-3.1/Thermal libraries and data tables. The processing steps and features are explained together with the Quality Assurance processes and records linked to the generation of such multipurpose libraries. (author)

Thermo-mechanical design windows for SiC/SiC composite first wall of A-SSTR2

International Nuclear Information System (INIS)

He Kaihui; Satoshi Nishio

2002-01-01

The finite element analysis and calculation is performed for the blanket first wall made of SiC/SiC composite material for Advanced Steady-state Tokamak Reactor 2, A-SSTR2, which is now conceptually designed in Naka Fusion Research Establishment, JAERI. Comparison analysis and design window is analyzed by using the finite element code ADINA 7.4. Through 2D calculation for various geometrical configurations and sensitive material properties, a fundamental guideline for first wall and blanket design is established with respect to maximum temperature, thermal and mechanical stress for many configurations. To satisfy hydrodynamic requirement, a4d4 (the dimension of coolant channel is 4 mm x 8 mm, and the distance between neighboring channels is 4 mm) is chosen as design point for high thermal conductivity up to 50 W/m·K

Use of a Boron Doped Spherical Phantom for the Investigation of Neutron Directional Properties: Comparison Between Experiment and MCNP Simulation

Energy Technology Data Exchange (ETDEWEB)

Drake, P.; Kierkegaard, J

1999-07-01

A boron doped 19 cm diameter spherical phantom was constructed to give information on the direction of neutrons inside the Ringhals 4 containment. The phantom was made of 40% paraffin and 60% boric acid. 10B contributes 2% of the total phantom weight. The phantom was tested for its angular sensitivity to neutrons. The response was tested with a {sup 252}Cf source and with a Monte Carlo calculation (MCNP) simulating a {sup 252}Cf source. In these investigations the phantom showed a strong directional response. However, there was only a fair correspondence between the experiment and the simulation. The discrepancies are, at least in part, due to the difference in energy and angular response of the dosemeters as compared with the idealised response characteristics in the MCNP calculation. In the MCNP calculation the experimental conditions were not fully simulated. The investigations also showed that the addition of boron to the phantom reduces the leakage of thermalised neutrons from the phantom, and the production of neutron induced photons in the phantom to insignificant levels. (author)

Use of a Boron Doped Spherical Phantom for the Investigation of Neutron Directional Properties: Comparison Between Experiment and MCNP Simulation

International Nuclear Information System (INIS)

Drake, P.; Kierkegaard, J.

1999-01-01

A boron doped 19 cm diameter spherical phantom was constructed to give information on the direction of neutrons inside the Ringhals 4 containment. The phantom was made of 40% paraffin and 60% boric acid. 10B contributes 2% of the total phantom weight. The phantom was tested for its angular sensitivity to neutrons. The response was tested with a 252 Cf source and with a Monte Carlo calculation (MCNP) simulating a 252 Cf source. In these investigations the phantom showed a strong directional response. However, there was only a fair correspondence between the experiment and the simulation. The discrepancies are, at least in part, due to the difference in energy and angular response of the dosemeters as compared with the idealised response characteristics in the MCNP calculation. In the MCNP calculation the experimental conditions were not fully simulated. The investigations also showed that the addition of boron to the phantom reduces the leakage of thermalised neutrons from the phantom, and the production of neutron induced photons in the phantom to insignificant levels. (author)

Development of C-band High-Power Mix-Mode RF Window

CERN Document Server

Michizono, S; Matsumoto, T; Nakao, K; Takenaka, T

2004-01-01

High power c-band (5712 MHz) rf system (40 MW, 2 μs, 50 Hz) is under consideration for the electron-linac upgrade aimed for the super KEKB project. An rf window, which isolates the vacuum and pass the rf power, is one of the most important components for the rf system. The window consists of a ceramic disk and a pill-box housing. The mix-mode rf window is designed so as to decrease the electric field on the periphery of the ceramic disk. A resonant ring is assembled in order to examine the high-power transmission test. The window was tested up to the transmission power of 160 MW. The rf losses are also measured during the rf operation.

Recent MCNP developments

International Nuclear Information System (INIS)

Hendricks, J.S.; Briesmeister, J.F.

1991-01-01

MCNP is a widely used and actively developed Monte Carlo radiation transport code. Many important features have recently been added and more are under development. Benchmark studies not only indicate that MCNP is accurate but also that modern computer codes can give answers basically as accurate as the physics data that goes in them. Even deep penetration problems can be correct to within a factor of two after 10 to 25 mean free paths of penetration. And finally, Monte Carlo calculations, once thought to be too expensive to run routinely, can now be run effectively on desktop computers which compete with the supercomputers of yesteryear. 21 refs., 3 tabs

MCNP capabilities for nuclear well logging calculations

International Nuclear Information System (INIS)

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

1990-01-01

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 (Monte Carlo neutron photon), 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 tally 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

Criticality calculations with MCNP trademark: A primer

International Nuclear Information System (INIS)

Harmon, C.D. II; Busch, R.D.; Briesmeister, J.F.; Forster, R.A.

1994-01-01

With the closure of many experimental facilities, the nuclear criticality safety analyst increasingly is required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, in many cases, the analyst has little experience with the specific codes available at his/her facility. This primer will help you, the analyst, understand and use the MCNP Monte Carlo code for nuclear criticality safety analyses. It assumes that you have a college education in a technical field. There is no assumption of familiarity with Monte Carlo codes in general or with MCNP in particular. Appendix A gives an introduction to Monte Carlo techniques. The primer is designed to teach by example, with each example illustrating two or three features of MCNP that are useful in criticality analyses. Beginning with a Quickstart chapter, the primer gives an overview of the basic requirements for MCNP input and allows you to run a simple criticality problem with MCNP. This chapter is not designed to explain either the input or the MCNP options in detail; but rather it introduces basic concepts that are further explained in following chapters. Each chapter begins with a list of basic objectives that identify the goal of the chapter, and a list of the individual MCNP features that are covered in detail in the unique chapter example problems. It is expected that on completion of the primer you will be comfortable using MCNP in criticality calculations and will be capable of handling 80 to 90 percent of the situations that normally arise in a facility. The primer provides a set of basic input files that you can selectively modify to fit the particular problem at hand

NaI(Tl) detectors modeling in MCNP-X and Gate/Geant4 codes

Energy Technology Data Exchange (ETDEWEB)

Affonso, Renato Raoni Werneck; Silva, Ademir Xavier da, E-mail: raoniwa@yahoo.com.br, E-mail: ademir@nuclear.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Salgado, Cesar Marques, E-mail: otero@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2017-07-01

NaI (Tl) detectors are widely used in gamma-ray densitometry, but their modeling in Monte Carlo codes, such as MCNP-X and Gate/Geant4, needs a lot of work and does not yield comparable results with experimental arrangements, possibly due to non-simulated physical phenomena, such as light transport within the scintillator. Therefore, it is necessary a methodology that positively impacts the results of the simulations while maintaining the real dimensions of the detectors and other objects to allow validating a modeling that matches up with the experimental arrangement. Thus, the objective of this paper is to present the studies conducted with the MCNPX and Gate/Geant4 codes, in which the comparisons of their results were satisfactory, showing that both can be used for the same purposes. (author)

A DRAGON-MCNP comparison of void reactivity calculations

Energy Technology Data Exchange (ETDEWEB)

Marleau, G [Ecole Polytechnique, Montreal, PQ (Canada). Inst. de Genie Nucleaire; Milgram, M S [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

1996-12-31

The determination of the reactivity coefficients associated with coolant voiding in a CANDU reactor is a subject which has attracted a large amount of interest in the last few years both from the theoretical and experimental point of view. One expects that deterministic codes such as DRAGON and WIMS-AECL or the MCNP4 Monte Carlo code should be able to adequately simulate the cell behaviour upon coolant voiding. However, the absence of an experimental database at equilibrium and discharge burnups has not permitted the full validation of any of these lattice codes, although a partial validation through comparison of two different computer codes has been considered. Here we present a comparison between DRAGON and MCNP4 of the void reactivity evaluation for fresh fuel. (author). 16 refs., 5 tabs.

A DRAGON-MCNP comparison of void reactivity calculations

International Nuclear Information System (INIS)

Marleau, G.

1995-01-01

The determination of the reactivity coefficients associated with coolant voiding in a CANDU reactor is a subject which has attracted a large amount of interest in the last few years both from the theoretical and experimental point of view. One expects that deterministic codes such as DRAGON and WIMS-AECL or the MCNP4 Monte Carlo code should be able to adequately simulate the cell behaviour upon coolant voiding. However, the absence of an experimental database at equilibrium and discharge burnups has not permitted the full validation of any of these lattice codes, although a partial validation through comparison of two different computer codes has been considered. Here we present a comparison between DRAGON and MCNP4 of the void reactivity evaluation for fresh fuel. (author). 16 refs., 5 tabs

SNPs of melanocortin 4 receptor (MC4R) associated with body weight in Beagle dogs.

Science.gov (United States)

Zeng, Ruixia; Zhang, Yibo; Du, Peng

2014-01-01

Melanocortin 4 receptor (MC4R), which is associated with inherited human obesity, is involoved in food intake and body weight of mammals. To study the relationships between MC4R gene polymorphism and body weight in Beagle dogs, we detected and compared the nucleotide sequence of the whole coding region and 3'- and 5'- flanking regions of the dog MC4R gene (1214 bp). In 120 Beagle dogs, two SNPs (A420C, C895T) were identified and their relation with body weight was analyzed with RFLP-PCR method. The results showed that the SNP at A420C was significantly associated with canine body weight trait when it changed amino acid 101 of the MC4R protein from asparagine to threonine, while canine body weight variations were significant in female dogs when MC4R nonsense mutation at C895T. It suggested that the two SNPs might affect the MC4R gene's function which was relative to body weight in Beagle dogs. Therefore, MC4R was a candidate gene for selecting different size dogs with the MC4R SNPs (A420C, C895T) being potentially valuable as a genetic marker.

Evaluation of Geometric Progression (GP Buildup Factors using MCNP Codes (MCNP6.1 and MCNP5-1.60

Directory of Open Access Journals (Sweden)

Kim Kyung-O

2016-01-01

Full Text Available The gamma-ray buildup factors of three-dimensional point kernel code (QAD-CGGP are re-evaluated by using MCNP codes (MCNP6.1 and MCNPX5-1.60 and ENDF/B-VI.8 photoatomic data, which cover an energy range of 0.015–15 MeV and an iron thickness of 0.5–40 Mean Free Path (MFP. These new data are fitted to the Geometric Progression (GP fitting function and are then compared with ANS standard data equipped with QAD-CGGP. In addition, a simple benchmark calculation was performed to compare the QAD-CGGP results applied with new and existing buildup factors based on the MCNP codes. In the case of the buildup factors of low-energy gamma-rays, new data are evaluated to be about 5% higher than the existing data. In other cases, these new data present a similar trend based on the specific penetration depth, while existing data continuously increase beyond that depth. In a simple benchmark, the calculations using the existing data were slightly underestimated compared to the reference data at a deep penetration depth. On the other hand, the calculations with new data were stabilized with an increasing penetration depth, despite a slight overestimation at a shallow penetration depth.

Potential of the MCNP computer code

International Nuclear Information System (INIS)

Kyncl, J.

1995-01-01

The MCNP code is designed for numerical solution of neutron, photon, and electron transport problems by the Monte Carlo method. The code is based on the linear transport theory of behavior of the differential flux of the particles. The code directly uses data from the cross section point data library for input. Experience is outlined, gained in the application of the code to the calculation of the effective parameters of fuel assemblies and of the entire reactor core, to the determination of the effective parameters of the elementary fuel cell, and to the numerical solution of neutron diffusion and/or transport problems of the fuel assembly. The agreement between the calculated and observed data gives evidence that the MCNP code can be used with advantage for calculations involving WWER type fuel assemblies. (J.B.). 4 figs., 6 refs

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

Science.gov (United States)

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.

Extensions of the MCNP5 and TRIPOLI4 Monte Carlo codes for transient reactor analysis

International Nuclear Information System (INIS)

Hoogenboom, J.E.

2013-01-01

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 branch-less 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 the 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 3*3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3*3 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. (authors)

Experimental determination of the weighting factor for the energy window subtraction-based downscatter correction for I-123 in brain SPECT studies

DEFF Research Database (Denmark)

de Nijs, Robin; Holm, Søren; Thomsen, Gerda

2010-01-01

Correction for downscatter in I-123 SPECT can be performed by the subtraction of a secondary energy window from the main window, as in the triple-energy window method. This is potentially noise sensitive. For studies with limited amount of counts (e.g. dynamic studies), a broad subtraction window...... were investigated in this study. Energy windows with a width of 32 keV were centered at 159 keV and 200 keV. The weighting factor was measured both with an I-123 point source and in a dopamine transporter brain SPECT study in 10 human subjects (5 healthy subjects and 5 patients) by minimizing...... the background outside the head. Weighting factors ranged from 1.11 to 1.13 for the point source and from 1.16 to 1.18 for human subjects. Point source measurements revealed no position dependence. After correction, the measured specific binding ratio (image contrast) increased significantly for healthy subjects...

Monte Carlo Simulation of Electron Beams for Radiotherapy - EGS4, MCNP4b and GEANT3 Intercomparison

CERN Document Server

Trindade, A; Alves, C M; Chaves, A; Lopes, C; Oliveira, C; Peralta, L

2000-01-01

In medical radiation physics, an increasing number of Monte Carlo codes are being used, which requires intercomparison between them to evaluated the accuracy of the simulated results against benchmark experiments. The Monte Carlo code EGS4, commonly used to simulate electron beams from medical linear accelerators, was compared with GEANT3 and MCNP4b. Intercomparison of electron energy spectra, angular and spatial distribution were carried out for the Siemens KD2 linear accelerator, at beam energies of 10 and 15 MeV for a field size of 10x10 cm2. Indirect validation was performed against electron depth doses curves and beam profiles measured in a MP3-PTW water phantom using a Markus planar chamber. Monte Carlo isodose lines were reconstructed and compared to those from commercial treatment planning systems (TPS's) and with experimental data.

MCNP Simulations of End Flux Peaking in ACR-1000, 2.4 wt % {sup 235}U Fuel Bundles

Energy Technology Data Exchange (ETDEWEB)

Hill, Ian; Donnelly, Jim [Atomic Energy of Canada Limited (AECL), 2251 Speakman Drive, Mississauga, ON, L5K 1B2 (Canada)

2008-07-01

This paper examines the end flux peaking in ACR-1000 fuel bundles. Reactor physics simulations are performed with MCNP to assess the steady state end-flux peaking in an infinite lattice of ACR fuel, as well as to quantify the peaking that occurs during refuelling. 3-dimensional MCNP models are created based on the detailed geometry of the fuel bundle. Detailed position-dependent fuel compositions are obtained from MONTEBURNS which couples MCNP and ORIGIN2.2. Axial and radial power profiles are obtained for both fresh and mid-burnup fuel bundles in an infinite lattice. Subsequently an assessment of the impact of a refuelling transient on the power profiles is performed. The refuelling transient is found to increase the end flux peaking in the region adjacent to light water. (authors)

Calculation of power density with MCNP in TRIGA reactor

International Nuclear Information System (INIS)

Snoj, L.; Ravnik, M.

2006-01-01

Modern Monte Carlo codes (e.g. MCNP) allow calculation of power density distribution in 3-D geometry assuming detailed geometry without unit-cell homogenization. To normalize MCNP calculation by the steady-state thermal power of a reactor, one must use appropriate scaling factors. The description of the scaling factors is not adequately described in the MCNP manual and requires detailed knowledge of the code model. As the application of MCNP for power density calculation in TRIGA reactors has not been reported in open literature, the procedure of calculating power density with MCNP and its normalization to the power level of a reactor is described in the paper. (author)

A- and C-weighted sound levels as predictors of the annoyance caused by shooting sounds, for various facade attenuation types

NARCIS (Netherlands)

Vos, J.

2003-01-01

In a previous study on the annoyance caused by a great variety of shooting sounds [J. Acoust. Soc. Am. 109, 244-253 (2001)], it was shown that the annoyance, as rated indoors with the windows closed, could be adequately predicted from the outdoor A-weighted and C-weighted sound-exposure levels [ASEL

MCNP: a general Monte Carlo code for neutron and photon transport

International Nuclear Information System (INIS)

1979-11-01

The general-purpose Monte Carlo code MCNP ca be used for neutron, photon, or coupled neutron-photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation are accounted for. Thermal neutrons are described by both the free-gas and S(α,β) models. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. MCNP includes an elaborate, interactive plotting capability that allows the user to view his input geometry to help check for setup errors. Standard features which are available to improve computational efficiency include geometry splitting and Russian roulette, weight cutoff with Russian roulette, correlated sampling, analog capture or capture by weight reduction, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point or ring detectors, deterministically transporting pseudo-particles to designated regions, track-length estimators, source biasing, and several parameter cutoffs. Extensive summary information is provided to help the user better understand the physics and Monte Carlo simulation of his problem. The standard, user-defined output of MCNP includes two-way current as a function of direction across any set of surfaces or surface segments in the problem. Flux across any set of surfaces or surface segments is available. 58 figures, 28 tables

Parallelization of MCNP 4, a Monte Carlo neutron and photon transport code system, in highly parallel distributed memory type computer

International Nuclear Information System (INIS)

Masukawa, Fumihiro; Takano, Makoto; Naito, Yoshitaka; Yamazaki, Takao; Fujisaki, Masahide; Suzuki, Koichiro; Okuda, Motoi.

1993-11-01

In order to improve the accuracy and calculating speed of shielding analyses, MCNP 4, a Monte Carlo neutron and photon transport code system, has been parallelized and measured of its efficiency in the highly parallel distributed memory type computer, AP1000. The code has been analyzed statically and dynamically, then the suitable algorithm for parallelization has been determined for the shielding analysis functions of MCNP 4. This includes a strategy where a new history is assigned to the idling processor element dynamically during the execution. Furthermore, to avoid the congestion of communicative processing, the batch concept, processing multi-histories by a unit, has been introduced. By analyzing a sample cask problem with 2,000,000 histories by the AP1000 with 512 processor elements, the 82 % of parallelization efficiency is achieved, and the calculational speed has been estimated to be around 50 times as fast as that of FACOM M-780. (author)

Accelerating Pseudo-Random Number Generator for MCNP on GPU

Science.gov (United States)

Gong, Chunye; Liu, Jie; Chi, Lihua; Hu, Qingfeng; Deng, Li; Gong, Zhenghu

2010-09-01

Pseudo-random number generators (PRNG) are intensively used in many stochastic algorithms in particle simulations, artificial neural networks and other scientific computation. The PRNG in Monte Carlo N-Particle Transport Code (MCNP) requires long period, high quality, flexible jump and fast enough. In this paper, we implement such a PRNG for MCNP on NVIDIA's GTX200 Graphics Processor Units (GPU) using CUDA programming model. Results shows that 3.80 to 8.10 times speedup are achieved compared with 4 to 6 cores CPUs and more than 679.18 million double precision random numbers can be generated per second on GPU.

Parallel MCNP Monte Carlo transport calculations with MPI

International Nuclear Information System (INIS)

Wagner, J.C.; Haghighat, A.

1996-01-01

The steady increase in computational performance has made Monte Carlo calculations for large/complex systems possible. However, in order to make these calculations practical, order of magnitude increases in performance are necessary. The Monte Carlo method is inherently parallel (particles are simulated independently) and thus has the potential for near-linear speedup with respect to the number of processors. Further, the ever-increasing accessibility of parallel computers, such as workstation clusters, facilitates the practical use of parallel Monte Carlo. Recognizing the nature of the Monte Carlo method and the trends in available computing, the code developers at Los Alamos National Laboratory implemented the message-passing general-purpose Monte Carlo radiation transport code MCNP (version 4A). The PVM package was chosen by the MCNP code developers because it supports a variety of communication networks, several UNIX platforms, and heterogeneous computer systems. This PVM version of MCNP has been shown to produce speedups that approach the number of processors and thus, is a very useful tool for transport analysis. Due to software incompatibilities on the local IBM SP2, PVM has not been available, and thus it is not possible to take advantage of this useful tool. Hence, it became necessary to implement an alternative message-passing library package into MCNP. Because the message-passing interface (MPI) is supported on the local system, takes advantage of the high-speed communication switches in the SP2, and is considered to be the emerging standard, it was selected

Solar orbiter/PHI full disk telescope entrance window mechanical mount

Science.gov (United States)

Barandiaran, J.; Zuluaga, P.; Fernandez, A. B.; Vera, I.; Garranzo, D.; Nuñez, A.; Bastide, L.; Royo, M. T.; Alvarez, A.

2017-11-01

PHI is a diffraction limited, wavelength tunable, quasi-monochromatic, and polarization sensitive imager. These capabilities are needed to infer the magnetic field and line-of-sight (LOS) velocity of the region targeted by the spacecraft (spacecraft (S/C)). PHI will consist of two telescopes: The High Resolution Telescope (HRT)[1] and the Full Disk Telescope (FDT). The HRT and the FDT will view the Sun through entrance windows located in the S/C heat shield. These windows act as heat rejecting filters with a transmission band of about 30 nm width centered on the science wavelength, such that the total transmittance (integral over the filter curve weighted with solar spectrum, including white leakage plus transmission profile of the pass band) does not exceed 4% of the total energy falling onto the window [2][3]. The HREW filter has been designed by SELEX in the framework of an ESA led technology development activity under original ESTEC contract No. 20018/06/NL/CP[4], and extensions thereof. For FDT HREW SLEX will provide the windows and it coatings. The HREW consists of two parallel-plane substrate plates (window 1 & window 2)[5] made of SUPRASIL 300 with a central thickness of 9 mm and a wedge of 30 arcsec each. These two substrates are each coated on both sides with four different coatings. These coatings and the choice of SUPRASIL help to minimize the optical absorptivity in the substrate and to radiatively decouple the HREW, which is expected to run at high temperatures during perihelion passages, from the PHI instrument cavity. The temperature distribution of the HREW is driven by two main factors: the mechanical mounting of the substrates to the feedthrough, and the radiative environment within the heat-shield/feedthrough assembly. The mechanical mount must ensure the correct integration of both suprasil substrates in its correct position and minimize the loads in windows due to thermal induced deformations and launching vibration environment. All the

Installation of MCNP on 64-bit parallel computers

International Nuclear Information System (INIS)

Meginnis, A.B.; Hendricks, J.S.; McKinney, G.W.

1995-01-01

The Monte Carlo radiation transport code MCNP has been successfully ported to two 64-bit workstations, the SGI and DEC Alpha. We found the biggest problem for installation on these machines to be Fortran and C mismatches in argument passing. Correction of these mismatches enabled, for the first time, dynamic memory allocation on 64-bit workstations. Although the 64-bit hardware is faster because 8-bytes are processed at a time rather than 4-bytes, we found no speed advantage in true 64-bit coding versus implicit double precision when porting an existing code to the 64-bit workstation architecture. We did find that PVM multiasking is very successful and represents a significant performance enhancement for scientific workstations

FENDL2/A-MCNP, FENDL2/A-VITJE and FENDL2/A-VITJFLAT. The processed FENDL-2 neutron activation cross-section data files. Summary documentation

International Nuclear Information System (INIS)

Pashchenko, A.B.; Wienke, H.

1997-01-01

This document summarizes the libraries of neutron activation cross-section data processed into the following three formats: continuous energy format as used by the Monte Carlo neutron/photon transport code MCNP4A; VITAMIN-J 175 multigroup format weighted with the VITAMIN-E weighting spectrum as used by the transmutation codes REAC*2/3 and FOUR ACES; VITAMIN-J 175 multigroup ENDF-6 format, with a flat weighting spectrum. The data are available from the IAEA Nuclear Data Section online via INTERNET by FTP command, or on magnetic tape. (author)

Calibration curves of a PGNAA system for cement raw material analysis using the MCNP code

International Nuclear Information System (INIS)

Oliveira, Carlos; Salgado, Jose

1998-01-01

In large samples, the γ-ray count rate of a prompt gamma neutron activation analysis system is a multi-variable function of the elemental dry composition, density, water content and thickness of the material. The experimental calibration curves require tremendous laboratory work, using a great number of standards with well-known compositions. Although a Monte Carlo simulation study does not avoid the experimental calibration work, it reduces the number of experimental calibration standards. This paper is part of a feasibility study for a PGNAA system for on-line continuous characterisation of cement raw material conveyed on a belt (Oliveira, C., Salgado, J. and Carvalho, F. G. (1997) Optimisation of PGNAA instrument design for cement raw materials using the MCNP code. J. Radioanal. Nucl. Chem. 216(2), 191-198; Oliveira, C., Salgado, J., Goncalves, I. F., Carvalho, F. G. and Leitao, F. (1997a) A Monte Carlo study of the influence of geometry arrangements and structural materials on a PGNAA system performance for cement raw materials analysis. Appl. Radiat. Isot. (accepted); Oliveira, C., Salgado, J. and Leitao, F. (1997b) Density and water content corrections in the gamma count rate of a PGNAA system for cement raw material analysis using the MCNP code. Appl. Radiat. Isot. (accepted).]. It reports on the influence of the density, mass water content and thickness on the calibration curves of the PGNAA system. The MCNP-4A code, running in a Pentium-PC and in a DEC workstation, was used to simulate the PGNAA configuration system

30 CFR 27.38 - Tests to determine adequacy of windows and lenses.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Tests to determine adequacy of windows and... Tests to determine adequacy of windows and lenses. Impact tests. A 4-pound cylindrical weight with a one-inch diameter hemispherical striking surface will be dropped (free fall) to strike the window or lens...

Validating MCNP5 libraries and tracking the reason for differences between libraries in criticality calculations

International Nuclear Information System (INIS)

Hossny, K.

2015-01-01

The purpose of this work is to validate MCNP5 libraries by simulating 4 detailed benchmark experiments and comparing MCNP5 results (each library) with the experimental results and also the previously validated codes for the same experiments MORET 4.A coupled with APOLLO2 (France), and MONK8 (UK). The reasons for difference between libraries are also investigated in this work. Investigating the reason for the differences between libraries will be done by specifying a different library for specific part (clad, fuel, light water) and checking the result deviation than the previously calculated result (with all parts of the same library). The investigated benchmark experiments are of single fuel rods arrays that are water-moderated and water-reflected. Rods contained low-enriched (4.738 wt.% 92 235 U)uranium dioxide (UO 2 ) fuel were clad with aluminum alloy AGS. These experiments were subcritical approaches extrapolated to critical, with the multiplication factor reached being very close to 1.000 (within 0.1%); the subcritical approach parameter was the water level. The studied four cases differ from each other in pitch, number of fuel rods and of course critical height of water. The results show that although library ENDF/B-IV lacks light water treatment card, however its results can be reliable as light water treatment library does not have significant differences from library to another, so it will not be necessary to specify light water treatment card. The main reason for differences between ENDF/B-V and ENDF/B-VI is light water material, especially the Hydrogen element. Specifying the library of Uranium is necessary in case of using library ENDF/B-IV. On the other hand it is not necessary to specify library of cladding material whatever the used library. Validated libraries are ENDF/BIV, ENDF/B-V and ENDF/B-VI with codes in MCNP 42C, 50C and 60C respectively. The presentation slides have been added to the article

Estimation and interpretation of keff confidence intervals in MCNP

International Nuclear Information System (INIS)

Urbatsch, T.J.

1995-01-01

MCNP has three different, but correlated, estimators for Calculating k eff in nuclear criticality calculations: collision, absorption, and track length estimators. The combination of these three estimators, the three-combined k eff estimator, is shown to be the best k eff estimator available in MCNP for estimating k eff confidence intervals. Theoretically, the Gauss-Markov Theorem provides a solid foundation for MCNP's three-combined estimator. Analytically, a statistical study, where the estimates are drawn using a known covariance matrix, shows that the three-combined estimator is superior to the individual estimator with the smallest variance. The importance of MCNP's batch statistics is demonstrated by an investigation of the effects of individual estimator variance bias on the combination of estimators, both heuristically with the analytical study and emprically with MCNP

Estimation and interpretation of keff confidence intervals in MCNP

International Nuclear Information System (INIS)

Urbatsch, T.J.

1995-01-01

The Monte Carlo code MCNP has three different, but correlated, estimators for calculating k eff in nuclear criticality calculations: collision, absorption, and track length estimators. The combination of these three estimators, the three-combined k eff estimator, is shown to be the best k eff estimator available in MCNP for estimating k eff confidence intervals. Theoretically, the Gauss-Markov theorem provides a solid foundation for MCNP's three-combined estimator. Analytically, a statistical study, where the estimates are drawn using a known covariance matrix, shows that the three-combined estimator is superior to the estimator with the smallest variance. Empirically, MCNP examples for several physical systems demonstrate the three-combined estimator's superiority over each of the three individual estimators and its correct coverage rates. Additionally, the importance of MCNP's statistical checks is demonstrated

Investigation of Anisotropy Caused by Cylinder Applicator on Dose Distribution around Cs-137 Brachytherapy Source using MCNP4C Code

Directory of Open Access Journals (Sweden)

Sedigheh Sina

2011-06-01

Full Text Available Introduction: Brachytherapy is a type of radiotherapy in which radioactive sources are used in proximity of tumors normally for treatment of malignancies in the head, prostate and cervix. Materials and Methods: The Cs-137 Selectron source is a low-dose-rate (LDR brachytherapy source used in a remote afterloading system for treatment of different cancers. This system uses active and inactive spherical sources of 2.5 mm diameter, which can be used in different configurations inside the applicator to obtain different dose distributions. In this study, first the dose distribution at different distances from the source was obtained around a single pellet inside the applicator in a water phantom using the MCNP4C Monte Carlo code. The simulations were then repeated for six active pellets in the applicator and for six point sources.  Results: The anisotropy of dose distribution due to the presence of the applicator was obtained by division of dose at each distance and angle to the dose at the same distance and angle of 90 degrees. According to the results, the doses decreased towards the applicator tips. For example, for points at the distances of 5 and 7 cm from the source and angle of 165 degrees, such discrepancies reached 5.8% and 5.1%, respectively.  By increasing the number of pellets to six, these values reached 30% for the angle of 5 degrees. Discussion and Conclusion: The results indicate that the presence of the applicator causes a significant dose decrease at the tip of the applicator compared with the dose in the transverse plane. However, the treatment planning systems consider an isotropic dose distribution around the source and this causes significant errors in treatment planning, which are not negligible, especially for a large number of sources inside the applicator.

Monte Carlo simulation for treatment planning optimization of the COMS and USC eye plaques using the MCNP4C code

International Nuclear Information System (INIS)

Jannati Isfahani, A.; Shokrani, P.; Raisali, Gh.

2010-01-01

Ophthalmic plaque radiotherapy using I-125 radioactive seeds in removable episcleral plaques is often used in management of ophthalmic tumors. Radioactive seeds are fixed in a gold bowl-shaped plaque and the plaque is sutured to the scleral surface corresponding to the base of the intraocular tumor. This treatment allows for a localized radiation dose delivery to the tumor with a minimum target dose of 85 Gy. The goal of this study was to develop a Monte Carlo simulation method for treatment planning optimization of the COMS and USC eye plaques. Material and Methods: The MCNP4C code was used to simulate three plaques: COMS-12mm, COMS-20mm, and USC ≠9 with I-125 seeds. Calculation of dose was performed in a spherical water phantom (radius 12 mm) using a 3D matrix with a size of 12 voxels in each dimension. Each voxel contained a sphere of radius 1 mm. Results: Dose profiles were calculated for each plaque. Isodose lines were created in 2 planes normal to the axes of the plaque, at the base of the tumor and at the level of the 85 Gy isodose in a 7 day treatment. Discussion and Conclusion: This study shows that it is necessary to consider the following tumor properties in design or selection of an eye plaque: the diameter of tumor base, its thickness and geometric shape, and the tumor location with respect to normal critical structures. The plaque diameter is selected by considering the tumor diameter. Tumor thickness is considered when selecting the seed parameters such as their number, activity and distribution. Finally, tumor shape and its location control the design of following parameters: the shape and material of the plaque and the need for collimation.

MCNP trademark Software Quality Assurance plan

International Nuclear Information System (INIS)

Abhold, H.M.; Hendricks, J.S.

1996-04-01

MCNP is a computer code that models the interaction of radiation with matter. MCNP is developed and maintained by the Transport Methods Group (XTM) of the Los Alamos National Laboratory (LANL). This plan describes the Software Quality Assurance (SQA) program applied to the code. The SQA program is consistent with the requirements of IEEE-730.1 and the guiding principles of ISO 900

Validation suite for MCNP

International Nuclear Information System (INIS)

Mosteller, Russell D.

2002-01-01

Two validation suites, one for criticality and another for radiation shielding, have been defined and tested for the MCNP Monte Carlo code. All of the cases in the validation suites are based on experiments so that calculated and measured results can be compared in a meaningful way. The cases in the validation suites are described, and results from those cases are discussed. For several years, the distribution package for the MCNP Monte Carlo code1 has included an installation test suite to verify that MCNP has been installed correctly. However, the cases in that suite have been constructed primarily to test options within the code and to execute quickly. Consequently, they do not produce well-converged answers, and many of them are physically unrealistic. To remedy these deficiencies, sets of validation suites are being defined and tested for specific types of applications. All of the cases in the validation suites are based on benchmark experiments. Consequently, the results from the measurements are reliable and quantifiable, and calculated results can be compared with them in a meaningful way. Currently, validation suites exist for criticality and radiation-shielding applications.

Calculation of spatial weight functions for WWER-440 ex-core neutron detectors

International Nuclear Information System (INIS)

Csom, Gy.; Czifrus, Sz.; Feher, S.; Berki, T.

2001-01-01

The objective of the work presented in this paper was determination of a spatial weight function for WWER-440 ex-core detectors to be used for the interpretation of reload startup rod drop measurements. In view of the complexity of the geometry of the core as well as the detector, furthermore the presence of a cavity between the vessel and the concrete shield, Monte Carlo calculations were applied. In spite of the fact that in the corresponding literature the use of adjoint methods dominates, in the present case the forward method was chosen and implemented using MCNP4C (Authors)

Status Report on the MCNP 2020 Initiative

Energy Technology Data Exchange (ETDEWEB)

Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

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.

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

Energy Technology Data Exchange (ETDEWEB)

Mashnik, Stepan Georgievich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

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 improve 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.

MCNP analysis of the nine-cell LWR gadolinium benchmark

International Nuclear Information System (INIS)

Arkuszewski, J.J.

1988-01-01

The Monte Carlo results for a 9-cell fragment of the light water reactor square lattice with a central gadolinium-loaded pin are presented. The calculations are performed with the code MCNP-3A and the ENDF-B/5 library and compared with the results obtained from the BOXER code system and the JEF-1 library. The objective of this exercise is to study the feasibility of BOXER for the analysis of a Gd-loaded LWR lattice in the broader framework of GAP International Benchmark Analysis. A comparison of results indicates that, apart from unavoidable discrepancies originating from different data evaluations, the BOXER code overestimates the multiplication factor by 1.4 % and underestimates the power release in a Gd cell by 4.66 %. It is hoped that further similar studies with use of the JEF-1 library for both BOXER and MCNP will help to isolate and explain these discrepancies in a cleaner way. (author) 4 refs., 9 figs., 10 tabs

Neutron-induced photon production in MCNP

International Nuclear Information System (INIS)

Little, R.C.; Seamon, R.E.

1983-01-01

An improved method of neutron-induced photon production has been incorporated into the Monte Carlo transport code MCNP. The new method makes use of all partial photon-production reaction data provided by ENDF/B evaluators including photon-production cross sections as well as energy and angular distributions of secondary photons. This faithful utilization of sophisticated ENDF/B evaluations allows more precise MCNP calculations for several classes of coupled neutron-photon problems

Development and application of MCNP auto-modeling tool: Mcam 3.0

International Nuclear Information System (INIS)

Liu Xiaoping; Luo Yuetong; Tong Lili

2005-01-01

Mcam is abbreviation of 'MCNP Automatic Modeling', which is a CAD interface program of MCNP geometry model based on CAD technology. Making use of existing CAD technology is Mcam's major characteristic. In rough, CAD technology is utilized in the following two ways: (1) Mcam makes it possible to create MCNP geometry model in some CAD software; (2) accelerate creation of MCNP geometry model by inheriting some existing 3D CAD model. The paper gives an introduction of Mcam's major ability: (1) ability to convert CAD model into MCNP geometry model; (2) ability to convert MCNP geometry model into CAD model; (3) ability to construct CAD model. At the end of the paper, several models are given to demonstrate Mcam's different ability respectively

Verification of the AZNHEX code v.1.4 with MCNP6 for different reference cases

International Nuclear Information System (INIS)

Galicia A, J.; Francois L, J. L.; Bastida O, G. E.; Del Valle G, E.

2017-09-01

The codes that make up the AZTLAN platform (AZTHECA, AZTRAN, AZKIND and AZNHEX) are currently in the testing phase simulating a variety of nuclear reactor assemblies and cores to compare and validate the results obtained for a particular case, with codes globally used in the nuclear area such as CASMO, Serpent and MCNP. The objective of this work is to continue improving the future versions of the codes of the AZTLAN platform so that accurate and reliable results can be obtained for the user. To test the current version of the AZNHEX code, 3 cases were taken into account, the first being the simulation of a VVER-440 reactor assembly; for the second case, the assembly of a fast reactor cooled with helium was simulated and for the third case it was decided to take up the case of the core of a fast reactor cooled with sodium, this because the previous versions of AZNHEX did not show adequate results and, in addition, they presented a considerable amount of limitations. The comparison and validation of the results (neutron multiplication factor, radial power, radial flow, axial power) for these three cases were made using the code MCNP6. The results obtained show that this version of AZNHEX produces values of the neutron multiplication factor and the neutron and power flow distributions very close to those of MCNP6. (Author)

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

Science.gov (United States)

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.

Implementation of 3D models in the Monte Carlo code MCNP

International Nuclear Information System (INIS)

Lopes, Vivaldo; Millian, Felix M.; Guevara, Maria Victoria M.; Garcia, Fermin; Sena, Isaac; Menezes, Hugo

2009-01-01

On the area of numerical dosimetry Applied to medical physics, the scientific community focuses on the elaboration of new hybrids models based on 3D models. But different steps of the process of simulation with 3D models needed improvement and optimization in order to expedite the calculations and accuracy using this methodology. This project was developed with the aim of optimize the process of introduction of 3D models within the simulation code of radiation transport by Monte Carlo (MCNP). The fast implementation of these models on the simulation code allows the estimation of the dose deposited on the patient organs on a more personalized way, increasing the accuracy with this on the estimates and reducing the risks to health, caused by ionizing radiations. The introduction o these models within the MCNP was made through a input file, that was constructed through a sequence of images, bi-dimensional in the 3D model, generated using the program '3DSMAX', imported by the program 'TOMO M C' and thus, introduced as INPUT FILE of the MCNP code. (author)

E language based on MCNP modeling software for autonomous

International Nuclear Information System (INIS)

Li Fei; Ge Liangquan; Zhang Qingxian

2010-01-01

MCNP (Monte Carlo N-Particle Code) is based on the Monte Carlo method for computing neutron, photon and other particles as the object of the movement simulation computer program. Because of its powerful computing simulation, flexible and universal features in many fields has been widely used, but due to a software professional in the operating area has been greatly restricted, so that in later development has been greatly hindered. E-language was used in order to develop the autonomy of MCNP modeling software, used to address users not familiar with MCNP and can not create object model, get rid of dull red tape 'notebook' type of program type and built a new MCNP modeling system. (authors)

Development of MCNP interface code in HFETR

International Nuclear Information System (INIS)

Qiu Liqing; Fu Rong; Deng Caiyu

2007-01-01

In order to describe the HFETR core with MCNP method, the interface code MCNPIP for HFETR and MCNP code is developed. This paper introduces the core DXSY and flowchart of MCNPIP code, and the handling of compositions of fuel elements and requirements on hardware and software. Finally, MCNPIP code is validated against the practical application. (authors)

Water vapor self-continuum absorption measurements in the 4.0 and 2.1 μm transparency windows

Science.gov (United States)

Richard, L.; Vasilchenko, S.; Mondelain, D.; Ventrillard, I.; Romanini, D.; Campargue, A.

2017-11-01

In a recent contribution [A. Campargue, S. Kassi, D. Mondelain, S. Vasilchenko, D. Romanini, Accurate laboratory determination of the near infrared water vapor self-continuum: A test of the MT_CKD model. J. Geophys. Res. Atmos., 121,13,180-13,203, doi:10.1002/2016JD025531], we reported accurate water vapor absorption continuum measurements by Cavity Ring-down Spectroscopy (CRDS) and Optical-Feedback-Cavity Enhanced Absorption Spectroscopy (OF-CEAS) at selected spectral points of 4 near infrared transparency windows. In the present work, the self-continuum cross-sections, CS, are determined for two new spectral points. The 2491 cm-1 spectral point in the region of maximum transparency of the 4.0 μm window was measured by OF-CEAS in the 23-52 °C temperature range. The 4435 cm-1 spectral point of the 2.1 μm window was measured by CRDS at room temperature. The self-continuum cross-sections were determined from the pressure squared dependence of the continuum absorption. Comparison to the literature shows a reasonable agreement with 1970 s and 1980 s measurements using a grating spectrograph in the 4.0 μm window and a very good consistency with our previous laser measurements in the 2.1 μm window. For both studied spectral points, our values are much smaller than previous room temperature measurements by Fourier Transform Spectroscopy. Significant deviations (up to about a factor 4) are noted compared to the widely used semi empirical MT_CKD model of the absorption continuum. The measured temperature dependence at 2491 cm-1 is consistent with previous high temperature measurements in the 4.0 μm window and follows an exp(D0/kT) law, D0 being the dissociation energy of the water dimer.

Application of MCNP in the criticality calculation for reactors

International Nuclear Information System (INIS)

Zhong Zhaopeng; Shi Gong; Hu Yongming

2003-01-01

The criticality calculation is carried out with 3-D Monte Carlo code (MCNP). The author focuses on the introduction of modelling of the core and reflector. The core description is simplified by using repetition structure function of MCNP. k eff in different control rods positions are calculated for the case of JRR3, and the results is consistent with that of the reference. This work shows that MCNP is applicable for reactor criticality calculation

MCNP6 fragmentation of light nuclei at intermediate energies

Energy Technology Data Exchange (ETDEWEB)

Mashnik, Stepan G., E-mail: mashnik@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Kerby, Leslie M. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); University of Idaho, Moscow, ID 83844 (United States)

2014-11-11

Fragmentation reactions induced on light target nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below are studied with the latest Los Alamos Monte Carlo transport code MCNP6 and with its cascade-exciton model (CEM) and Los Alamos version of the quark-gluon string model (LAQGSM) event generators, version 03.03, used as stand-alone codes. Such reactions are involved in different applications, like cosmic-ray-induced single event upsets (SEU's), radiation protection, and cancer therapy with proton and ion beams, among others; therefore, it is important that MCNP6 simulates them as well as possible. CEM and LAQGSM assume that intermediate-energy fragmentation reactions on light nuclei occur generally in two stages. The first stage is the intranuclear cascade (INC), followed by the second, Fermi breakup disintegration of light excited residual nuclei produced after the INC. Both CEM and LAQGSM account also for coalescence of light fragments (complex particles) up to {sup 4}He from energetic nucleons emitted during INC. We investigate the validity and performance of MCNP6, CEM, and LAQGSM in simulating fragmentation reactions at intermediate energies and discuss possible ways of further improving these codes.

CTEx Beowulf cluster for MCNP performance

International Nuclear Information System (INIS)

Gonzaga, Roberto N.; Amorim, Aneuri S. de; Balthar, Mario Cesar V.

2011-01-01

This work is an introduction to the CTEx Nuclear Defense Department's Beowulf Cluster. Building a Beowulf Cluster is a complex learning process that greatly depends upon your hardware and software requirements. The feasibility and efficiency of performing MCNP5 calculations with a small, heterogeneous computing cluster built in Red Hat's Fedora Linux operating system personal computers (PC) are explored. The performance increases that may be expected with such clusters are estimated for cases that typify general radiation transport calculations. Our results show that the speed increase from additional slave PCs is nearly linear up to 10 processors. The pre compiled parallel binary version of MCNP uses the Message-Passing Interface (MPI) protocol. The use of this pre compiled parallel version of MCNP5 with the MPI protocol on a small, heterogeneous computing cluster built from Red Hat's Fedora Linux operating system PCs is the subject of this work. (author)

Search of the VAXintosh customizing VMS V4.0 for DCL windows

International Nuclear Information System (INIS)

Downward, J.G.

1985-01-01

This paper will describe methods of implementing such windows for: (1) DCL and MAIL; (2) Command Procedures; and (3) the VMS V4.0 editors EDT, LSE and TPU. While VMS and a number of utilities have the SPAWN command available to the user, the window interface to be described has been found in practice to be simpler to use because: (1) windows are consistently invoked with a single key command; and (2) confusion is minimized because one always knows if one is in a window

Benchmarking comparison and validation of MCNP photon interaction data

Directory of Open Access Journals (Sweden)

Colling Bethany

2017-01-01

Full Text Available 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.

Benchmarking comparison and validation of MCNP photon interaction data

Science.gov (United States)

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.

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

Science.gov (United States)

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.

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

Science.gov (United States)

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.

MCNP HPGe detector benchmark with previously validated Cyltran model.

Science.gov (United States)

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 wide-gap a-SiC:H PV-powered electrochromic window coating

Energy Technology Data Exchange (ETDEWEB)

Gao, W.; Lee, S.H.; Xu, Y.; Benson, D.K.; Deb, S.K.; Branz, H.M. [National Renewable Energy Lab., Golden, CO (United States)

1998-09-01

The authors report on the first monolithic, amorphous-silicon-based, photovoltaic-powered electrochromic window coating. The coating employs a wide bandgap a-Si{sub 1{minus}x}C{sub x}:H n-i-p photovoltaic (PV) cell as a semitransparent power supply, and a Li{sub y}WO{sub 3}/LiAlF{sub 4}/V{sub 2}O{sub 5} electrochromic (EC) device as an optical-transmittance modulator. The EC device is deposited directly on top of a PV cell that coats a glass substrate. The a-Si{sub 1{minus}x}C{sub x}:H PV cell has a Tauc gap of 2.2 eV and a transmittance of 60--80% over a large portion of the visible light spectrum. The authors reduced the thickness of the device to about 600 {angstrom} while maintaining a 1-sun open-circuit voltage of 0.9 V and short-circuit current of 2 mA/cm{sup 2}. The prototype 16 cm{sup 2} PV/EC device modulates the transmittance by more than 60% over a large portion of the visible spectrum. The coloring and bleaching times of the EC device are approximately 1 minute under normal operating conditions ({+-} 1 volt). A brief description of photoelectrochromic windows study is also given.

Simulations for the neutron detector TETRA with MCNP

International Nuclear Information System (INIS)

Testov, D.; Kuznetsova, E.; Wilson, Jh.

2013-01-01

To study the nuclear structure of β-delayed neutron precursors at ALTO ISOL-facility at IPN (Orsay), the high efficiency 4π neutron detector TETRA with 3 He filled counters built at JINR (Dubna) was modified. The MCNP simulations to optimize the future configuration were necessary. The details of the calculations and the major results obtained are discussed

Low Enrichment Uranium (LEU)-fueled SLOWPOKE-2 nuclear reactor simulation with the Monte-Carlo based MCNP 4A code

International Nuclear Information System (INIS)

Pierre, J.R.M.

1996-01-01

Following the commissioning of the Low Enrichment Uranium (LEU) Fuelled SLOWPOKE-2 research reactor at the Royal Military College-College Militaire Royal (RMC-CMR), excess reactivity measurements were conducted over a range of temperature and power. The results showed a maximum excess reactivity of 3.37 mk at 33 o C. Several deterministic models using computer codes like WIMS-CRNL, CITATION, TRIVAC and DRAGON have been used to try to reproduce the excess reactivity and temperature trend of both the LEU and HEU SLOWPOKE-2 reactors. The best simulations had been obtained at Ecole Polytechnique de Montreal. They were able to reproduce the temperature trend of their HEU-fuelled reactor using TRIVAC calculations, but this model over-estimated the absolute value of the excess reactivity by 119 mk. Although calculations using DRAGON did not reproduce the temperature trend as well as TRIVAC, these calculations represented a significant improvement on the absolute value at 20 o C reducing the discrepancy to 13 mk. Given the advance in computer technology, a probabilistic approach was tried in this work, using the Monte-Carlo N-Particle Transport Code System MCNP 4A, to model the RMC-CMR SLOWPOKE-2 reactor.

New Neutron, Proton, and S(α,β) MCNP Data Libraries Based on ENDF/B-VII

International Nuclear Information System (INIS)

Little, Robert C.; Trellue, Holly R.; MacFarlane, Robert E.; Kahler, A.C.; Lee, Mary Beth; White, Morgan C.

2008-01-01

The general-purpose Evaluated Nuclear Data File ENDF/B-VII.0 was released in December 2006. A number of sub-libraries were included in ENDF/B-VII.0 such that data were provided for incident neutrons, photons, and charged particles. This paper describes the creation of MCNP data libraries at Los Alamos National Laboratory based on three ENDF/B-VII.0 sub-libraries: neutrons, protons, and thermal scattering. An ACE-formatted continuous-energy neutron data library called ENDF70 for MCNP has been produced. This library provides data for 390 materials at five temperatures: 293.6, 600, 900, 1200, and 2500 K. The library was processed primarily with Version 248 of NJOY99. Extensive checking and quality-assurance tests were applied to the data. Improvements to the processing code were made and certain evaluations were modified as a result of these tests. ENDF/B-VII.0 included proton evaluations for 48 target materials. Forty-seven proton evaluations (all except for 13 C) were processed at room temperature and combined into the MCNP library ENDF70PROT. Neutron thermal S(α,β) scattering data exist for twenty different materials in ENDF/B-VII.0. All twenty of these evaluations were processed at all applicable temperatures (these vary for each evaluation), and combined into the MCNP library ENDF70SAB. All of these ENDF/B-VII.0 based MCNP libraries (ENDF70, ENDF70PROT, and ENDF70SAB) are available as part of the MCNP5 1.50 release. (authors)

MCNP6 Fission Cross Section Calculations at Intermediate and High Energies

OpenAIRE

Mashnik, Stepan G.; Sierk, Arnold J.; Prael, Richard E.

2013-01-01

MCNP6 has been Validated and Verified (V&V) against intermediate- and high-energy fission cross-section experimental data. An error in the calculation of fission cross sections of 181Ta and a few nearby target nuclei by the CEM03.03 event generator in MCNP6 and a "bug: in the calculation of fission cross sections with the GENXS option of MCNP6 while using the LAQGSM03.03 event generator were detected during our V&V work. After fixing both problems, we find that MCNP6 using CEM03.03 and LAQGSM...

Monte Carlo parameter studies and uncertainty analyses with MCNP5

International Nuclear Information System (INIS)

Brown, F. B.; Sweezy, J. E.; Hayes, R.

2004-01-01

A software tool called mcnp p study has been developed to automate the setup, execution, and collection of results from a series of MCNP5 Monte Carlo calculations. This tool provides a convenient means of performing parameter studies, total uncertainty analyses, parallel job execution on clusters, stochastic geometry modeling, and other types of calculations where a series of MCNP5 jobs must be performed with varying problem input specifications. (authors)

MCNP code

International Nuclear Information System (INIS)

Cramer, S.N.

1984-01-01

The MCNP code is the major Monte Carlo coupled neutron-photon transport research tool at the Los Alamos National Laboratory, and it represents the most extensive Monte Carlo development program in the United States which is available in the public domain. The present code is the direct descendent of the original Monte Carlo work of Fermi, von Neumaum, and Ulam at Los Alamos in the 1940s. Development has continued uninterrupted since that time, and the current version of MCNP (or its predecessors) has always included state-of-the-art methods in the Monte Carlo simulation of radiation transport, basic cross section data, geometry capability, variance reduction, and estimation procedures. The authors of the present code have oriented its development toward general user application. The documentation, though extensive, is presented in a clear and simple manner with many examples, illustrations, and sample problems. In addition to providing the desired results, the output listings give a a wealth of detailed information (some optional) concerning each state of the calculation. The code system is continually updated to take advantage of advances in computer hardware and software, including interactive modes of operation, diagnostic interrupts and restarts, and a variety of graphical and video aids

Benchmark of WIMS-IST against MCNP for CANDU pressure tube fast fluxes

International Nuclear Information System (INIS)

Donders, R.E.; Douglas, S.R.

2002-01-01

Pressure tube fast-flux data in CANDU are currently calculated using the multi-group neutron transport code WIMS-IST. In this study, the WIMS-IST fast flux calculations are benchmarked against MCNP calculations (a Monte Carlo particle transport code), over the range of fuel burnup and coolant density in CANDU. The comparison shows good agreement between WIMS and MCNP, with WIMS fast fluxes being 1.5% to 4% lower than the MCNP values. The difference is smallest for fresh fuel, and increases with burnup. The fast flux gradient across the pressure tube (factor of 1.23 from inner edge to outer edge) is accurately calculated by WIMS. When reporting fast fluxes in pressure tubes, these are generally given as >1.000 MeV fluxes. For WIMS, this requires an extra conversion step, since the WIMS ENDF/B libraries do not have a group boundary at 1 MeV. The conversion step is based on a fictitious isotope ONEMEV in the WIMS nuclear data library. The conversion factor in WIMS was found to be about one percent too high. When providing >1 MeV fluxes from WIMS, this partially compensates for the slight under prediction of the fast flux. Pressure tube >1 MeV fluxes from WIMS are therefore 0.5% to 3% lower than MCNP values. To obtain accurate fast flux data, neutron transport calculations must be performed on a critical cell. For this study, all calculations were performed with radial albedo boundary conditions giving a critical cell. This required the use of an albedo version of MCNP, developed at AECL. (author)

Learning Windows Azure Mobile Services for Windows 8 and Windows Phone 8

CERN Document Server

Webber-Cross, Geoff

2014-01-01

This book is based around a case study game which was written for the book. This means that the chapters progress in a logical way and build upon lessons learned as we go. Real-world examples are provided for each topic that are practical and not given out-of-context so they can be applied directly to other applications.If you are a developer who wishes to build Windows 8 and Phone 8 applications and integrate them with Windows Azure Mobile Services, this book is for you. Basic C# and JavaScript skills are advantageous, as well as some knowledge of building Windows 8 or Windows Phone 8 applica

Evaluation of various energy windows at different radionuclides for scatter and attenuation correction in nuclear medicine.

Science.gov (United States)

Asgari, Afrouz; Ashoor, Mansour; Sohrabpour, Mostafa; Shokrani, Parvaneh; Rezaei, Ali

2015-05-01

Improving signal to noise ratio (SNR) and qualified images by the various methods is very important for detecting the abnormalities at the body organs. Scatter and attenuation of photons by the organs lead to errors in radiopharmaceutical estimation as well as degradation of images. The choice of suitable energy window and the radionuclide have a key role in nuclear medicine which appearing the lowest scatter fraction as well as having a nearly constant linear attenuation coefficient as a function of phantom thickness. The energy windows of symmetrical window (SW), asymmetric window (ASW), high window (WH) and low window (WL) using Tc-99m and Sm-153 radionuclide with solid water slab phantom (RW3) and Teflon bone phantoms have been compared, and Matlab software and Monte Carlo N-Particle (MCNP4C) code were modified to simulate these methods and obtaining the amounts of FWHM and full width at tenth maximum (FWTM) using line spread functions (LSFs). The experimental data were obtained from the Orbiter Scintron gamma camera. Based on the results of the simulation as well as experimental work, the performance of WH and ASW display of the results, lowest scatter fraction as well as constant linear attenuation coefficient as a function of phantom thickness. WH and ASW were optimal windows in nuclear medicine imaging for Tc-99m in RW3 phantom and Sm-153 in Teflon bone phantom. Attenuation correction was done for WH and ASW optimal windows and for these radionuclides using filtered back projection algorithm. Results of simulation and experimental show that very good agreement between the set of experimental with simulation as well as theoretical values with simulation data were obtained which was nominally less than 7.07 % for Tc-99m and less than 8.00 % for Sm-153. Corrected counts were not affected by the thickness of scattering material. The Simulated results of Line Spread Function (LSF) for Sm-153 and Tc-99m in phantom based on four windows and TEW method were

UNR. A code for processing unresolved resonance data for MCNP

International Nuclear Information System (INIS)

Hogenbirk, A.

1994-09-01

In neutron transport problems the correct treatment of self-shielding is important for those nuclei present in large concentrations. Monte Carlo calculations using continuous-energy cross section data, such as calculations with the code MCNP, offer the advantage that neutron transport is calculated in a very accurate way. Self-shielding in the resolved resonance region is taken into account exactly in MCNP. However, self-shielding in the unresolved resonance region can not be taken into account by MCNP, although the effect of it may be important in many applications. In this report a description is given of the computer code UNR. With this code problem-dependent cross section libraries can be produced for MCNP. In these libraries self-shielded cross section data in the unresolved resonance range are given, which are produced by NJOY-module UNRESR. It is noted, that the treatment for resonance self-shielding presented in this report is approximate. However, the current version of MCNP does not allow the use of probability tables, which would be a general solution. (orig.)

A comparison study for mass attenuation coefficients of some amino acids using MCNP code

Energy Technology Data Exchange (ETDEWEB)

Vahabi, Seyed Milad; Bahreynipour, Mostean; Shamsaie-Zafarghandi, Mojtaba [Amirkabir Univ. of Technology, Tehran (Iran, Islamic Republic of). Dept. of Energy Engineering and Physics

2017-07-15

In this study, a novel model of MCNP4C code reported recently was used to determine the photon mass attenuation coefficients of some amino acids at energies, 123, 360, 511, 662, 1170, 1280 and 1330 keV. The simulation results were compared with the XCOM data. It was indicated that the results were highly close to the calculated XCOM values. Obtained results were used to calculate the molar extinction coefficient. All the results showed the convenience and usefulness of the model in calculation of mass attenuation coefficients of amino acids.

MCNP modelling of a combined neutron/gamma counter

CERN Document Server

Bourva, L C A; Ottmar, H; Weaver, D R

1999-01-01

A series of Monte Carlo neutron calculations for a combined gamma/passive neutron coincidence counter has been performed. This type of device, part of a suite of non-destructive assay instruments utilised for the enforcement of the Euratom nuclear safeguards within the European Union, is to be used for high accuracy measurements of the plutonium content of small samples of nuclear materials. The multi-purpose Monte Carlo N-particle (MCNP) code version 4B has been used to model in detail the neutron coincidence detector and to investigate the leakage self-multiplication of PuO sub 2 and mixed U-Pu oxide (MOX) reference samples used to calibrate the instrument. The MCNP calculations have been used together with a neutron coincidence counting interpretative model to determine characteristic parameters of the detector. A comparative study to both experimental and previous numerical results has been performed. Sensitivity curves of the variation of the detector's efficiency, epsilon, to, alpha, the ratio of (alpha...

Spectral measurements in critical assemblies: MCNP specifications and calculated results

Energy Technology Data Exchange (ETDEWEB)

Stephanie C. Frankle; Judith F. Briesmeister

1999-12-01

Recently, a suite of 86 criticality benchmarks for the Monte Carlo N-Particle (MCNP) transport code was developed, and the results of testing the ENDF/B-V and ENDF/B-VI data (through Release 2) were published. In addition to the standard k{sub eff} measurements, other experimental measurements were performed on a number of these benchmark assemblies. In particular, the Cross Section Evaluation Working Group (CSEWG) specifications contain experimental data for neutron leakage and central-flux measurements, central-fission ratio measurements, and activation ratio measurements. Additionally, there exists another set of fission reaction-rate measurements performed at the National Institute of Standards and Technology (NIST) utilizing a {sup 252}Cf source. This report will describe the leakage and central-flux measurements and show a comparison of experimental data to MCNP simulations performed using the ENDF/B-V and B-VI (Release 2) data libraries. Central-fission and activation reaction-rate measurements will be described, and the comparison of experimental data to MCNP simulations using available data libraries for each reaction of interest will be presented. Finally, the NIST fission reaction-rate measurements will be described. A comparison of MCNP results published previously with the current MCNP simulations will be presented for the NIST measurements, and a comparison of the current MCNP simulations to the experimental measurements will be presented.

Spectral measurements in critical assemblies: MCNP specifications and calculated results

International Nuclear Information System (INIS)

Frankle, Stephanie C.; Briesmeister, Judith F.

1999-01-01

Recently, a suite of 86 criticality benchmarks for the Monte Carlo N-Particle (MCNP) transport code was developed, and the results of testing the ENDF/B-V and ENDF/B-VI data (through Release 2) were published. In addition to the standard k eff measurements, other experimental measurements were performed on a number of these benchmark assemblies. In particular, the Cross Section Evaluation Working Group (CSEWG) specifications contain experimental data for neutron leakage and central-flux measurements, central-fission ratio measurements, and activation ratio measurements. Additionally, there exists another set of fission reaction-rate measurements performed at the National Institute of Standards and Technology (NIST) utilizing a 252 Cf source. This report will describe the leakage and central-flux measurements and show a comparison of experimental data to MCNP simulations performed using the ENDF/B-V and B-VI (Release 2) data libraries. Central-fission and activation reaction-rate measurements will be described, and the comparison of experimental data to MCNP simulations using available data libraries for each reaction of interest will be presented. Finally, the NIST fission reaction-rate measurements will be described. A comparison of MCNP results published previously with the current MCNP simulations will be presented for the NIST measurements, and a comparison of the current MCNP simulations to the experimental measurements will be presented

Development and improvement for MCNP-3B interactive plotter

International Nuclear Information System (INIS)

Gao Yanfeng

1996-01-01

The author briefly explains the development and improvement for the MCNP-3B interactive plotter. It describes the functions of geometry visualization and tally result plot, and introduces the progresses in user interface, process display and surface matching. The construction of MCNP-3B/PC is given

Building a Laboratory Information Management System Using Windows4GL

International Nuclear Information System (INIS)

Pickens, M.A.; Shaieb, M.R.

1996-05-01

The system discussed is currently implemented at LLNL in the Environmental Services program which operates out of the Chemistry ampersand Materials Science (C ampersand MS) directorate. Responsibility is to provide the C ampersand MS Environmental Services (CES) program with an enterprise-wide information system which will aid CES. The specific portion of the information system is the Sample Tracking, Analysis and Reporting System (STARS). Since CES was formed by merging two analytical laboratory organizations in May 1995, a new Laboratory Information Management System (LIMS) had to be developed. The development of a LIMS in Windows4GL was found to be satisfactory. The product STARS was well received by the user community, and it has improved business practices and efficiency in CES. The CES management staff has seen increased personnel productivity since STARS was release. We look forward to upgrading to CA-OpenROAD and taking advantage of its many improved and innovative features to further enhance STARS

Building a Laboratory Information Management System Using Windows4GL

Energy Technology Data Exchange (ETDEWEB)

Pickens, M.A.; Shaieb, M.R.

1996-05-01

The system discussed is currently implemented at LLNL in the Environmental Services program which operates out of the Chemistry & Materials Science (C&MS) directorate. Responsibility is to provide the C&MS Environmental Services (CES) program with an enterprise-wide information system which will aid CES. The specific portion of the information system is the Sample Tracking, Analysis and Reporting System (STARS). Since CES was formed by merging two analytical laboratory organizations in May 1995, a new Laboratory Information Management System (LIMS) had to be developed. The development of a LIMS in Windows4GL was found to be satisfactory. The product STARS was well received by the user community, and it has improved business practices and efficiency in CES. The CES management staff has seen increased personnel productivity since STARS was release. We look forward to upgrading to CA-OpenROAD and taking advantage of its many improved and innovative features to further enhance STARS.

Simulation of reactor noise analysis measurement for light-water critical assembly TCA using MCNP-DSP

International Nuclear Information System (INIS)

Yamamoto, Toshihiro; Sakurai, Kiyoshi; Tonoike, Kotaro; Miyoshi, Yoshinori

2001-01-01

Reactor noise analysis methods using Monte Carlo technique have been proposed and developed in the field of nuclear criticality safety. The Monte Carlo simulation for noise analysis can be made by simulating physical phenomena in the course of neutron transport in a nuclear fuel as practically as possible. MCNP-DSP was developed by T. Valentine of ORNL for this purpose and it is a modified version of MCNP-4A. The authors applied this code to frequency analysis measurements performed in light-water critical assembly TCA. Prompt neutron generation times for critical and subcritical cores were measured by doing the frequency analysis of detector signals. The Monte Carlo simulations for these experiments were carried out using MCNP-DSP, and prompt neutron generation times were calculated. (author)

Development of automatic editing system for MCNP library 'autonj'

International Nuclear Information System (INIS)

Maekawa, Fujio; Sakurai, Kiyoshi; Kume, Etsuo; Nomura, Yasushi; Kosako, Kazuaki; Kawasaki, Nobuo; Naito, Yoshitaka

1999-12-01

As an activity of the MCNP High-Temperature Library Production Working Group under the Nuclear Code Evaluation Special Committee of Nuclear Code Committee, the automatic editing system for MCNP library 'autonj' was developed. The autonj includes the NJOY-97 code as its main body, and is a system that enables us to easily produce cross section libraries for MCNP from evaluated nuclear data files such as JENDL-3.2. A temperature dependent library at six temperature points based on JENDL-3.2 was produced by using autonj. The autonj system and the temperature dependent library were installed on the JAERI AP3000 computer. (author)

A review of radiation dosimetry applications using the MCNP Monte Carlo code

Energy Technology Data Exchange (ETDEWEB)

Solberg, T.D.; DeMarco, J.J.; Chetty, I.J.; Mesa, A.V.; Cagnon, C.H.; Li, A.N.; Mather, K.K.; Medin, P.M.; Arellano, A.R.; Smathers, J.B. [California Univ., Los Angeles, CA (United States). Dept. of Radiation Oncology

2001-07-01

The Monte Carlo code MCNP (Monte Carlo N-Particle) has a significant history dating to the early years of the Manhattan Project. More recently, MCNP has been used successfully to solve many problems in the field of medical physics. In radiotherapy applications MCNP has been used successfully to calculate the bremsstrahlung spectra from medical linear accelerators, for modeling the dose distributions around high dose rate brachytherapy sources, and for evaluating the dosimetric properties of new radioactive sources used in intravascular irradiation for prevention of restenosis following angioplasty. MCNP has also been used for radioimmunotherapy and boron neutron capture therapy applications. It has been used to predict fast neutron activation of shielding and biological materials. One area that holds tremendous clinical promise is that of radiotherapy treatment planning. In diagnostic applications, MCNP has been used to model X-ray computed tomography and positron emission tomography scanners, to compute the dose delivered from CT procedures, and to determine detector characteristics of nuclear medicine devices. MCNP has been used to determine particle fluxes around radiotherapy treatment devices and to perform shielding calculations in radiotherapy treatment rooms. This manuscript is intended to provide to the reader a comprehensive summary of medical physics applications of the MCNP code. (orig.)

A review of radiation dosimetry applications using the MCNP Monte Carlo code

International Nuclear Information System (INIS)

Solberg, T.D.; DeMarco, J.J.; Chetty, I.J.; Mesa, A.V.; Cagnon, C.H.; Li, A.N.; Mather, K.K.; Medin, P.M.; Arellano, A.R.; Smathers, J.B.

2002-01-01

The Monte Carlo code MCNP (Monte Carlo N-Particle) has a significant history dating to the early years of the Manhattan Project. More recently, MCNP has been used successfully to solve many problems in the field of medical physics. In radiotherapy applications MCNP has been used successfully to calculate the bremsstrahlung spectra from medical linear accelerators, for modeling the dose distributions around high dose rate brachytherapy sources, and for evaluating the dosimetric properties of new radioactive sources used in intravascular irradiation for prevention of restenosis following angioplasty. MCNP has also been used for radioimmunotherapy and boron neutron capture therapy applications. It has been used to predict fast neutron activation of shielding and biological materials. One area that holds tremendous clinical promise is that of radiotherapy treatment planning. In diagnostic applications, MCNP has been used to model X-ray computed tomography and positron emission tomography scanners, to compute the dose delivered from CT procedures, and to determine detector characteristics of nuclear medicine devices. MCNP has been used to determine particle fluxes around radiotherapy treatment devices and to perform shielding calculations in radiotherapy treatment rooms. This manuscript is intended to provide to the reader a comprehensive summary of medical physics applications of the MCNP code. (author)

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

Science.gov (United States)

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.

About the application of MCNP4 code in nuclear reactor core design calculations

International Nuclear Information System (INIS)

Svarny, J.

2000-01-01

This paper provides short review about application of MCNP code for reactor physics calculations performed in SKODA JS. Problems of criticality safety analysis of spent fuel systems for storage and transport of spent fuel are discussed and relevant applications are presented. Application of standard Monte Carlo code for accelerator driven system for LWR waste destruction is shown and conclusions are reviewed. Specific heterogeneous effects in neutron balance of WWER nuclear cores are solved for adjusting standard design codes. (Authors)

Adjoint-Based Uncertainty Quantification with MCNP

Energy Technology Data Exchange (ETDEWEB)

Seifried, Jeffrey E. [Univ. of California, Berkeley, CA (United States)

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 in the simulation is acquired.

A Temperature Window for the Synthesis of Single-Walled Carbon Nanotubes by Catalytic Chemical Vapor Deposition of CH4over Mo2-Fe10/MgO Catalyst

Directory of Open Access Journals (Sweden)

Yu Ouyang

2009-01-01

Full Text Available Abstract A temperature window for the synthesis of single-walled carbon nanotubes by catalytic chemical vapor deposition of CH4over Mo2-Fe10/MgO catalyst has been studied by Raman spectroscopy. The results showed that when the temperature is lower than 750 °C, there were few SWCNTs formed, and when the temperature is higher than 950 °C, mass amorphous carbons were formed in the SWCNTs bundles due to the self-decomposition of CH4. The temperature window of SWCNTs efficient growth is between 800 and 950 °C, and the optimum growth temperature is about 900 °C. These results were supported by transmission electron microscope images of samples formed under different temperatures. The temperature window is important for large-scale production of SWCNTs by catalytic chemical vapor deposition method.

Monte Carlo simulation using MCNP4B for an optimal shielding design of a 252 Cf source

International Nuclear Information System (INIS)

Silva, Ademir X. da; Crispim, Verginia R.

2001-01-01

This study aim to investigate an optimum shielding design against neutrons and gamma-rays from a source of 252 Cf, using Monte Carlo simulation. The shielding materials studied were: borated polyethylene, borated-lead polyethylene and stainless steel. The Monte Carlo code MCNP, version 4B, was used to design shielding for 252 Cf based neutron irradiator systems. By normalizing the dose equivalent rate values presented to the neutron production rate of the source, the resulting calculations are independents of the intensity of actual 252 Cf source. The results shown what the total dose equivalent rates were reduced significantly by the shielding system optimization. (author)

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

Science.gov (United States)

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.

Systems guide to MCNP (Monte Carlo Neutron and Photon Transport Code)

International Nuclear Information System (INIS)

Kirk, B.L.; West, J.T.

1984-06-01

The subject of this report is the implementation of the Los Alamos National Laboratory Monte Carlo Neutron and Photon Transport Code - Version 3 (MCNP) on the different types of computer systems, especially the IBM MVS system. The report supplements the documentation of the RSIC computer code package CCC-200/MCNP. Details of the procedure to follow in executing MCNP on the IBM computers, either in batch mode or interactive mode, are provided

Use of McCad for the conversion of ITER CAD data to MCNP geometry

International Nuclear Information System (INIS)

Tsige-Tamirat, H.; Fischer, U.; Serikov, A.; Stickel, S.

2008-01-01

The program McCad provides a CAD interface for the Monte Carlo transport code MCNP. It is able to convert CAD data into MCNP input geometry description and provides GUI components for modeling, visualization, and data exchange. It performs sequences of tests on CAD data to check its validity and neutronics appropriateness including completion of the final MCNP model by void geometries. McCad has been used to convert a 40 deg. ITER torus sector CAD model to a suitable MCNP geometry model. Results of MCNP calculations performed to validate the converted geometry are presented

Generating and verification of ACE-multigroup library for MCNP

International Nuclear Information System (INIS)

Chen Chaobin; Hu Zehua; Chen Yixue; Wu Jun; Yang Shouhai

2012-01-01

The Monte Carlo code MCNP can handle multigroup calculations and a sample multigroup set based on ENDF/B-V, MGXSNP, is available for MCNP for coupled neutron-photon transport. However, this library is not suit- able for all problems, and there is a need for users to be able to generate multigroup libraries tailored to their specific applications. For these purposes CSPT (cross section processing tool) is created to generate multigroup library for MCNP from deterministic multigroup cross sections (GENDF or ANISN format at present). Several ACE-multigroup libraries based on ENDF/B-VII.0 converted and verified in this work, we drawn the conclusion that the CSPT code works correctly and the libraries produced are credible. (authors)

Validation and verification of MCNP6 against intermediate and high-energy experimental data and results by other codes

International Nuclear Information System (INIS)

Mashnik, Stepan G.

2011-01-01

MCNP6, the latest and most advanced LANL transport code representing a recent merger of MCNP5 and MCNPX, has been Validated and Verified (V and V) against a variety of intermediate and high-energy experimental data and against results by different versions of MCNPX and other codes. In the present work, we V and V MCNP6 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 CEM03.02 and LAQGSM03.03. We found that MCNP6 describes reasonably well various reactions induced by particles and nuclei at incident energies from 18 MeV to about 1 TeV per nucleon measured on thin and thick targets and agrees very well with similar results obtained with MCNPX and calculations by CEM03.02, LAQGSM03.01 (03.03), INCL4 + ABLA, and Bertini INC + Dresner evaporation, EPAX, ABRABLA, HIPSE, and AMD, used as stand alone codes. Most of several computational bugs and more serious physics problems observed in MCNP6/X during our V and V have been fixed; we continue our work to solve all the known problems before MCNP6 is distributed to the public. (author)

Weighted combination of LOD values oa splitted into frequency windows

Science.gov (United States)

Fernandez, L. I.; Gambis, D.; Arias, E. F.

In this analysis a one-day combined time series of LOD(length-of-day) estimates is presented. We use individual data series derived by 7 GPS and 3 SLR analysis centers, which routinely contribute to the IERS database over a recent 27-month period (Jul 1996 - Oct 1998). The result is compared to the multi-technique combined series C04 produced by the Central Bureau of the IERS that is commonly used as a reference for the study of the phenomena of Earth rotation variations. The Frequency Windows Combined Series procedure brings out a time series, which is close to C04 but shows an amplitude difference that might explain the evident periodic behavior present in the differences of these two combined series. This method could be useful to generate a new time series to be used as a reference in the high frequency variations of the Earth rotation studies.

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

Science.gov (United States)

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.

Comparison of the Window-Frame RHIC-abort kicker with C-type Kicker

International Nuclear Information System (INIS)

Tsoupas, N.; McMahan, Brandon

2014-01-01

The high intensity proton bunches (~2.5x10 11 p/bunch ) circulating in RHIC increase the temperature of the ferrite-made RHIC-abort-kickers above the Curie point; as a result, the kickers cannot provide the required field to abort the beam at the beam dump. A team of experts in the CAD department worked on modifying the design of the window-frame RHIC-abort kicker to minimize the hysteresis losses responsible for the increase of the ferrite's temperature. In this technical note we report some results from the study of two possible modifications of the window-frame RHIC-abort kicker, and we compare these results with those of a propose C-type RHIC-abort kicker. We also include an Appendix where we describe a method which may further reduce the hysteresis losses of the window-frame kicker.

Development of automatic cross section compilation system for MCNP

International Nuclear Information System (INIS)

Maekawa, Fujio; Sakurai, Kiyoshi

1999-01-01

A development of a code system to automatically convert cross-sections for MCNP is in progress. The NJOY code is, in general, used to convert the data compiled in the ENDF format (Evaluated Nuclear Data Files by BNL) into the cross-section libraries required by various reactor physics codes. While the cross-section library: FSXLIB-J3R2 was already converted from the JENDL-3.2 version of Japanese Evaluated Nuclear Data Library for a continuous energy Monte Carlo code MCNP, the library keeps only the cross-sections at room temperature (300 K). According to the users requirements which want to have cross-sections at higher temperature, say 600 K or 900 K, a code system named 'autonj' is under development to provide a set of cross-section library of arbitrary temperature for the MCNP code. This system can accept any of data formats adopted JENDL that may not be treated by NJOY code. The input preparation that is repeatedly required at every nuclide on NJOY execution is greatly reduced by permitting the conversion process of as many nuclides as the user wants in one execution. A few MCNP runs were achieved for verification purpose by using two libraries FSXLIB-J3R2 and the output of autonj'. The almost identical MCNP results within the statistical errors show the 'autonj' output library is correct. In FY 1998, the system will be completed, and in FY 1999, the user's manual will be published. (K. Tsuchihashi)

Frit Development Efforts for Sludge Batch 4 (SB4): Operating Window Assessments of Scenarios Leading Up to the Selected Preparation Plan for SB4

International Nuclear Information System (INIS)

Peeler, D

2006-01-01

The objective of this report is to document technical information that has been provided to Defense Waste Processing Facility (DWPF) and Closure Business Unit (CBU) personnel as part of the frit development support for Sludge Batch 4 (SB4). The information presented in this report includes projected operating windows (expressed in terms of waste loading) for various sludge blending and/or washing options coupled with candidate frits of interest. Although the Nominal Stage assessment serves as the primary tool for these evaluations, select systems were also evaluated using a Variation Stage assessment in which compositional variations were introduced. In addition, assessments of the impacts of nepheline formation potential and the SO 4 - solubility limit on the projected operating windows are also provided. Although this information was used as part of the technical basis leading to CBU's development of the preferred SB4 preparation plan, none of the options presented in this report was selected as the preferred plan. Therefore, the information is presented without significant interpretation of the resulting operating windows, but the projected windows are provided so additional insight can be explored if desired. Detailed assessments of the projected operating windows (using both Nominal and Variation Stage assessments) of the preferred sludge preparation plan with candidate frits are to be documented elsewhere. The information provided in this report is focused solely on model-based projections of the operating windows for various SB4 blending strategies of interest. Although nepheline formation potential is monitored via model predictions as a part of this assessment, experimental work investigating the impact of nepheline on glass quality is also being addressed in a parallel study. The results of this paper study and the experimental assessments of melt rate, SO 4 solubility, and/or nepheline formation potential are all critical components of the inputs into

Thermal Analysis of the Al Window for a New CESR-c Luminosity Monitor

CERN Document Server

He, Yun; Palmer, Mark A; Rice, David

2005-01-01

A luminosity monitor using photons from radiative bhabha events at the CLEO interaction point (IP) has been installed in the Cornell Electron Storage Ring (CESR). A key vacuum and detector component is the photon window/converter whose uniformity and thickness are critical for determining the resolution of the total energy deposited in the segmented luminosity monitor. The window design must accommodate the operational requirements of the new monitor at CLEO-c beam energies of 1.5-2.5 GeV and also provide sufficient safety margin for operation at 5.3 GeV beam energies for Cornell High Energy Synchrotron Source (CHESS) running. During 5.3 GeV operation, intense stripes of synchrotron radiation from the interaction region superconducting quadrupole magnets as well as nearby bending magnets strike the window. During the course of window development, several materials and designs were evaluated. Thermal stresses were calculated using the finite element code ANSYS for various beam conditions to guide the cooling d...

Data analysis and visualization in MCNP trademark

International Nuclear Information System (INIS)

Waters, L.S.

1994-01-01

There are many situations where the user may wish to go beyond current MCNP capabilities. For example, data produced by the code may need formatting for input into an external graphics package. Limitations on disk space may hinder writing out large PTRAK files. Specialized data analysis routines may be needed to model complex experimental results. One may wish to produce particle histories in a format not currently available in the code. To address these and other similar concerns a new capability in MCNP is being tested. A number of real, integer, logical and character variables describing the current and past characteristics of a particle are made available online to the user in three subroutines. The type of data passed can be controlled by cards in the INP file. The subroutines otherwise are empty, and the user may code in any desired analysis. A new MCNP executable is produced by compiling these subroutines and linking to a library which contains the object files for the rest of the code

Improved photon production data for MCNP trademark

International Nuclear Information System (INIS)

Adams, A.A.; Frankle, S.C.; Little, R.C.

1998-04-01

Computer simulations with MCNP are often used to obtain information from measurements of neutron induced gamma-ray spectra. For such simulations to be useful, the complicated spectra produced by a wide variety of nuclides must be reproduced, requiring high quality nuclear data. A previous assessment of the neutron induced photon production data in the MCNP data libraries indicated a need for improvement. The photon production data were often based on outdated experiments and binned in such wide energy groups as to be of limited value for some applications. This paper describes the work that is underway at Los Alamos National Laboratory to improve the photon production data for thermal neutron capture reactions. To date, high quality photon production data for each stable isotope of chlorine, chromium, iron, copper, and nickel have been obtained. The improved spectra have been incorporated into ENDF formatted evaluations and processed into corresponding MCNP data files. Similar improvements for aluminum, manganese, silicon, calcium, and vanadium are also planned. The methodology used to produce the spectra is discussed, and sample results for chlorine are presented

A graphical user interface for diagnostic radiology dosimetry using Monte Carlo (MCNP) simulation

International Nuclear Information System (INIS)

Collins, P.J.; Gorbatkov, D.; Schultz, F.W.

2000-01-01

Monte Carlo methods (for example, MCNP, EGGS4) are the 'gold standard' for both external and internal dosimetry in humans. These powerful simulation tools are, however, general-purpose codes and consequently do not provide a simple user interface for specific dosimetry tasks. We have developed a graphical user interface, for external radiation dosimetry (diagnostic radiology) using MCNP and an anthropomorphic mathematical phantom (Adam/Eva), which enables convenient modification and processing of the MCNP input and output files. The input form displays a colour coded, 3D representation of the phantom with a superimposed 'beam' for the required x-ray projection. The phantom can be rotated through 360 degrees and a transverse section at the level of the mid-point of the beam is also displayed. Text fields enable entry of input data (beam dimensions, source position, kVp, total filtration, focus-to-skin distance). A pull-down menu enables the user to select from 22 standard radiographic views. A standard projection can be modified, or new projection data entered if required. The input program modifies the MCNP input file and initiates processing. An output form displays the organ doses, normalised to unit skin entrance dose (with backscatter) (SED). The user can also enter the SED (calculated or measured) for a particular machine, to obtain the effective dose. To validate the program, the results for a PA Chest study (80 kVp, 2.5 mm Al total filtration) were compared with NRPB data (Jones and Wall, 1985). In conclusion, a convenient and reliable graphical user interface has been developed for MCNP, which enables dosimetry calculation for a full range of diagnostic radiological studies. (author)

Pebble bed reactors simulation using MCNP: The Chinese HTR-10 reactor

Directory of Open Access Journals (Sweden)

SA Hosseini

2013-09-01

Full Text Available   Given the role of Gas-Graphite reactors as the fourth generation reactors and their recently renewed importance, in 2002 the IAEA proposed a set of Benchmarking problems. In this work, we propose a model both efficient in time and resources and exact to simulate the HTR-10 reactor using MCNP-4C code. During the present work, all of the pressing factors in PBM reactor design such as the inter-pebble leakage, fuel particle distribution and fuel pebble packing fraction effects have been taken into account to obtain an exact and easy to run model. Finally, the comparison between the results of the present work and other calculations made at INEEL proves the exactness of the proposed model.

MCNP-REN a Monte Carlo tool for neutron detector design

CERN Document Server

Abhold, M E

2002-01-01

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 code developed at Los Alamos National Laboratory, Monte Carlo N-Particle (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, predicts neutron detector response without using the point 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 mixed oxide fresh fuel w...

Modeling the irradiation facility in the Deir Al-Hajar area to calculate the spatial gamma dose distribution using the MCNP code

International Nuclear Information System (INIS)

Khattab, K.; Bush, M; Kassery, H.

2009-03-01

A 3-D model for the irradiation plant which belongs to the Atomic Energy Commission, Department of Radiation Technology in the Deir Al-Hajar area near Damascus, is presented in this work using the MCNP-4C code. This model is used to calculate the spatial gamma ray dose in the (x, y, z) coordinate. Good agreements are noticed between the measured and the calculated results. (author)

First results of saturation curve measurements of heat-resistant steel using GEANT4 and MCNP5 codes

International Nuclear Information System (INIS)

Hoang, Duc-Tam; Tran, Thien-Thanh; Le, Bao-Tran; Vo, Hoang-Nguyen; Chau, Van-Tao; Tran, Kim-Tuyet; Huynh, Dinh-Chuong

2015-01-01

A gamma backscattering technique is applied to calculate the saturation curve and the effective mass attenuation coefficient of material. A NaI(Tl) detector collimated by collimator of large diameter is modeled by Monte Carlo technique using both MCNP5 and GEANT4 codes. The result shows a good agreement in response function of the scattering spectra for the two codes. Based on such spectra, the saturation curve of heat-resistant steel is determined. The results represent a strong confirmation that it is appropriate to use the detector collimator of large diameter to obtain the scattering spectra and this work is also the basis of experimental set-up for determining the thickness of material. (author)

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

Science.gov (United States)

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.

MCNP6 Simulation of Light and Medium Nuclei Fragmentation at Intermediate Energies

Energy Technology Data Exchange (ETDEWEB)

Mashnik, Stepan Georgievich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kerby, Leslie Marie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

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 using 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.

MCNP-REN: a Monte Carlo tool for neutron detector design

International Nuclear Information System (INIS)

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

2002-01-01

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 code developed at Los Alamos National Laboratory, Monte Carlo N-Particle (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, predicts neutron detector response without using the point 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 mixed oxide fresh fuel were taken with the Underwater Coincidence Counter, and measurements of highly enriched uranium reactor fuel were taken with the active neutron interrogation Research Reactor Fuel Counter and compared to calculation. Simulations completed for other detector design applications are described. 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

The use of the MCNP code for the quantitative analysis of elements in geological formations

Energy Technology Data Exchange (ETDEWEB)

Cywicka-Jakiel, T.; Woynicka, U. [The Henryk Niewodniczanski Institute of Nuclear Physics, Krakow (Poland); Zorski, T. [University of Mining and Metallurgy, Faculty of Geology, Geophysics and Environmental Protection, Krakow (Poland)

2003-07-01

The Monte Carlo modelling calculations using the MCNP code have been performed, which support the spectrometric neutron-gamma (SNGL) borehole logging. The SNGL enables the lithology identification through the quantitative analysis of the elements in geological formations and thus can be very useful for the oil and gas industry as well as for prospecting of the potential host rocks for radioactive waste disposal. In the SNGL experiment, gamma-rays induced by the neutron interactions with the nuclei of the rock elements are detected using the gamma-ray probe of complex mechanical and electronic construction. The probe has to be calibrated for a wide range of the elemental concentrations, to assure the proper quantitative analysis. The Polish Calibration Station in Zielona Gora is equipped with a limited number of calibration standards. An extension of the experimental calibration and the evaluation of the effect of the so-called side effects (for example the borehole and formation salinity variation) on the accuracy of the SNGL method can be done by the use of the MCNP code. The preliminary MCNP results showing the effect of the borehole and formation fluids salinity variations on the accuracy of silicon (Si), calcium (Ca) and iron (Fe) content determination are presented in the paper. The main effort has been focused on a modelling of the complex SNGL probe situated in a fluid filled borehole, surrounded by a geological formation. Track length estimate of the photon flux from the (n,gamma) interactions as a function of gamma-rays energy was used. Calculations were run on the PC computer with AMD Athlon 1.33 GHz processor. Neutron and photon cross-sections libraries were taken from the MCNP4c package and based mainly on the ENDF/B-6, ENDF/B-5 and MCPLIB02 data. The results of simulated experiment are in conformity with results of the real experiment performed with the use of the main lithology models (sandstones, limestones and dolomite). (authors)

The use of the MCNP code for the quantitative analysis of elements in geological formations

International Nuclear Information System (INIS)

Cywicka-Jakiel, T.; Woynicka, U.; Zorski, T.

2003-01-01

The Monte Carlo modelling calculations using the MCNP code have been performed, which support the spectrometric neutron-gamma (SNGL) borehole logging. The SNGL enables the lithology identification through the quantitative analysis of the elements in geological formations and thus can be very useful for the oil and gas industry as well as for prospecting of the potential host rocks for radioactive waste disposal. In the SNGL experiment, gamma-rays induced by the neutron interactions with the nuclei of the rock elements are detected using the gamma-ray probe of complex mechanical and electronic construction. The probe has to be calibrated for a wide range of the elemental concentrations, to assure the proper quantitative analysis. The Polish Calibration Station in Zielona Gora is equipped with a limited number of calibration standards. An extension of the experimental calibration and the evaluation of the effect of the so-called side effects (for example the borehole and formation salinity variation) on the accuracy of the SNGL method can be done by the use of the MCNP code. The preliminary MCNP results showing the effect of the borehole and formation fluids salinity variations on the accuracy of silicon (Si), calcium (Ca) and iron (Fe) content determination are presented in the paper. The main effort has been focused on a modelling of the complex SNGL probe situated in a fluid filled borehole, surrounded by a geological formation. Track length estimate of the photon flux from the (n,gamma) interactions as a function of gamma-rays energy was used. Calculations were run on the PC computer with AMD Athlon 1.33 GHz processor. Neutron and photon cross-sections libraries were taken from the MCNP4c package and based mainly on the ENDF/B-6, ENDF/B-5 and MCPLIB02 data. The results of simulated experiment are in conformity with results of the real experiment performed with the use of the main lithology models (sandstones, limestones and dolomite). (authors)

Utilization of new 150-MeV neutron and proton evaluations in MCNP

International Nuclear Information System (INIS)

Little, R.C.; Frankle, S.C.; Hughes, H.G. III; Prael, R.E.

1997-01-01

MCNP trademark and LAHET trademark are two of the codes included in the LARAMIE (Los Alamos Radiation Modeling Interactive Environment) code system. Both MCNP and LAHET are three-dimensional continuous-energy Monte Carlo radiation transport codes. The capabilities of MCNP and LAHET are currently being merged into one code for the Accelerator Production of Tritium (APT) program at Los Alamos National Laboratory. Concurrently, a significant effort is underway to improve the accuracy of the physics in the merged code. In particular, full nuclear-data evaluations (in ENDF6 format) for many materials of importance to APT are being produced for incident neutrons and protons up to an energy of 150-MeV. After processing, cross-section tables based on these new evaluations will be available for use fin the merged code. In order to utilize these new cross-section tables, significant enhancements are required for the merged code. Neutron cross-section tables for MCNP currently specify emission data for neutrons and photons only; the new evaluations also include complete neutron-induced data for protons, deuterons, tritons, and alphas. In addition, no provision in either MCNP or LAHET currently exists for the use of incident charged-particle tables other than for electrons. To accommodate the new neutron-induced data, it was first necessary to expand the format definition of an MCNP neutron cross-section table. The authors have prepared a 150-MeV neutron cross-section library in this expanded format for 15 nuclides. Modifications to MCNP have been implemented so that this expanded neutron library can be utilized

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

Science.gov (United States)

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.

Comparison between correlated sampling and the perturbation technique of MCNP5 for fixed-source problems

International Nuclear Information System (INIS)

He Tao; Su Bingjing

2011-01-01

Highlights: → The performance of the MCNP differential operator perturbation technique is compared with that of the MCNP correlated sampling method for three types of fixed-source problems. → In terms of precision, the MCNP perturbation technique outperforms correlated sampling for one type of problem but performs comparably with or even under-performs correlated sampling for the other two types of problems. → In terms of accuracy, the MCNP perturbation calculations may predict inaccurate results for some of the test problems. However, the accuracy can be improved if the midpoint correction technique is used. - Abstract: Correlated sampling and the differential operator perturbation technique are two methods that enable MCNP (Monte Carlo N-Particle) to simulate small response change between an original system and a perturbed system. In this work the performance of the MCNP differential operator perturbation technique is compared with that of the MCNP correlated sampling method for three types of fixed-source problems. In terms of precision of predicted response changes, the MCNP perturbation technique outperforms correlated sampling for the problem involving variation of nuclide concentrations in the same direction but performs comparably with or even underperforms correlated sampling for the other two types of problems that involve void or variation of nuclide concentrations in opposite directions. In terms of accuracy, the MCNP differential operator perturbation calculations may predict inaccurate results that deviate from the benchmarks well beyond their uncertainty ranges for some of the test problems. However, the accuracy of the MCNP differential operator perturbation can be improved if the midpoint correction technique is used.

Monte Carlo importance sampling for the MCNP trademark general source

International Nuclear Information System (INIS)

Lichtenstein, H.

1996-01-01

Research was performed to develop an importance sampling procedure for a radiation source. The procedure was developed for the MCNP radiation transport code, but the approach itself is general and can be adapted to other Monte Carlo codes. The procedure, as adapted to MCNP, relies entirely on existing MCNP capabilities. It has been tested for very complex descriptions of a general source, in the context of the design of spent-reactor-fuel storage casks. Dramatic improvements in calculation efficiency have been observed in some test cases. In addition, the procedure has been found to provide an acceleration to acceptable convergence, as well as the benefit of quickly identifying user specified variance-reduction in the transport that effects unstable convergence

A temperature dependent ENDF/B-VI.8 ACE library for UO2, ThO2, Zirc4, SS AISI-348, H2O, B4C and Ag-In-Cd alloy

International Nuclear Information System (INIS)

Dalle, Hugo M.; Bianchini, Mario; Gomes, Paulo Cezar

2009-01-01

Most MCNP standard neutron ACE libraries are processed at room temperature, 293,6 deg K. The temperature enters into the processing of the evaluation of a data file through the Doppler broadening of cross-sections. The nuclear fuel burnup usually takes place at reactor core temperatures much higher than room temperature, consequently, Monte Carlo burnup calculations should not only use the best cross-sections evaluations available but also use evaluations that are at temperatures approximating the temperatures of the application. In order to face the scarcity of temperature dependent MCNP cross-sections data to most isotopes, CDTN is developing an in-house temperature dependent neutron library for those nuclides commonly necessary in the systems simulated for the ongoing projects demanding Monte Carlo burnup. This paper describes the data processing of the ENDF/B-VI, release 8, using the NJOY99 code, towards provides this temperature dependent ACE library. Up to now fifty one elements and isotopes of the materials uranium oxide, thorium oxide, zircaloy-4, stainless steel AISI-348, light water, boron carbide and the silver-indium-cadmium alloy were processed at temperatures range from 293,6 deg K to 1200 deg K. Some benchmarks for thorium cycle described in the 'OECD/NEA International Handbook of Evaluated Criticality Safety Benchmark Experiments' were simulated using MCNP5 and the data set of this in-house library and the results usually agree with those obtained for the .60c standard MCNP neutron library for room temperature. (author)

Multi-canister overpack project - verification and validation, MCNP 4A

International Nuclear Information System (INIS)

Goldmann, L.H.

1997-01-01

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 the 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

Development of an interface between MCNP and ORIGEN codes for calculations of fuel evolution in nuclear systems. Initial project

International Nuclear Information System (INIS)

Campolina, Daniel de Almeida Magalhaes

2009-01-01

In Many situations of nuclear system study, it is necessary to know the detailed particle flux in a geometry. Deterministic 1-D and 2-D methods aren't suitable to represent some strong 3-D behavior configurations, for example in cores where the neutron flux varies considerably in the space and Monte Carlo analysis are necessary. The majority of Monte Carlo transport calculation codes, performs time static simulations, in terms of fuel isotopic composition. This work is a initial project to incorporate depletion capability to the MCNP code, by means of a connection with ORIGEN2.1 burnup code. The method to develop the program proposed followed the methodology of other programs used to the same purpose. Essentially, MCNP data library are used to generate one group microscopic cross sections that override default ORIGEN libraries. To verify the actual implemented part, comparisons which MCNPX (version 2.6.0) results were made. The neutron flux and criticality value of core agree. The neutron flux and criticality value of the core agree, especially in beginning of burnup when the influence of fission products are not very considerable. The small difference encountered was probably caused by the difference in the number of isotopes considered in the transport models (89 MCNPX x 25 GB). Next step of this work is to adapt MCNP version 4C to work with a memory higher than its standard value (4MB), in order to allow a greater number of isotopes in the transport model. (author)

MCNP calculation for calibration curve of X-ray fluorescence analysis

International Nuclear Information System (INIS)

Tan Chunming; Wu Zhifang; Guo Xiaojing; Xing Guilai; Wang Zhentao

2011-01-01

Due to the compositional variation of the sample, linear relationship between the element concentration and fluorescent intensity will not be well maintained in most X-ray fluorescence analysis. To overcome this, we use MCNP program to simulate fluorescent intensity of Fe (0∼100% concentration range) within binary mixture of Cr and O which represent typical strong absorption and weak absorption conditions respectively. The theoretic calculation shows that the relationship can be described as a curve determined by parameter p and value of p can be obtained with given absorption coefficient of substrate elements and element under detection. MCNP simulation results are consistent with theoretic calculation. Our research reveals that MCNP program can calculate the Calibration Curve of X-ray fluorescence very well. (authors)

Application of the NJOY code for unresolved resonance treatment in the MCNP utility code

International Nuclear Information System (INIS)

Milosevic, M.; Greenspan, E.; Vujic, J. . E-mail addresses of corresponding authors: mmilos@vin.bg.ac.yu , vujic@nuc.berkeley.edu ,; Milosevic, M.; Vujic, J.)

2005-01-01

There are numerous uncertainties in the prediction of neutronic characteristics of reactor cores, particularly in the case of innovative reactor designs, arising from approximations used in the solution of the transport equation, and in nuclear data processing and cross section libraries generation. This paper describes the problems encountered in the analysis of the Encapsulated Nuclear Heat Source (ENHS) benchmark core and the new procedures and cross section libraries developed to overcome these problems. The ENHS is a new lead-bismuth or lead cooled novel reactor concept that is fuelled with metallic alloy of Pu, U and Zr, and it is designed to operate for 20 effective full power years without refuelling and with very small burnup reactivity swing. The computational tools benchmarked include: MOCUP - a coupled MCNP-4C and ORIGEN2.1 utility codes with MCNP data libraries based on the ENDF/B-VI evaluations; and KWO2 - a coupled KENO-V.a and ORIGEN2.1 code with ENDFB-V.2 based 238 group library. Calculations made for the ENHS benchmark have shown that the differences between the results obtained using different code systems and cross section libraries are significant and should be taken into account in assessing the quality of nuclear data libraries. (author)

MCNP evaluation of top node control rod depletion below the core in KKL

International Nuclear Information System (INIS)

Beran, Tâm; Seltborg, Per; Lindahl, Sten-Örjan; Bieli, Roger; Ledergerber, Guido

2014-01-01

In previous studies, there has been identified a significant discrepancy in the BWR control rod top node depletion between the two core simulator nodal codes POLCA7 and PRESTO-2, which indicates that there is a large general uncertainty in nodal codes in calculating the top node depletion of fully withdrawn control rods. In this study, the stochastic Monte Carlo code MCNP has been used to calculate the top node control rod depletion for benchmarking the nodal codes. By using the TIP signal obtained from an extended TIP campaign below the core performed in the KKL reactor, the MCNP model has been verified by comparing the axial profile between the TIP data and the gamma flux calculated by MCNP. The MCNP results have also been compared with calculations from POLCA7, which was found to yield slightly higher depletion rates than MCNP. It was also found that the 10 B depletion in the top node is very sensitive to the exact axial location of the control rod top when it is fully withdrawn. By using the MCNP results, the neutron flux model below the core in the nodal codes can be improved by implementing an exponential function for the neutron flux. (author)

4C-ker: A Method to Reproducibly Identify Genome-Wide Interactions Captured by 4C-Seq Experiments.

Science.gov (United States)

Raviram, Ramya; Rocha, Pedro P; Müller, Christian L; Miraldi, Emily R; Badri, Sana; Fu, Yi; Swanzey, Emily; Proudhon, Charlotte; Snetkova, Valentina; Bonneau, Richard; Skok, Jane A

2016-03-01

4C-Seq has proven to be a powerful technique to identify genome-wide interactions with a single locus of interest (or "bait") that can be important for gene regulation. However, analysis of 4C-Seq data is complicated by the many biases inherent to the technique. An important consideration when dealing with 4C-Seq data is the differences in resolution of signal across the genome that result from differences in 3D distance separation from the bait. This leads to the highest signal in the region immediately surrounding the bait and increasingly lower signals in far-cis and trans. Another important aspect of 4C-Seq experiments is the resolution, which is greatly influenced by the choice of restriction enzyme and the frequency at which it can cut the genome. Thus, it is important that a 4C-Seq analysis method is flexible enough to analyze data generated using different enzymes and to identify interactions across the entire genome. Current methods for 4C-Seq analysis only identify interactions in regions near the bait or in regions located in far-cis and trans, but no method comprehensively analyzes 4C signals of different length scales. In addition, some methods also fail in experiments where chromatin fragments are generated using frequent cutter restriction enzymes. Here, we describe 4C-ker, a Hidden-Markov Model based pipeline that identifies regions throughout the genome that interact with the 4C bait locus. In addition, we incorporate methods for the identification of differential interactions in multiple 4C-seq datasets collected from different genotypes or experimental conditions. Adaptive window sizes are used to correct for differences in signal coverage in near-bait regions, far-cis and trans chromosomes. Using several datasets, we demonstrate that 4C-ker outperforms all existing 4C-Seq pipelines in its ability to reproducibly identify interaction domains at all genomic ranges with different resolution enzymes.

4C-ker: A Method to Reproducibly Identify Genome-Wide Interactions Captured by 4C-Seq Experiments.

Directory of Open Access Journals (Sweden)

Ramya Raviram

2016-03-01

Full Text Available 4C-Seq has proven to be a powerful technique to identify genome-wide interactions with a single locus of interest (or "bait" that can be important for gene regulation. However, analysis of 4C-Seq data is complicated by the many biases inherent to the technique. An important consideration when dealing with 4C-Seq data is the differences in resolution of signal across the genome that result from differences in 3D distance separation from the bait. This leads to the highest signal in the region immediately surrounding the bait and increasingly lower signals in far-cis and trans. Another important aspect of 4C-Seq experiments is the resolution, which is greatly influenced by the choice of restriction enzyme and the frequency at which it can cut the genome. Thus, it is important that a 4C-Seq analysis method is flexible enough to analyze data generated using different enzymes and to identify interactions across the entire genome. Current methods for 4C-Seq analysis only identify interactions in regions near the bait or in regions located in far-cis and trans, but no method comprehensively analyzes 4C signals of different length scales. In addition, some methods also fail in experiments where chromatin fragments are generated using frequent cutter restriction enzymes. Here, we describe 4C-ker, a Hidden-Markov Model based pipeline that identifies regions throughout the genome that interact with the 4C bait locus. In addition, we incorporate methods for the identification of differential interactions in multiple 4C-seq datasets collected from different genotypes or experimental conditions. Adaptive window sizes are used to correct for differences in signal coverage in near-bait regions, far-cis and trans chromosomes. Using several datasets, we demonstrate that 4C-ker outperforms all existing 4C-Seq pipelines in its ability to reproducibly identify interaction domains at all genomic ranges with different resolution enzymes.

MCNP Code in Assessment of Variations of Effective Dose with Torso Adipose Tissue Thickness

International Nuclear Information System (INIS)

Massoud, E.

2005-01-01

The effective dose is the unite used in the field of radiation protection. It is a well defined doubly weighted uantity involving both physical and biological variables. Several factors may induce variation in the effective dose in different individuals of similar exposure data. One of these factors is the variation of adipose tissue thickness in different exposed individuals. This study essentially concenrs the assessment of the possible variation in the effective dose due to variation in the thickness of adipose tissue. The study was done using MCNP4b code to perform mathematical model of the human body depending on that given to the reference man developed by International Commission of Radiological Protection (ICRP), and calculate the effective dose with different thicknessess of adipose tissues. The study includes a comprehensive appraisal of the Monte Cario simulation, the Medical Internal Radiation Dose (MIRD) model for the human body, and the various mathematical considerations involved in the radiation dose calculations for the various pertinent parts of the human body. The radiation energies considered were 80 KeV, 300 KeV and I MeV, applying two exposure positions; anteroposterior (AP), postero-anterior (PA) with different adipose tissue thickness. This study is a theoretical approach based on detailed mathematical calculations of great precision that deals with all considerations involved in the mechanisms of radiation energy absorption in biological system depending on the variation in the densities of the particular in biological system depending on the variation in the densities of the particular tissues. The results obtained indicate that maximum decrease in effective dose occures with the lowest energy at 5cm adipose tissues thickeness for both AP and PA exposure positions. The results obtained were compared to similar work previsouly done using MCNP4 b showing very good agreement

Irreversible thermochromic response of RF sputtered nanocrystalline BaWO4 films for smart window applications

Directory of Open Access Journals (Sweden)

C. Anil Kumar

2015-10-01

Full Text Available We report irreversible thermochromic behaviour of BaWO4 (BWO films for the first time. BWO films have been deposited at different substrate temperatures (RT, 200, 400, 600 and 800 °C using RF magnetron sputtering in pure argon plasma. BWO films deposited at 800 °C exhibit crystalline nature. Also, BWO films deposited in the temperature range of 400 - 600 °C exhibit WO3 as a secondary phase and its weight percentage decreases with an increase in deposition temperature, whereas the films deposited at 800 °C exhibited pure tetragonal phase. FESEM images revealed that as the average particle sizes of the films are higher as compared with the thickness of the films and is explained based on Avrami type nucleation and growth. The transmittance of the films decreases with an increase in deposition temperature up to 600 °C and increases thereafter. Films deposited at 600 °C show ≤ 20% transmittance, looking at the films deposited at room temperature and 800 °C exhibits 90 and 70%, respectively. The refractive index and extinction coefficient of the films show profound dependence on crystallinity and packing density. The optical bandgap of BWO films increases significantly with an increase in O2% during the deposition. The optical bandgap of the BWO films deposited at different temperatures in pure argon plasma, are in the range of 3.7 to 3.94 eV whereas the films deposited at 600 °C under different O2 plasma are in the range of 3.6 - 4.5 eV. The formations of colour centres are associated with the oxygen vacancies, which are clearly seen from the optical bandgap studies. The observed irreversible thermochromic behaviour in BWO films is attributed to the presence of oxygen vacancies that arises due to the electrons trapped at oxygen vacancies causing an inter valence charge transfer of W5+ to W6+ and is confirmed through the change in the optical density (ΔOD. Further, the Raman spectra are being used to quantify the presence

RELAP5-3D Developmental Assessment. Comparison of Version 4.3.4i on Linux and Windows

International Nuclear Information System (INIS)

Bayless, Paul David

2015-01-01

Figures have been generated comparing the parameters used in the developmental assessment of the RELAP5-3D code, version 4.3i, compiled on Linux and Windows platforms. The figures, which are the same as those used in Volume III of the RELAP5-3D code manual, compare calculations using the semi-implicit solution scheme with available experiment data. These figures provide a quick, visual indication of how the code predictions differ between the Linux and Windows versions.

An experimental test on large animals of MCNP application for whole body counting

International Nuclear Information System (INIS)

Borisov, N.; Yatsenko, V.; Kochetkov, O.; Gusev, I.; Vlasov, P.; Kalistratova, V.; Nisimov, P.; Levochkin, F.; Borovkov, M.; Stolyarov, V.; Tsedish, S.; Tyurin, I.; Franck, D.; Carlan, L. de

2005-01-01

Measurements of actinide body burden using whole body counting spectrometry is hampered due to intensive absorption of γ-rays inside the patient's body, which depends on the anatomy of a patient. To establish the correspondence between pulse-height-spectra intensity and radionuclide activity, Monte Carlo calculations are widely used. For such calculations, the radiation transport geometry is usually described in terms of small rectangular boxes (voxels) retrieved from computed tomography or magnetic resonance images. The software for Monte Carlo-assisted calibration of whole body counting, which performs automatic creation of individual MCNP voxel phantoms, was checked in a quasi-in vivo experiment on large animals. During the experiment, pigs of 35-40 kg body mass were used as phantoms for measurement of actinides body burden. 241 Am was administered (via injection of a radioactive solution or via implantation of plastic capsules containing the radioactive material) into the lungs of pigs. The pigs were measured using the pure germanium low-energy γ-spectrometers. The images of animals were obtained using the computed tomography machine. On the base of these tomograms, MCNP4c2 calculations were done to obtain the pulse-height-spectra of the whole body counters. The experimental results were reproduced in calculations with error of less than 30% for 241 Am administered via injection and less than 10% for 241 Am administered inside the capsules. (authors)

Determination of the detection efficiency of a HPGe detector by means of the MCNP 4A simulation code; Determinacion de la eficiencia de deteccion de un detector HPGe mediante el codigo de simulacion MCNP 4A

Energy Technology Data Exchange (ETDEWEB)

Leal, B. [Centro Regional de Estudios Nucleares, A.P. 579C, 98068 Zacatecas (Mexico)

2004-07-01

In the majority of the laboratories, the calibration in efficiency of the detector is carried out by means of the standard sources measurement of gamma photons that have a determined activity, or for matrices that contain a variety of radionuclides that can embrace the energy range of interest. Given the experimental importance that has the determination from the curves of efficiency to the effects of establishing the quantitative results, is appealed to the simulation of the response function of the detector used in the Regional Center of Nuclear Studies inside the energy range of 80 keV to 1400 keV varying the density of the matrix, by means of the application of the Monte Carlo code MCNP-4A. The adjustment obtained shows an acceptance grade in the range of 100 to 600 keV, with a smaller percentage discrepancy to 5%. (Author)

Window observers for linear systems

Directory of Open Access Journals (Sweden)

Utkin Vadim

2000-01-01

Full Text Available Given a linear system x ˙ = A x + B u with output y = C x and a window function ω ( t , i.e., ∀ t , ω ( t ∈ {0,1 }, and assuming that the window function is Lebesgue measurable, we refer to the following observer, x ˆ = A x + B u + ω ( t L C ( x − x ˆ as a window observer. The stability issue is treated in this paper. It is proven that for linear time-invariant systems, the window observer can be stabilized by an appropriate design under a very mild condition on the window functions, albeit for linear time-varying system, some regularity of the window functions is required to achieve observer designs with the asymptotic stability. The corresponding design methods are developed. An example is included to illustrate the possible applications

NMTC/JAERI-MCNP4Aコードシステムによる熱・冷中性子輸送計算

OpenAIRE

伊賀 公紀; 高田 弘; 永尾 忠司

1998-01-01

水冷タンタルターゲット・減速材・反射体体系について、NMTC/JAERI-MCNP4Aコードシステムを用いた熱・冷中性子輸送計算を行い、核破砕中性子源の核設計への本コードシステムの適用性を調べた。計算では、減速材に軽水及び液体水素を用いた場合に放出される中性子のエネルギースペクトルについて、B4Cデカップラーの有無による強度変化、外部中性子線源位置による強度変化等の計算結果が実験結果に基づく半実験式に物理的に妥当な値のパラメータを用いて再現できることを確認した。しかし、放出中性子の時間スペクトルを精度良く評価するためには、MCNP4Aによる計算で反射体やデカップラー領域に適当なインポータンスを設定する工夫が必要であることがわかった。本計算によって、今後の核破砕値源の核設計にNMTC/JAERI-MCNP4Aコードシステムを適用できることを確認した。...

Gamma knife simulation using the MCNP4C code and the zubal phantom and comparison with experimental data

International Nuclear Information System (INIS)

Gholami, S.; Kamali Asl, A.; Aghamiri, M.; Allahverdi, M.

2010-01-01

Gamma Knife is an instrument specially designed for treating brain disorders. In Gamma Knife, there are 201 narrow beams of cobalt-60 sources that intersect at an isocenter point to treat brain tumors. The tumor is placed at the isocenter and is treated by the emitted gamma rays. Therefore, there is a high dose at this point and a low dose is delivered to the normal tissue surrounding the tumor. Material and Method: In the current work, the MCNP simulation code was used to simulate the Gamma Knife. The calculated values were compared to the experimental ones and previous works. Dose distribution was compared for different collimators in a water phantom and the Zubal brain-equivalent phantom. The dose profiles were obtained along the x, y and z axes. Result: The evaluation of the developed code was performed using experimental data and we found a good agreement between our simulation and experimental data. Discussion: Our results showed that the skull bone has a high contribution to both scatter and absorbed dose. In other words, inserting the exact material of brain and other organs of the head in digital phantom improves the quality of treatment planning. This work is regarding the measurement of absorbed dose and improving the treatment planning procedure in Gamma-Knife radiosurgery in the brain.

Gamma Knife Simulation Using the MCNP4C Code and the Zubal Phantom and Comparison with Experimental Data

Directory of Open Access Journals (Sweden)

Somayeh Gholami

2010-06-01

Full Text Available Introduction: Gamma Knife is an instrument specially designed for treating brain disorders. In Gamma Knife, there are 201 narrow beams of cobalt-60 sources that intersect at an isocenter point to treat brain tumors. The tumor is placed at the isocenter and is treated by the emitted gamma rays. Therefore, there is a high dose at this point and a low dose is delivered to the normal tissue surrounding the tumor. Material and Method: In the current work, the MCNP simulation code was used to simulate the Gamma Knife. The calculated values were compared to the experimental ones and previous works. Dose distribution was compared for different collimators in a water phantom and the Zubal brain-equivalent phantom. The dose profiles were obtained along the x, y and z axes. Result: The evaluation of the developed code was performed using experimental data and we found a good agreement between our simulation and experimental data. Discussion: Our results showed that the skull bone has a high contribution to both scatter and absorbed dose. In other words, inserting the exact material of brain and other organs of the head in digital phantom improves the quality of treatment planning. This work is regarding the measurement of absorbed dose and improving the treatment planning procedure in Gamma-Knife radiosurgery in the brain.

Shielding property of bismuth glass based on MCNP 5 and WINXCOM simulated calculation

International Nuclear Information System (INIS)

Zhang Zhicheng; Zhang Jinzhao; Liu Ze; Lu Chunhai; Chen Min

2013-01-01

Background: Currently, lead glass is widely used as observation window, while lead is toxic heavy metal. Purpose: Non-toxic materials and their shielding effects are researched in order to find a new material to replace lead containing material. Methods: The mass attenuation coefficients of bismuth silicate glass were investigated with gamma-ray's energy at 0.662 MeV, 1.17 MeV and 1.33 MeV, respectively, by MCNP 5 (Monte Carlo) and WINXCOM program, and compared with those of the lead glass. Results: With attenuation factor K, shielding and mechanical properties taken into consideration bismuth glass containing 50% bismuth oxide might be selected as the right material. Dose rate distributions of water phantom were calculated with 2-cm and 10-cm thick glass, respectively, irradiated by 137 Cs and 60 Co in turn. Conclusion: Results show that the bismuth glass may replace lead glass for radiation shielding with appropriate energy. (authors)

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

Energy Technology Data Exchange (ETDEWEB)

Mashnik, Stepan G [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2014-09-08

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 MCNP6 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.

Generation of Anaphylatoxins by Human β-Tryptase from C3, C4, and C51

Science.gov (United States)

Fukuoka, Yoshihiro; Xia, Han-Zhang; Sanchez-Muñoz, Laura B.; Dellinger, Anthony L.; Escribano, Luis; Schwartz, Lawrence B.

2009-01-01

Both mast cells and complement participate in innate and acquired immunity. The current study examines whether β-tryptase, the major protease of human mast cells, can directly generate bioactive complement anaphylatoxins. Important variables included pH, monomeric vs tetrameric forms of β-tryptase, and the β-tryptase-activating polyanion. The B12 mAb was used to stabilize β-tryptase in its monomeric form. C3a and C4a were best generated from C3 and C4, respectively, by monomeric β-tryptase in the presence of low molecular weight dextran sulfate or heparin at acidic pH. High molecular weight polyanions increased degradation of these anaphylatoxins. C5a was optimally generated from C5 at acidic pH by β-tryptase monomers in the presence of high molecular weight dextran sulfate and heparin polyanions, but also was produced by β-tryptase tetramers under these conditions. Mass spectrometry verified that the molecular mass of each anaphylatoxin was correct. Both β-tryptase-generated C5a and C3a (but not C4a) were potent activators of human skin mast cells. These complement anaphylatoxins also could be generated by β-tryptase in releasates of activated skin mast cells. Of further biologic interest, β-tryptase also generated C3a from C3 in human plasma at acidic pH. These results suggest β-tryptase might generate complement anaphylatoxins in vivo at sites of inflammation, such as the airway of active asthma patients where the pH is acidic and where elevated levels of β-tryptase and complement anaphylatoxins are detected. PMID:18424754

Production of neutronic discrete equations for a cylindrical geometry in one group energy and benchmark the results with MCNP-4B code with one group energy library

International Nuclear Information System (INIS)

Salehi, A. A.; Vosoughi, N.; Shahriari, M.

2002-01-01

In reactor core neutronic calculations, we usually choose a control volume and investigate about the input, output, production and absorption inside it. Finally, we derive neutron transport equation. This equation is not easy to solve for simple and symmetrical geometry. The objective of this paper is to introduce a new direct method for neutronic calculations. This method is based on physics 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 equation series without production of neutron transport differential equation and mandatory passing form differential equation bridge. This method, which is named Direct Discrete Method, was applied in static state, for a cylindrical geometry in one group energy. The validity of the results from this new method are tested with MCNP-4B code with a one group energy library. One energy group direct discrete equation produces excellent results, which can be compared with the results of MCNP-4B

Developing an interface between MCNP and McStas for simulation of neutron moderators

DEFF Research Database (Denmark)

Klinkby, Esben Bryndt; Lauritzen, Bent; Nonbøl, Erik

2012-01-01

Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using MCNP/X whereas simulations of neutron transport and instrument performance are carried out by neutron ray tracing codes such as McStas. The coupling between the two simulations suites...... typically consists of providing analytical fits from MCNP/X neutron spectra to McStas. This method is generally successful, but as will be discussed in the this paper, there are limitations and a more direct coupling between MCNP/X andMcStas could allow for more accurate simulations of e.g. complex...... moderator geometries, interference between beamlines as well as shielding requirements along the neutron guides. In this paper different possible interfaces between McStas and MCNP/X are discussed and first preliminary performance results are shown....

MCNP6 Simulation of Light and Medium Nuclei Fragmentation at Intermediate Energies

Energy Technology Data Exchange (ETDEWEB)

Mashnik, Stepan Georgievich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kerby, Leslie Marie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Univ. of Idaho, Moscow, ID (United States)

2015-08-24

Fragmentation reactions induced on light and medium nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below are studied with the Los Alamos transport code MCNP6 and with its CEM03.03 and LAQGSM03.03 event generators. CEM and LAQGSM assume that intermediate-energy fragmentation reactions on light nuclei occur generally in two stages. The first stage is the intranuclear cascade (INC), followed by the second, Fermi breakup disintegration of light excited residual nuclei produced after the INC. CEM and LAQGSM account also for coalescence of light fragments (complex particles) up to sup>4He from energetic nucleons emitted during INC. We investigate the validity and performance of MCNP6, CEM, and LAQGSM in simulating fragmentation reactions at intermediate energies and discuss possible ways of further improving these codes.

Program for the Generation of MCNP Inputs from State Files of CAREM

International Nuclear Information System (INIS)

Leszczynski, Francisco; Lopasso, Edmundo; Villarino, E

2000-01-01

The objective of this work is the development and tests of detailed input data for the Monte Carlo program MCNP, to be able of model the core of CAREM reactor, with the detail included on the updated models, for having available a calculation system that allow the production of confident results to be compared with results obtained with the system used today for designing the CAREM reactor core (CONDOR-CITVAP).The model includes the possibility of temperature and coolant density, and temperature and numeric densities of fuel.The detail consists of 21 different fuel elements (symmetry 3) and 14 axial zones.Results of comparisons of reactivity and power pick factors are presented, between MCNP and CONDOR-CITVAP.On average, these results show an acceptable agreement for all the compared parameters.It is described, also, the interface CONDOR-CITVAP-MCNP program, that has been developed for generating inputs of materials for MCNP, from outputs of CONDOR and CITVAP, for different reactor states

Relationship between HCO_3"- concentration to weight of C_6H_6 of environmental isotop "1"4C analysis and its relationship with sampling in the field

International Nuclear Information System (INIS)

Satrio; Rasi Prasetio

2016-01-01

It has been done the groundwater sampling process of deep aquifer in Jakarta and surrounding areas for the analysis of environmental isotope "1"4C. Groundwater sampling was preceded by calculating the concentration of HCO_3"- (bicarbonate ion) through titration in the field. The number of repetitions of sampling is determined by the concentration data of HCO_3"- which obtained. The Repetition of this sampling will determine the acquisition of a solution of C_6H_6 (benzene) during the synthesis process benzene. In the field, the sampling is done by extracting of 60 liters of water to precipitate BaCO_3. The sampling process is repeated based on data from the bicarbonate ion concentration. The purpose of this study to determine the relationship between the concentration of HCO_3"- to the weights C_6H_6 which obtained in the analysis of environmental isotope "1"4C and evaluate the number of repetitions of the sampling that should be done. Based on the analysis of titration in the field, shows that concentration HCO_3"- ranged between 180 - 600 ppm with the acquisition of benzene between 1.84 to 4.5 grams. There is a strong relationship between the concentration of HCO_3"- and C_6H_6 weights obtained in the process of synthesis of benzene with a correlation of about 0.900. This correlation can be improved by measuring the concentration of HCO_3"- in advance in the laboratory tend to be more accurate than in the field. (author)

Radiation calculations using LAHET/MCNP/CINDER90

International Nuclear Information System (INIS)

Waters, L.

1994-01-01

The LAHET monte carlo code system has recently been expanded to include high energy hadronic interactions via the FLUKA code, while retaining the original Los Alamos versions of HETC and ISABEL at lower energies. Electrons and photons are transported with EGS4 or ITS, while the MCNP coupled neutron/photon monte carlo code provides analysis of neutrons with kinetic energies less than 20 MeV. An interface with the CINDER activation code is now in common use. Various other changes have been made to facilitate analysis of high energy accelerator radiation environments and experimental physics apparatus, such as those found at SSC and RHIC. Current code developments and applications are reviewed

Utilization of MCNP code in the research and design for China advanced research reactor

International Nuclear Information System (INIS)

Shen Feng

2006-01-01

MCNP, which is the internationalized neutronics code, is used for nuclear research and design in China Advanced Research Reactor (CARR). MCNP is an important neutronics code in the research and design for CARR since many calculation tasks could be undertaken by it. Many nuclear parameters on reactor core, the design and optimization research for many reactor utilizations, much verification for other nuclear calculation code and so on are conducted with help of MCNP. (author)

Improving computational efficiency of Monte Carlo simulations with variance reduction

International Nuclear Information System (INIS)

Turner, A.; Davis, A.

2013-01-01

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 to 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)

Using Machine Learning to Predict MCNP Bias

Energy Technology Data Exchange (ETDEWEB)

Grechanuk, Pavel Aleksandrovi [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2018-01-09

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, 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.

Comparison of MCNP5 and experimental results on neutron shielding effects for materials

Energy Technology Data Exchange (ETDEWEB)

Torres, D. A. (Daniel A.); Mosteller, R. D. (Russell D.); Sweezy, J. E. (Jeremy E.)

2004-01-01

The MCNP Radiation-Shielding Validation Suite was created to assess the impact on dose rates and attenuation factors of future improvements in the MCNP Monte Carlo code or its nuclear data libraries. However, it does not currently contain any deep-penetration cases. For this reason, a set of deep-penetration benchmarks has been investigated for possible inclusion in the Suite. Overall, the MCNP5 results match the measured values quite well. Furthermore, with the exception of Resin-F, there is no systematic trend in the ratio of calculated to measured results.

Current status of ACE format libraries for MCNP at nuclear date center of KAERI

Energy Technology Data Exchange (ETDEWEB)

Kim, Do Heon; Gil, Choong Sup; Lee, Young Ouk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-09-15

The current status of ACE format MCNP/MCNPX libraries by NDC of KAERI is presented with a short description of each library. Validation calculations with recent nuclear data evaluations ENDF/B-VII.0, ENDF/B-VII.1, JEFF-3.2, and JENDL-4.0 have been carried out by the MCNP5 code for 119 criticality benchmark problems taken from the expanded criticality validation suite supplied by LANL. The overall performances of the ACE format KN-libraries have been analyzed in comparison with the results calculated with the ENDF/B-VII.0-based ENDF70 library of LANL. It was confirmed that the ENDF/B-VII.1-based KNE71 library showed better performances than the others by comparing the RMS errors and χ2 values for five benchmark categories as well as whole benchmark problems. ENDF/B-VII.1 and JEFF-3.2 have a tendency to yield more reliable MCNP calculation results within certain confidence intervals regarding the total uncertainties for the keff values. It is found that the adoption of the latest evaluated nuclear data might ensure better outcomes in various research and development areas.

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

Science.gov (United States)

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.

Irreversible thermochromic response of RF sputtered nanocrystalline BaWO{sub 4} films for smart window applications

Energy Technology Data Exchange (ETDEWEB)

Anil Kumar, C.; Santhosh Kumar, T.; Pamu, D., E-mail: pamu@iitg.ernet.in [Department of Physics, Indian Institute of Technology Guwahati, Guwahati -781039 (India)

2015-10-15

We report irreversible thermochromic behaviour of BaWO{sub 4} (BWO) films for the first time. BWO films have been deposited at different substrate temperatures (RT, 200, 400, 600 and 800 °C) using RF magnetron sputtering in pure argon plasma. BWO films deposited at 800 °C exhibit crystalline nature. Also, BWO films deposited in the temperature range of 400 - 600 °C exhibit WO{sub 3} as a secondary phase and its weight percentage decreases with an increase in deposition temperature, whereas the films deposited at 800 °C exhibited pure tetragonal phase. FESEM images revealed that as the average particle sizes of the films are higher as compared with the thickness of the films and is explained based on Avrami type nucleation and growth. The transmittance of the films decreases with an increase in deposition temperature up to 600 °C and increases thereafter. Films deposited at 600 °C show ≤ 20% transmittance, looking at the films deposited at room temperature and 800 °C exhibits 90 and 70%, respectively. The refractive index and extinction coefficient of the films show profound dependence on crystallinity and packing density. The optical bandgap of BWO films increases significantly with an increase in O{sub 2}% during the deposition. The optical bandgap of the BWO films deposited at different temperatures in pure argon plasma, are in the range of 3.7 to 3.94 eV whereas the films deposited at 600 °C under different O{sub 2} plasma are in the range of 3.6 - 4.5 eV. The formations of colour centres are associated with the oxygen vacancies, which are clearly seen from the optical bandgap studies. The observed irreversible thermochromic behaviour in BWO films is attributed to the presence of oxygen vacancies that arises due to the electrons trapped at oxygen vacancies causing an inter valence charge transfer of W{sup 5+} to W{sup 6+} and is confirmed through the change in the optical density (ΔOD). Further, the Raman spectra are being used to

Impact of patient weight on tumor visibility based on human-shaped phantom simulation study in PET imaging system

Science.gov (United States)

Musarudin, M.; Saripan, M. I.; Mashohor, S.; Saad, W. H. M.; Nordin, A. J.; Hashim, S.

2015-10-01

Energy window technique has been implemented in all positron emission tomography (PET) imaging protocol, with the aim to remove the unwanted low energy photons. Current practices in our institution however are performed by using default energy threshold level regardless of the weight of the patient. Phantom size, which represents the size of the patient's body, is the factor that determined the level of scatter fraction during PET imaging. Thus, the motivation of this study is to determine the optimum energy threshold level for different sizes of human-shaped phantom, to represent underweight, normal, overweight and obese patients. In this study, the scanner was modeled by using Monte Carlo code, version MCNP5. Five different sizes of elliptical-cylinder shaped of human-sized phantoms with diameter ranged from 15 to 30 cm were modeled. The tumor was modeled by a cylindrical line source filled with 1.02 MeV positron emitters at the center of the phantom. Various energy window widths, in the ranged of 10-50% were implemented to the data. In conclusion, the phantom mass volume did influence the scatter fraction within the volume. Bigger phantom caused more scattering events and thus led to coincidence counts lost. We evaluated the impact of phantom sizes on the sensitivity and visibility of the simulated models. Implementation of wider energy window improved the sensitivity of the system and retained the coincidence photons lost. Visibility of the tumor improved as an appropriate energy window implemented for the different sizes of phantom.

Organ dose conversion coefficients based on a voxel mouse model and MCNP code for external photon irradiation.

Science.gov (United States)

Zhang, Xiaomin; Xie, Xiangdong; Cheng, Jie; Ning, Jing; Yuan, Yong; Pan, Jie; Yang, Guoshan

2012-01-01

A set of conversion coefficients from kerma free-in-air to the organ absorbed dose for external photon beams from 10 keV to 10 MeV are presented based on a newly developed voxel mouse model, for the purpose of radiation effect evaluation. The voxel mouse model was developed from colour images of successive cryosections of a normal nude male mouse, in which 14 organs or tissues were segmented manually and filled with different colours, while each colour was tagged by a specific ID number for implementation of mouse model in Monte Carlo N-particle code (MCNP). Monte Carlo simulation with MCNP was carried out to obtain organ dose conversion coefficients for 22 external monoenergetic photon beams between 10 keV and 10 MeV under five different irradiation geometries conditions (left lateral, right lateral, dorsal-ventral, ventral-dorsal, and isotropic). Organ dose conversion coefficients were presented in tables and compared with the published data based on a rat model to investigate the effect of body size and weight on the organ dose. The calculated and comparison results show that the organ dose conversion coefficients varying the photon energy exhibits similar trend for most organs except for the bone and skin, and the organ dose is sensitive to body size and weight at a photon energy approximately <0.1 MeV.

Nonperturbative results for two-index conformal windows

Energy Technology Data Exchange (ETDEWEB)

Bergner, Georg [Albert Einstein Center for Fundamental Physics,Institute for Theoretical Physics, University of Bern,Sidlerstrasse 5, CH-3012 Bern (Switzerland); Ryttov, Thomas A.; Sannino, Francesco [CP-Origins and the Danish IAS, University of Southern Denmark,5230 Odense M (Denmark)

2015-12-10

Via large and small N{sub c} relations we derive nonperturbative results about the conformal window of two-index theories. Using Schwinger-Dyson methods as well as four-loops results we estimate subleading corrections and show that naive large number of colors extrapolations are unreliable when N{sub c} is less than about six. Nevertheless useful nonperturbative inequalities for the size of the conformal windows, for any number of colors, can be derived. By further observing that the adjoint conformal window is independent of the number of colors we argue, among other things, that: the large N{sub c} two-index conformal window is twice the conformal window of the adjoint representation (which can be determined at small N{sub c}) expressed in terms of Dirac fermions; lattice results for adjoint matter can be used to provide independent information on the conformal dynamics of two-index theories such as SU(N{sub c}) with two and four symmetric Dirac flavors.

Nonperturbative results for two-index conformal windows

International Nuclear Information System (INIS)

Bergner, Georg; Ryttov, Thomas A.; Sannino, Francesco

2015-01-01

Via large and small N c relations we derive nonperturbative results about the conformal window of two-index theories. Using Schwinger-Dyson methods as well as four-loops results we estimate subleading corrections and show that naive large number of colors extrapolations are unreliable when N c is less than about six. Nevertheless useful nonperturbative inequalities for the size of the conformal windows, for any number of colors, can be derived. By further observing that the adjoint conformal window is independent of the number of colors we argue, among other things, that: the large N c two-index conformal window is twice the conformal window of the adjoint representation (which can be determined at small N c ) expressed in terms of Dirac fermions; lattice results for adjoint matter can be used to provide independent information on the conformal dynamics of two-index theories such as SU(N c ) with two and four symmetric Dirac flavors.

Implementasi Filter Finite Impulse Response (FIR Window Hamming dan Blackman menggunakan DSK TMS320C6713

Directory of Open Access Journals (Sweden)

LITA LIDYAWATI

2016-02-01

Full Text Available ABSTRAK Filter didefinisikan sebagai proses atau rangkaian yang melewatkan pita frekuensi tertentu yang diinginkan dan meredam pita frekuensi lainnya. Salah satu metode perancangan filter digital Finite Impulse Response (FIR adalah metode windowing. Dalam penelitian ini digunakan jenis window Hamming dan Blackman. Simulasi dilakukan dengan menggunakan software Matlab dengan memasukan frekuensi passband, frekuensi stopband, ripple passband, dan stopband attenuation. Dengan frekuensi sampling sebesar 15000 Hz, frekuensi passband sebesar 3000 Hz, frekuensi stopband sebesar 5000 Hz. Setelah simulasi dilakukan implementasi filter dengan parameter yang sama menggunakan DSK TMS 320C6713 dengan bantuan software CCS. Simulasi dan implementasi dilakukan pada semua band frekuensi. Hasil pengujian terhadap implementasi filter adalah respon magnitude, frekuensi cut-off, bandwidth, dan faktor kualitas dengan hasil simulasi tidak menunjukkan perbedaan yang signifikan. Kata kunci: filter digital, windowing, Hamming, Blackman, frekuensi cut-off . ABSTRACT Filter is defined as a process or series that skip certain desired frequency band and other frequency bands drown. One method of designing a digital filter Finite Impulse Response (FIR is a windowing method. This study used the type of window Hamming and Blackman. Simulations performed using Matlab software by inserting a frequency passband, stopband frequency, passband ripple, and stopband attenuation. With a sampling frequency of 15,000 Hz, a frequency of 3000 Hz passband, stopband frequency of 5000 Hz. After the simulation is completed, implementation of the filter with the same parameters using TMS 320C6713 DSK with the help of software CCS. Simulation and implmentasi performed on all frequency bands. The test results of the implementation of the filter is the Magnitude response, the cut-off frequency, bandwidth, and quality factor with simulation results showed no significant difference. Keywords: digital

An Electron/Photon/Relaxation Data Library for MCNP6

Energy Technology Data Exchange (ETDEWEB)

Hughes, III, H. Grady [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-08-07

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.

Estimation and interpretation of keff confidence intervals in MCNP

International Nuclear Information System (INIS)

Urbatsch, T.J.

1995-11-01

MCNP's criticality methodology and some basic statistics are reviewed. Confidence intervals are discussed, as well as how to build them and their importance in the presentation of a Monte Carlo result. The combination of MCNP's three k eff estimators is shown, theoretically and empirically, by statistical studies and examples, to be the best k eff estimator. The method of combining estimators is based on a solid theoretical foundation, namely, the Gauss-Markov Theorem in regard to the least squares method. The confidence intervals of the combined estimator are also shown to have correct coverage rates for the examples considered

Evaluation of computational models and cross sections used by MCNP6 for simulation of electron backscattering

Energy Technology Data Exchange (ETDEWEB)

Poškus, Andrius, E-mail: andrius.poskus@ff.vu.lt

2016-02-01

This work evaluates the accuracy of the single-event (SE) and condensed-history (CH) models of electron transport in Monte Carlo simulations of electron backscattering from thick layers of Be, C, Al, Cu, Ag, Au and U at incident electron energies from 200 eV to 15 MeV. The CH method is used in simulations performed with MCNP6.1, and the SE method is used in simulations performed with an open-source single-event code MCNelectron written by the author of this paper. Both MCNP6.1 and MCNelectron use mainly ENDF/B-VI.8 library data, but MCNelectron allows replacing cross sections of certain types of interactions by alternative datasets from other sources. The SE method is evaluated both using only ENDF/B-VI.8 cross sections (the “SE-ENDF/B method”, which is equivalent to using MCNP6.1 in SE mode) and with an alternative set of elastic scattering cross sections obtained from relativistic (Dirac) partial-wave (DPW) calculations (the “SE-DPW method”). It is shown that at energies from 200 eV to 300 keV the estimates of the backscattering coefficients obtained using the SE-DPW method are typically within 10% of the experimental data, which is approximately the same accuracy that is achieved using MCNP6.1 in CH mode. At energies below 1 keV and above 300 keV, the SE-DPW method is much more accurate than the SE-ENDF/B method due to lack of angular distribution data in the ENDF/B library in those energy ranges. At energies from 500 keV to 15 MeV, the CH approximation is roughly twice more accurate than the SE-DPW method, with the average relative errors equal 7% and 14%, respectively. The energy probability density functions (PDFs) of backscattered electrons for Al and Cu, calculated using the SE method with DPW cross sections when energy of incident electrons is 20 keV, have an average absolute error as low as 4% of the average PDF. This error is approximately twice less than the error of the corresponding PDF calculated using the CH approximation. It is concluded

Use experiences of MCNP in nuclear energy study. 2. Review of variance reduction techniques

Energy Technology Data Exchange (ETDEWEB)

Sakurai, Kiyoshi; Yamamoto, Toshihiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; eds.

1998-03-01

`MCNP Use Experience` Working Group was established in 1996 under the Special Committee on Nuclear Code Evaluation. This year`s main activity of the working group has been focused on the review of variance reduction techniques of Monte Carlo calculations. This working group dealt with the variance reduction techniques of (1) neutron and gamma ray transport calculation of fusion reactor system, (2) concept design of nuclear transmutation system using accelerator, (3) JMTR core calculation, (4) calculation of prompt neutron decay constant, (5) neutron and gamma ray transport calculation for exposure evaluation, (6) neutron and gamma ray transport calculation of shielding system, etc. Furthermore, this working group started an activity to compile `Guideline of Monte Carlo Calculation` which will be a standard in the future. The appendices of this report include this `Guideline`, the use experience of MCNP 4B and examples of Monte Carlo calculations of high energy charged particles. The 11 papers are indexed individually. (J.P.N.)

Use experiences of MCNP in nuclear energy study. 2. Review of variance reduction techniques

International Nuclear Information System (INIS)

Sakurai, Kiyoshi; Yamamoto, Toshihiro

1998-03-01

''MCNP Use Experience'' Working Group was established in 1996 under the Special Committee on Nuclear Code Evaluation. This year''s main activity of the working group has been focused on the review of variance reduction techniques of Monte Carlo calculations. This working group dealt with the variance reduction techniques of (1) neutron and gamma ray transport calculation of fusion reactor system, (2) concept design of nuclear transmutation system using accelerator, (3) JMTR core calculation, (4) calculation of prompt neutron decay constant, (5) neutron and gamma ray transport calculation for exposure evaluation, (6) neutron and gamma ray transport calculation of shielding system, etc. Furthermore, this working group started an activity to compile ''Guideline of Monte Carlo Calculation'' which will be a standard in the future. The appendices of this report include this ''Guideline'', the use experience of MCNP 4B and examples of Monte Carlo calculations of high energy charged particles. The 11 papers are indexed individually. (J.P.N.)

Acceleration of calculation of nuclear heating distributions in ITER toroidal field coils using hybrid Monte Carlo/deterministic techniques

International Nuclear Information System (INIS)

Ibrahim, Ahmad M.; Polunovskiy, Eduard; Loughlin, Michael J.; Grove, Robert E.; Sawan, Mohamed E.

2016-01-01

Highlights: • Assess the detailed distribution of the nuclear heating among the components of the ITER toroidal field coils. • Utilize the FW-CADIS method to dramatically accelerate the calculation of detailed nuclear analysis. • Compare the efficiency and reliability of the FW-CADIS method and the MCNP weight window generator. - Abstract: Because the superconductivity of the ITER toroidal field coils (TFC) must be protected against local overheating, detailed spatial distribution of the TFC nuclear heating is needed to assess the acceptability of the designs of the blanket, vacuum vessel (VV), and VV thermal shield. Accurate Monte Carlo calculations of the distributions of the TFC nuclear heating are challenged by the small volumes of the tally segmentations and by the thick layers of shielding provided by the blanket and VV. To speed up the MCNP calculation of the nuclear heating distribution in different segments of the coil casing, ground insulation, and winding packs of the ITER TFC, the ITER Organization (IO) used the MCNP weight window generator (WWG). The maximum relative uncertainty of the tallies in this calculation was 82.7%. In this work, this MCNP calculation was repeated using variance reduction parameters generated by the Oak Ridge National Laboratory AutomateD VAriaNce reducTion Generator (ADVANTG) code and both MCNP calculations were compared in terms of computational efficiency and reliability. Even though the ADVANTG MCNP calculation used less than one-sixth of the computational resources of the IO calculation, the relative uncertainties of all the tallies in the ADVANTG MCNP calculation were less than 6.1%. The nuclear heating results of the two calculations were significantly different by factors between 1.5 and 2.3 in some of the segments of the furthest winding pack turn from the plasma neutron source. Even though the nuclear heating in this turn may not affect the ITER design because it is much smaller than the nuclear heating in the

Acceleration of calculation of nuclear heating distributions in ITER toroidal field coils using hybrid Monte Carlo/deterministic techniques

Energy Technology Data Exchange (ETDEWEB)

Ibrahim, Ahmad M., E-mail: ibrahimam@ornl.gov [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Polunovskiy, Eduard; Loughlin, Michael J. [ITER Organization, Route de Vinon Sur Verdon, 13067 St. Paul Lez Durance (France); Grove, Robert E. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Sawan, Mohamed E. [University of Wisconsin-Madison, 1500 Engineering Dr., Madison, WI 53706 (United States)

2016-11-01

Highlights: • Assess the detailed distribution of the nuclear heating among the components of the ITER toroidal field coils. • Utilize the FW-CADIS method to dramatically accelerate the calculation of detailed nuclear analysis. • Compare the efficiency and reliability of the FW-CADIS method and the MCNP weight window generator. - Abstract: Because the superconductivity of the ITER toroidal field coils (TFC) must be protected against local overheating, detailed spatial distribution of the TFC nuclear heating is needed to assess the acceptability of the designs of the blanket, vacuum vessel (VV), and VV thermal shield. Accurate Monte Carlo calculations of the distributions of the TFC nuclear heating are challenged by the small volumes of the tally segmentations and by the thick layers of shielding provided by the blanket and VV. To speed up the MCNP calculation of the nuclear heating distribution in different segments of the coil casing, ground insulation, and winding packs of the ITER TFC, the ITER Organization (IO) used the MCNP weight window generator (WWG). The maximum relative uncertainty of the tallies in this calculation was 82.7%. In this work, this MCNP calculation was repeated using variance reduction parameters generated by the Oak Ridge National Laboratory AutomateD VAriaNce reducTion Generator (ADVANTG) code and both MCNP calculations were compared in terms of computational efficiency and reliability. Even though the ADVANTG MCNP calculation used less than one-sixth of the computational resources of the IO calculation, the relative uncertainties of all the tallies in the ADVANTG MCNP calculation were less than 6.1%. The nuclear heating results of the two calculations were significantly different by factors between 1.5 and 2.3 in some of the segments of the furthest winding pack turn from the plasma neutron source. Even though the nuclear heating in this turn may not affect the ITER design because it is much smaller than the nuclear heating in the

Understanding Windows | Efficient Windows Collaborative

Science.gov (United States)

Foundry Foundry New Construction Windows Window Selection Tool Selection Process Design Guidance Installation Replacement Windows Window Selection Tool Assessing Options Selection Process Design Guidance Installation Understanding Windows Benefits Design Considerations Measuring Performance Performance Standards

Calculation of the X-Ray Spectrum of a Mammography System with Various Voltages and Different Anode-Filter Combinations Using MCNP Code

Directory of Open Access Journals (Sweden)

Lida Gholamkar

2016-09-01

Full Text Available Introduction One of the best methods in the diagnosis and control of breast cancer is mammography. The importance of mammography is directly related to its value in the detection of breast cancer in the early stages, which leads to a more effective treatment. The purpose of this article was to calculate the X-ray spectrum in a mammography system with Monte Carlo codes, including MCNPX and MCNP5. Materials and Methods The device, simulated using the MCNP code, was Planmed Nuance digital mammography device (Planmed Oy, Finland, equipped with an amorphous selenium detector. Different anode/filter materials, such as molybdenum-rhodium (Mo-Rh, molybdenum-molybdenum (Mo-Mo, tungsten-tin (W-Sn, tungsten-silver (W-Ag, tungsten-palladium (W-Pd, tungsten-aluminum (W-Al, tungsten-molybdenum (W-Mo, molybdenum-aluminum (Mo-Al, tungsten-rhodium (W-Rh, rhodium-aluminum (Rh-Al, and rhodium-rhodium (Rh-Rh, were simulated in this study. The voltage range of the X-ray tube was between 24 and 34 kV with a 2 kV interval. Results The charts of changing photon flux versus energy were plotted for different types of anode-filter combinations. The comparison with the findings reported by others indicated acceptable consistency. Also, the X-ray spectra, obtained from MCNP5 and MCNPX codes for W-Ag and W-Rh combinations, were compared. We compared the present results with the reported data of MCNP4C and IPEM report No. 78 for Mo-Mo, Mo-Rh, and W-Al combinations. Conclusion The MCNPX calculation outcomes showed acceptable results in a low-energy X-ray beam range (10-35 keV. The obtained simulated spectra for different anode/filter combinations were in good conformity with the finding of previous research.

Evaluation of a 50-MV photon therapy beam from a racetrack microtron using MCNP4B Monte Carlo code

International Nuclear Information System (INIS)

Gudowska, I.; Svensson, R.

2001-01-01

High energy photon therapy beam from the 50 MV racetrack microtron has been evaluated using the Monte Carlo code MCNP4B. The spatial and energy distribution of photons, radial and depth dose distributions in the phantom are calculated for the stationary and scanned photon beams from different targets. The calculated dose distributions are compared to the experimental data using a silicon diode detector. Measured and calculated depth-dose distributions are in fairly good agreement, within 2-3% for the positions in the range 2-30 cm in the phantom, whereas the larger discrepancies up to 10% are observed in the dose build-up region. For the stationary beams the differences in the calculated and measured radial dose distributions are about 2-10%. (orig.)

Simplification of an MCNP model designed for dose rate estimation

Science.gov (United States)

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.

Simplification of an MCNP model designed for dose rate estimation

Directory of Open Access Journals (Sweden)

Laptev Alexander

2017-01-01

Full Text Available 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.

Semi-Analytical Benchmarks for MCNP6

Energy Technology Data Exchange (ETDEWEB)

Grechanuk, Pavel Aleksandrovi [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-11-07

Code verification is an extremely important process that involves proving or disproving the validity of code algorithms by comparing them against analytical results of the underlying physics or mathematical theory on which the code is based. Monte Carlo codes such as MCNP6 must undergo verification and testing upon every release to ensure that the codes are properly simulating nature. Specifically, MCNP6 has multiple sets of problems with known analytic solutions that are used for code verification. Monte Carlo codes primarily specify either current boundary sources or a volumetric fixed source, either of which can be very complicated functions of space, energy, direction and time. Thus, most of the challenges with modeling analytic benchmark problems in Monte Carlo codes come from identifying the correct source definition to properly simulate the correct boundary conditions. The problems included in this suite all deal with mono-energetic neutron transport without energy loss, in a homogeneous material. The variables that differ between the problems are source type (isotropic/beam), medium dimensionality (infinite/semi-infinite), etc.

38 CFR 4.112 - Weight loss.

Science.gov (United States)

2010-07-01

.... The term “inability to gain weight” means that there has been substantial weight loss with inability... 38 Pensions, Bonuses, and Veterans' Relief 1 2010-07-01 2010-07-01 false Weight loss. 4.112... DISABILITIES Disability Ratings The Digestive System § 4.112 Weight loss. For purposes of evaluating conditions...

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

Science.gov (United States)

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.

Transparent Wood Smart Windows: Polymer Electrochromic Devices Based on Poly(3,4-Ethylenedioxythiophene):Poly(Styrene Sulfonate) Electrodes.

Science.gov (United States)

Lang, Augustus W; Li, Yuanyuan; De Keersmaecker, Michel; Shen, D Eric; Österholm, Anna M; Berglund, Lars; Reynolds, John R

2018-03-09

Transparent wood composites, with their high strength and toughness, thermal insulation, and excellent transmissivity, offer a route to replace glass for diffusely transmitting windows. Here, conjugated-polymer-based electrochromic devices (ECDs) that switch on-demand are demonstrated using transparent wood coated with poly(3,4-ethylenedioxythiophene):poly(styrene sulfonate) (PEDOT:PSS) as a transparent conducting electrode. These ECDs exhibit a vibrant magenta-to-clear color change that results from a remarkably colorless bleached state. Furthermore, they require low energy and power inputs of 3 mWh m -2 at 2 W m -2 to switch due to a high coloration efficiency (590 cm 2  C -1 ) and low driving voltage (0.8 V). Each device component is processed with high-throughput methods, which highlights the opportunity to apply this approach to fabricate mechanically robust, energy-efficient smart windows on a large scale. © 2018 The Authors. Published by Wiley-VCH Verlag GmbH & Co. KGaA.

Influence of scanning time window on the binding potentials of dopamine transporter in the brain of healthy volunteers with 11C-CFT PET imaging

International Nuclear Information System (INIS)

Qiu Chun; Zuo Chuantao; Zhang Zhengwei; Wu Ping; Zhang Huiwei; Guan Yihui

2013-01-01

Objective: To find the optimal scanning time window and then set up the normal binding potentials (BPs) of 2β-carbomethoxy-3β-(4-fluorophenyl)-(N- 11 C-methyl) tropane ( 11 C-CFT) DAT PET/CT imaging. Methods: Thirty-one healthy volunteers (20 males, 11 females, average age: (55.7±2.3) years), who all gave written informed consent, were divided into three age and gender-matched groups according to block randomization. Each group underwent static PET/CT scan in different time windows from 40-60 min, 60-80 min to 80-100 min after 11 C-CFT injection. To determine the best scanning time window, the ratios of caudate and putamen of all volunteers were analyzed using automatic ROI method (caudate (putamen)/parieto-occipital cotex-1) and compared by one-way analysis of variance and the least significant difference (LSD) t test. The ratio of the same area between different age-groups and gender-groups was compared with independent two-sample t test. Results: Ratios of left caudate (2.08±0.06, 1.75±0.07 and 1.77±0.12 respectively), right anterior putamen (2.33±0.06, 1.95±0.09 and 2.08±0.12 respectively) and bilateral posterior putamen (left: 1.88±0.66, 1.55±0.88 and 1.72±0.09; right: 1.98±0.07, 1.61±0.09 and 1.69±0.12) were all different in three time windows (F=3.588, 3.345, 4.479, 3.557, all P<0.05). There were significant differences in ratios of left caudate, right anterior and bilateral posterior putamen between 40-60 min and the 60-80 min (all P<0.05), as well as the ratios of left caudate between 40-60 min and the 80-100 min group (P<0.05). While no valid differences in ratios of those areas were shown between the groups of 60-80 min and 80-100 min scanning time window (all P>0.05). DAT densities in right and left side of caudate, anterior and posterior putamen were significantly lower in the group over 60 years of age than those under 60 years (t=-3.260, -3.090, -3.270, -3.190, -2.270, -3.110, all P<0.05), but were not different between gender

Use of MCNP for characterization of reactor vessel internals waste from decommissioned nuclear reactors

International Nuclear Information System (INIS)

Love, E.F.; Pauley, K.A.; Reid, B.D.

1995-09-01

This study describes the use of the Monte Carlo Neutron-Photon (MCNP) code for determining activation levels of irradiated reactor vessel internals hardware. The purpose of the analysis is to produce data for the Department of Energy's Greater-Than-Class C Low-Level Radioactive Waste Program. An MCNP model was developed to analyze the Yankee Rowe reactor facility. The model incorporates reactor geometry, material compositions, and operating history data acquired from Yankee Atomic Electric Company. In addition to the base activation analysis, parametric studies were performed to determine the sensitivity of activation to specific parameters. A component sampling plan was also developed to validate the model results, although the plan was not implemented. The calculations for the Yankee Rowe reactor predict that only the core baffle and the core support plates will be activated to levels above the Class C limits. The parametric calculations show, however, that the large uncertainties in the material compositions could cause errors in the estimates that could also increase the estimated activation level of the core barrel to above the Class C limits. Extrapolation of the results to other reactor facilities indicates that in addition to the baffle and support plates, core barrels may also be activated to above Class C limits; however the classification will depend on the specific operating conditions of the reactor and the specific material compositions of the metal, as well as the use of allowable concentration averaging practices in packaging and classifying the waste

Processing and benchmarking of evaluated nuclear data file/b-viii.0β4 cross-section library by analysis of a series of critical experimental benchmark using the monte carlo code MCNP(X and NJOY2016

Directory of Open Access Journals (Sweden)

Kabach Ouadie

2017-12-01

Full Text Available To validate the new Evaluated Nuclear Data File (ENDF/B-VIII.0β4 library, 31 different critical cores were selected and used for a benchmark test of the important parameter keff. The four utilized libraries are processed using Nuclear Data Processing Code (NJOY2016. The results obtained with the ENDF/B-VIII.0β4 library were compared against those calculated with ENDF/B-VI.8, ENDF/B-VII.0, and ENDF/B-VII.1 libraries using the Monte Carlo N-Particle (MCNP(X code. All the MCNP(X calculations of keff values with these four libraries were compared with the experimentally measured results, which are available in the International Critically Safety Benchmark Evaluation Project. The obtained results are discussed and analyzed in this paper.

Validation of the MCNP-DSP Monte Carlo code for calculating source-driven noise parameters of subcritical systems

International Nuclear Information System (INIS)

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

1995-01-01

This paper describes calculations performed to validate the modified version of the MCNP code, the MCNP-DSP, used for: the neutron and photon spectra of the spontaneous fission of californium 252; the representation of the detection processes for scattering detectors; the timing of the detection process; and the calculation of the frequency analysis parameters for the MCNP-DSP code

Criticality calculations of the HTR-10 pebble-bed reactor with SCALE6/CSAS6 and MCNP5

International Nuclear Information System (INIS)

Wang, Meng-Jen; Sheu, Rong-Jiun; Peir, Jinn-Jer; Liang, Jenq-Horng

2014-01-01

Highlights: • Comparisons of the HTR-10 criticality calculations with SCALE6/CSAS6 and MCNP5 were performed. • The DOUBLEHET unit-cell treatment provides the best k eff estimation among PBR criticality calculations using SCALE6. • The continuous-energy SCALE6 calculations present a non-negligible discrepancy with MCNP5 in three PBR cases. - Abstract: HTR-10 is a 10 MWt prototype pebble-bed reactor (PBR) that presents a doubly heterogeneous geometry for neutronics calculations. An appropriate unit-cell treatment for the associated fuel elements is vital for creating problem-dependent multigroup cross sections. Considering four unit-cell options for resonance self-shielding correction in SCALE6, a series of HTR-10 core models were established using the CSAS6 sequence to systematically investigate how they affected the computational accuracy and efficiency of PBR criticality calculations. Three core configurations, which ranged from simplified infinite lattices to a detailed geometry, were examined. Based on the same ENDF/B-VII.0 cross-section library, multigroup results were evaluated by comparing with continuous-energy SCALE6/CSAS6 and MCNP5 calculations. The comparison indicated that the INFHOMMEDIUM results overestimated the effective multiplication factor (k eff ) by about 2800 pcm, whereas the LATTICECELL and MULTIREGION treatments overestimated k eff values with similar biases at approximately 470–680 pcm. The DOUBLEHET results attained further improvement, reducing the k eff overestimation to approximately 280 pcm. The comparison yielded two unexpected problems from using SCALE6/CSAS6 in HTR-10 criticality calculations. In particular, the continuous-energy CSAS6 calculations in this study present a non-negligible discrepancy with MCNP5, potentially causing a k eff value overestimate of approximately 680 pcm. Notably, using a cell-weighted mixture instead of an explicit model of individual TRISO particles in the pebble fuel zone does not shorten the

Comparison of TITAN hybrid deterministic transport code and MCNP5 for simulation of SPECT

International Nuclear Information System (INIS)

Royston, K.; Haghighat, A.; Yi, C.

2010-01-01

Traditionally, Single Photon Emission Computed Tomography (SPECT) simulations use Monte Carlo methods. The hybrid deterministic transport code TITAN has recently been applied to the simulation of a SPECT myocardial perfusion study. The TITAN SPECT simulation uses the discrete ordinates formulation in the phantom region and a simplified ray-tracing formulation outside of the phantom. A SPECT model has been created in the Monte Carlo Neutral particle (MCNP)5 Monte Carlo code for comparison. In MCNP5 the collimator is directly modeled, but TITAN instead simulates the effect of collimator blur using a circular ordinate splitting technique. Projection images created using the TITAN code are compared to results using MCNP5 for three collimator acceptance angles. Normalized projection images for 2.97 deg, 1.42 deg and 0.98 deg collimator acceptance angles had maximum relative differences of 21.3%, 11.9% and 8.3%, respectively. Visually the images are in good agreement. Profiles through the projection images were plotted to find that the TITAN results followed the shape of the MCNP5 results with some differences in magnitude. A timing comparison on 16 processors found that the TITAN code completed the calculation 382 to 2787 times faster than MCNP5. Both codes exhibit good parallel performance. (author)

Numerical analysis of the feasibility of a beam window for TEF target

International Nuclear Information System (INIS)

Obayashi, H.; Takei, H.; Iwamoto, H.; Sasa, T.

2015-01-01

In the various research activities for ADS (accelerator-driven systems), the construction of the Transmutation Experimental Facility (TEF) is planned within the framework of the J-PARC project as a preceding step before the construction of a pilot ADS. The ADS proposed by JAEA is a tank-type subcritical reactor with a thermal power of 800 MWth which uses lead-bismuth eutectic (LBE) alloy as a target material and a coolant, driven by a 30 MW superconducting proton linac. The beam window of ADS, which separates the proton accelerator and the LBE subcritical core vessel, is exposed to the high-temperature environment induced by the incidence of proton beam. Therefore, the feasibility of the beam window is the most important factor for the realisation of TEF. The objective of this study is to evaluate the feasibility of a designed beam window of TEF target by the numerical analysis with a three-dimensional model. The analysis was performed by considering: (1) the current density and shape of the incident beam, (2) the thermal-fluid behaviour of LBE around the beam window as a function of the flow rate and inlet temperature, (3) the material and the thickness of the beam window, (4) the structural strength of the beam window. In the reference case, the current density and the profile of the proton beam were 20 μA/cm 2 and a Gaussian shape, respectively. The flow rate of LBE and temperature at the inlet were 1 l/s and 350 C. degrees. The material of a beam window was SUS316 stainless steel 2 mm thick. In this reference case, the maximum velocity of LBE and the maximum temperature located at the top of the beam window were about 1.2 m/sec and 477 C. degrees. By increasing the flow rate of LBE up to 4 l/s, the maximum temperature of a beam window was reduced to around 420 C. degrees. The maximum shear stress was 194 MPa, which was observed at the centre on the outside surface of the beam window. The analysed stress in the reference case was lower than the tolerance level

4,4'-Dichlorodiphenyltrichloroethane (DDT) and 4,4'-dichlorodiphenyldichloroethylene (DDE) inhibit myogenesis in C2C12 myoblasts.

Science.gov (United States)

Kim, Jonggun; Park, Min Young; Kim, Yoo; Yoon, Kyong Sup; Clark, John Marshall; Park, Yeonhwa; Whang, Kwang-Youn

2017-12-01

Most countries have banned the use of 4,4'-dichlorodiphenyltrichloroethane (DDT). However, owing to its extremely high lipophilic characteristics, DDT and its metabolite 4,4'-dichlorodiphenyldichloroethylene (DDE) are ubiquitous in the environment and in many types of food. The positive correlation between exposure to insecticides, including DDT and DDE, and weight gain, resulting in impaired energy metabolism in offspring following perinatal DDT and DDE exposure, was previously reported. Therefore the influence of DDT and DDE on myogenesis using C2C12 myoblasts was investigated in this study. DDT and DDE decreased myotube formation dose- and time-dependently. Among myogenic regulatory factors, DDT and DDE mainly decreased MyoD1 and Myf5 expression. DDT and DDE treatment also altered Myostatin expression, phosphorylation of protein kinase B, p70 ribosomal protein S6 kinase, forkhead box O protein 3 and mammalian target of rapamycin, resulting in attenuation of myotube formation. These results may have significant implications for understanding the effects of developmental exposure of DDT and DDE on myogenesis and development of obesity and type 2 diabetes later in life. © 2017 Society of Chemical Industry. © 2017 Society of Chemical Industry.

Window Glazing Types | Efficient Windows Collaborative

Science.gov (United States)

Foundry Foundry New Construction Windows Window Selection Tool Selection Process Design Guidance Installation Replacement Windows Window Selection Tool Assessing Options Selection Process Design Guidance Installation Understanding Windows Benefits Design Considerations Measuring Performance Performance Standards

Window Frame Types | Efficient Windows Collaborative

Science.gov (United States)

Foundry Foundry New Construction Windows Window Selection Tool Selection Process Design Guidance Installation Replacement Windows Window Selection Tool Assessing Options Selection Process Design Guidance Installation Understanding Windows Benefits Design Considerations Measuring Performance Performance Standards

Window Operator Types | Efficient Windows Collaborative

Science.gov (United States)

Foundry Foundry New Construction Windows Window Selection Tool Selection Process Design Guidance Installation Replacement Windows Window Selection Tool Assessing Options Selection Process Design Guidance Installation Understanding Windows Benefits Design Considerations Measuring Performance Performance Standards

Testing of the ENDF/B-VI neutron data library ENDF60 for use with MCNP trademark

International Nuclear Information System (INIS)

Frankle, S.C.; MacFarlane, R.E.

1995-01-01

The continuous-energy neutron data library ENDF60, for use with the Monte Carlo N-Particle radiation transport code MCNP4A, was released in the fall of 1994. It is comprised of 124 nuclide data files based on the ENDF/B-Vi evaluations through Release 2. Forty-eight percent of these materials are new or modified evaluations, while the balance are translations from ENDF/B-V. The new evaluations include most of the important materials for criticality safety calculations, and include significant enhancements such as more isotopic evaluations, better resonance-range representations, and the new correlated energy-angle distributions for emitted particles. As part of the overall quality assurance testing of the ENDF60 library, calculations for well known benchmark assemblies were performed. The results of these calculations help the user to know how the combination of ENDF60 and MCNP4A will perform for real problems

Radiation field characterization of a BNCT research facility using Monte Carlo method - code MCNP-4B

International Nuclear Information System (INIS)

Hernandez, Antonio Carlos

2002-01-01

Boron Neutron Capture Therapy - BNCT - is a selective cancer treatment and arises as an alternative therapy to treat cancer when usual techniques - surgery, chemotherapy or radiotherapy - show no satisfactory results. The main proposal of this work is to project a facility to BNCT studies. This facility relies on the use of an Am Be neutron source and on a set of moderators, filters and shielding which will provide the best neutron/gamma beam characteristic for these Becton studies, i.e., high intensity thermal and/or epithermal neutron fluxes and with the minimum feasible gamma rays and fast neutrons contaminants. A computational model of the experiment was used to obtain the radiation field in the sample irradiation position. The calculations have been performed with the MCNP 4B Monte Carlo Code and the results obtained can be regarded as satisfactory, i.e., a thermal neutron fluencyN T = 1,35x10 8 n/cm , a fast neutron dose of 5,86x10 -10 Gy/N T and a gamma ray dose of 8,30x10 -14 Gy/N T . (author)

Radiation field characterization of a BNCT research facility using Monte Carlo Method - Code MCNP-4B

International Nuclear Information System (INIS)

Hernandes, Antonio Carlos

2002-01-01

Boron Neutron Capture Therapy - BNCT- is a selective cancer treatment and arises as an alternative therapy to treat cancer when usual techniques - surgery, chemotherapy or radiotherapy - show no satisfactory results. The main proposal of this work is to project a facility to BNCT studies. This facility relies on the use of an AmBe neutron source and on a set of moderators, filters and shielding which will provide the best neutron/gamma beam characteristic for these BNCT studies, i.e., high intensity thermal and/or epithermal neutron fluxes and with the minimum feasible gamma rays and fast neutrons contaminants. A computational model of the experiment was used to obtain the radiation field in the sample irradiation position. The calculations have been performed with the MCNP 4B Monte Carlo Code and the results obtained can be regarded as satisfactory, i.e., a thermal neutron fluency Ν Τ = 1,35x10 8 n/cm 2 , a fast neutron dose of 5,86x -1 0 Gy/Ν Τ and a gamma ray dose of 8,30x -14 Gy/Ν Τ . (author)

Studies on the liquid fluoride thorium reactor: Comparative neutronics analysis of MCNP6 code with SRAC95 reactor analysis code based on FUJI-U3-(0)

Energy Technology Data Exchange (ETDEWEB)

Jaradat, S.Q., E-mail: sqjxv3@mst.edu; Alajo, A.B., E-mail: alajoa@mst.edu

2017-04-01

Highlights: • The verification for FUJI-U3-(0)—a molten salt reactor—was performed. • The MCNP6 was used to study the reactor physics characteristics for FUJI-U3 type. • The results from the MCNP6 were comparable with the ones obtained from literature. - Abstract: The verification for FUJI-U3-(0)—a molten salt reactor—was performed. The reactor used LiF-BeF2-ThF4-UF4 as the mixed liquid fuel salt, and the core was graphite moderated. The MCNP6 code was used to study the reactor physics characteristics for the FUJI-U3-(0) reactor. Results for reactor physics characteristic of the FUJI-U3-(0) exist in literature, which were used as reference. The reference results were obtained using SRAC95 (a reactor analysis code) coupled with ORIGEN2 (a depletion code). Some modifications were made in the reconstruction of the FUJI-U3-(0) reactor in MCNP due to unavailability of more detailed description of the reactor core. The assumptions resulted in two representative models of the reactor. The results from the MCNP6 models were compared with the reference results obtained from literature. The results were comparable with each other, but with some notable differences. The differences are because of the approximations that were done on the SRAC95 model of the FUJI-U3 to simplify the simulation. Based on the results, it is concluded that MCNP6 code predicts well the overall simulation of neutronics analysis to the previous simulation works using SRAC95 code.

Calculation of the importance-weighted neutron generation time using MCNIC method

International Nuclear Information System (INIS)

Feghhi, S.A.H.; Shahriari, M.; Afarideh, H.

2008-01-01

In advanced nuclear power systems, such as ADS, the need for reliable kinetics parameters is of considerable importance because of the lower value for β eff due to the large amount of transuranic elements loaded in the core of those systems. All reactor kinetic parameters are weighted quantities. In other words each neutron with a given position and energy is weighted with its importance. Neutron generation time as an important kinetic parameter, in all nuclear power systems has a significant role in the analysis of fast transients. The difference between non-weighted neutron generation time; Λ; standard in most Monte Carlo codes; and the weighted one Λ + can be quite significant depending on the type of the system. In previous work, based on the physical concept of neutron importance, a new method; MCNIC; using the MCNP code has been introduced for the calculation of neutron importance in fissionable assemblies for all criticality states. In the present work the applicability of MCNIC method has been extended for the calculation of the importance-weighted neutron generation time. The influence of reflector thickness on importance-weighted neutron generation time has been investigated by the development of an auxiliary code, IWLA, for a hypothetic assembly. The results of these calculations were compared with the non-weighted neutron generation times calculated using the Monte Carlo code MCNP. The difference between the importance-weighted and non-weighted quantity is more significant in a reflected system and increases with reflector thickness

MCNP calculations for the HCPB submodules in-pile test

Energy Technology Data Exchange (ETDEWEB)

Pijlgroms, B.J. [Section Nuclear and Reactor Physics, ECN Nuclear Research, Petten (Netherlands)

1998-11-01

This report describes the MCNP calculations that have been performed for the Helium Cooled Pebble Bed (HCPB) Submodules In-pile Test that has been planned for irradiation in the materials testing High Flux Reactor (HFR) at Petten. In this test, four HSM-8 submodules will be placed at core position H4. The report presents the neutron flux and power density profiles to be expected in the submodules. For the gamma induced heating only a rough estimation could be made. In the HCPB submodules the total specific heating does not exceed (36.7 {+-} 2.9)[W/cc]. 8 refs.

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

Science.gov (United States)

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.

EJ2-MCNPlib. Contents of the JEF-2.2 based neutron cross-section library for MCNP4A

International Nuclear Information System (INIS)

Hogenbirk, A.; Oppe, J.

1995-05-01

In this report a description is given of the EJ2-MCNPlib library. The EJ2-MCNPlib library is to be used for reactivity/critically calculations and general neutron/photon transport calculations with the Monte Carlo code MCNP4A. The library is based on the European JEF-2.2 nuclear data evaluation and contains data for all (i.e. 313) nuclides available on this evaluation.The cross-section data were generated using the NJOY cross-section processing code system, version 91.118. For easy reference cross-section plots are given in this report for the total, elastic and absorption cross sections for all nuclides on the EJ2-MCNPlib library. Furthermore, for verification purposes a graphical intercomparison is given of the results of standard benchmark calculations performed with JEF-2.2 cross-section data and with ENDF/B-V cross-section data (whenever available). 6 refs

Determination of the detection efficiency of a HPGe detector by means of the MCNP 4A simulation code

International Nuclear Information System (INIS)

Leal, B.

2004-01-01

In the majority of the laboratories, the calibration in efficiency of the detector is carried out by means of the standard sources measurement of gamma photons that have a determined activity, or for matrices that contain a variety of radionuclides that can embrace the energy range of interest. Given the experimental importance that has the determination from the curves of efficiency to the effects of establishing the quantitative results, is appealed to the simulation of the response function of the detector used in the Regional Center of Nuclear Studies inside the energy range of 80 keV to 1400 keV varying the density of the matrix, by means of the application of the Monte Carlo code MCNP-4A. The adjustment obtained shows an acceptance grade in the range of 100 to 600 keV, with a smaller percentage discrepancy to 5%. (Author)

Spark plasma sintering of bulk SrAl2O4-Sr3Al2O6 eutectic glass with wide-band optical window

Science.gov (United States)

Liu, Jiaxi; Lu, Nan; He, Gang; Li, Xiaoyu; Li, Jianqiang; Li, Jiangtao

2018-06-01

SrAl2O4-Sr3Al2O6 eutectic glass was prepared by using an aerodynamic levitator equipped with a CO2 laser device. A bulk transparent amorphous sample was obtained by the spark plasma sintering (SPS) of the prepared eutectic glass. XRD, a UV–vis-NIR spectrophotometer and FT-IR were employed to characterize the phase evolution and optical properties. The results show that the bulk SrAl2O4-Sr3Al2O6 samples fabricated by the containerless process and SPS between 852 °C–857 °C were fully amorphous. The amorphous sample has a wide transparent window between 270 nm and 6.2 μm. The average refractive index in the visible light region is 1.680 and the Abbe number is 27.4. The prepared bulk SrAl2O4-Sr3Al2O6 eutectic glass with the wide-band optical window may be a promising candidate for optical applications.

Generating Importance Map for Geometry Splitting using Discrete Ordinates Code in Deep Shielding Problem

International Nuclear Information System (INIS)

Kim, Jong Woon; Lee, Young Ouk

2016-01-01

When we use MCNP code for a deep shielding problem, we prefer to use variance reduction technique such as geometry splitting, or weight window, or source biasing to have relative error within reliable confidence interval. To generate importance map for geometry splitting in MCNP calculation, we should know the track entering number and previous importance on each cells since a new importance is calculated based on these information. If a problem is deep shielding problem such that we have zero tracks entering on a cell, we cannot generate new importance map. In this case, discrete ordinates code can provide information to generate importance map easily. In this paper, we use AETIUS code as a discrete ordinates code. Importance map for MCNP is generated based on a zone average flux of AETIUS calculation. The discretization of space, angle, and energy is not necessary for MCNP calculation. This is the big merit of MCNP code compared to the deterministic code. However, deterministic code (i.e., AETIUS) can provide a rough estimate of the flux throughout a problem relatively quickly. This can help MCNP by providing variance reduction parameters. Recently, ADVANTG code is released. This is an automated tool for generating variance reduction parameters for fixed-source continuous-energy Monte Carlo simulations with MCNP5 v1.60

New data for MCNP

International Nuclear Information System (INIS)

Hendricks, J.S.; Frankle, S.C.; Court, J.D.

1994-01-01

We report here for the first time the availability of an official set of ENDF/B-VI neutron data for MCNP(trademark). The LANL Radiation Transport group engaged the Nuclear Theory and Applications Group to construct a complete library based on ENDF/B-VI Release in the Spring of 1994. A new and thorough set of quality assurance tests was established and data passing those tests were subject only to a limited set of benchmarking tests. All nuclides were subjected to infinite medium calculations. The fissionable materials were benchmarked against critical assemblies, and 28 nuclides were benchmarked against the LLNL pulsed sphere experiments

Enhancement of thermal neutron attenuation of nano-B4C, -BN dispersed neutron shielding polymer nanocomposites

International Nuclear Information System (INIS)

Kim, Jaewoo; Lee, Byung-Chul; Uhm, Young Rang; Miller, William H.

2014-01-01

Highlights: • Preparation of B 4 C and BN nanopowders using a simple ball milling process. • Homogeneous dispersion and strong adhesion of nano-B 4 C and -BN with polymer matrix. • Enhancement of mechanical properties of the nanocomposites compared to their micro counterparts. • Enhancement of thermal neutron attenuation of the nanocomposites. - Abstract: Nano-sized boron carbide (B 4 C) and boron nitride (BN) powder were prepared using ball milling. Micro- and milled nano-powders were melt blended with high density polyethylene (HDPE) using a polymer mixer followed by hot pressing to fabricate sheet composites. The tensile and flexural strengths of HDPE nanocomposites were ∼20% higher than their micro counterparts, while those for latter decreased compared to neat HDPE. Thermal neutrons attenuation of the prepared HDPE nanocomposites was evaluated using a monochromatic ∼0.025 eV neutron beam. Thermal neutron attenuation of the HDPE nanocomposites was greatly enhanced compared to their micro counterparts at the same B-10 areal densities. Monte Carlo n-Particles (MCNP) simulations based on the lattice structure modeling also shows the similar filler size dependent thermal neutron absorption

Flux at a point in MCNP

International Nuclear Information System (INIS)

Cashwell, E.D.; Schrandt, R.G.

1980-01-01

The current state of the art of calculating flux at a point with MCNP is discussed. Various techniques are touched upon, but the main emphasis is on the fast improved version of the once-more-collided flux estimator, which has been modified to treat neutrons thermalized by the free gas model. The method is tested on several problems on interest and the results are presented

MCNPX{trademark} -- The LAHET{trademark}/MCNP{trademark} code merger

Energy Technology Data Exchange (ETDEWEB)

Hughes, H.G.; Adams, K.J.; Chadwick, M.B. [and others

1997-08-01

The MCNP code is written and maintained by Group X-TM at Los Alamos National Laboratory. In response to the demands of the accelerator community, the authors have undertaken a major effort to expand the capabilities of MCNP to increase the set of transportable particles; to make use of newly evaluated high-energy nuclear data tables for neutrons, protons, and potentially other particles; and to incorporate physics models for use where tabular data are unavailable. A preliminary version of the expanded code, called MCNPX, has now been issued for testing. The new code includes all existing LAHET physics modules, and has the ability to utilize the 150-MeV data libraries that have recently been released by LANL Group T-2.

Convergence testing for MCNP5 Monte Carlo eigenvalue calculations

International Nuclear Information System (INIS)

Brown, F.; Nease, B.; Cheatham, J.

2007-01-01

Determining convergence of Monte Carlo criticality problems is complicated by the statistical noise inherent in the random, walks of the neutrons in each generation. The latest version of MCNP5 incorporates an important new tool for assessing convergence: the Shannon entropy of the fission source distribution, H src . Shannon entropy is a well-known concept from information theory and provides a single number for each iteration to help characterize convergence trends for the fission source distribution. MCNP5 computes H src for each iteration, and these values may be plotted to examine convergence trends. Convergence testing should include both k eff and H src , since the fission distribution will converge more slowly than k eff , especially when the dominance ratio is close to 1.0. (authors)

Use the nuclear code MCNP4X in the study of the behavior of nuclear probe in soils with variation of Mg, Ca, Fe

International Nuclear Information System (INIS)

Braga, Mario R.M.S.S.; Oliveira, Arno H.; Lima, Claubia P.B.

2013-01-01

The aim of this work is to evaluate the behavior of the variation the elements: Mg, Ca, Fe in the soils composition on a nuclear probe to measure the density of porous materials nondestructive in testing based on coherent Compton Effect, the effect Rayleigh. To study the effect of composition in soil was used nuclear code MCNP4X where was simulated two sources, a source 14mCi americium-241 and other source 4mCi cesium-137, lead shielding and volume scintillator. To avoid problems with geometries were simulated spheres with 1.00 meters of diameter filled with soil to be evaluated. Data analysis allowed establishing correction parameters for nuclear probe. (author)

Support for ITER ECRF design. Substask 3. Design and optimzation of the window unit. Final Report

International Nuclear Information System (INIS)

Thumm, M.; Braz, O.; Iatrou, C.T.; Heidinger, S.; Henry, S.; Makowski, M.; Spoerl, R.; Szczesny, J.

1999-02-01

The proposed ITER ECH window unit employs a single, edge-cooled (water, e.g. 20 C) CVD-diamond disk in a corrugated HE 11 waveguide with 52 mm inner diameter, with an outer disk diameter of 77 mm and a thickness of 1.482 mm (4 λ/2). Thermal computations show that for larger outer disk diameters the peak temperature is unaffected. Thus due to the high thermal conductivity of the diamond, the exposed window edge area does not have to be large to obtain significant heat transfer. This implies that the window diameter can be minimized which has the added benefit of reducing the cost. For a power of 1 MW at 170 GHz, a loss tangent of 1.10 -5 , a thermal conductivity of 1800 W/mK (at room temperature) and a heat transfer coefficient of 12 kW/m 2 K (water flow: 13.5 1/min, water velocity: 2 m/s, room temperature) to the cooling water the central window temperature will not be higher than approx. 45 C and the edge temperature is about 30 C. The absorbed power is 176 W. Simulations also show that steady state conditions are generally achieved in under 3 s and that a 2 MW window should be feasible. Owing to the negligible temperature dependence of the CVD-diamond loss tangent, also the approximately 100 C hot torus cooling water could be used to cool the window. Simulations of an ''encased'' window, a window in which the edge of the disk has been covered with a 0.4 mm thick layer of electrodeposited copper (tritium barrier in case of broken window disk), show that this is feasible without a significant decrease in heat transfer rate. Neutron irradiation tests were extended to fluences of 10 21 n/m 2 (E > 0.1 MeV)

First line shape analysis and spectroscopic parameters for the ν11 band of 12C2H4

KAUST Repository

Es-sebbar, Et-touhami

2016-08-11

An accurate knowledge of line intensities, collisional broadening coefficients and narrowing parameters is necessary for the interpretation of high-resolution infrared spectra of the Earth and other planetary atmospheres. One of the most promising spectral domains for (C2H4)-C-12 monitoring in such environments is located near the 336 gm window, through its v(11) C-H stretching mode. In this paper, we report an extensive study in which we precisely determine spectroscopic parameters of (C2H4)-C-12 v(11) band at 297 +/- 1 K, using a narrow Difference-Frequency-Generation (DFG) laser with 10(-4) cm(-1) resolution. Absorption measurements were performed in the 2975-2980 cm(-1) spectral window to investigate 32 lines corresponding to where, J\\'ka\\',kc\\'<- Jka,kc, 5 <= J <= 7; 0.5 <= K-a <= 6 and 1 <= K-c <= 14. Spectroscopic parameters are retrieved using either Voigt or appropriate Galatry profile to simulate the measured (C2H4)-C-12 line shape. Line intensities along with self-broadening coefficients are reported for all lines. Narrowing coefficients for each isolated line are also derived. To our knowledge, the current study reports the first extensive spectroscopic parameter measurements of the (C2H4)-C-12 v(11) band in the 2975-2980 cm(-1) range. (C) 2016 Elsevier Ltd. All rights reserved.

First line shape analysis and spectroscopic parameters for the ν11 band of 12C2H4

KAUST Repository

Es-sebbar, Et-touhami; Mantzaras, John; Benilan, Yves; Farooq, Aamir

2016-01-01

An accurate knowledge of line intensities, collisional broadening coefficients and narrowing parameters is necessary for the interpretation of high-resolution infrared spectra of the Earth and other planetary atmospheres. One of the most promising spectral domains for (C2H4)-C-12 monitoring in such environments is located near the 336 gm window, through its v(11) C-H stretching mode. In this paper, we report an extensive study in which we precisely determine spectroscopic parameters of (C2H4)-C-12 v(11) band at 297 +/- 1 K, using a narrow Difference-Frequency-Generation (DFG) laser with 10(-4) cm(-1) resolution. Absorption measurements were performed in the 2975-2980 cm(-1) spectral window to investigate 32 lines corresponding to where, J'ka',kc'<- Jka,kc, 5 <= J <= 7; 0.5 <= K-a <= 6 and 1 <= K-c <= 14. Spectroscopic parameters are retrieved using either Voigt or appropriate Galatry profile to simulate the measured (C2H4)-C-12 line shape. Line intensities along with self-broadening coefficients are reported for all lines. Narrowing coefficients for each isolated line are also derived. To our knowledge, the current study reports the first extensive spectroscopic parameter measurements of the (C2H4)-C-12 v(11) band in the 2975-2980 cm(-1) range. (C) 2016 Elsevier Ltd. All rights reserved.

Comparative studies of actinide and sub-actinide fission cross section calculation from MCNP6 and TALYS

International Nuclear Information System (INIS)

Perkasa, Y. S.; Waris, A.; Kurniadi, R.; Su'ud, Z.

2014-01-01

Comparative studies of actinide and sub-actinide fission cross section calculation from MCNP6 and TALYS have been conducted. In this work, fission cross section resulted from MCNP6 prediction will be compared with result from TALYS calculation. MCNP6 with its event generator CEM03.03 and LAQGSM03.03 have been validated and verified for several intermediate and heavy nuclides fission reaction data and also has a good agreement with experimental data for fission reaction that induced by photons, pions, and nucleons at energy from several ten of MeV to about 1 TeV. The calculation that induced within TALYS will be focused mainly to several hundred MeV for actinide and sub-actinide nuclides and will be compared with MCNP6 code and several experimental data from other evaluator

An MCNP parametric study of George C. Laurence's subcritical pile experiment

International Nuclear Information System (INIS)

Dranga, R.; Blomeley, L.; Carrington, R.

2014-01-01

In the early 1940s at the National Research Council (NRC) Laboratories in Ottawa, Canada, Dr. George Laurence conducted several experiments to determine if a sustained nuclear fission chain reaction in a carbon-uranium arrangement (or 'pile') was possible. Although Dr. Laurence did not achieve criticality, these pioneering experiments marked a significant historical event in nuclear science, and they provided a valuable reference for subsequent experiments that led to the design of Canada's first heavy-water reactors at the Chalk River Nuclear Laboratories. This paper summarizes the results of a recent collaborative project between Atomic Energy of Canada Limited and the Deep River Science Academy undertaken to numerically explore the experiments carried out at the NRC Laboratories by Dr. Laurence, while teaching high school students about nuclear science and technology. In this study, a modern Monte Carlo reactor physics code, MCNP6, was utilized to identify and study the key parameters impacting the subcritical pile's neutron multiplication factor (e.g., moderation, geometry, material impurities) and quantify their effect on the extent of subcriticality. The findings presented constitute the first endeavour to model, using a current computational reactor physics tool, the seminal experiment that provided the foundation of Canada's nuclear science and technology program. (author)

Investigating the Impact of Maternal Residential Mobility on Identifying Critical Windows of Susceptibility to Ambient Air Pollution During Pregnancy.

Science.gov (United States)

Warren, Joshua L; Son, Ji-Young; Pereira, Gavin; Leaderer, Brian P; Bell, Michelle L

2018-05-01

Identifying periods of increased vulnerability to air pollution during pregnancy with respect to the development of adverse birth outcomes can improve understanding of possible mechanisms of disease development and provide guidelines for protection of the child. Exposure to air pollution during pregnancy is typically based on the mother's residence at delivery, potentially resulting in exposure misclassification and biasing the estimation of critical windows of pregnancy. In this study, we determined the impact of maternal residential mobility during pregnancy on defining weekly exposure to particulate matter less than or equal to 10 μm in aerodynamic diameter (PM10) and estimating windows of susceptibility to term low birth weight. We utilized data sets from 4 Connecticut birth cohorts (1988-2008) that included information on all residential addresses between conception and delivery for each woman. We designed a simulation study to investigate the impact of increasing levels of mobility on identification of critical windows. Increased PM10 exposure during pregnancy weeks 16-18 was associated with an increased probability of term low birth weight. Ignoring residential mobility when defining weekly exposure had only a minor impact on the identification of critical windows for PM10 and term low birth weight in the data application and simulation study. Identification of critical pregnancy windows was robust to exposure misclassification caused by ignoring residential mobility in these Connecticut birth cohorts.

MCNP benchmark analyses of critical experiments for the Space Nuclear Thermal Propulsion program

International Nuclear Information System (INIS)

Selcow, E.C.; Cerbone, R.J.; Ludewig, H.; Mughabghab, S.F.; Schmidt, E.; Todosow, M.; Parma, E.J.; Ball, R.M.; Hoovler, G.S.

1993-01-01

Benchmark analyses have been performed of Particle Bed Reactor (PBR) critical experiments (CX) using the MCNP radiation transport code. The experiments have been conducted at the Sandia National Laboratory reactor facility in support of the Space Nuclear Thermal Propulsion (SNTP) program. The test reactor is a nineteen element water moderated and reflected thermal system. A series of integral experiments have been carried out to test the capabilities of the radiation transport codes to predict the performance of PBR systems. MCNP was selected as the preferred radiation analysis tool for the benchmark experiments. Comparison between experimental and calculational results indicate close agreement. This paper describes the analyses of benchmark experiments designed to quantify the accuracy of the MCNP radiation transport code for predicting the performance characteristics of PBR reactors

MCNP benchmark analyses of critical experiments for the Space Nuclear Thermal Propulsion program

Science.gov (United States)

Selcow, Elizabeth C.; Cerbone, Ralph J.; Ludewig, Hans; Mughabghab, Said F.; Schmidt, Eldon; Todosow, Michael; Parma, Edward J.; Ball, Russell M.; Hoovler, Gary S.

1993-01-01

Benchmark analyses have been performed of Particle Bed Reactor (PBR) critical experiments (CX) using the MCNP radiation transport code. The experiments have been conducted at the Sandia National Laboratory reactor facility in support of the Space Nuclear Thermal Propulsion (SNTP) program. The test reactor is a nineteen element water moderated and reflected thermal system. A series of integral experiments have been carried out to test the capabilities of the radiation transport codes to predict the performance of PBR systems. MCNP was selected as the preferred radiation analysis tool for the benchmark experiments. Comparison between experimental and calculational results indicate close agreement. This paper describes the analyses of benchmark experiments designed to quantify the accuracy of the MCNP radiation transport code for predicting the performance characteristics of PBR reactors.

Performance of MPI parallel processing implemented by MCNP5/ MCNPX for criticality benchmark problems

International Nuclear Information System (INIS)

Mark Dennis Usang; Mohd Hairie Rabir; Mohd Amin Sharifuldin Salleh; Mohamad Puad Abu

2012-01-01

MPI parallelism are implemented on a SUN Workstation for running MCNPX and on the High Performance Computing Facility (HPC) for running MCNP5. 23 input less obtained from MCNP Criticality Validation Suite are utilized for the purpose of evaluating the amount of speed up achievable by using the parallel capabilities of MPI. More importantly, we will study the economics of using more processors and the type of problem where the performance gain are obvious. This is important to enable better practices of resource sharing especially for the HPC facilities processing time. Future endeavours in this direction might even reveal clues for best MCNP5/ MCNPX coding practices for optimum performance of MPI parallelisms. (author)

Production and characterization of AA6061-B4C stir cast composite

International Nuclear Information System (INIS)

Kalaiselvan, K.; Murugan, N.; Parameswaran, Siva

2011-01-01

Highlights: → Stir casting of AA6061-B 4 C Composite. Color metallographic of composites → Enhanced wetting of B 4 C particles by K 2 TiF 6 flux. → Effect of B 4 C particles on mechanical properties of AA6061. -- Abstract: This work focuses on the fabrication of aluminum (6061-T6) matrix composites (AMCs) reinforced with various weight percentage of B 4 C particulates by modified stir casting route. The wettability of B 4 C particles in the matrix has been improved by adding K 2 TiF 6 flux into the melt. The microstructure and mechanical properties of the fabricated AMCs are analyzed. The optical microstructure and scanning electron microscope (SEM) images reveal the homogeneous dispersion of B 4 C particles in the matrix. The reinforcement dispersion has also been identified with X-ray diffraction (XRD). The mechanical properties like hardness and tensile strength have improved with the increase in weight percentage of B 4 C particulates in the aluminum matrix.

Proper Installation of Replacement Windows | Efficient Windows

Science.gov (United States)

Collaborative Foundry Foundry New Construction Windows Window Selection Tool Selection Process Design Guidance Installation Replacement Windows Window Selection Tool Assessing Options Selection Process Design Guidance Installation Understanding Windows Benefits Design Considerations Measuring

Performance Standards for Windows | Efficient Windows Collaborative

Science.gov (United States)

Foundry Foundry New Construction Windows Window Selection Tool Selection Process Design Guidance Installation Replacement Windows Window Selection Tool Assessing Options Selection Process Design Guidance Installation Understanding Windows Benefits Design Considerations Measuring Performance Performance Standards

Benefits of Efficient Windows | Efficient Windows Collaborative

Science.gov (United States)

Foundry Foundry New Construction Windows Window Selection Tool Selection Process Design Guidance Installation Replacement Windows Window Selection Tool Assessing Options Selection Process Design Guidance Installation Understanding Windows Benefits Design Considerations Measuring Performance Performance Standards

Assessing Window Replacement Options | Efficient Windows Collaborative

Science.gov (United States)

Foundry Foundry New Construction Windows Window Selection Tool Selection Process Design Guidance Installation Replacement Windows Window Selection Tool Assessing Options Selection Process Design Guidance Installation Understanding Windows Benefits Design Considerations Measuring Performance Performance Standards

Windows for New Construction | Efficient Windows Collaborative

Science.gov (United States)

Foundry Foundry New Construction Windows Window Selection Tool Selection Process Design Guidance Installation Replacement Windows Window Selection Tool Assessing Options Selection Process Design Guidance Installation Understanding Windows Benefits Design Considerations Measuring Performance Performance Standards

Track 4: basic nuclear science variance reduction for Monte Carlo criticality simulations. 2. Assessment of MCNP Statistical Analysis of keff Eigenvalue Convergence with an Analytical Criticality Verification Test Set

International Nuclear Information System (INIS)

Sood, Avnet; Forster, R. Arthur; Parsons, D. Kent

2001-01-01

Monte Carlo simulations of nuclear criticality eigenvalue problems are often performed by general purpose radiation transport codes such as MCNP. MCNP performs detailed statistical analysis of the criticality calculation and provides feedback to the user with warning messages, tables, and graphs. The purpose of the analysis is to provide the user with sufficient information to assess spatial convergence of the eigenfunction and thus the validity of the criticality calculation. As a test of this statistical analysis package in MCNP, analytic criticality verification benchmark problems have been used for the first time to assess the performance of the criticality convergence tests in MCNP. The MCNP statistical analysis capability has been recently assessed using the 75 multigroup criticality verification analytic problem test set. MCNP was verified with these problems at the 10 -4 to 10 -5 statistical error level using 40 000 histories per cycle and 2000 active cycles. In all cases, the final boxed combined k eff answer was given with the standard deviation and three confidence intervals that contained the analytic k eff . To test the effectiveness of the statistical analysis checks in identifying poor eigenfunction convergence, ten problems from the test set were deliberately run incorrectly using 1000 histories per cycle, 200 active cycles, and 10 inactive cycles. Six problems with large dominance ratios were chosen from the test set because they do not achieve the normal spatial mode in the beginning of the calculation. To further stress the convergence tests, these problems were also started with an initial fission source point 1 cm from the boundary thus increasing the likelihood of a poorly converged initial fission source distribution. The final combined k eff confidence intervals for these deliberately ill-posed problems did not include the analytic k eff value. In no case did a bad confidence interval go undetected. Warning messages were given signaling that

Comparative Analysis of the Dalat Nuclear Research Reactor with HEU Fuel Using SRAC and MCNP5

Directory of Open Access Journals (Sweden)

Giang Phan

2017-01-01

Full Text Available Neutronics analysis has been performed for the 500 kW Dalat Nuclear Research Reactor loaded with highly enriched uranium fuel using the SRAC code system. The effective multiplication factors, keff, were analyzed for the core at criticality conditions and in two cases corresponding to the complete withdrawal and the full insertion of control rods. MCNP5 calculations were also conducted and compared to that obtained with the SRAC code. The results show that the difference of the keff values between the codes is within 55 pcm. Compared to the criticality conditions established in the experiments, the maximum differences of the keff values obtained from the SRAC and MCNP5 calculations are 119 pcm and 64 pcm, respectively. The radial and axial power peaking factors are 1.334 and 1.710, respectively, in the case of no control rod insertion. At the criticality condition these values become 1.445 and 1.832 when the control rods are partially inserted. Compared to MCNP5 calculations, the deviation of the relative power densities is less than 4% at the fuel bundles in the middle of the core, while the maximum deviation is about 7% appearing at some peripheral bundles. This agreement indicates the verification of the analysis models.

Acceleration of the MCNP branch of the OCTOPUS depletion code system

Energy Technology Data Exchange (ETDEWEB)

Pijlgroms, B.J.; Hogenbirk, A.; Oppe, J. [Section Nuclear and Reactor Physics, ECN Nuclear Research, Petten (Netherlands)

1998-09-01

OCTOPUS depletion calculations using the 3D Monte Carlo spectrum code MCNP (Monte Carlo Code for Neutron and Photon Transport) require much computing time. In a former implementation, the time required by OCTOPUS to perform multi-zone calculations, increased roughly proportional to the number of burnable zones. By using a different method the situation has improved considerably. In the new implementation described here, the dependence of the computing time on the number of zones has been moved from the MCNP code to a faster postprocessing code. By this, the overall computing time will reduce substantially. 11 refs.

Acceleration of the MCNP branch of the OCTOPUS depletion code system

International Nuclear Information System (INIS)

Pijlgroms, B.J.; Hogenbirk, A.; Oppe, J.

1998-09-01

OCTOPUS depletion calculations using the 3D Monte Carlo spectrum code MCNP (Monte Carlo Code for Neutron and Photon Transport) require much computing time. In a former implementation, the time required by OCTOPUS to perform multi-zone calculations, increased roughly proportional to the number of burnable zones. By using a different method the situation has improved considerably. In the new implementation described here, the dependence of the computing time on the number of zones has been moved from the MCNP code to a faster postprocessing code. By this, the overall computing time will reduce substantially. 11 refs

Efficient Windows Collaborative

Energy Technology Data Exchange (ETDEWEB)

Nils Petermann

2010-02-28

The project goals covered both the residential and commercial windows markets and involved a range of audiences such as window manufacturers, builders, homeowners, design professionals, utilities, and public agencies. Essential goals included: (1) Creation of 'Master Toolkits' of information that integrate diverse tools, rating systems, and incentive programs, customized for key audiences such as window manufacturers, design professionals, and utility programs. (2) Delivery of education and outreach programs to multiple audiences through conference presentations, publication of articles for builders and other industry professionals, and targeted dissemination of efficient window curricula to professionals and students. (3) Design and implementation of mechanisms to encourage and track sales of more efficient products through the existing Window Products Database as an incentive for manufacturers to improve products and participate in programs such as NFRC and ENERGY STAR. (4) Development of utility incentive programs to promote more efficient residential and commercial windows. Partnership with regional and local entities on the development of programs and customized information to move the market toward the highest performing products. An overarching project goal was to ensure that different audiences adopt and use the developed information, design and promotion tools and thus increase the market penetration of energy efficient fenestration products. In particular, a crucial success criterion was to move gas and electric utilities to increase the promotion of energy efficient windows through demand side management programs as an important step toward increasing the market share of energy efficient windows.

Enhancement of thermal neutron attenuation of nano-B{sub 4}C, -BN dispersed neutron shielding polymer nanocomposites

Energy Technology Data Exchange (ETDEWEB)

Kim, Jaewoo, E-mail: kimj@kaeri.re.kr [Nuclear Materials Research Division, Korea Atomic Energy Research Institute, 111-989 Daeduck-daero, Yuseong-gu, Daejeon-si 305-353 (Korea, Republic of); WCI Quantum Beam based Radiation Research Center, Korea Atomic Energy Research Institute, 111-989 Daeduck-daero, Yuseong-gu, Daejeon-si 305-353 (Korea, Republic of); Missouri University Research Reactor, University of Missouri-Columbia, Columbia, MO 65211 (United States); Lee, Byung-Chul [Nuclear Reactor Core Design Division, Korea Atomic Energy Research Institute, 111-989 Daeduck-daero, Yuseong-gu, Daejeon-si 305-353 (Korea, Republic of); Uhm, Young Rang [Radioisotopes Research Division, Korea Atomic Energy Research Institute, 111-989 Daeduck-daero, Yuseong-gu, Daejeon-si 305-353 (Korea, Republic of); Miller, William H. [Missouri University Research Reactor, University of Missouri-Columbia, Columbia, MO 65211 (United States)

2014-10-15

Highlights: • Preparation of B{sub 4}C and BN nanopowders using a simple ball milling process. • Homogeneous dispersion and strong adhesion of nano-B{sub 4}C and -BN with polymer matrix. • Enhancement of mechanical properties of the nanocomposites compared to their micro counterparts. • Enhancement of thermal neutron attenuation of the nanocomposites. - Abstract: Nano-sized boron carbide (B{sub 4}C) and boron nitride (BN) powder were prepared using ball milling. Micro- and milled nano-powders were melt blended with high density polyethylene (HDPE) using a polymer mixer followed by hot pressing to fabricate sheet composites. The tensile and flexural strengths of HDPE nanocomposites were ∼20% higher than their micro counterparts, while those for latter decreased compared to neat HDPE. Thermal neutrons attenuation of the prepared HDPE nanocomposites was evaluated using a monochromatic ∼0.025 eV neutron beam. Thermal neutron attenuation of the HDPE nanocomposites was greatly enhanced compared to their micro counterparts at the same B-10 areal densities. Monte Carlo n-Particles (MCNP) simulations based on the lattice structure modeling also shows the similar filler size dependent thermal neutron absorption.

Modification to the Monte Carlo N-Particle (MCNP) Visual Editor (MCNPVised) to Read in Computer Aided Design (CAD) Files

International Nuclear Information System (INIS)

Randolph Schwarz; Leland L. Carter; Alysia Schwarz

2005-01-01

Monte Carlo N-Particle Transport Code (MCNP) is the code of choice for doing complex neutron/photon/electron transport calculations for the nuclear industry and research institutions. The Visual Editor for Monte Carlo N-Particle is internationally recognized as the best code for visually creating and graphically displaying input files for MCNP. The work performed in this grant was used to enhance the capabilities of the MCNP Visual Editor to allow it to read in both 2D and 3D Computer Aided Design (CAD) files, allowing the user to electronically generate a valid MCNP input geometry

Modification to the Monte N-Particle (MCNP) Visual Editor (MCNPVised) to read in Computer Aided Design (CAD) files

International Nuclear Information System (INIS)

Schwarz, Randy A.; Carter, Leeland L.

2004-01-01

Monte Carlo N-Particle Transport Code (MCNP) (Reference 1) is the code of choice for doing complex neutron/photon/electron transport calculations for the nuclear industry and research institutions. The Visual Editor for Monte Carlo N-Particle (References 2 to 11) is recognized internationally as the best code for visually creating and graphically displaying input files for MCNP. The work performed in this grant enhanced the capabilities of the MCNP Visual Editor to allow it to read in a 2D Computer Aided Design (CAD) file, allowing the user to modify and view the 2D CAD file and then electronically generate a valid MCNP input geometry with a user specified axial extent

Hydration reactions in pastes C3S+C3A+CaSO4.2aq+H20 at 25°C.I

NARCIS (Netherlands)

Corstanje, W.A.; Stein, H.N.; Stevels, J.M.

1973-01-01

A characteristic retardation of the hydration of C3A is found in pastes C3S+C3A+CaSO4.2aq+H2O of weight ratios 1:3:z:4 at certain values of z, when sulphate concentration becomes insufficient for monosulphate formation. This retardation is ascribed to precipitation of amorphous Al(OH)3, when C3A

Nuclear reactor multi-physics simulations with coupled MCNP5 and STAR-CCM+

International Nuclear Information System (INIS)

Cardoni, Jeffrey Neil; Rizwan-uddin

2011-01-01

The MCNP5 Monte Carlo particle transport code has been coupled to the computational fluid dynamics code, STAR-CCM+, to provide a high fidelity multi-physics simulation tool for pressurized water nuclear reactors. The codes are executed separately and coupled externally through a Perl script. The Perl script automates the exchange of temperature, density, and volumetric heating information between the codes using ASCII text data files. Fortran90 and Java utility programs assist job automation with data post-processing and file management. The MCNP5 utility code, MAKXSF, pre-generates temperature dependent cross section libraries for the thermal feedback calculations. The MCNP5–STAR-CCM+ coupled simulation tool, dubbed MULTINUKE, was applied to a steady state, PWR cell model to demonstrate its usage and capabilities. The demonstration calculation showed reasonable results that agree with PWR values typically reported in literature. Temperature and fission reaction rate distributions were realistic and intuitive. Reactivity coefficients were also deemed reasonable in comparison to historically reported data. The demonstration problem consisted of 9,984 CFD cells and 7,489 neutronic cells. MCNP5 tallied fission energy deposition over 3,328 UO_2 cells. The coupled solution converged within eight hours and in three MULTINUKE iterations. The simulation was carried out on a 64 bit, quad core, Intel 2.8 GHz microprocessor with 1 GB RAM. The simulations on a quad core machine indicated that a massively parallelized implementation of MULTINUKE can be used to assess larger multi-million cell models. (author)

Tank 101-SY Window E core sample: Interpretation of results

International Nuclear Information System (INIS)

Reynolds, D.A.

1993-02-01

A full depth core sample was taken for tank 241-SY-101 in December 1991 during a time period called ''Window E.'' This was the second full depth core sample from this tank during the year. The core had two major portions that are known as the convective zone and the nonconvective zone. A crust was on the top of tank but as poorly sampled. The analysis of the Window E core sample stressed segment composite chemical analysis instead of segment by segment as in Window C. Adiabatic calorimetry on samples from both cores showed a slow self heating reaction above 150 degrees C on dried samples. The exothermic events were milder than similar synthetic samples. The chemical and physical properties complemented the information from Window C. The Window E material from the convective zone was more viscous than the Window C convective zone material. The nonconvective zone viscosities were similar for both cores. Heating and dilution tests were made to test mitigation concepts

Tank 101-SY Window E core sample: Interpretation of results

Energy Technology Data Exchange (ETDEWEB)

Reynolds, D.A.

1993-02-01

A full depth core sample was taken for tank 241-SY-101 in December 1991 during a time period called ``Window E.`` This was the second full depth core sample from this tank during the year. The core had two major portions that are known as the convective zone and the nonconvective zone. A crust was on the top of tank but as poorly sampled. The analysis of the Window E core sample stressed segment composite chemical analysis instead of segment by segment as in Window C. Adiabatic calorimetry on samples from both cores showed a slow self heating reaction above 150{degrees}C on dried samples. The exothermic events were milder than similar synthetic samples. The chemical and physical properties complemented the information from Window C. The Window E material from the convective zone was more viscous than the Window C convective zone material. The nonconvective zone viscosities were similar for both cores. Heating and dilution tests were made to test mitigation concepts.

Modeling of LVRF critical experiments in ZED-2 using WIMS9A/PANTHER and MCNP5

International Nuclear Information System (INIS)

Sissaoui, M.T.; Carlson, P.A.; Lebenhaft, J.R.

2009-01-01

The accuracy of WIMS9A/PANTHER and MCNP5 in modeling D 2 O-moderated, and H 2 O-, D 2 O- or air-cooled, doubly heterogeneous lattices of fuel clusters was demonstrated using Low Void Reactivity Fuel (LVRF) substitution experiments in the ZED-2 critical facility. MCNP5 with ENDF/B-VI (Release 5) underpredicted k eff but gave excellent coolant void reactivity (CVR) bias values. WIMS9A/PANTHER with JEF-2.2 overpredicted k eff and underpredicted the CVR bias relative to MCNP5 by 100-200 pcm. Both codes reproduced the measured axial and radial flux shapes accurately

A Monte Carlo burnup code linking MCNP and REBUS

International Nuclear Information System (INIS)

Hanan, N.A.; Olson, A.P.; Pond, R.B.; Matos, J.E.

1998-01-01

The REBUS-3 burnup code, used in the anl RERTR Program, is a very general code that uses diffusion theory (DIF3D) to obtain the fluxes required for reactor burnup analyses. Diffusion theory works well for most reactors. However, to include the effects of exact geometry and strong absorbers that are difficult to model using diffusion theory, a Monte Carlo method is required. MCNP, a general-purpose, generalized-geometry, time-dependent, Monte Carlo transport code, is the most widely used Monte Carlo code. This paper presents a linking of the MCNP code and the REBUS burnup code to perform these difficult analyses. The linked code will permit the use of the full capabilities of REBUS which include non-equilibrium and equilibrium burnup analyses. Results of burnup analyses using this new linked code are also presented. (author)

A Monte Carlo burnup code linking MCNP and REBUS

International Nuclear Information System (INIS)

Hanan, N. A.

1998-01-01

The REBUS-3 burnup code, used in the ANL RERTR Program, is a very general code that uses diffusion theory (DIF3D) to obtain the fluxes required for reactor burnup analyses. Diffusion theory works well for most reactors. However, to include the effects of exact geometry and strong absorbers that are difficult to model using diffusion theory, a Monte Carlo method is required. MCNP, a general-purpose, generalized-geometry, time-dependent, Monte Carlo transport code, is the most widely used Monte Carlo code. This paper presents a linking of the MCNP code and the REBUS burnup code to perform these difficult burnup analyses. The linked code will permit the use of the full capabilities of REBUS which include non-equilibrium and equilibrium burnup analyses. Results of burnup analyses using this new linked code are also presented

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

International Nuclear Information System (INIS)

Crawford, C.; Palmer, B.

1992-11-01

This report discusses critical experiments involving SPERT-D 1,2 fuel elements and BORAX-V 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 assemblies 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

An improved MCNP version of the NORMAN voxel phantom for dosimetry studies.

Science.gov (United States)

Ferrari, P; Gualdrini, G

2005-09-21

In recent years voxel phantoms have been developed on the basis of tomographic data of real individuals allowing new sets of conversion coefficients to be calculated for effective dose. Progress in radiation studies brought ICRP to revise its recommendations and a new report, already circulated in draft form, is expected to change the actual effective dose evaluation method. In the present paper the voxel phantom NORMAN developed at HPA, formerly NRPB, was employed with MCNP Monte Carlo code. A modified version of the phantom, NORMAN-05, was developed to take into account the new set of tissues and weighting factors proposed in the cited ICRP draft. Air kerma to organ equivalent dose and effective dose conversion coefficients for antero-posterior and postero-anterior parallel photon beam irradiations, from 20 keV to 10 MeV, have been calculated and compared with data obtained in other laboratories using different numerical phantoms. Obtained results are in good agreement with published data with some differences for the effective dose calculated employing the proposed new tissue weighting factors set in comparison with previous evaluations based on the ICRP 60 report.

Utilization of the MCNP-3A code for criticality safety analysis

International Nuclear Information System (INIS)

Maragni, M.G.; Moreira, J.M.L.

1996-01-01

In the last decade, Brazil started to operate facilities for processing and storing uranium in different forms. The necessity of criticality safety analysis appeared in the design phase of the uranium pilot process plants and also in the licensing of transportation and storage of fissile materials. The 2-MW research reactor and the Angra I power plant also required criticality safety assessments because their spent-fuel storage was approaching full-capacity utilization. The criticality safety analysis in Brazil has been based on KENO IV code calculations, which present some difficulties for correct geometry representation. The MCNP-3A code is not reported to be used frequently for criticality safety analysis in Brazil, but its good geometry representation makes it a possible tool for treating problems of complex geometry. A set of benchmark tests was performed to verify its applicability for criticality safety analysis in Brazil. This paper presents several benchmark tests aimed at selecting a set of options available in the MCNP-3A code that would be adequate for criticality safety analysis. The MCNP-3A code is also compared with the KENO-IV code regarding its performance for criticality safety analysis

Improved response function calculations for scintillation detectors using an extended version of the MCNP code

CERN Document Server

Schweda, K

2002-01-01

The analysis of (e,e'n) experiments at the Darmstadt superconducting electron linear accelerator S-DALINAC required the calculation of neutron response functions for the NE213 liquid scintillation detectors used. In an open geometry, these response functions can be obtained using the Monte Carlo codes NRESP7 and NEFF7. However, for more complex geometries, an extended version of the Monte Carlo code MCNP exists. This extended version of the MCNP code was improved upon by adding individual light-output functions for charged particles. In addition, more than one volume can be defined as a scintillator, thus allowing the simultaneous calculation of the response for multiple detector setups. With the implementation of sup 1 sup 2 C(n,n'3 alpha) reactions, all relevant reactions for neutron energies E sub n <20 MeV are now taken into consideration. The results of these calculations were compared to experimental data using monoenergetic neutrons in an open geometry and a sup 2 sup 5 sup 2 Cf neutron source in th...

Replacement Windows for Existing Homes Homes | Efficient Windows

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Collaborative Foundry Foundry New Construction Windows Window Selection Tool Selection Process Design Guidance Installation Replacement Windows Window Selection Tool Assessing Options Selection Process Design Guidance Installation Understanding Windows Benefits Design Considerations Measuring

Duplicating MC-15 Output with Python and MCNP

Energy Technology Data Exchange (ETDEWEB)

McSpaden, Alexander Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-08-23

Two Python scripts have been written that process the output files of MCNP6 into a format that mimics the list-mode output of Los Alamos National Laboratory’s MC-15 and NPOD neutron detection systems. This report details the methods implemented in these scripts and instructions on their use.

Evaluation of the WIMS (KAERI) - VENTURE code system for peak power prediction of KMRR core using MCNP

International Nuclear Information System (INIS)

Park, W.S.; Lee, K.M.; Lee, C.S.; Lee, J.T.; Oh, S.K.

1992-01-01

In this work, the validity and quantitative uncertainty of WIMS (KAERI) - VENTURE code system for the design and analysis of KMRR core was tried to be inferred using a well known benchmark code, MCNP. WIMS (KAERI) showed an excellent agreement with MCNP code. For three different control rod positions at a simulated core which has a quarter symmetry, total peaking factors and three sub-factors (radial, axial, and local) obtained from VENTURE were compared with those of MCNP. The comparison proved the validity of VENTURE and showed better agreement in the order of radial, axial, and local factors. The uncertainty of WIMS (KAERI) - VENTURE system was inferred using the 2σ band of total peaking obtained by MCNP. The uncertainty of WIMS (KAERI) - VENTURE system were found to be 18.5 % for the operating condition. (author)

Design Guidance for New Windows | Efficient Windows Collaborative

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Foundry Foundry New Construction Windows Window Selection Tool Selection Process Design Guidance Installation Replacement Windows Window Selection Tool Assessing Options Selection Process Design Guidance Installation Understanding Windows Benefits Design Considerations Measuring Performance Performance Standards

Selection Process for New Windows | Efficient Windows Collaborative

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Foundry Foundry New Construction Windows Window Selection Tool Selection Process Design Guidance Installation Replacement Windows Window Selection Tool Assessing Options Selection Process Design Guidance Installation Understanding Windows Benefits Design Considerations Measuring Performance Performance Standards

Selection Process for Replacement Windows | Efficient Windows Collaborative

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Foundry Foundry New Construction Windows Window Selection Tool Selection Process Design Guidance Installation Replacement Windows Window Selection Tool Assessing Options Selection Process Design Guidance Installation Understanding Windows Benefits Design Considerations Measuring Performance Performance Standards

Design Guidance for Replacement Windows | Efficient Windows Collaborative

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Foundry Foundry New Construction Windows Window Selection Tool Selection Process Design Guidance Installation Replacement Windows Window Selection Tool Assessing Options Selection Process Design Guidance Installation Understanding Windows Benefits Design Considerations Measuring Performance Performance Standards

Proper Installation of New Windows | Efficient Windows Collaborative

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Foundry Foundry New Construction Windows Window Selection Tool Selection Process Design Guidance Installation Replacement Windows Window Selection Tool Assessing Options Selection Process Design Guidance Installation Understanding Windows Benefits Design Considerations Measuring Performance Performance Standards

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

Science.gov (United States)

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.

Exponential smoothing weighted correlations

Science.gov (United States)

Pozzi, F.; Di Matteo, T.; Aste, T.

2012-06-01

In many practical applications, correlation matrices might be affected by the "curse of dimensionality" and by an excessive sensitiveness to outliers and remote observations. These shortcomings can cause problems of statistical robustness especially accentuated when a system of dynamic correlations over a running window is concerned. These drawbacks can be partially mitigated by assigning a structure of weights to observational events. In this paper, we discuss Pearson's ρ and Kendall's τ correlation matrices, weighted with an exponential smoothing, computed on moving windows using a data-set of daily returns for 300 NYSE highly capitalized companies in the period between 2001 and 2003. Criteria for jointly determining optimal weights together with the optimal length of the running window are proposed. We find that the exponential smoothing can provide more robust and reliable dynamic measures and we discuss that a careful choice of the parameters can reduce the autocorrelation of dynamic correlations whilst keeping significance and robustness of the measure. Weighted correlations are found to be smoother and recovering faster from market turbulence than their unweighted counterparts, helping also to discriminate more effectively genuine from spurious correlations.

Computerized tomography magnified bone windows are superior to standard soft tissue windows for accurate measurement of stone size: an in vitro and clinical study.

Science.gov (United States)

Eisner, Brian H; Kambadakone, Avinash; Monga, Manoj; Anderson, James K; Thoreson, Andrew A; Lee, Hang; Dretler, Stephen P; Sahani, Dushyant V

2009-04-01

We determined the most accurate method of measuring urinary stones on computerized tomography. For the in vitro portion of the study 24 calculi, including 12 calcium oxalate monohydrate and 12 uric acid stones, that had been previously collected at our clinic were measured manually with hand calipers as the gold standard measurement. The calculi were then embedded into human kidney-sized potatoes and scanned using 64-slice multidetector computerized tomography. Computerized tomography measurements were performed at 4 window settings, including standard soft tissue windows (window width-320 and window length-50), standard bone windows (window width-1120 and window length-300), 5.13x magnified soft tissue windows and 5.13x magnified bone windows. Maximum stone dimensions were recorded. For the in vivo portion of the study 41 patients with distal ureteral stones who underwent noncontrast computerized tomography and subsequently spontaneously passed the stones were analyzed. All analyzed stones were 100% calcium oxalate monohydrate or mixed, calcium based stones. Stones were prospectively collected at the clinic and the largest diameter was measured with digital calipers as the gold standard. This was compared to computerized tomography measurements using 4.0x magnified soft tissue windows and 4.0x magnified bone windows. Statistical comparisons were performed using Pearson's correlation and paired t test. In the in vitro portion of the study the most accurate measurements were obtained using 5.13x magnified bone windows with a mean 0.13 mm difference from caliper measurement (p = 0.6). Measurements performed in the soft tissue window with and without magnification, and in the bone window without magnification were significantly different from hand caliper measurements (mean difference 1.2, 1.9 and 1.4 mm, p = 0.003, window settings with magnification. For uric acid calculi the measurement error was observed only in standard soft tissue window settings. In vivo 4.0x

Whole core burnup calculations using `MCNP`

Energy Technology Data Exchange (ETDEWEB)

Haran, O; Shaham, Y [Israel Atomic Energy Commission, Beersheba (Israel). Nuclear Research Center-Negev

1996-12-01

Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors).

Whole core burnup calculations using 'MCNP'

International Nuclear Information System (INIS)

Haran, O.; Shaham, Y.

1996-01-01

Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors)

Modeling of LVRF Critical Experiments in ZED-2 Using WIMS9A/PANTHER and MCNP5

International Nuclear Information System (INIS)

Sissaoui, M.T.; Lebenhaft, J.R; Carlson, P.A.

2008-01-01

The accuracy of WIMS9A/PANTHER and MCNP5 in modeling D 2 O-moderated, and H 2 O-, D 2 O- or air-cooled, doubly heterogeneous lattices of fuel clusters was demonstrated using Low Void Reactivity Fuel (LVRF) substitution experiments in the ZED-2 critical facility. MCNP5 with ENDF/B-VI (Release 5) under-predicted k eff but gave excellent coolant void reactivity (CVR) bias values. WIMS9A/PANTHER with JEF-2.2 over-predicted k eff and under-predicted the CVR bias relative to MCNP5 by 100 pcm to 200 pcm. Both codes reproduced the measured axial and radial flux shapes accurately. (authors)

Fast reactor fuel reprocessing plant D1206: disassembly cave window 4 replacement

International Nuclear Information System (INIS)

Sutherland, H.G.; Beckitt, S.; Potts, A.B.

1996-01-01

At UKAEA's fast reactor reprocessing plant at Dounreay, the containment glass on the zinc bromide cave viewing window tank failed after 13 years active use. External shielding was fitted and the window tank subsequently drained to make it safe. Fuel cropping operations carried out behind the window were resited to enable cave work to continue whilst a project team made arrangements and plans to replace the damaged window. Because of the complexity of the task and high (alpha, beta, gamma and neutron) radiation levels in excess of 500 Sv/hr a rehearsal facility was built to develop the remote handling techniques to be employed in the task. (UK)

Comparison of thermal scattering processing options for S(α,β) cards in MCNP

International Nuclear Information System (INIS)

Čerba, Štefan; Damian, Jose Ignacio Marquez; Lüley, Jakub; Vrban, Branislav; Farkas, Gabriel; Nečas, Vladimír; Haščík, Jan

2013-01-01

Highlights: ► Determination of MCNP calculation bias for WWER-440. ► Specific scattering law S(α,β). ► Benchmark cases investigated. ► Three methods to process material cards for hydrogen bound in light water. - Abstract: The MCNP distributions include sets of pre-calculated thermal scattering libraries but these libraries are available for several temperature steps only. In order to achieve reliable results it is suitable to process the cross section libraries for the desired temperature. In general, there are three methods to process these thermal scattering libraries for the desired temperatures. This paper deals with the comparison of these three methods on the basis of several benchmarks and on the basis of a thermal transient experiment of a WWER-440 reactor. The choice is up to the MCNP user but unfortunately very few studies concerning the comparison have been published so far. Therefore conclusions and results presented in this paper may help the user to choose the most appropriate method for his calculation

Application of preprocessing filtering on Decision Tree C4.5 and rough set theory

Science.gov (United States)

Chan, Joseph C. C.; Lin, Tsau Y.

2001-03-01

This paper compares two artificial intelligence methods: the Decision Tree C4.5 and Rough Set Theory on the stock market data. The Decision Tree C4.5 is reviewed with the Rough Set Theory. An enhanced window application is developed to facilitate the pre-processing filtering by introducing the feature (attribute) transformations, which allows users to input formulas and create new attributes. Also, the application produces three varieties of data set with delaying, averaging, and summation. The results prove the improvement of pre-processing by applying feature (attribute) transformations on Decision Tree C4.5. Moreover, the comparison between Decision Tree C4.5 and Rough Set Theory is based on the clarity, automation, accuracy, dimensionality, raw data, and speed, which is supported by the rules sets generated by both algorithms on three different sets of data.

Design windows and cost analysis on helical reactors

International Nuclear Information System (INIS)