MCNP capabilities for nuclear well logging calculations
Forster, R.A.; Little, R.C.; Briesmeister, J.F.; Hendricks, J.S. . Applied Theoretical Physics Div.)
1990-06-01
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 ({und M}onte {und C}arlo {und n}eutron {und p}hoton), part of the LARTCS, provides a computational predictive capability for many applications of interest to the nuclear well logging community. The generalized three-dimensional geometry of MCNP is well suited for borehole-tool models. SABRINA, another component of the LARTCS, is a graphics code that can be used to interactively create a complex MCNP geometry. Users can define many source and 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.
MCNP (Monte Carlo Neutron Photon) capabilities for nuclear well logging calculations
Forster, R.A.; Little, R.C.; Briesmeister, J.F.
1989-01-01
The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. 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. A rich collections of variance reduction features can greatly increase the efficiency of a calculation. MCNP is written in FORTRAN 77 and has been run on variety of computer systems from scientific workstations to supercomputers. The next production version of MCNP will include features such as continuous-energy electron transport and a multitasking option. Areas of ongoing research of interest to the well logging community include angle biasing, adaptive Monte Carlo, improved discrete ordinates capabilities, and discrete ordinates/Monte Carlo hybrid development. Los Alamos has requested approval by the Department of Energy to create a Radiation Transport Computational Facility under their User Facility Program to increase external interactions with industry, universities, and other government organizations. 21 refs.
MCNP benchmark analyses of critical experiments for the Space Nuclear Thermal Propulsion program
Selcow, E.C.; Cerbone, R.J.; Ludewig, H.; Mughabghab, S.F.; Schmidt, E.; Todosow, M. ); Parma, E.J. ); Ball, R.M.; Hoovler, G.S. )
1993-01-15
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
Selcow, E.C.; Cerbone, R.J.; Ludewig, H.; Mughabghab, S.F.; Schmidt, E.; Todosow, M.; Parma, E.J.; Ball, R.M.; Hoovler, G.S.
1993-06-01
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 very good 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.
NASA Astrophysics Data System (ADS)
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.; Waters, L.; Wilcox, T.; Zukaitis, T.
2014-06-01
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.
MCNP Progress & Performance Improvements
Brown, Forrest B.; Bull, Jeffrey S.; Rising, Michael Evan
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.
Kiedrowski, B. C.; Brown, F. B.
2013-07-01
A continuous-energy sensitivity coefficient capability has been introduced into MCNP6. The methods for generating energy-resolved and energy-integrated sensitivity profiles are discussed. Results from the verification exercises that were performed are given, and these show that MCNP6 compares favorably with analytic solutions, direct density perturbations, and comparisons to TSUNAMI-3D and MONK. Run-time and memory requirements are assessed for typical applications, and these are shown to be reasonable with modern computing resources. (authors)
MCNP Super Lattice Method for VHTR ORIGEN2.2 Nuclear Library Improvement Based on ENDF/B-VII
G. S. Chang; J. R. Parry
2010-10-01
The advanced Very High Temperature gas-cooled Reactor (VHTR) achieves simplification of safety through reliance on innovative features and passive systems. One of the VHTRs innovative features is the reliance on ceramic-coated fuel particles to retain the fission products under extreme accident conditions. The effect of the random fuel kernel distribution in the fuel prismatic block creates a double-heterogeneous lattice, which needs to be addressed through the use of the newly developed prismatic super Kernel-by-Kernel Fuel (KbKF) lattice model method. Based on the new ENDF/B-VII nuclear cross section evaluated data, the developed KbKF super lattice model was then used with MCNP to calculate the material isotopes neutron reaction rates, such as, (n,?); (n,n’); (n,2n’); (n,f); (n,p); (n,?). Then, the MCNP-calculated results are rearranged to generate a set of new libraries “VHTRXS.lib,” for the ORIGEN2.2 isotopes depletion and build-up analysis code. The libraries contain one group cross section data for the structural light elements, actinides, and fission products that can be applied in the VHTR related fuel burnup and material transmutation analysis codes. The efficiency and ease of use of the MCNP method to generate and update the ORIGEN2.2 one-group spectrum weighed cross section library for VHTR was demonstrated.
Verification of MCNP5-1.60 and MCNP6-Beta-2 for Criticality Safety Applications
Brown, Forrest B.; Kiedrowski, Brian C.; Bull, Jeffrey S.
2012-05-01
To verify that both MCNP5-1.60 and MCNP6-Beta-2 are performing correctly for criticality safety applications, several suites of verification/validation benchmark problems were run in early 2012. Results from these benchmark suites were compared with results from previously verified versions of MCNP5. The goals of this verification testing were: (1) Verify that MCNP5-1.60 works correctly for nuclear criticality safety applications, producing the same results as for the previous verification performed in 2010; (2) Determine the sensitivity to computer roundoff using different Fortran-90 compilers for building MCNP5 and MCNP6, to support moving to current versions of the compilers; and (3) Verify that MCNP6-Beta-2 works correctly for nuclear criticality safety applications, producing the same results as for MCNP5-1.60. This provides support for eventual migration of users and applications to MCNP6. The current production version of MCNP5 included in the RSICC release package is MCNP5-1.60. This version was first distributed by RSICC in October 2010. While there were subsequent RSICC distributions of the MCNP package in July 2011 and February 2012, no changes were made to MCNP5-1.60. The RSICC release package in February 2012 included both MCNP5-1.60 and the current beta version of MCNP6, MCNP6-Beta-2. MCNP6 is the merger of MCNP5 and MCNPX capabilities. The current release of MCNP6 available from RSICC as of February 2012 is MCNP6-Beta-2. This version includes all of the features for criticality safety calculations that are available in MCNP5-1.60, and many new features largely unrelated to nuclear criticality safety calculations. This release is a 'beta' release to allow intermediate and advanced users to begin testing the merged code in their field of expertise. It should not be used for production calculations.
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 {sup 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.
MCNP Perturbation Capability for Monte Carlo Criticality Calculations
Hendricks, J.S.; Carter, L.L.; McKinney, G.W.
1999-09-20
The differential operator perturbation capability in MCNP4B has been extended to automatically calculate perturbation estimates for the track length estimate of k{sub 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.
Nuclear heat load calculations for the NBSR cold neutron source using MCNP
Blau, M. ); Rowe, J.M.; Williams, R.E. )
1993-01-01
A liquid-hydrogen (LH2) cold neutron source is being designed for installation in the 20-MW National Bureau of Standards reactor (NBSR) at National Institute of Standards and Technology to replace the D[sub 2]O-ice cold source currently in use. An accurate estimate of the heat deposited in the cold source is needed to ensure that it can be adequately cooled for successful operation. The heat load is caused by the interaction of neutrons and gamma rays with the cold moderator and the walls of the moderator chamber. The Monte Carlo code, MCNP (Version 4.2), was used to model the entire NBSR core and both the existing and the proposed cold sources. The model was used to calculate not only the heat load but also the reactivity and neutron gain of each source.
Goorley, John T.
2012-06-25
We, the development teams for MCNP, NJOY, and parts of ENDF, would like to invite you to a proposed 3 day workshop October 30, 31 and November 1 2012, to be held at Los Alamos National Laboratory. At this workshop, we will review new and developing missions that MCNP6 and the underlying nuclear data are being asked to address. LANL will also present its internal plans to address these missions and recent advances in these three capabilities and we will be interested to hear your input on these topics. Additionally we are interested in hearing from you additional technical advances, 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.
MCNP benchmark analyses of critical experiments for space nuclear thermal propulsion
Selcow, E.C.; Cerbone, R.J.; Ludewig, H. )
1993-01-01
The particle-bed reactor (PBR) system is being developed for use in the Space Nuclear Thermal Propulsion (SNTP) Program. This reactor system is characterized by a highly heterogeneous, compact configuration with many streaming pathways. The neutronics analyses performed for this system must be able to accurately predict reactor criticality, kinetics parameters, material worths at various temperatures, feedback coefficients, and detailed fission power and heating distributions. The latter includes coupled axial, radial, and azimuthal profiles. These responses constitute critical inputs and interfaces with the thermal-hydraulics design and safety analyses of the system.
William Martin
2012-11-16
A new method to obtain Doppler broadened cross sections has been implemented into MCNP, removing the need to generate cross sections for isotopes at problem temperatures. Previous work had established the scientific feasibility of obtaining Doppler-broadened cross sections "on-the-fly" (OTF) during the random walk of the neutron. Thus, when a neutron of energy E enters a material region that is at some temperature T, the cross sections for that material at the exact temperature T are immediately obtained by interpolation using a high order functional expansion for the temperature dependence of the Doppler-broadened cross section for that isotope at the neutron energy E. A standalone Fortran code has been developed that generates the OTF library for any isotope that can be processed by NJOY. The OTF cross sections agree with the NJOY-based cross sections for all neutron energies and all temperatures in the range specified by the user, e.g., 250K - 3200K. The OTF methodology has been successfully implemented into the MCNP Monte Carlo code and has been tested on several test problems by comparing MCNP with conventional ACE cross sections versus MCNP with OTF cross sections. The test problems include the Doppler defect reactivity benchmark suite and two full-core VHTR configurations, including one with multiphysics coupling using RELAP5-3D/ATHENA for the thermal-hydraulic analysis. The comparison has been excellent, verifying that the OTF libraries can be used in place of the conventional ACE libraries generated at problem temperatures. In addition, it has been found that using OTF cross sections greatly reduces the complexity of the input for MCNP, especially for full-core temperature feedback calculations with many temperature regions. This results in an order of magnitude decrease in the number of input lines for full-core configurations, thus simplifying input preparation and reducing the potential for input errors. Finally, for full-core problems with multiphysics
MCNP: Photon benchmark problems
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.
NASA Astrophysics Data System (ADS)
Pauzi, A. M.
2013-06-01
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-batteryTM", 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.
Adjoint-Based Uncertainty Quantification with MCNP
Seifried, Jeffrey E.
2011-09-01
This work serves to quantify the instantaneous uncertainties in neutron transport simulations born from nuclear data and statistical counting uncertainties. Perturbation and adjoint theories are used to derive implicit sensitivity expressions. These expressions are transformed into forms that are convenient for construction with MCNP6, creating the ability to perform adjoint-based uncertainty quantification with MCNP6. These new tools are exercised on the depleted-uranium hybrid LIFE blanket, quantifying its sensitivities and uncertainties to important figures of merit. Overall, these uncertainty estimates are small (< 2%). Having quantified the sensitivities and uncertainties, physical understanding of the system is gained and some confidence in the simulation is acquired.
Possible Improvements to MCNP6 and its CEM/LAQGSM Event-Generators
Mashnik, Stepan Georgievich
2015-08-04
This report is intended to the MCNP6 developers and sponsors of MCNP6. It presents a set of suggested possible future improvements to MCNP6 and to its CEM03.03 and LAQGSM03.03 event-generators. A few suggested modifications of MCNP6 are quite simple, aimed at avoiding possible problems with running MCNP6 on various computers, i.e., these changes are not expected to change or improve any results, but should make the use of MCNP6 easier; such changes are expected to require limited man-power resources. On the other hand, several other suggested improvements require a serious further development of nuclear reaction models, are expected to 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.
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 (trademark) ENDF/B-VI iron benchmark calculations
NASA Astrophysics Data System (ADS)
Court, J. D.; Hendricks, J. S.
Four iron shielding benchmarks have been calculated for, we believe the first time, with MCNP4A and its new ENDF/B-VI library. These calculations are part of the Hiroshima/Nagasaki dose re-evaluation for the National Academy of Sciences and the Defense Nuclear Agency. We believe these calculations are significant because they validate MCNP and the new ENDF/B-VI libraries. These calculations are compared to ENDF/B-V, experiment, and in some cases the recommended MCNP data library (a T-2 evaluation) and ENDF/IV.
Mosteller, R. 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.
MCNP: Multigroup/adjoint capabilities
Wagner, J.C.; Redmond, E.L. II; Palmtag, S.P.; Hendricks, J.S.
1994-04-01
This report discusses various aspects related to the use and validity of the general purpose Monte Carlo code MCNP for multigroup/adjoint calculations. The increased desire to perform comparisons between Monte Carlo and deterministic codes, along with the ever-present desire to increase the efficiency of large MCNP calculations has produced a greater user demand for the multigroup/adjoint capabilities. To more fully utilize these capabilities, we review the applications of the Monte Carlo multigroup/adjoint method, describe how to generate multigroup cross sections for MCNP with the auxiliary CRSRD code, describe how to use the multigroup/adjoint capability in MCNP, and provide examples and results indicating the effectiveness and validity of the MCNP multigroup/adjoint treatment. This information should assist users in taking advantage of the MCNP multigroup/adjoint capabilities.
Stevens, J. G.; Nuclear Engineering Division
2009-04-24
The Reduced Enrichment Research and Test Reactor (RERTR) Program uses the REBUS-PC computer code to provide reactor physics and core design information such as neutron flux distributions in space, energy, and time, and to track isotopic changes in fuel and neutron absorbers with burnup. REBUS-PC models the complete fuel cycle including shuffling capability. REBUS-PC evolved using the neutronic capabilities of multi-group diffusion theory code DIF3D 9.0, but was extended to apply the continuous energy Monte Carlo code MCNP for one-group fluxes and cross-sections. The linkage between REBUS-PC and MCNP has recently been modernized and extended, as described in this manual. REBUS-PC now calls MCNP via a system call so that the user can apply any valid MCNP executable. The interface between REBUS-PC and MCNP requires minimal changes to an existing MCNP model, and little additional input. The REBUS-MCNP interface can also be used in conjunction with DIF3D neutronics to update an MCNP model with fuel compositions predicted using a DIF3D based depletion.
Fischer, Noah A.
2012-08-14
The reactor core input generator allows for MCNP input files to be tailored to design specifications and generated in seconds. Full reactor models can now easily be created by specifying a small set of parameters and generating an MCNP input for a full reactor core. Axial zoning of the core will allow for density variation in the fuel and moderator, with pin-by-pin fidelity, so that BWR cores can more accurately be modeled. LWR core work in progress: (1) Reflectivity option for specifying 1/4, 1/2, or full core simulation; (2) Axial zoning for moderator densities that vary with height; (3) Generating multiple types of assemblies for different fuel enrichments; and (4) Parameters for specifying BWR box walls. Fuel pin work in progress: (1) Radial and azimuthal zoning for generating further unique materials in fuel rods; (2) Options for specifying different types of fuel for MOX or multiple burn assemblies; (3) Additional options for replacing fuel rods with burnable poison rods; and (4) Control rod/blade modeling.
Criticality calculations with MCNP{trademark}: A primer
Harmon, C.D. II; Busch, R.D.; Briesmeister, J.F.; Forster, R.A.
1994-06-06
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.
Energy Science and Technology Software Center (ESTSC)
2014-09-01
Version 01 MCNP6 is a general-purpose, continuous-energy, generalized-geometry, time-dependent, Monte Carlo radiation-transport code designed to track many particle types over broad ranges of energies. This MCNP6.1.1Beta is a follow-on to the MCNP6.1 production version which itself was the culmination of a multi-year effort to merge the MCNP5 [X-503] and MCNPX [PEL11] codes into a single product. This MCNP6.1.1 beta has been released in order to provide the radiation transport community with the latest feature developmentsmore » and bug fixes in the code. MCNP6.1.1 has taken input from a group of people, residing in the Los Alamos National Laboratory's (LANL) X Computational Physics Division, Radiation Transport Group (XCP-3), and Nuclear Engineering and Nonproliferation Division, Systems Design and Analysis Group (NEN-5). They have combined their code development efforts to produce this next evolution of MCNP. For those familiar with previous versions of MCNP, you will discover the code has been expanded to handle a multitude of particles and to include model physics options for energies above the cross-section table range, a material burnup feature, and delayed particle production. Expanded and/or new tally, source, and variance-reduction options are available to the user as well as an improved plotting capability. The capability to calculate keff eigenvalues for fissile systems remains a standard feature. Although MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, the result is much more than the sum of these two computer codes. MCNP6 is the result of five years of effort by the MCNP5 and MCNPX code development teams.« less
2014-09-01
Version 01 MCNP6 is a general-purpose, continuous-energy, generalized-geometry, time-dependent, Monte Carlo radiation-transport code designed to track many particle types over broad ranges of energies. This MCNP6.1.1Beta is a follow-on to the MCNP6.1 production version which itself was the culmination of a multi-year effort to merge the MCNP5 [X-503] and MCNPX [PEL11] codes into a single product. This MCNP6.1.1 beta has been released in order to provide the radiation transport community with the latest feature developments and bug fixes in the code. MCNP6.1.1 has taken input from a group of people, residing in the Los Alamos National Laboratory's (LANL) X Computational Physics Division, Radiation Transport Group (XCP-3), and Nuclear Engineering and Nonproliferation Division, Systems Design and Analysis Group (NEN-5). They have combined their code development efforts to produce this next evolution of MCNP. For those familiar with previous versions of MCNP, you will discover the code has been expanded to handle a multitude of particles and to include model physics options for energies above the cross-section table range, a material burnup feature, and delayed particle production. Expanded and/or new tally, source, and variance-reduction options are available to the user as well as an improved plotting capability. The capability to calculate keff eigenvalues for fissile systems remains a standard feature. Although MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, the result is much more than the sum of these two computer codes. MCNP6 is the result of five years of effort by the MCNP5 and MCNPX code development teams.
MCNP6. Simulating Correlated Data in Fission Events
Rising, Michael Evan; Sood, Avneet
2015-12-03
This report is a series of slides discussing the MCNP6 code and its status in simulating fission. Applications of interest include global security and nuclear nonproliferation, detection of special nuclear material (SNM), passive and active interrogation techniques, and coincident neutron and photon leakage.
Criticality calculations with MCNP{sup TM}: A primer
Mendius, P.W.; Harmon, C.D. II; Busch, R.D.; Briesmeister, J.F.; Forster, R.A.
1994-08-01
The purpose of this Primer is to assist the nuclear criticality safety analyst to perform computer calculations using the Monte Carlo code MCNP. Because of the closure of many experimental facilities, reliance on computer simulation is increasing. Often the analyst has little experience with specific codes available at his/her facility. This Primer helps the analyst understand and use the MCNP Monte Carlo code for nuclear criticality analyses. It assumes no knowledge of or particular experience with Monte Carlo codes in general or with MCNP in particular. The document begins with a Quickstart chapter that introduces the basic concepts of using MCNP. The following chapters expand on those ideas, presenting a range of problems from simple cylinders to 3-dimensional lattices for calculating keff confidence intervals. Input files and results for all problems are included. The Primer can be used alone, but its best use is in conjunction with the MCNP4A manual. After completing the Primer, a criticality analyst should be capable of performing and understanding a majority of the calculations that will arise in the field of nuclear criticality safety.
A new MCNP{trademark} test set
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.
Application of MCNP{trademark} to storage facility dose rate assessment
Urban, W.T.; Roberts, R.R.; Estes, G.P.; Taylor, W.M.
1996-12-31
The MCNP code is widely used in the determination of neutral particle dose rate analyses. In this paper we examine the application of MCNP to several storage facilities containing special nuclear material, SNM, wherein the neutron dose rate is the primary quantity of interest. In particular, we describe the special geometry, modeling assumptions, and physics considerations encountered in each of three applications.
Potential MCNP enhancements for NCT
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.
Potential MCNP enhancements for NCT
Estes, G.P.; Taylor, W.M.
1992-12-01
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.
Performance of scientific computing platforms running MCNP4B
McLaughlin, H.E.; Hendricks, J.S.
1997-11-01
A performance study has been made for the MCNP4B Monte Carlo radiation transport code on a wide variety of scientific computing platforms ranging from personal computers to Cray mainframes. We present the timing study results using MCNP4B and its new test set and libraries. This timing study is unlike other timing studies because of its widespread reproducibility, its direct comparability to the predecessor study in 1993, and its focus upon a nuclear engineering code. Our results, derived from using the new 29-problem test set for MCNP4B, (1) use a highly versatile and readily available physics code; (2) show that timing studies are very problem dependent; (3) present the results as raw data allowing comparisons of performance to other computing platforms not included in this study to those platforms that were included; (4) are reproducible; and (5) provide a measure of improvement in performance with the MCNP code due to advancements in software and hardware over the past 4 years. In the 1993 predecessor study using MCNP4A, the performances were based on a 25 problem test set. We present our data based on MCNP4B`s new 29 problem test set which cover 97% of all the FORTRAN physics code lines in MCNP4B. Like the previous study the new test problems and the test data library are available from the Radiation Shielding and Information Computational Center (RSICC) in Oak Ridge, Tennessee. Our results are reproducible because anyone with the same workstation, compiler, and operating system can duplicate the results presented here. The computing platforms included in this study are the Sun Sparc2, Sun Sparc5, Cray YMP 8/128, HP C180,SGI origin 2000, DBC 3000/600, DBC AiphaStation 500(300 MHz), IBM RS/6000-590, HP /9000-735, Micron Milienia Pro 200 MHz PC, and the Cray T94/128.
Cox, L. J.; Barrett, R. F.; Booth, Thomas Edward; Briesmeister, Judith F.; Brown, F. B.; Bull, J. S.; Giesler, G. C.; Goorley, J. T.; Mosteller, R. D.; Forster, R. A.; Post, S. E.; Prael, R. E.; Selcow, Elizabeth Carol,; Sood, A.
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).
Depletion analysis of the UMLRR reactor core using MCNP6
NASA Astrophysics Data System (ADS)
Odera, Dim Udochukwu
Accurate knowledge of the neutron flux and temporal nuclide inventory in reactor physics calculations is necessary for a variety of application in nuclear engineering such as criticality safety, safeguards, and spent fuel storage. The Monte Carlo N- Particle (MCNP6) code with integrated buildup depletion code (CINDER90) provides a high-fidelity tool that can be used to perform 3D, full core simulation to evaluate fissile material utilization, and nuclide inventory calculations as a function of burnup. The University of Massachusetts Lowell Research Reactor (UMLRR) reactor has been modeled with the deterministic based code, VENTURE and with an older version of MCNP (MCNP5). The MIT developed MCODE (MCNP ORIGEN DEPLETION CODE) was used previously to perform some limited depletion calculations. This work chronicles the use of MCNP6, released in June 2013, to perform coupled neutronics and depletion calculation. The results are compared to previously benchmarked results. Furthermore, the code is used to determine the ratio of fission products 134Cs and 137Cs (burnup indicators), and the resultant ratio is compared to the burnup of the UMLRR.
Hot Cell Window Shielding Analysis Using MCNP
Chad L. Pope; Wade W. Scates; J. Todd Taylor
2009-05-01
The Idaho National Laboratory Materials and Fuels Complex nuclear facilities are undergoing a documented safety analysis upgrade. In conjunction with the upgrade effort, shielding analysis of the Fuel Conditioning Facility (FCF) hot cell windows has been conducted. This paper describes the shielding analysis methodology. Each 4-ft thick window uses nine glass slabs, an oil film between the slabs, numerous steel plates, and packed lead wool. Operations in the hot cell center on used nuclear fuel (UNF) processing. Prior to the shielding analysis, shield testing with a gamma ray source was conducted, and the windows were found to be very effective gamma shields. Despite these results, because the glass contained significant amounts of lead and little neutron absorbing material, some doubt lingered regarding the effectiveness of the windows in neutron shielding situations, such as during an accidental criticality. MCNP was selected as an analysis tool because it could model complicated geometry, and it could track gamma and neutron radiation. A bounding criticality source was developed based on the composition of the UNF. Additionally, a bounding gamma source was developed based on the fission product content of the UNF. Modeling the windows required field inspections and detailed examination of drawings and material specifications. Consistent with the shield testing results, MCNP results demonstrated that the shielding was very effective with respect to gamma radiation, and in addition, the analysis demonstrated that the shielding was also very effective during an accidental criticality.
Benchmarking MCNP and TRIPOLI with PGNAA measurements
NASA Astrophysics Data System (ADS)
Carasco, C.; Perot, B.; Sikora, A.; Mauerhofer, E.; Havenith, A.; Payan, E.; Kettler, J.; Kring, T.; Ma, J. L.
2014-06-01
The French Alternative Energies and Atomic Energy Commission (CEA Cadarache), the Forschungszentrum Jülich GmbH (FZJ), and the RWTH Aachen University (RWTH) are involved in a cooperation aiming at characterizing toxic and reactive elements in radioactive waste packages by means of Prompt Gamma Neutron Activation Analysis (PGNAA). The design of an optimized measurement system and the assessment of its performances for realistic scenarios can be conveniently studied by numerical Monte Carlo simulation, provided the model and nuclear data offer a sufficient precision. Previous studies performed with MCNP have shown that when the nuclear data libraries lack of precision, relevant results can still be obtained by performing calculations in multiple steps (by first determining the radiative capture rate, and transporting the induced gamma toward the detector) and by injecting valid gamma-ray production data in-between [1]. In such cases, it is interesting to compare the results obtained with different codes. In the present paper, we propose to compare the MCNP and TRIPOLI codes with measurements obtained in MEDINA (Multi Element Detection based on Instrumental Neutron Activation), which is the new FZJ PGNAA facility [2]. The aim of the measurement campaign is to assess capture gamma rays of toxic elements that can be found in 200 L waste drums which are expected for geological repository.
SUPERIMPOSED MESH PLOTTING IN MCNP
J. HENDRICKS
2001-02-01
The capability to plot superimposed meshes has been added to MCNP{trademark}. MCNP4C featured a superimposed mesh weight window generator which enabled users to set up geometries without having to subdivide geometric cells for variance reduction. The variance reduction was performed with weight windows on a rectangular or cylindrical mesh superimposed over the physical geometry. Experience with the new capability was favorable but also indicated that a number of enhancements would be very beneficial, particularly a means of visualizing the mesh and its values. The mathematics for plotting the mesh and its values is described here along with a description of other upgrades.
Valentine, T.E.
2001-01-19
The Monte Carlo code MCNP-DSP was developed from the Los Alamos MCNP4a code to calculate the time and frequency response statistics obtained from subcritical measurements. The code can be used to simulate a variety of subcritical measurements including source-driven noise analysis, Rossi-{alpha}, pulsed source, passive frequency analysis, multiplicity, and Feynman variance measurements. This code can be used to validate Monte Carlo methods and cross section data sets with subcritical measurements and replaces the use of point kinetics models for interpreting subcritical measurements.
MCNP{sup TM} criticality primer and training experiences
Briesmeister, J.; Forster, R.A.; Busch, R.
1995-09-01
With the closure of many experimental facilities, the nuclear criticality safety analyst is increasingly required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, the analyst may have little experience with the specific codes available at his or her facility. Usually, the codes are quite complex, black boxes capable of analyzing numerous problems with a myriad of input options. Documentation for these codes is designed to cover all the possible configurations and types of analyses but does not give much detail on any particular type of analysis. For criticality calculations, the user of a code is primarily interested in the value of the effective multiplication factor for a system (k{sub eff}). Most codes will provide this, and truckloads of other information that may be less pertinent to criticality calculations. Based on discussions with code users in the nuclear criticality safety community, it was decided that a simple document discussing the ins and outs of criticality calculations with specific codes would be quite useful. The Transport Methods Group, XTM, at Los Alamos National Laboratory (LANL) decided to develop a primer for criticality calculations with their Monte Carlo code, MCNP. This was a joint task between LANL with a knowledge and understanding of the nuances and capabilities of MCNP and the University of New Mexico with a knowledge and understanding of nuclear criticality safety calculations and educating first time users of neutronics calculations. The initial problem was that the MCNP manual just contained too much information. Almost everything one needs to know about MCNP can be found in the manual; the problem is that there is more information than a user requires to do a simple k{sub eff} calculation. The basic concept of the primer was to distill the manual to create a document whose only focus was criticality calculations using MCNP.
Randolph Schwarz; Leland L. Carter; Alysia Schwarz
2005-08-23
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.
MCNP(TM) Release 6.1.1 beta: Creating and Testing the Code Distribution
Cox, Lawrence J.; Casswell, Laura
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.
GOORLEY, TIM
2013-07-16
Version 00 US DOE 10CFR810 Jurisdiction. MCNP6 is a general-purpose, continuous-energy, generalized-geometry, time-dependent, Monte Carlo radiation-transport code designed to track many particle types over broad ranges of energies. MCNP6 represents the culmination of a multi-year effort to merge the MCNP5 [X-503] and MCNPX [PEL11] codes into a single product comprising all features of both. For those familiar with previous versions of MCNP, you will discover the code has been expanded to handle a multitude of particles and to include model physics options for energies above the cross-section table range, a material burnup feature, and delayed particle production. Expanded and/or new tally, source, and variance-reduction options are available to the user as well as an improved plotting capability. The capability to calculate keff eigenvalues for fissile systems remains a standard feature. Although MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, the result is much more than the sum of these two computer codes. MCNP6 is the result of five years of effort by the MCNP5 and MCNPX code development teams. These groups of people, residing in the Los Alamos National Laboratory's (LANL) X Computational Physics Division, Monte Carlo Codes Group (XCP-3), and Nuclear Engineering and Nonproliferation Division, Systems Design and Analysis Group (NEN-5, formerly D-5), have combined their code development efforts to produce the next evolution of MCNP. While maintenance and bug fixes will continue for MCNP5 v.1.60 and MCNPX v.2.7.0 for upcoming years, new code development capabilities will be developed and released only in MCNP6. In fact, this initial production release of MCNP6 (v. 1.0) contains 16 new features not previously found in either code. These new features include (among others) the abilities to import unstructured mesh geometries from the finite element code Abaqus, to transport photons down to 1.0 eV, to model complete atomic
GOORLEY, TIM
2013-07-16
Version 01 US DOE 10CFR810 Jurisdiction. MCNP6 is a general-purpose, continuous-energy, generalized-geometry, time-dependent, Monte Carlo radiation-transport code designed to track many particle types over broad ranges of energies. MCNP6 represents the culmination of a multi-year effort to merge the MCNP5 [X-503] and MCNPX [PEL11] codes into a single product comprising all features of both. For those familiar with previous versions of MCNP, you will discover the code has been expanded to handle a multitude of particles and to include model physics options for energies above the cross-section table range, a material burnup feature, and delayed particle production. Expanded and/or new tally, source, and variance-reduction options are available to the user as well as an improved plotting capability. The capability to calculate keff eigenvalues for fissile systems remains a standard feature. Although MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, the result is much more than the sum of these two computer codes. MCNP6 is the result of five years of effort by the MCNP5 and MCNPX code development teams. These groups of people, residing in the Los Alamos National Laboratory's (LANL) X Computational Physics Division, Monte Carlo Codes Group (XCP-3), and Nuclear Engineering and Nonproliferation Division, Systems Design and Analysis Group (NEN-5, formerly D-5), have combined their code development efforts to produce the next evolution of MCNP. While maintenance and bug fixes will continue for MCNP5 v.1.60 and MCNPX v.2.7.0 for upcoming years, new code development capabilities will be developed and released only in MCNP6. In fact, this initial production release of MCNP6 (v. 1.0) contains 16 new features not previously found in either code. These new features include (among others) the abilities to import unstructured mesh geometries from the finite element code Abaqus, to transport photons down to 1.0 eV, to model complete atomic
Energy Science and Technology Software Center (ESTSC)
2013-07-16
Version 01 US DOE 10CFR810 Jurisdiction. MCNP6 is a general-purpose, continuous-energy, generalized-geometry, time-dependent, Monte Carlo radiation-transport code designed to track many particle types over broad ranges of energies. MCNP6 represents the culmination of a multi-year effort to merge the MCNP5 [X-503] and MCNPX [PEL11] codes into a single product comprising all features of both. For those familiar with previous versions of MCNP, you will discover the code has been expanded to handle a multitude ofmore » particles and to include model physics options for energies above the cross-section table range, a material burnup feature, and delayed particle production. Expanded and/or new tally, source, and variance-reduction options are available to the user as well as an improved plotting capability. The capability to calculate keff eigenvalues for fissile systems remains a standard feature. Although MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, the result is much more than the sum of these two computer codes. MCNP6 is the result of five years of effort by the MCNP5 and MCNPX code development teams. These groups of people, residing in the Los Alamos National Laboratory's (LANL) X Computational Physics Division, Monte Carlo Codes Group (XCP-3), and Nuclear Engineering and Nonproliferation Division, Systems Design and Analysis Group (NEN-5, formerly D-5), have combined their code development efforts to produce the next evolution of MCNP. While maintenance and bug fixes will continue for MCNP5 v.1.60 and MCNPX v.2.7.0 for upcoming years, new code development capabilities will be developed and released only in MCNP6. In fact, this initial production release of MCNP6 (v. 1.0) contains 16 new features not previously found in either code. These new features include (among others) the abilities to import unstructured mesh geometries from the finite element code Abaqus, to transport photons down to 1.0 eV, to model
Energy Science and Technology Software Center (ESTSC)
2013-07-16
Version 00 US DOE 10CFR810 Jurisdiction. MCNP6 is a general-purpose, continuous-energy, generalized-geometry, time-dependent, Monte Carlo radiation-transport code designed to track many particle types over broad ranges of energies. MCNP6 represents the culmination of a multi-year effort to merge the MCNP5 [X-503] and MCNPX [PEL11] codes into a single product comprising all features of both. For those familiar with previous versions of MCNP, you will discover the code has been expanded to handle a multitude ofmore » particles and to include model physics options for energies above the cross-section table range, a material burnup feature, and delayed particle production. Expanded and/or new tally, source, and variance-reduction options are available to the user as well as an improved plotting capability. The capability to calculate keff eigenvalues for fissile systems remains a standard feature. Although MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, the result is much more than the sum of these two computer codes. MCNP6 is the result of five years of effort by the MCNP5 and MCNPX code development teams. These groups of people, residing in the Los Alamos National Laboratory's (LANL) X Computational Physics Division, Monte Carlo Codes Group (XCP-3), and Nuclear Engineering and Nonproliferation Division, Systems Design and Analysis Group (NEN-5, formerly D-5), have combined their code development efforts to produce the next evolution of MCNP. While maintenance and bug fixes will continue for MCNP5 v.1.60 and MCNPX v.2.7.0 for upcoming years, new code development capabilities will be developed and released only in MCNP6. In fact, this initial production release of MCNP6 (v. 1.0) contains 16 new features not previously found in either code. These new features include (among others) the abilities to import unstructured mesh geometries from the finite element code Abaqus, to transport photons down to 1.0 eV, to model
McKinney, Gregg W; Armstrong, Hirotatsu; James, Michael R; Clem, John; Goldhagen, Paul
2012-06-19
MCNP is a Monte Carlo radiation transport code that has been under development for over half a century. Over the last decade, the development team of a high-energy offshoot of MCNP, called MCNPX, has implemented several physics and algorithm improvements important for modeling galactic cosmic-ray (GCR) interactions with matter. In this presentation, we discuss the latest of these improvements, a new Cosmic-Source option, that has been implemented in MCNP6.
MCNP6 Fission Multiplicity with FMULT Card
Wilcox, Trevor; Fensin, Michael Lorne; Hendricks, John S.; James, Michael R.; McKinney, Gregg W.
2012-06-18
With the merger of MCNPX and MCNP5 into MCNP6, MCNP6 now provides all the capabilities of both codes allowing the user to access all the fission multiplicity data sets. Detailed in this paper is: (1) the new FMULT card capabilities for accessing these different data sets; (2) benchmark calculations, as compared to experiment, detailing the results of selecting these separate data sets for thermal neutron induced fission on U-235.
Performance upgrades to the MCNP6 burnup capability for large scale depletion calculations
Fensin, M. L.; Galloway, J. D.; James, M. R.
2015-04-11
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. With the merger of MCNPX and MCNP5, MCNP6 combined the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. The new MCNP6 depletion capability was first showcased at the International Congress for Advancements in Nuclear Power Plants (ICAPP) meeting in 2012. At that conference the new capabilities addressed included the combined distributive and shared memory parallel architecture for the burnup capability, improved memory management, physics enhancements, and newmore » predictability as compared to the H.B Robinson Benchmark. At Los Alamos National Laboratory, a special purpose cluster named “tebow,” was constructed such to maximize available RAM per CPU, as well as leveraging swap space with solid state hard drives, to allow larger scale depletion calculations (allowing for significantly more burnable regions than previously examined). As the MCNP6 burnup capability was scaled to larger numbers of burnable regions, a noticeable slowdown was realized.This paper details two specific computational performance strategies for improving calculation speedup: (1) retrieving cross sections during transport; and (2) tallying mechanisms specific to burnup in MCNP. To combat this slowdown new performance upgrades were developed and integrated into MCNP6 1.2.« less
Performance upgrades to the MCNP6 burnup capability for large scale depletion calculations
Fensin, M. L.; Galloway, J. D.; James, M. R.
2015-04-11
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. With the merger of MCNPX and MCNP5, MCNP6 combined the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. The new MCNP6 depletion capability was first showcased at the International Congress for Advancements in Nuclear Power Plants (ICAPP) meeting in 2012. At that conference the new capabilities addressed included the combined distributive and shared memory parallel architecture for the burnup capability, improved memory management, physics enhancements, and new predictability as compared to the H.B Robinson Benchmark. At Los Alamos National Laboratory, a special purpose cluster named “tebow,” was constructed such to maximize available RAM per CPU, as well as leveraging swap space with solid state hard drives, to allow larger scale depletion calculations (allowing for significantly more burnable regions than previously examined). As the MCNP6 burnup capability was scaled to larger numbers of burnable regions, a noticeable slowdown was realized.This paper details two specific computational performance strategies for improving calculation speedup: (1) retrieving cross sections during transport; and (2) tallying mechanisms specific to burnup in MCNP. To combat this slowdown new performance upgrades were developed and integrated into MCNP6 1.2.
The New MCNP6 Depletion Capability
Fensin, Michael Lorne; James, Michael R.; Hendricks, John S.; Goorley, John T.
2012-06-19
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.
The new MCNP6 depletion capability
Fensin, M. L.; James, M. R.; Hendricks, J. S.; Goorley, J. T.
2012-07-01
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)
MCNP4A: Features and philosophy
Hendricks, J.S.
1993-05-01
This paper describes MCNP, states its philosophy, introduces a number of new features becoming available with version MCNP4A, and answers a number of questions asked by participants in the workshop. MCNP is a general-purpose three-dimensional neutron, photon and electron transport code. Its philosophy is ``Quality, Value and New Features.`` Quality is exemplified by new software quality assurance practices and a program of benchmarking against experiments. Value includes a strong emphasis on documentation and code portability. New features are the third priority. MCNP4A is now available at Los Alamos. New features in MCNP4A include enhanced statistical analysis, distributed processor multitasking, new photon libraries, ENDF/B-VI capabilities, X-Windows graphics, dynamic memory allocation, expanded criticality output, periodic boundaries, plotting of particle tracks via SABRINA, and many other improvements. 23 refs.
The MCNP{trademark}/LCS{trademark} merger project
Hughes, H.G.; Adams, K.J.; Chadwick, M.B.
1997-10-01
The MCNP code is written and maintained by Group X-TM at Los Alamos National Laboratory. In response to a variety of needs, and particularly in support of the Accelerator Production of Tritium (APT) program, the authors have recently 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.
MCNP capabilities at the dawn of the 21st century: Neutron-gamma applications
Selcow, E.C.; McKinney, G.W.
2000-10-01
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.
MCNP Capabilities at the Dawn of the 21st Century: Neutron-Gamma Applications
NASA Astrophysics Data System (ADS)
Selcow, E. C.; McKinney, G. W.; Booth, T. E.; Briesmeister, J. F.; Cox, L. J.; Forster, R. A.; Hendricks, J. S.; Mosteller, R. D.; Prael, R. E.; Sood, A.; White, S. W.
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 oil-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 [1], was released to the Radiation Safety Information Computational Center (RSICC) in February 2000. This paper describes the new features and capabilities of the code, and discusses the specific applicability to neutron-gamma problems. We will also discuss some of the future directions for MCNP code development.
MatMCNP: A Code for Producing Material Cards for MCNP
DePriest, Kendall Russell; Saavedra, Karen C.
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.
MCNP APPLICATIONS FOR THE 21ST CENTURY
G. MCKINNEY; T. BOOTH; ET AL
2000-10-01
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.
MCNP application for the 21 century
McKinney, M.C.
2000-08-01
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.
Visualizing MCNP Tally Segment Geometry and Coupling Results with ABAQUS
J. R. Parry; J. A. Galbraith
2007-11-01
The Advanced Graphite Creep test, AGC-1, is planned for irradiation in the Advanced Test Reactor (ATR) in support of the Next Generation Nuclear Plant program. The experiment requires very detailed neutronics and thermal hydraulics analyses to show compliance with programmatic and ATR safety requirements. The MCNP model used for the neutronics analysis required hundreds of tally regions to provide the desired detail. A method for visualizing the hundreds of tally region geometries and the tally region results in 3 dimensions has been created to support the AGC-1 irradiation. Additionally, a method was created which would allow ABAQUS to access the results directly for the thermal analysis of the AGC-1 experiment.
Utilization of new 150-MeV neutron and proton evaluations in MCNP
Little, R.C.; Frankle, S.C.; Hughes, H.G. III; Prael, R.E.
1997-10-01
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.
MCNP-ORIGEN2 Coupling Utility Program
Energy Science and Technology Software Center (ESTSC)
1997-07-30
The MOCUP code system is a series of pre- and post-processor modules to connect the MCNP Monte Carlo transport code and the ORIGEN2.1 depletion and isotopics code into a generalized transport/depletion package for use on non-lattice or non-uniform lattice reactor calculations. No modifications were made to either MCNP or ORIGEN2.1, permitting standard versions of each code to be used. MOCUP contains a simple graphical user interface to allow the user to easily execute the modulesmore » governing MCNP and ORIGEN2.1 input assembly, output processing, and execution, and to perform various file housekeeping tass. Flux and reaction rate calculations are performed in MCNP, with the results extracted by the menpPRO module and passed to the ORIGEN2.1 code by the origenPRO module for deletion. The resulting new isotopic inventories are used to modify the MCNP input in the compPRO module for use in the next timestep. MOCUP permits an arbitary number of depletable cells, different depletable cell types (fuel, targets, etc.) and isotopes that may be tracked. anticipated applications are to test and research reactor physics analyses; isotope production; fuel, target, filter, control, and/or burnable absorber depletion; structural material transmutation; and verification of lattice code calculations.« less
MCNP analyses of criticality calculation results
Forster, R.A.; Booth, T.E.
1995-05-01
Careful assessment of the results of a calculation by the code itself can reduce mistakes in the problem setup and execution. MCNP has over four hundred error messages that inform the user of FATAL or WARNING errors that have been discovered during the processing of just the input file. The latest version, MCNP4A, now performs a self assessment of the calculated results to aid the user in determining the quality of the Monte Carlo results. MCNP4A, which was released to RSIC in October 1993, contains new analyses of the MCNP Monte Carlo calculation that provide simple user WARNINGs for both criticality and fixed source calculations. The goal of the new analyses is to provide the MCNP criticality practitioner with enough information in the output to assess the validity of the k{sub eff} calculation and any associated tallies. The results of these checks are presented in the k{sub eff} results summary page, several k{sub eff} tables and graphs, and tally tables and graphs. Plots of k{sub eff} at the workstation are also available as the problem is running or in a postprocessing mode to assess problem performance and results.
Progress with On-The-Fly Neutron Doppler Broadening in MCNP
Brown, Forrest B.; Martin, William R.; Yesilyurt, Gokhan; Wilderman, Scott
2012-06-18
The University of Michigan, ANL, and LANL have been collaborating on a US-DOE-NE University Programs project 'Implementation of On-the-Fly Doppler Broadening in MCNP5 for Multiphysics Simulation of Nuclear Reactors.' This talk describes the project and provides results from the initial implementation of On-The-Fly Doppler broadening (OTF) in MCNP and testing. The OTF methodology involves high precision fitting of Doppler broadened cross-sections over a wide temperature range (the target for reactor calculations is 250-3200K). The temperature dependent fits are then used within MCNP during the neutron transport, for OTF broadening based on cell temperatures. It is straightforward to extend this capability to cover any temperature range of interest, allowing the Monte Carlo simulation to account for a continuous distribution of temperature ranges throughout the problem geometry.
Benchmark analysis of MCNP{trademark} ENDF/B-VI iron
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.
SABRINA - an interactive geometry modeler for MCNP
West, J.T.; Murphy, J. )
1988-01-01
One of the most difficult tasks when analyzing a complex three-dimensional system with Monte Carlo is geometry model development. SABRINA attempts to make the modeling process more user-friendly and less of an obstacle. It accepts both combinatorial solid bodies and MCNP surfaces and produces MCNP cells. The model development process in SABRINA is highly interactive and gives the user immediate feedback on errors. Users can view their geometry from arbitrary perspectives while the model is under development and interactively find and correct modeling errors. An example of a SABRINA display is shown. It represents a complex three-dimensional shape.
Application of MCNP{trademark} to computed tomography in medicine
Brockhoff, R.C.; Estes, G.P.; Hills, C.R.; Demarco, J.J.; Solberg, T.D.
1996-03-01
The MCNP{trademark} code has been used to simulate CT scans of the MIRD human phantom. In addition. an actual CT scan of a patient was used to create an MCNP geometry, and this geometry was computationally ``CT scanned`` using MCNP to reconstruct CT images. The results show that MCNP can be used to model the human body based on data obtained from CT scans and to simulate CT scans that are based on these or other models.
Addressing Fission Product Validation in MCNP Burnup Credit Criticality Calculations
Mueller, Don; Bowen, Douglas G; Marshall, William BJ J
2015-01-01
The US Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation issued Interim Staff Guidance (ISG) 8, Revision 3 in September 2012. This ISG provides guidance for NRC staff members’ review of burnup credit (BUC) analyses supporting transport and dry storage of pressurized water reactor spent nuclear fuel (SNF) in casks. The ISG includes guidance for addressing validation of criticality (k_{eff}) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MAs). Based on previous work documented in NRC Regulatory Guide (NUREG) Contractor Report (CR)-7109, the ISG recommends that NRC staff members accept the use of either 1.5 or 3% of the FP&MA worth—in addition to bias and bias uncertainty resulting from validation of k_{eff} calculations for the major actinides in SNF—to conservatively account for the bias and bias uncertainty associated with the specified unvalidated FP&MAs. The ISG recommends (1) use of 1.5% of the FP&MA worth if a modern version of SCALE and its nuclear data are used and (2) 3% of the FP&MA worth for well qualified, industry standard code systems other than SCALE with the Evaluated Nuclear Data Files, Part B (ENDF/B),-V, ENDF/B-VI, or ENDF/B-VII cross sections libraries. The work presented in this paper provides a basis for extending the use of the 1.5% of the FP&MA worth bias to BUC criticality calculations performed using the Monte Carlo N-Particle (MCNP) code. The extended use of the 1.5% FP&MA worth bias is shown to be acceptable by comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII–based nuclear data. The comparison supports use of the 1.5% FP&MA worth bias when the MCNP code is used for criticality calculations, provided that the cask design is similar to the hypothetical generic BUC-32 cask model and that the credited FP&MA worth is no more than 0.1 Δk_{eff} (ISG-8, Rev. 3, Recommendation 4).
MCNP Analysis of a Phoswich Detector
Nathan Childress; William H. Miller
2002-06-12
A series of triple crystal phosphor sandwich detectors have been developed and constructed for testing at the University of Missouri-Columbia [1-7]. These detectors can simultaneously measure alpha, beta, and gamma radiation and utilize digital pulse shape discrimination to identify and separate radiation events coming from each of the separate phosphors. The research reported here uses Monte Carlo [8] software analysis to determine operating parameters for this detector system and optimizes its design for measuring trace amounts of alpha, beta and gamma-ray activity in effluent streams from nuclear waste cleanup processes. The previously designed, fabricated and tested phoswich detector [5] consisted of three scintillators placed on top of each other with a common diameter of 5.08 cm and viewed with a single photomultiplier tube. The scintillators (ZnS-0.00376 cm, CaF{sub 2}-0.254 cm and NaI-2.54 cm) interact preferentially with alpha, beta and gamma-ray radiation, respectively. This design allows preferential, but not exclusive, interaction of various radiations with specific layers. Taking into account and correcting for events that can occur in the ''wrong'' phosphor, this system was experimentally shown to have a 99% accuracy for properly identifying radiation coming from a mixed alpha/beta/gamma-ray source. In an attempt to better understand this system and provide design guidance for a detector system to be used in monitoring effluents from nuclear waste treatment facilities, this detector was modeled using MCNP [8]. This analysis [9] indicated that the thin ZnS layer adequately stops alpha particle energy, but greatly reduces beta detection efficiency to essentially zero at beta E{sub max} energies below 300 keV. The CaF{sub 2} layer, designed to keep any beta particle energy from entering the NaI detector results in an incorrect gamma-ray response that is approximately 23% of the NaI's response and is variable with energy. High energy beta events in the Ca
New methods for neutron response calculations with MCNP
Hendricks, J.S.
1997-05-01
MCNP4B was released for international distribution in February, 1997. The author summarized the new MCNP4B features since the release of MCNP4A over three years earlier and compare some results. Then he describes new methods being developed for future code releases. The focus is methods and applications of ex-core neutron response calculations.
Testing the Delayed Gamma Capability in MCNP6
Weldon, Robert A.; Fensin, Michael L.; McKinney, Gregg W.
2015-10-28
The mission of the Domestic Nuclear Detection Office is to quickly and reliably detect unauthorized attempts to import or transport special nuclear material for use against the United States. Developing detection equipment to meet this objective requires accurate simulation of both the detectable signature and detection mechanism. A delayed particle capability was initially added to MCNPX 2.6.A in 2005 to sample the radioactive fission product parents and emit decay particles resulting from the decay chain. To meet the objectives of detection scenario modeling, the capability was designed to sample a particular time for emitting particular multiplicity of a particular energy. Because the sampling process of selecting both time and energy is interdependent, to linearize the time and emission sampling, atom densities are computed at several discrete time steps, and the time-integrated production is computed by multiplying the atom density by the decay constant and time step size to produce a cumulative distribution function for sampling the emission time, energy, and multiplicity. The delayed particle capability was initially given a time-bin structure to help reasonably reproduce, from a qualitative sense, a fission benchmark by Beddingfield, which examined the delayed gamma emission. This original benchmark was only qualitative and did not contain the magnitudes of the actual measured data but did contain relative graphical representation of the spectra. A better benchmark with measured data was later provided by Hunt, Mozin, Reedy, Selpel, and Tobin at the Idaho Accelerator Center; however, because of the complexity of the benchmark setup, sizable systematic errors were expected in the modeling, and initial results compared to MCNPX 2.7.0 showed errors outside of statistical fluctuation. Presented in this paper is a more simplified approach to benchmarking, utilizing closed form analytic solutions to the granddaughter equations for particular sets of decay systems
Evaluation of Geometric Progression (GP) Buildup Factors using MCNP Codes (MCNP6.1 and MCNP5-1.60)
NASA Astrophysics Data System (ADS)
Kim, Kyung-O.; Roh, Gyuhong; Lee, Byungchul
2016-02-01
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.
The MCNP5 Random number generator
Brown, F. B.; Nagaya, Y.
2002-01-01
MCNP and other Monte Carlo particle transport codes use random number generators to produce random variates from a uniform distribution on the interval. These random variates are then used in subsequent sampling from probability distributions to simulate the physical behavior of particles during the transport process. This paper describes the new random number generator developed for MCNP Version 5. The new generator will optionally preserve the exact random sequence of previous versions and is entirely conformant to the Fortran-90 standard, hence completely portable. In addition, skip-ahead algorithms have been implemented to efficiently initialize the generator for new histories, a capability that greatly simplifies parallel algorithms. Further, the precision of the generator has been increased, extending the period by a factor of 10{sup 5}. Finally, the new generator has been subjected to 3 different sets of rigorous and extensive statistical tests to verify that it produces a sufficiently random sequence.
Experimental validation of lead cross sections for scale and MCNP
Henrikson, D.J.
1995-12-01
Moving spent nuclear fuel between facilities often requires the use of lead-shielded casks. Criticality safety that is based upon calculations requires experimental validation of the fuel matrix and lead cross section libraries. A series of critical experiments using a high-enriched uranium-aluminum fuel element with a variety of reflectors, including lead, has been identified. Twenty-one configurations were evaluated in this study. The fuel element was modelled for KENO V.a and MCNP 4a using various cross section sets. The experiments addressed in this report can be used to validate lead-reflected calculations. Factors influencing calculated k{sub eff} which require further study include diameters of styrofoam inserts and homogenization.
Present and future capabilities of MCNP
Hendricks; Adam; Booth; Briesmeister; Carter; Cox; Favorite; Forster; McKinney; Prael
2000-10-01
Several new capabilities have been added to MCNP4C including: (1) macrobody surfaces; (2) the superimposed mesh importance functions, so that it is no longer necessary to subdivide geometries for variance reduction; and (3) Xlib graphics and DVF Fortran 90 for PCs. There are also improvements in neutron physics, electron physics, perturbations, and parallelization. In the more distant future we are working on adaptive Monte Carlo code modernization, more parallelization, visualization, and charged particles. PMID:11003531
MCNP Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1365K.
Energy Science and Technology Software Center (ESTSC)
2001-04-19
Version 00 UTXS is a project whereby continuous-energy cross section libraries in ACE format suitable for the MCNP code were generated using the NJOY94.105 processing code. Libraries for various materials were generated at typical operating temperatures of the US Pressurized Water Reactor (PWR), Boiling Water Reactor (BWR), and the Russian PWR (VVER) as well as libraries for other non-reactor applications such as nuclear medicine.
JEF 2.2 Cross Section Library for the MCNP Monte Carlo Code.
Energy Science and Technology Software Center (ESTSC)
2003-11-24
Version 01 This continuous energy cross-section data library for MCNP is based on the JEF-2.2 evaluated nuclear data library (ACE format). The present library was satisfactorily tested in thermal and fast criticality benchmarks. For analyses below 20 MeV, MCJEF22NEA.BOLlB was applied also in cell and core calculations dedicated to the study of the subcritical accelerator driven systems (ADS).
MCNP4B{sup {trademark}} verification and validation
Hendricks, J.S.; Court, J.D.
1996-08-01
Several new features and bug fixes have been incorporated into the new release of MCNP. As required by the MCNP Software Quality Assurance Plan, these changes to the code and the test set are documented here for user reference. This document summarizes the new MCNP4B features and corrections, separated into major and minor groupings. Also included are a code cleanup section and a section delineating problems identified in LA-12839 which have not been corrected. Finally, we document the MCNP4B test set modifications and explain how test set coverage has been improved.
MCNP Output Data Analysis with ROOT (MODAR)
NASA Astrophysics Data System (ADS)
Carasco, C.
2010-06-01
MCNP Output Data Analysis with ROOT (MODAR) is a tool based on CERN's ROOT software. MODAR has been designed to handle time-energy data issued by MCNP simulations of neutron inspection devices using the associated particle technique. MODAR exploits ROOT's Graphical User Interface and functionalities to visualize and process MCNP simulation results in a fast and user-friendly way. MODAR allows to take into account the detection system time resolution (which is not possible with MCNP) as well as detectors energy response function and counting statistics in a straightforward way. Program summaryProgram title: MODAR Catalogue identifier: AEGA_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEGA_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 155 373 No. of bytes in distributed program, including test data, etc.: 14 815 461 Distribution format: tar.gz Programming language: C++ Computer: Most Unix workstations and PC Operating system: Most Unix systems, Linux and windows, provided the ROOT package has been installed. Examples where tested under Suse Linux and Windows XP. RAM: Depends on the size of the MCNP output file. The example presented in the article, which involves three two-dimensional 139×740 bins histograms, allocates about 60 MB. These data are running under ROOT and include consumption by ROOT itself. Classification: 17.6 External routines: ROOT version 5.24.00 ( http://root.cern.ch/drupal/) Nature of problem: The output of an MCNP simulation is an ASCII file. The data processing is usually performed by copying and pasting the relevant parts of the ASCII file into Microsoft Excel. Such an approach is satisfactory when the quantity of data is small but is not efficient when the size of the simulated data is large, for example when time
MCNP output data analysis with ROOT (MODAR)
NASA Astrophysics Data System (ADS)
Carasco, C.
2010-12-01
MCNP Output Data Analysis with ROOT (MODAR) is a tool based on CERN's ROOT software. MODAR has been designed to handle time-energy data issued by MCNP simulations of neutron inspection devices using the associated particle technique. MODAR exploits ROOT's Graphical User Interface and functionalities to visualize and process MCNP simulation results in a fast and user-friendly way. MODAR allows to take into account the detection system time resolution (which is not possible with MCNP) as well as detectors energy response function and counting statistics in a straightforward way. New version program summaryProgram title: MODAR Catalogue identifier: AEGA_v1_1 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEGA_v1_1.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 150 927 No. of bytes in distributed program, including test data, etc.: 4 981 633 Distribution format: tar.gz Programming language: C++ Computer: Most Unix workstations and PCs Operating system: Most Unix systems, Linux and windows, provided the ROOT package has been installed. Examples where tested under Suse Linux and Windows XP. RAM: Depends on the size of the MCNP output file. The example presented in the article, which involves three two dimensional 139×740 bins histograms, allocates about 60 MB. These data are running under ROOT and include consumption by ROOT itself. Classification: 17.6 Catalogue identifier of previous version: AEGA_v1_0 Journal reference of previous version: Comput. Phys. Comm. 181 (2010) 1161 External routines: ROOT version 5.24.00 ( http://root.cern.ch/drupal/) Does the new version supersede the previous version?: Yes Nature of problem: The output of a MCNP simulation is an ascii file. The data processing is usually performed by copying and pasting the relevant parts of the ascii
Use of MCNP + GADRAS in Generating More Realistic Gamma-Ray Spectra for Plutonium and HEU Objects
Rawool-Sullivan, Mohini; Mattingly, John; Mitchell, Dean
2012-08-07
The ability to accurately simulate high-resolution gamma spectra from materials that emit both neutrons and gammas is very important to the analysis of special nuclear materials (SNM), e.g., uranium and plutonium. One approach under consideration has been to combine MCNP and GADRAS. This approach is expected to generate more accurate gamma ray spectra for complex three-dimensional geometries than can be obtained from one-dimensional deterministic transport simulations (e.g., ONEDANT). This presentation describes application of combining MCNP and GADRAS in simulating plutonium and uranium spectra.
MCNP simulations of material exposure experiments (u)
Temple, Brian A
2010-12-08
Simulations of proposed material exposure experiments were performed using MCNP6. The experiments will expose ampules containing different materials of interest with radiation to observe the chemical breakdown of the materials. Simulations were performed to map out dose in materials as a function of distance from the source, dose variation between materials, dose variation due to ampule orientation, and dose variation due to different source energy. This write up is an overview of the simulations and will provide guidance on how to use the data in the spreadsheet.
Mattingly, John K.; Pozzi, Sara A.; Clarke, Shaun D.; Dennis, Ben D.; Miller, Eric C.; Padovani, E.
2010-06-01
With increasing concern over the ability to detect and characterize special nuclear materials, the need for computer codes that can successfully predict the response of detector systems to various measurement scenarios is extremely important. These computer algorithms need to be benchmarked against a variety of experimental configurations to ensure their accuracy and understand their limitations. The Monte Carlo code MCNP-PoliMi is a modified version of the MCNP-4c code. Recently these modifications have been ported into the new MCNPX 2.6.0 code, which gives the new MCNPX-PoliMi a wider variety of options and abilities, taking advantage of the improvements made to MCNPX. To verify the ability of the MCNPX-PoliMi code to simulate the response of a neutron multiplicity detector simulated results were compared to experimental data. The experiment consisted of a 4.5-kg sphere of alpha-phase plutonium that was moderated with various thicknesses of polyethylene. The results showed that our code system can simulate the multiplicity distributions with relatively good agreement with measured data. The enhancements made to MCNP since the release of MCNP-4c have had little to no effect on the ability of the MCNP-PoliMi to resolve the discrepancies observed in the simulated neutron multiplicity distributions when compared experimental data.
MCNP6 Cosmic & Terrestrial Background Particle Fluxes -- Release 4
McMath, Garrett E.; McKinney, Gregg W.; Wilcox, Trevor
2015-01-23
Essentially a set of slides, the presentation begins with the MCNP6 cosmic-source option, then continues with the MCNP6 transport model (atmospheric, terrestrial) and elevation scaling. It concludes with a few slides on results, conclusions, and suggestions for future work.
An assessment of the MCNP4C weight window
Christopher N. Culbertson; John S. Hendricks
1999-12-01
A new, enhanced weight window generator suite has been developed for MCNP version 4C. The new generator correctly estimates importances in either a user-specified, geometry-independent, orthogonal grid or in MCNP geometric cells. The geometry-independent option alleviates the need to subdivide the MCNP cell geometry for variance reduction purposes. In addition, the new suite corrects several pathologies in the existing MCNP weight window generator. The new generator is applied in a set of five variance reduction problems. The improved generator is compared with the weight window generator applied in MCNP4B. The benefits of the new methodology are highlighted, along with a description of its limitations. The authors also provide recommendations for utilization of the weight window generator.
Validation of MCNP: SPERT-D and BORAX-V fuel
Crawford, C.; Palmer, B.
1992-11-01
This report discusses critical experiments involving SPERT-D[sup 1,2] fuel elements and BORAX-V[sup 3-8] fuel which have been modeled and calculations performed with MCNP. MCNP is a Monte Carlo based transport code. For this study continuous-energy nuclear data from the ENDF/B-V cross section library was used. The SPERT-D experiments consisted of various arrays of fuel elements moderated and reflected with either water or a uranyl nitrate solution. Some SPERT-D experiments used cadmium as a fixed neutron poison, while others were poisoned with various concentrations of boron in the moderating/reflecting solution. ne BORAX-V experiments were arrays of either boiling fuel rod 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.
Validation of MCNP: SPERT-D and BORAX-V fuel
Crawford, C.; Palmer, B.
1992-11-01
This report discusses critical experiments involving SPERT-D{sup 1,2} fuel elements and BORAX-V{sup 3-8} fuel which have been modeled and calculations performed with MCNP. MCNP is a Monte Carlo based transport code. For this study continuous-energy nuclear data from the ENDF/B-V cross section library was used. The SPERT-D experiments consisted of various arrays of fuel elements moderated and reflected with either water or a uranyl nitrate solution. Some SPERT-D experiments used cadmium as a fixed neutron poison, while others were poisoned with various concentrations of boron in the moderating/reflecting solution. ne BORAX-V experiments were arrays of either boiling fuel rod 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 MCNP model of glove boxes in a plutonium processing facility
Dooley, D.E.; Kornreich, D.E.
1998-12-31
Nuclear material processing usually occurs simultaneously in several glove boxes whose primary purpose is to contain radioactive materials and prevent inhalation or ingestion of radioactive materials by workers. A room in the plutonium facility at Los Alamos National Laboratory has been slated for installation of a glove box for storing plutonium metal in various shapes during processing. This storage glove box will be located in a room containing other glove boxes used daily by workers processing plutonium parts. An MCNP model of the room and glove boxes has been constructed to estimate the neutron flux at various locations in the room for two different locations of the storage glove box and to determine the effect of placing polyethylene shielding around the storage glove box. A neutron dose survey of the room with sources dispersed as during normal production operations was used as a benchmark to compare the neutron dose equivalent rates calculated by the MCNP model.
MCNPX{trademark} -- The LAHET{trademark}/MCNP{trademark} code merger
Hughes, H.G.; Adams, K.J.; Chadwick, M.B.
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.
Performance of MCNP4A on seven computing platforms
Hendricks, J.S.; Brockhoff, R.C.
1994-12-31
The performance of seven computer platforms has been evaluated with the MCNP4A Monte Carlo radiation transport code. For the first time we report timing results using MCNP4A and its new test set and libraries. Comparisons are made on platforms not available to us in previous MCNP timing studies. By using MCNP4A and its 325-problem test set, a widely-used and readily-available physics production code is used; the timing comparison is not limited to a single ``typical`` problem, demonstrating the problem dependence of timing results; the results are reproducible at the more than 100 installations around the world using MCNP; comparison of performance of other computer platforms to the ones tested in this study is possible because we present raw data rather than normalized results; and a measure of the increase in performance of computer hardware and software over the past two years is possible. The computer platforms reported are the Cray-YMP 8/64, IBM RS/6000-560, Sun Sparc10, Sun Sparc2, HP/9000-735, 4 processor 100 MHz Silicon Graphics ONYX, and Gateway 2000 model 4DX2-66V PC. In 1991 a timing study of MCNP4, the predecessor to MCNP4A, was conducted using ENDF/B-V cross-section libraries, which are export protected. The new study is based upon the new MCNP 25-problem test set which utilizes internationally available data. MCNP4A, its test problems and the test data library are available from the Radiation Shielding and Information Center in Oak Ridge, Tennessee, or from the NEA Data Bank in Saclay, France. Anyone with the same workstation and compiler can get the same test problem sets, the same library files, and the same MCNP4A code from RSIC or NEA and replicate our results. And, because we report raw data, comparison of the performance of other compute platforms and compilers can be made.
Performance analysis of the Monte Carlo code MCNP4A for photon-based radiotherapy applications
DeMarco, J.J.; Solberg, T.D.; Wallace, R.E.; Smathers, J.B.
1995-12-31
The Los Alamos code MCNP4A (Monte Carlo M-Particle version 4A) is currently used to simulate a variety of problems ranging from nuclear reactor analysis to boron neutron capture therapy. This study is designed to evaluate MCNP4A as the dose calculation system for photon-based radiotherapy applications. A graphical user interface (MCNP Radiation Therapy) has been developed which automatically sets up the geometry and photon source requirements for three-dimensional simulations using Computed Tomography (CT) data. Preliminary results suggest the code is capable of calculating satisfactory dose distributions in a variety of simulated homogeneous and heterogeneous phantoms. The major drawback for this dosimetry system is the amount of time to obtain a statistically significant answer. MCNPRT allows the user to analyze the performance of MCNP4A as a function of material, geometry resolution and MCNP4A photon and electron physics parameters. A typical simulation geometry consists of a 10 MV photon point source incident on a 15 x 15 x 15 cm{sup 3} phantom composed of water voxels ranging in size from 10 x 10 x 10 mm{sup 3} to 2 x 2 x 2 mm{sup 3}. As the voxel size is decreased, a larger percentage of time is spent tracking photons through the voxelized geometry as opposed to the secondary electrons. A PRPR Patch file is under development that will optimize photon transport within the simulation phantom specifically for radiotherapy applications. MCNP4A also supports parallel processing capabilities via the Parallel Virtual Machine (PVM) message passing system. A dedicated network of five SUN SPARC2 processors produced a wall-clock speedup of 4.4 based on a simulation phantom containing 5 x 5 x 5 mm{sup 3} water voxels. The code was also tested on the 80 node IBM RS/6000 cluster at the Maui High Performance Computing Center (NHPCC). A non-dedicated system of 75 processors produces a wall clock speedup of 29 relative to one SUN SPARC2 computer.
MCNP modelling of the PBMR equilibrium core
Albornoz, F.; Korochinsky, S.
2006-07-01
A complete MCNP model of the PBMR equilibrium core is presented, which accounts for the same fuel regions defined in the PBMR core management code, as well as for complete fuel and reflector temperature distributions. This comprehensive 3D model is the means to calculate and characterize the neutron and photon boundary sources of the equilibrium core, and is also used to support some specific core neutronic studies needing detailed geometry modelling. Due to the geometrical modelling approach followed, an unrealistic partial cutting of fuel kernels and pebbles is introduced in the model. The variations introduced by this partial cutting both on the packing fraction and on the uranium load of the modelled core and its corresponding effect on core reactivity and flux levels, have been investigated and quantified. A complete set of high-temperature cross-section data was applied to the calculation of the PBMR equilibrium core, and its effect on the calculated core reactivity is also reported. (authors)
Enhancements to the MCNP6 background source
McMath, Garrett E.; McKinney, Gregg W.
2015-10-19
The particle transport code MCNP has been used to produce a background radiation data file on a worldwide grid that can easily be sampled as a source in the code. Location-dependent cosmic showers were modeled by Monte Carlo methods to produce the resulting neutron and photon background flux at 2054 locations around Earth. An improved galactic-cosmic-ray feature was used to model the source term as well as data from multiple sources to model the transport environment through atmosphere, soil, and seawater. A new elevation scaling feature was also added to the code to increase the accuracy of the cosmic neutronmore » background for user locations with off-grid elevations. Furthermore, benchmarking has shown the neutron integral flux values to be within experimental error.« less
Enhancements to the MCNP6 background source
McMath, Garrett E.; McKinney, Gregg W.
2015-10-19
The particle transport code MCNP has been used to produce a background radiation data file on a worldwide grid that can easily be sampled as a source in the code. Location-dependent cosmic showers were modeled by Monte Carlo methods to produce the resulting neutron and photon background flux at 2054 locations around Earth. An improved galactic-cosmic-ray feature was used to model the source term as well as data from multiple sources to model the transport environment through atmosphere, soil, and seawater. A new elevation scaling feature was also added to the code to increase the accuracy of the cosmic neutron background for user locations with off-grid elevations. Furthermore, benchmarking has shown the neutron integral flux values to be within experimental error.
Computational radiology and imaging with the MCNP Monte Carlo code
Estes, G.P.; Taylor, W.M.
1995-05-01
MCNP, a 3D coupled neutron/photon/electron Monte Carlo radiation transport code, is currently used in medical applications such as cancer radiation treatment planning, interpretation of diagnostic radiation images, and treatment beam optimization. This paper will discuss MCNP`s current uses and capabilities, as well as envisioned improvements that would further enhance MCNP role in computational medicine. It will be demonstrated that the methodology exists to simulate medical images (e.g. SPECT). Techniques will be discussed that would enable the construction of 3D computational geometry models of individual patients for use in patient-specific studies that would improve the quality of care for patients.
Pozzi, S. A.; Mihalczo, J. T.
2002-06-03
In the past, passive Nuclear Materials Identification System (NMIS) measurements on plutonium metal shells at VNIIEF have shown the sensitivity of the acquired covariance functions to shell mass and thickness for a variety of shell thicknesses from 6 to 30 mm and masses varying from 1829 to 4468g. The technique acquires the time-dependent coincidence distribution between plastic scintillators detecting radiation from the Pu. The measurements showed the sensitivity of the acquired signature to the different spontaneous emission, attenuation, and multiplication properties of the shells. In this work, the MCNP-POLIMI neutron and photon transport code was used to simulate passive measurements on plutonium metal and oxide. The code is a modified version of MCNP, which attempts to calculate more correctly quantities that depend on the second moment of the neutron and gamma distributions, and attempts to model detector pulses as closely as possible. MCNP-POLIMI, together with a post-processing code, can simulate all the time-dependent coincidence distributions measured by NMIS. In particular, the simulations evaluate the time-dependent coincidence distributions between detectors for plutonium samples having mass 2 and 4 kg, in metal and oxide form. This work shows that the time-dependent coincidence distributions between two scintillators measured by NMIS can be used to distinguish metal from oxide.
Automated variance reduction for Monte Carlo shielding analyses with MCNP
NASA Astrophysics Data System (ADS)
Radulescu, Georgeta
Variance reduction techniques are employed in Monte Carlo analyses to increase the number of particles in the space phase of interest and thereby lower the variance of statistical estimation. Variance reduction parameters are required to perform Monte Carlo calculations. It is well known that adjoint solutions, even approximate ones, are excellent biasing functions that can significantly increase the efficiency of a Monte Carlo calculation. In this study, an automated method of generating Monte Carlo variance reduction parameters, and of implementing the source energy biasing and the weight window technique in MCNP shielding calculations has been developed. The method is based on the approach used in the SAS4 module of the SCALE code system, which derives the biasing parameters from an adjoint one-dimensional Discrete Ordinates calculation. Unlike SAS4 that determines the radial and axial dose rates of a spent fuel cask in separate calculations, the present method provides energy and spatial biasing parameters for the entire system that optimize the simulation of particle transport towards all external surfaces of a spent fuel cask. The energy and spatial biasing parameters are synthesized from the adjoint fluxes of three one-dimensional Discrete Ordinates adjoint calculations. Additionally, the present method accommodates multiple source regions, such as the photon sources in light-water reactor spent nuclear fuel assemblies, in one calculation. With this automated method, detailed and accurate dose rate maps for photons, neutrons, and secondary photons outside spent fuel casks or other containers can be efficiently determined with minimal efforts.
Validating MCNP for LEU Fuel Design via Power Distribution Comparisons
Primm, Trent; Maldonado, G Ivan; Chandler, David
2008-11-01
The mission of the Reduced Enrichment for Research and Test Reactors (RERTR) Program is to minimize and, to the extent possible, eliminate the use of highly enriched uranium (HEU) in civilian nuclear applications by working to convert research and test reactors, as well as radioisotope production processes, to low enriched uranium (LEU) fuel and targets. Oak Ridge National Lab (ORNL) is reviewing the design bases and key operating criteria including fuel operating parameters, enrichment-related safety analyses, fuel performance, and fuel fabrication in regard to converting the fuel of the High Flux Isotope Reactor (HFIR) from HEU to LEU. The purpose of this study is to validate Monte Carlo methods currently in use for conversion analyses. The methods have been validated for the prediction of flux values in the reactor target, reflector, and beam tubes, but this study focuses on the prediction of the power density profile in the core. A current 3-D Monte Carlo N-Particle (MCNP) model was modified to replicate the HFIR Critical Experiment 3 (HFIRCE-3) core of 1965. In this experiment, the power profile was determined by counting the gamma activity at selected locations in the core. Foils (chunks of fuel meat and clad) were punched out of the fuel elements in HFIRCE-3 following irradiation and experimental relative power densities were obtained by measuring the activity of these foils and comparing each foil s activity to the activity of a normalizing foil. The current work consisted of calculating corresponding activities by inserting volume tallies into the modified MCNP model to represent the punchings. The average fission density was calculated for each foil location and then normalized to the normalizing foil. Power distributions were obtained for the clean core (no poison in moderator and symmetrical rod position at 17.5 inches) and fully poisoned-moderator (1.35 g B/liter in moderator and rods fully withdrawn) conditions. The observed deviations between the
MCNP-model for the OAEP Thai Research Reactor
Gallmeier, F.X.; Tang, J.S.; Primm, R.T. III
1998-06-01
An MCNP input was prepared for the Thai Research Reactor, making extensive use of the MCNP geometry`s lattice feature that allows a flexible and easy rearrangement of the core components and the adjustment of the control elements. The geometry was checked for overdefined or undefined zones by two-dimensional plots of cuts through the core configuration with the MCNP geometry plotting capabilities, and by a three-dimensional view of the core configuration with the SABRINA code. Cross sections were defined for a hypothetical core of 67 standard fuel elements and 38 low-enriched uranium fuel elements--all filled with fresh fuel. Three test calculations were performed with the MCNP4B-code to obtain the multiplication factor for the cases with control elements fully inserted, fully withdrawn, and at a working position.
MCWO - Linking MCNP And ORIGEN2 For Fuel Burnup Analysis
Gray S Chang
2005-04-01
The UNIX BASH (Bourne Again Shell) script MCWO has been developed at the Idaho National Engineering and Environment Laboratory (INEEL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN2. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN2. MCWO can handle a large number of fuel burnup and material loading specifications, Advanced Test Reactor (ATR) powers, and irradiation time intervals. The program processes input from the user that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Calculated results from MCNP, ORIGEN2, and data process module calculations are then output successively as the code runs. The principal function of MCWO is to transfer one-group cross-section and flux values from MCNP to ORIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN2 back to MCNP in a repeated, cyclic fashion. The basic requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with ORIGEN2 and other calculations are performed by UNIX BASH script MCWO. This paper presents the MCWO-calculated results of the RERTR-1 and -2 , and the Weapons-Grade Mixed Oxiide fuel (Wg-MOX) fuel experiments in ATR and compares the MCWO-calculated results with the measured data.
Spectral measurements in critical assemblies: MCNP specifications and calculated results
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.
Goorley, T.; Bull, J. S.; Brown, F. B.; Booth, Thomas Edward; Hughes, H. G.; Mosteller, R. D.; Forster, R. A.; Post, S. E.; Prael, R. E.; Selcow, Elizabeth Carol,; Sood, A.; Sweezy, J. E.
2004-01-01
In July of 2004, an updated version of MCNP5{trademark} (MCNP5-RSICC-1.30) was released to the Radiation Shielding Information Computational Center. This updated version has three new features, thirteen bug fixes and several minor coding improvements. The new features are: support for 8 byte integers, specialized tally treatment of large lattices, and mesh tally enhancements. Of the thirteen bug fixes, only four resulted in incorrect answers in specific circumstances. In addition to the standard RSICC distribution of the MCNP5 source, executables and patches, the patch file (only) is available on the MCNP website: http://www-xdiv.lanl.gov/x5/MCNP/theresources.html. The three new MCNP5 features are discussed. Several new improvements have also been made to the manual and development environment. All of the features, bug fixes, coding improvement issues and related documentation are now maintained in Sourceforge. Fortran and C source code and regression test problems are now under version control with CVS.
The comparison of two MCNP models used for prompt gamma in vivo detection of cadmium and mercury
NASA Astrophysics Data System (ADS)
Atanackovic, J.; Grinyer, J.; Chettle, D. R.; Byun, S. H.
2007-10-01
In vivo detection of trace elements is one of the most important research areas at the Medical Physics and Applied Radiation Sciences Department at McMaster University. Prompt gamma neutron activation analysis (PGNAA) used for detection of cadmium and mercury takes place simultaneously at two different experimental sites; the McMaster Nuclear Reactor (MNR) and the 238Pu/Be neutron source site. This particular study consists of two parts. In the first part the water phantoms (125 mL) were used in MCNP simulations. The water phantoms were doped with different concentrations of Cd, Hg and HCl. This is done in order to compare the (n, γ) prompt gamma reaction rate; in fact, the rate of neutron capture by the nuclides of interest; 113Cd, 199Hg and 35Cl. The second part involves, the neutron and photon dosimetry calculations that were performed for both sites using MCNP compatible body builder software developed in Los Alamos. The output of this program is the actual MCNP geometry description for various human anthropomorphic phantoms (different sex and ages). This phantom geometry output is incorporated into the original MCNP geometry and the dosimetry calculations were performed for various organs at risk.
ENDF/B-VI Release 3 Cross Section Library for Use with the MCNP Monte Carlo Code.
Energy Science and Technology Software Center (ESTSC)
2003-12-16
Version 00 This continuous energy cross-section data library for MCNP is in ACE format. The present library was satisfactorily tested in thermal and fast criticality benchmarks. For analyses below 20 MeV, MCB63NEA.BOLlB was applied also in cell and core calculations dedicated to the study of the subcritical accelerator driven systems (ADS). This library provides users an additional ENDF/B-VI based, continuous-energy and multi-temperature library for MCNP with an important feature: there is a perfect consistency withmore » the twin library MCJEFF22NEA.BOLIB already released, in terms of nuclear data processing calculation methodology. Both libraries are based on the NJOY-94.66 data processing system. This may be important, in particular, for the users involved in nuclear data validation who have already used the MCJEF22NEA.BOLIB library.« less
Characterization of the NPOD3 Detectors in MCNP5 and MCNP6
Clark, Kimberly L.; Hutchinson, Jesson D.; Sood, Avneet
2014-01-21
Researchers performed a series of measurements in May 2012 to characterize the NPOD3 detector systems. The detectors were placed in varying states of disassembly to determine the effect of individual components on detection efficiency. A 4.5 kg α-phase Pu sphere known as the Los Alamos BeRP Ball was used as the SNM source in both a bare configuration and reflected by varying thicknesses of polyethylene. A set of simulations matching the experimental setups were run and the data were compared to the measured data. The total and leakage multiplication and the inferred k values were determined for both the simulations and the measurements. Table 3 shows a comparison of the results from MCNP6 and MCNP5 with the list-mode patch to the measured results. The count rates for the calculated results were obtained by dividing the total line count in the list-mode file (equivalent to the total number of absorptions in the NPOD detectors) by the total run time. The count rates are identical for both codes, and they both produce the same multiplicity and inferred k values regardless of measurement time as expected.
Verification of the MCNP (TM) Perturbation Correction Feature for Cross-Section Dependent Tallies
A. K. Hess; G. W. McKinney; J. S. Hendricks; L. L. Carter
1998-10-01
The Monte Carlo N-Particle Transport Code MCNP version 4B perturbation capability has been extended to cross-section dependent tallies and to the track-length estimate of Iqff in criticality problems. We present the complete theory of the MCNP perturbation capability including the correction to MCNP4B which enables cross-section dependent perturbation tallies. We also present the MCNP interface as an upgrade to the MCNP4B manual. Finally, we present test results demonstrating the validity of the perturbation capability in MCNP, particularly cross-section dependent problems.
Standard Neutron, Photon, and Electron Data Libraries for MCNP4C.
Energy Science and Technology Software Center (ESTSC)
2004-02-16
Version 03 US DOE 10CFR810 Jurisdiction. DLC-200/MCNPDATA is for use with Versions 4C and and 4C2 of the MCNP transport code. This data library provides a comprehensive set of cross sections for a wide range of radiation transport applications using the Monte Carlo code package CCC-700/MCNP4C. See Appendix G of the MCNP report LA-13709-M for information on the libraries and how to select specific nuclides for use in MCNP. Newer MCNP cross sections from LANLmore » are included in CCC-710/MCNP5.« less
Visualization and analyses of MCNP criticality calculation results
Urbatsch, T.J.; Forster, R.A.; Booth, T.E.; Van Riper, K.A.; Waters, L.S.
1995-07-01
Careful assessment of the results of a calculation by the code itself can detect mistakes in the problem setup and execution. MCNP has over four hundred error messages that inform the user of FATAL or WARNING errors that have been discovered during processing of just the input file. MCNP4A performs a self assessment of the calculated results to aid the user in determining the quality of the Monte Carlo results. MCNP4A contains new built-in sensitivity analyses of the Monte Carlo calculation that provide the user with simple WARNING messages for both criticality and fixed source calculations. The goal of the new analyses described in this paper is to provide the MCNP criticality practitioner with enough information in the output to assess the validity of the k{sub eff} calculation and any associated tallies. The results of these checks are presented in the k{sub eff} results summary, several k{sub eff} tables and graphs, and tally tables and graphs. Plots of k{sub eff} at the workstation are also available as the problem is running or in a postprocessing mode to assess problem performance and results. Plots of the fission source by cycle supply valuable visual information, although they are not yet available in the production version of MCNP.
TRIPOLI-4® - MCNP5 ITER A-lite neutronic model benchmarking
NASA Astrophysics Data System (ADS)
Jaboulay, J.-C.; Cayla, P.-Y.; Fausser, C.; Lee, Y.-K.; Trama, J.-C.; Li-Puma, A.
2014-06-01
The aim of this paper is to present the capability of TRIPOLI-4®, the CEA Monte Carlo code, to model a large-scale fusion reactor with complex neutron source and geometry. In the past, numerous benchmarks were conducted for TRIPOLI-4® assessment on fusion applications. Experiments (KANT, OKTAVIAN, FNG) analysis and numerical benchmarks (between TRIPOLI-4® and MCNP5) on the HCLL DEMO2007 and ITER models were carried out successively. In this previous ITER benchmark, nevertheless, only the neutron wall loading was analyzed, its main purpose was to present MCAM (the FDS Team CAD import tool) extension for TRIPOLI-4®. Starting from this work a more extended benchmark has been performed about the estimation of neutron flux, nuclear heating in the shielding blankets and tritium production rate in the European TBMs (HCLL and HCPB) and it is presented in this paper. The methodology to build the TRIPOLI-4® A-lite model is based on MCAM and the MCNP A-lite model (version 4.1). Simplified TBMs (from KIT) have been integrated in the equatorial-port. Comparisons of neutron wall loading, flux, nuclear heating and tritium production rate show a good agreement between the two codes. Discrepancies are mainly included in the Monte Carlo codes statistical error.
Recent Developments in the MCNP-POLIMI Postprocessing Code
Pozzi, S.A.
2004-12-17
The design and analysis of measurements performed with organic scintillators rely on the use of Monte Carlo codes to simulate the interaction of neutrons and photons, originating from fission and other reactions, with the materials present in the system and the radiation detectors. MCNP-PoliMi is a modification of the MCNP-4c code that models the physics of secondary particle emission from fission and other processes realistically. This characteristic allows for the simulation of the higher moments of the distribution of the number of neutrons and photons in a multiplying system. The present report describes the recent additions to the MCNP-PoliMi post-processing code. These include the simulation of detector dead time, multiplicity, and third order statistics.
MCNP load balancing and fault tolerance with PVM
McKinney, G.W.
1995-07-01
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.
Features of MCNP6 Relevant to Medical Radiation Physics
Hughes, H. Grady III; Goorley, John T.
2012-08-29
MCNP (Monte Carlo N-Particle) is a general-purpose Monte Carlo code for simulating the transport of neutrons, photons, electrons, positrons, and more recently other fundamental particles and heavy ions. Over many years MCNP has found a wide range of applications in many different fields, including medical radiation physics. In this presentation we will describe and illustrate a number of significant recently-developed features in the current version of the code, MCNP6, having particular utility for medical physics. Among these are major extensions of the ability to simulate large, complex geometries, improvement in memory requirements and speed for large lattices, introduction of mesh-based isotopic reaction tallies, advances in radiography simulation, expanded variance-reduction capabilities, especially for pulse-height tallies, and a large number of enhancements in photon/electron transport.
MCNP speed advances for boron neutron capture therapy
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.
Accelerating Pseudo-Random Number Generator for MCNP on GPU
NASA Astrophysics Data System (ADS)
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.
Geometry creation for MCNP by Sabrina and XSM
Van Riper, K.A.
1994-02-01
The Monte Carlo N-Particle transport code MCNP is based on a surface description of 3-dimensional geometry. Cells are defined in terms of boolean operations on signed quadratic surfaces. MCNP geometry is entered as a card image file containing coefficients of the surface equations and a list of surfaces and operators describing cells. Several programs are available to assist in creation of the geometry specification, among them Sabrina and the new ``Smart Editor`` code XSM. We briefly describe geometry creation in Sabrina and then discuss XSM in detail. XSM is under development; our discussion is based on the state of XSM as of January 1, 1994.
Cold neutron gain calculations for the NBSR using MCNP
Williams, R.E.; Rowe, J.M. ); Blau, M. )
1993-01-01
The existing D[sub 2]O-ice cold neutron source in the 20-MW National Bureau of Standards reactor (NBSR) will be replaced in 1994 with a liquid-hydrogen (LH2) source, to increase the yield of cold neutrons (X > 0.4 nm). Two series of Monte Carlo calculations using MCNP were performed to determine the optimum cold moderator geometry and to verify its performance. Only the region near the cryostat was modeled for the first series of calculations, leading to the choice of a spherical annulus for the LH[sub 2] source. A complete MCNP model of the core was subsequently developed.
Calculation of cell volumes and surface areas in MCNP
Hendricks, J.S.
1980-01-01
MCNP is a general Monte Carlo neutron-photon particle transport code which treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces, and some special fourth-degree surfaces. It is necessary to calculate cell volumes and surface areas so that cell masses, fluxes, and other important information can be determined. The volume/area calculation in MCNP computes cell volumes and surface areas for cells and surfaces rotationally symmetric about any arbitrary axis. 5 figures, 1 table.
Reactor physics verification of the MCNP6 unstructured mesh capability
Burke, T. P.; Kiedrowski, B. C.; Martz, R. L.; Martin, W. R.
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)
An Electron/Photon/Relaxation Data Library for MCNP6
Hughes, III, H. Grady
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.
Criticality Benchmark Results Using Various MCNP Data Libraries
Stephanie C. Frankle
1999-07-01
A suite of 86 criticality benchmarks has been recently implemented in MCNP{trademark} as part of the nuclear data validation effort. These benchmarks have been run using two sets of MCNP continuous-energy neutron data: ENDF/B-VI based data through Release 2 (ENDF60) and the ENDF/B-V based data. New evaluations were completed for ENDF/B-VI for a number of the important nuclides such as the isotopes of H, Be, C, N, O, Fe, Ni, {sup 235,238}U, {sup 237}Np, and {sup 239,240}Pu. When examining the results of these calculations for the five manor categories of {sup 233}U, intermediate-enriched {sup 235}U (IEU), highly enriched {sup 235}U (HEU), {sup 239}Pu, and mixed metal assembles, we find the following: (1) The new evaluations for {sup 9}Be, {sup 12}C, and {sup 14}N show no net effect on k{sub eff}; (2) There is a consistent decrease in k{sub eff} for all of the solution assemblies for ENDF/B-VI due to {sup 1}H and {sup 16}O, moving k{sub eff} further from the benchmark value for uranium solutions and closer to the benchmark value for plutonium solutions; (3) k{sub eff} decreased for the ENDF/B-VI Fe isotopic data, moving the calculated k{sub eff} further from the benchmark value; (4) k{sub eff} decreased for the ENDF/B-VI Ni isotopic data, moving the calculated k{sub eff} closer to the benchmark value; (5) The W data remained unchanged and tended to calculate slightly higher than the benchmark values; (6) For metal uranium systems, the ENDF/B-VI data for {sup 235}U tends to decrease k{sub eff} while the {sup 238}U data tends to increase k{sub eff}. The net result depends on the energy spectrum and material specifications for the particular assembly; (7) For more intermediate-energy systems, the changes in the {sup 235,238}U evaluations tend to increase k{sub eff}. For the mixed graphite and normal uranium-reflected assembly, a large increase in k{sub eff} due to changes in the {sup 238}U evaluation moved the calculated k{sub eff} much closer to the benchmark value. (8
NASA Astrophysics Data System (ADS)
Joo, Hanyoung; Lee, Arim; Kim, Rinah; Park, Chan Hee; Moon, Joo Hyun
2015-09-01
In this paper, a fiber-optic radiation sensor (FORS) was developed to measure gamma rays from the radionuclides frequently found in radioactively contaminated soil. The sensing probe of the FORS was made of an inorganic (Lu,Y)2SiO5:Ce (LYSO:Ce) scintillator, a mixture of epoxy resin and hardener and a plastic fiber. The FORS was applied to measure gamma rays from Cs-137 source (1.1 μCi) in a disk shape. Also, MCNP simulation was performed for the same geometry as that in the experimental setup. Comparison between measurements by the FORS and MCNP simulation showed that the detection efficiency of the fiber-optic sensor was about 19.2%. The FORS is expected to be useful in measuring gamma rays from the radioactive soil at nuclear facility site.
Impact of MCNP unresolved resonance probability-table treatment on uranium and plutonium benchmarks
Mosteller, R.D.; Little, R.C.
1998-12-31
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.
Fission Matrix Capability for MCNP Monte Carlo
Carney, Sean E.; Brown, Forrest B.; Kiedrowski, Brian C.; Martin, William R.
2012-09-05
In a Monte Carlo criticality calculation, before the tallying of quantities can begin, a converged fission source (the fundamental eigenvector of the fission kernel) is required. Tallies of interest may include powers, absorption rates, leakage rates, or the multiplication factor (the fundamental eigenvalue of the fission kernel, k{sub eff}). Just as in the power iteration method of linear algebra, if the dominance ratio (the ratio of the first and zeroth eigenvalues) is high, many iterations of neutron history simulations are required to isolate the fundamental mode of the problem. Optically large systems have large dominance ratios, and systems containing poor neutron communication between regions are also slow to converge. The fission matrix method, implemented into MCNP[1], addresses these problems. When Monte Carlo random walk from a source is executed, the fission kernel is stochastically applied to the source. Random numbers are used for: distances to collision, reaction types, scattering physics, fission reactions, etc. This method is used because the fission kernel is a complex, 7-dimensional operator that is not explicitly known. Deterministic methods use approximations/discretization in energy, space, and direction to the kernel. Consequently, they are faster. Monte Carlo directly simulates the physics, which necessitates the use of random sampling. Because of this statistical noise, common convergence acceleration methods used in deterministic methods do not work. In the fission matrix method, we are using the random walk information not only to build the next-iteration fission source, but also a spatially-averaged fission kernel. Just like in deterministic methods, this involves approximation and discretization. The approximation is the tallying of the spatially-discretized fission kernel with an incorrect fission source. We address this by making the spatial mesh fine enough that this error is negligible. As a consequence of discretization we get a
Gas Core Reactor Numerical Simulation Using a Coupled MHD-MCNP Model
NASA Technical Reports Server (NTRS)
Kazeminezhad, F.; Anghaie, S.
2008-01-01
Analysis is provided in this report of using two head-on magnetohydrodynamic (MHD) shocks to achieve supercritical nuclear fission in an axially elongated cylinder filled with UF4 gas as an energy source for deep space missions. The motivation for each aspect of the design is explained and supported by theory and numerical simulations. A subsequent report will provide detail on relevant experimental work to validate the concept. Here the focus is on the theory of and simulations for the proposed gas core reactor conceptual design from the onset of shock generations to the supercritical state achieved when the shocks collide. The MHD model is coupled to a standard nuclear code (MCNP) to observe the neutron flux and fission power attributed to the supercritical state brought about by the shock collisions. Throughout the modeling, realistic parameters are used for the initial ambient gaseous state and currents to ensure a resulting supercritical state upon shock collisions.
Certification of MCNP version 4A for WHC computer platforms
Carter, L.L., Westinghouse Hanford
1996-05-07
MCNP is a general-purpose Monte Carlo code that can 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).
MCNP6 enhancements of delayed-particle production
McKinney, G. W.
2012-07-01
Over the last decade, there has been an increased interest in the production of delayed-particle signatures from neutron and photon interactions with matter. To address this interest, various radiation transport codes have developed a wide range of delayed-particle physics packages. With the recent merger of the Monte Carlo transport codes MCNP5 and MCNPX, MCNP6 inherited the comprehensive model-based delayed-particle production capabilities developed in MCNPX over the last few years. An integral part of this capability consists of the depletion code CINDER90 which was incorporated into MCNPX in 2004. During this last year, significant improvements have been made to the MCNP6 physics and algorithms associated with delayed-particle production, including the development of a delayed-beta capability, an algorithm enhancement for the delayed-neutron treatment, and a database enhancement for delayed-gamma emission. The delayed-beta feature represents an important component in modeling background signals produced by active interrogation sources. Combined, these improvements provide MCNP6 with a flexible state-of-the-art physics package for generating high-fidelity signatures from fission and activation. This paper provides details of these enhancements and presents results for a variety of fission and activation examples. (authors)
Preliminary Benchmarking and MCNP Simulation Results for Homeland Security
Robert Hayes
2008-03-01
The purpose of this article is to create Monte Carlo N-Particle (MCNP) input stacks for benchmarked measurements sufficient for future perturbation studies and analysis. The approach was to utilize historical experimental measurements to recreate the empirical spectral results in MCNP, both qualitatively and quantitatively. Results demonstrate that perturbation analysis of benchmarked MCNP spectra can be used to obtain a better understanding of field measurement results which may be of national interest. If one or more spectral radiation measurements are made in the field and deemed of national interest, the potential source distribution, naturally occurring radioactive material shielding, and interstitial materials can only be estimated in many circumstances. The effects from these factors on the resultant spectral radiation measurements can be very confusing. If benchmarks exist which are sufficiently similar to the suspected configuration, these benchmarks can then be compared to the suspect measurements. Having these benchmarks with validated MCNP input stacks can substantially improve the predictive capability of experts supporting these efforts.
Implementation of on-the-fly doppler broadening in MCNP
Martin, W. R.; Wilderman, S.; Brown, F. B.; Yesilyurt, G.
2013-07-01
A new method to obtain Doppler broadened cross sections has been implemented into MCNP, removing the need to generate cross sections for isotopes at problem temperatures. When a neutron of energy E enters a material region that is at some temperature T, the cross sections for that material at temperature T are immediately obtained 'on-the-fly' (OTF) by interpolation using a high order functional expansion for the temperature dependence of the Doppler-broadened cross section for that isotope at the neutron energy E. The OTF cross sections agree with the NJOY-based cross sections for all neutron energies and all temperatures in the range specified by the user, e.g., 250 K - 3200 K. The OTF methodology has been successfully implemented into the MCNP Monte Carlo code and has been tested on several test problems by comparing MCNP with conventional ACE cross sections versus MCNP with OTF cross sections. The test problems include the Doppler defect reactivity benchmark suite and two full-core VHTR configurations, including one with multiphysics coupling using RELAP5-3D/ATHENA for the thermal-hydraulic analysis. The comparison has been excellent, verifying that the OTF libraries can be used in place of the conventional ACE libraries generated at problem temperatures. In addition, it has been found that the OTF methodology greatly reduces the complexity of the input for MCNP, resulting in an order of magnitude decrease in the number of input lines for full-core configurations. Finally, for full-core problems with multiphysics feedback, the memory required to store the cross section data is considerably reduced with OTF cross sections and the additional computational effort with OTF is modest, on the order of 10-15%. (authors)
Physics and Algorithm Enhancements for a Validated MCNP/X Monte Carlo Simulation Tool, Phase VII
McKinney, Gregg W
2012-07-17
Currently the US lacks an end-to-end (i.e., source-to-detector) radiation transport simulation code with predictive capability for the broad range of DHS nuclear material detection applications. For example, gaps in the physics, along with inadequate analysis algorithms, make it difficult for Monte Carlo simulations to provide a comprehensive evaluation, design, and optimization of proposed interrogation systems. With the development and implementation of several key physics and algorithm enhancements, along with needed improvements in evaluated data and benchmark measurements, the MCNP/X Monte Carlo codes will provide designers, operators, and systems analysts with a validated tool for developing state-of-the-art active and passive detection systems. This project is currently in its seventh year (Phase VII). This presentation will review thirty enhancements that have been implemented in MCNPX over the last 3 years and were included in the 2011 release of version 2.7.0. These improvements include 12 physics enhancements, 4 source enhancements, 8 tally enhancements, and 6 other enhancements. Examples and results will be provided for each of these features. The presentation will also discuss the eight enhancements that will be migrated into MCNP6 over the upcoming year.
Conversion of Input Data between KENO V.a and MCNP File Formats, Version 5L.
Energy Science and Technology Software Center (ESTSC)
2007-10-31
Version 00 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 SCALEmore » 5.0 or SCALE 5.1 cross section files will work with this release. This version of KENO2MCNP was tested with CCC-730/MCNP5 1.40 and with CCC-725/SCALE5.0 and CCC-732/SCALE 5.1. Note that this distribution does not include either MCNP or SCALE, which are available separately through either RSICC or the NEA Data Bank.« less
Results of the MCNP analysis of 20/20 LEU fuel for the Oregon State University TRIGA Reactor
Dodd, B.; Klein, A.C.; Lewis, B.R.; Merritt, P.A.
1995-12-31
The Monte Carlo Neutron/Photon (MCNP) code has been used to perform the neutronics analysis required to support revision of the Oregon State University TRIGA Reactor (OSTR) Safety Analysis Report (SAR). The SAR revision is a necessary part of the preparation of the application for authorization to convert the OSTR core from High Enriched Uranium (HEU) FLIP fuel to a Low Enriched Uranium (LEU) fuel. Before MCNP was applied to LEU-fueled cores, it was first validated by comparing MCNP calculations on FLIP cores to historical, measured values for these cores. The LEU fuel considered was the 20 wt%, 20% enriched (20/20) TRIGA fuel approved by the Nuclear Regulatory Commission (NRC) in NUREG 1282. The results show that the 20/20 fuel is much more reactive than FLIP fuel. A just-critical OSTR FLIP core contains 65 elements, while a just-critical 20/20 core only needs 51 elements. Similarly, the current operational FLIP core consists of 88 elements, whereas a 20/20 core giving the same core excess only requires 65 elements. This presents a significant problem for the OSTR because of potentially significant neutron flux loss in experimental facilities. Further analysis shows that to achieve a full size operational core of about 90 LEU elements the erbium content of the LEU fuel would need to be increased from 0.47wt% to about 0.85 wt%.
Results of the MCNP analysis of 20/20 LEU fuel for the Oregon State University TRIGA reactor
Dodd, B.; Klein, A.C.; Lewis, B.R.; Merritt, P.A
1994-07-01
The Monte Carlo Neutron/Photon (MCNP) code has been used to perform the neutronics analysis required to support revision of the Oregon State University TRIGA Reactor (OSTR) Safety Analysis Report (SAR). The SAR revision is a necessary part of the preparation of the application for authorization to convert the OSTR core from High Enriched Uranium (HEU) FLIP fuel to a Low Enriched Uranium (LEU) fuel. Before MCNP was applied to LEU-fueled cores, it was first validated by comparing MCNP calculations on FLIP cores to historical, measured values for these cores. The LEU fuel considered was the 20 wt%, 20 % enriched (20/20) TRIGA fuel approved by the Nuclear Regulatory Commission (NRC) in NUREG 1282. The results show that the 20/20 fuel is much more reactive than FLIP fuel. A just-critical OSTR FLIP core contains 65 elements, while a just-critical 20/20 core only needs 51 elements. Similarly, the current operational FLIP core consists of 88 elements, whereas a 20/20 core giving the same core excess only requires 65 elements. This presents a significant problem for the OSTR because of potentially significant neutron flux loss in experimental facilities. Further analysis shows that to achieve a full size operational core of about 90 LEU elements the erbium content of the LEU fuel would need to be increased from 0.47 wt% to about 0.85 wt%. (author)
Benchmarking ENDF/B-VII.1, JENDL-4.0 and JEFF-3.1.1 with MCNP6
NASA Astrophysics Data System (ADS)
van der Marck, Steven C.
2012-12-01
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 6Li, 7Li, 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
Benchmarking ENDF/B-VII.1, JENDL-4.0 and JEFF-3.1.1 with MCNP6
Marck, Steven C. van der
2012-12-15
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 {sup 6}Li, {sup 7}Li, Be, C, N, O, Mg, Al, Ti, Fe, Pb, D2O, H2O, concrete, polyethylene and teflon). The new functionality in MCNP6 to calculate the effective delayed neutron fraction was tested by comparison with more than thirty measurements in widely varying systems. Among these were measurements in the Tank Critical Assembly (TCA in Japan) and IPEN/MB-01 (Brazil), both with a thermal spectrum, two cores in Masurca (France) and three cores in the Fast Critical Assembly (FCA, Japan), all with fast spectra. The performance of the three libraries, in combination with MCNP6, is shown to be good. The results for the LEU-COMP-THERM category are on average very close to the benchmark value. Also for most other categories the results are satisfactory. Deviations from the benchmark values do occur in certain benchmark series, or in isolated cases within benchmark series
A Verification of MCNP6 FMESH Tally Capabilities
Swift, Alicia L.; McKigney, Edward A.; Schirato, Richard C.; Robinson, Alex Philip; Temple, Brian Allen
2015-02-10
This work serves to verify the MCNP6 FMESH capability through comparison to two types of data. FMESH tallies, binned in time, were generated on an ideal detector face for neutrons undergoing a single scatter in a graphite target. For verification, FMESH results were compared to analytic calculations of the nonrelativistic TOF for elastic and inelastic single neutron scatters (TOF for the purposes of this paper is the time for a neutron to travel from its scatter location in the graphite target to the detector face). FMESH tally results were also compared to F4 tally results, an MNCP tally that calculates fluence in the same way as the FMESH tally. The FMESH tally results agree well with the analytic results and the F4 tally; hence, it is believed that, for simple geometries, MCNP6 FMESH tallies represent the physics of neutron scattering very well.
Multi-canister overpack project -- verification and validation, MCNP 4A
Goldmann, L.H.
1997-11-10
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.
Neutron and photon shielding benchmark calculations by MCNP on the LR-0 experimental facility.
Hordósy, G
2005-01-01
In the framework of the REDOS project, the space-energy distribution of the neutron and photon flux has been calculated over the pressure vessel simulator thickness of the LR-0 experimental reactor, Rez, Czech Republic. The results calculated by the Monte Carlo code MCNP4C are compared with the measurements performed in the Nuclear Research Institute, Rez. The spectra have been measured at the barrel, in front of, inside and behind the pressure vessel in different configurations. The neutron measurements were performed in the energy range 0.1-10 MeV. This work has been done in the frame of the 5th Frame Work Programme of the European Community 1998-2002. PMID:16604591
MCNP-DSP calculations of measurements with uranyl nitrate solution system
Valentine, T.E.
1998-09-01
The {sup 252}Cf-source-driven noise analysis method has been used to determine the subcriticality of various configurations of fissile materials. In the past, the application of this method was limited because point-kinetics models had to be used to interpret the data; however, with the development of the Monte Carlo code MCNP-DSP, the measurements can be analyzed using the more general Monte Carlo models. The results of the Monte carlo calculations will be dependent on the ability to model the experiment accurately and on the nuclear data used to perform the calculations. This paper presents a comparison of the measured and calculated ratio of spectral densities for a subset of measurements performed with a uranyl nitrate solution tank filled to various heights. The results presented are for calculations that were performed with both ENDF/B-IV and ENDF/B-V cross-section data sets.
Comparison of discrete and continuous thermal neutron scattering treatments in MCNP5
Pavlou, A. T.; Brown, F. B.; Martin, W. R.; Kiedrowski, B. C.
2012-07-01
The standard discrete thermal neutron S({alpha},{beta}) scattering treatment in MCNP5 is compared with a continuous S({alpha},{beta}) scattering treatment using a criticality suite of 119 benchmark cases and ENDF/B-VII.0 nuclear data. In the analysis, six bound isotopes are considered: beryllium metal, graphite, hydrogen in water, hydrogen in polyethylene, beryllium in beryllium oxide and oxygen in beryllium oxide. Overall, there are only small changes in the eigenvalue (k{sub eff}) between discrete and continuous treatments. In the comparison of 64 cases that utilize S({alpha},{beta}) scattering, 62 agreed at the 95% confidence level, and the 2 cases with differences larger than 3 {sigma} agreed within 1 {sigma} when more neutrons were run in the calculations. The results indicate that the changes in eigenvalue between continuous and discrete treatments are random, small, and well within the uncertainty of measured data for reactor criticality experiments. (authors)
MCNP/X TRANSPORT IN THE TABULAR REGIME
HUGHES, H. GRADY
2007-01-08
The authors review the transport capabilities of the MCNP and MCNPX Monte Carlo codes in the energy regimes in which tabular transport data are available. Giving special attention to neutron tables, they emphasize the measures taken to improve the treatment of a variety of difficult aspects of the transport problem, including unresolved resonances, thermal issues, and the availability of suitable cross sections sets. They also briefly touch on the current situation in regard to photon, electron, and proton transport tables.
Characteristics of multiprocessing MCNP5 on small personal computer clusters
Robinson, Sean M.; McConn, Ronald J.; Pagh, Richard T.; Schweppe, John E.; Siciliano, Edward R.
2006-06-05
The feasibility and efficiency of performing MCNP5 calculations with a small, heterogeneous computing cluster built from Microsoft® Windows™ Personal Computers (PCs) are explored. The performance increases that may be expected with such clusters are estimated. 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.
MCNP6 fragmentation of light nuclei at intermediate energies
NASA Astrophysics Data System (ADS)
Mashnik, Stepan G.; Kerby, Leslie M.
2014-11-01
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 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.
Yankee Rowe isotopics benchmark using MCNP-XT
Xu, Z.; Whitmer, C.
2013-07-01
The Yankee Rowe spent fuel isotopic data provides a valuable source to benchmark the burnup calculations as part of verification and validation (V and V) efforts for the TerraPower's Monte Carlo depletion code, MCNP-XT. A total of 71 fuel rods were selected in the Yankee Rowe isotopic measurements covering a burnup range up to 44 MWd/kg ({approx}4.4%) under both the asymptotic spectrum and the non-asymptotic spectrum. The MCNP-XT pin cell depletion provides a comparison against the asymptotic spectrum measurement; and full assembly depletion with 322 depletion materials provides comparisons against various non-asymptotic depletion conditions. All calculations are performed based on the recent ENDF/B-VII.O data. Furthermore, the Monte Carlo depletion uncertainties and biases were examined showing their effect as insignificant. The set of burnup calculations cover the scattered experimental measurements demonstrating excellent agreement with the measured values. This benchmark exercise demonstrates the depletion analysis capability of the MCNP-XT code and validates the low burnup range. (authors)
Code System for Generation of Input Data for MCNP.
Energy Science and Technology Software Center (ESTSC)
1998-07-16
Version 00 The MSM-SOURCE code was designed for quick and easy estimations of basic stopping characteristics of proton transmission, for generation of the source definition (SDEF) portion of the input data for MCNP (for 3b- and 4- versions) [2], simulating the set of single neutron sources, produced in the sample during the proton transmission. It does not generate the ful MCNP input file. The results of calculations well reproduce the experimental data [3]. It permitsmore » one to extend the possibilities of the MCNP code for consideration of secondary neutrons from the proton interaction with nuclei of the sample substance. The MSM-SOURCE code is applicable for calculations of the proton transport for the incident energies from 0.1 to 1 GeV and various targets 12 < A < 238. This code is based of the Moving Source Model (MSM) (using the original parametrization [3],[4]) and Bethe stopping theory with the relativistic corrections for protons. It allows the estimations of the proton range, the changes of the proton current and the neutron production versus the depth. The double differential spectra and the multiplicities of nucleons, produced in the primary proton-induced reactions, are obtained. For the evaluation of inelastic cross section the original parametrization is used [4].« less
ACTI - An MCNP Data Library for Prompt Gamma-Ray Spectroscopy.
Frankle, S. C.; Reedy, R. C.; Young, P. G. ,
2002-01-01
Prompt gamma-ray spectroscopy is used in a wide variety of applications for determining material compositions. High-quality photon-production data from thermal-neutron capture reactions are essential for these applications. Radiation transport codes, such as MCNP{trademark}, are often used to design detector systems, determine minimum detection thresholds, etc. These transport codes rely on evaluated nuclear databases such as ENDF (Evaluated Nuclear Data File) to provide the fundamental data used in the transport calculations. Often the photon-production data from incident neutron reactions in the evaluations are of relatively poor quality. We have compiled the best experimental data for thermal-neutron capture for the naturally occurring isotopes for elements from H through Zn as well as for {sup 70,72,73,74,76}Ge, {sup 149}Sm, {sup 155,157}Gd, {sup 181}Ta and {sup 182,183,184,186}W. This compilation has been used to update the ENDF evaluations for {sup 1}H, {sup 4}He, {sup 9}Be, {sup 14}N, {sup 16}O, {sup 19}F, Na, Mg, {sup 27}Al, {sup 32}S, S, {sup 35,37}Cl, K, Ca, {sup 45}Sc, Ti, {sup 51}V, {sup 50,52,53,54}Cr, {sup 55}Mn, {sup 54,56,57,58}Fe, {sup 58,60,61,62,64}Ni, {sup 63,65}Cu and {sup 182,183,184,186}W. In addition, the inelastic cross sections and corresponding secondary-photon distributions were updated for {sup 16}O. Complete new evaluations were submitted to ENDF for {sup 35,37}Cl. This paper will discuss the evaluation effort and the production of the MCNP data library, ACTI, based on the new evaluations. Data from the ENDF evaluations for {sup 28-30}Si were also included in the ACTI library for completeness. The silicon evaluations were updated in 1997 and include the latest experimental data for radiative capture.
Negron, S.B.; Lee, B.L. Jr.; Tayloe, R.W. Jr.
1996-01-01
This document has been prepared to allow use of the Radiation Shielding and Information Center (RSIC) release of MCNP4a, which has been installed on the Battelle Process Safety and Risk Management (PSRM) IBM RS/6000 workstation, for production calculations for the Portsmouth Gaseous Diffusion Plant (PORTS). This hardware/software configuration is under the configuration control plan listed in Reference 1. The first portion of this document outlines basic information with regard to validation of MCNP4a using the supplied cross sections and the additional MCNPDAT cross sections. A basic discussion of MCNP is provided, along with discussions of the validation database in general. A description of the statistical analysis then follows. The results of this validation indicate that the software and data libraries examined may be used with confidence for criticality calculations at the Portsmouth Gaseous Diffusion Plant (PORTS). When the validation results are treated as a single group, there is a 95% confidence that 99.9% of future calculations of similar critical systems will have a calculated k{sub eff} > 0.95. Based on this result, the Battelle PSRM Nuclear Safety Group has adopted the calculational acceptance criteria that a calculated k{sub eff} + 2{sigma}, {le} 0.95 is safely subcritical. The conclusion of this document is that MCNP4a and all associated cross section libraries installed on the PSRM IBM RS/6000 are acceptable for use in performing production criticality safety calculations for the Portsmouth Gaseous Diffusion Plant.
Technical Note: Improved implementation of doppler broadening in MCNP5
Bartol, Laura J.; DeWerd, Larry A.
2012-09-15
Purpose: Incoherent scattering has a substantial effect on spectroscopic measurements and simulations. Many general-purpose Monte Carlo codes include models that account for the effects of bound electrons on incoherent scattering, including Doppler broadening (DB). This work investigates the DB model used in the Monte Carlo N-particle transport code (MCNP5). Methods: Simulations were run with three versions of MCNP5: v1.51, v1.60, and a modified form of v1.60 (v1.60m). All simulations used the MCPLIB04 photon data library, which presents the electron subshell data for incoherent scattering in the form of a probability density function. In v1.60m, the source code was altered to sample the electron subshell from a cumulative density function instead. Each version of the code was tested using an identical set of simulations that investigated DB in a slab of silicon at scattering angles of 15 Degree-Sign , 30 Degree-Sign , and 45 Degree-Sign . For each angle, simulations were run for multiple energies between 200 keV and 800 keV. The spectrum of singly-scattered photons at the exit of the slab was scored. Spectra were analytically calculated for comparison. Results: In v1.51, DB was modeled for incident photon energies below 760 keV, 384 keV, and 260 keV at scattering angles of 15 Degree-Sign , 30 Degree-Sign , and 45 Degree-Sign , respectively. Above these energy thresholds, v1.51 did not model DB. The spectra calculated using v1.60 and v1.60m exhibited DB for all energy-angle combinations; however, v1.60m, exhibited more energy broadening than did v1.60. The spectra calculated with v1.60m agreed with the analytical calculations. Conclusions: MCNP5 v1.51 and v1.60 model partial broadening when used with the MCPLIB04 data library. MCNP5 v1.60m models DB more accurately due to the form of the electron subshell data. In response to these results, Los Alamos National Laboratory has released a new photon data library, MCPLIB84, that presents the electron subshell data in
Radiation calculations using LAHET/MCNP/CINDER90
Waters, L.
1994-10-01
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.
Radiation calculations using LAHET/MCNP/CINDER90
Waters, L.S.
1993-08-01
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.
Modelling of dynamic experiments in MCNP5 environment.
Mosorov, Volodymyr; Zych, Marcin; Hanus, Robert; Petryka, Leszek
2016-06-01
The design of radiation measurement systems includes a modelling phase which ascertains the best 3D geometry for a projected gauge. To simulate measured counts by a detector, the widely-used rigorous phenomenological model is used. However, this model does not consider possible source or/and detector movement during a measurement interval. Therefore, the phenomenological model has been successfully modified in order to consider such a displacement during the time sampling interval in dynamic experiments. To validate the proposed model, a simple radiation system was accurately implemented in the MCNP5 code. The experiments confirmed the accuracy of the proposed model. PMID:27058321
Heart simulation with surface equations for using on MCNP code
Rezaei-Ochbelagh, D.; Salman-Nezhad, S.; Asadi, A.; Rahimi, A.
2011-12-26
External photon beam radiotherapy is carried out in a way to achieve an 'as low as possible' a dose in healthy tissues surrounding the target. One of these surroundings can be heart as a vital organ of body. As it is impossible to directly determine the absorbed dose by heart, using phantoms is one way to acquire information around it. The other way is Monte Carlo method. In this work we have presented a simulation of heart geometry by introducing of different surfaces in MCNP code. We used 14 surface equations in order to determine human heart modeling. Those surfaces are borders of heart walls and contents.
Heart simulation with surface equations for using on MCNP code
NASA Astrophysics Data System (ADS)
Rezaei-Ochbelagh, D.; Salman-Nezhad, S.; Asadi, A.; Rahimi, A.
2011-12-01
External photon beam radiotherapy is carried out in a way to achieve an "as low as possible" a dose in healthy tissues surrounding the target. One of these surroundings can be heart as a vital organ of body. As it is impossible to directly determine the absorbed dose by heart, using phantoms is one way to acquire information around it. The other way is Monte Carlo method. In this work we have presented a simulation of heart geometry by introducing of different surfaces in MCNP code. We used 14 surface equations in order to determine human heart modeling. Those surfaces are borders of heart walls and contents.
MCNP5 CALCULATIONS REPLICATING ARH-600 NITRATE DATA
FINFROCK SH
2011-10-25
This report serves to extend the previous document: 'MCNP Calculations Replicating ARH-600 Data' by replicating the nitrate curves found in ARH-600. This report includes the MCNP models used, the calculated critical dimension for each analyzed parameter set, and the resulting data libraries for use with the CritView code. As with the ARH-600 data, this report is not meant to replace the analysis of the fissile systems by qualified criticality personnel. The M CNP data is presented without accounting for the statistical uncertainty (although this is typically less than 0.001) or bias and, as such, the application of a reasonable safety margin is required. The data that follows pertains to the uranyl nitrate and plutonium nitrate spheres, infinite cylinders, and infinite slabs of varying isotopic composition, reflector thickness, and molarity. Each of the cases was modeled in MCNP (version 5.1.40), using the ENDF/B-VI cross section set. Given a molarity, isotopic composition, and reflector thickness, the fissile concentration and diameter (or thicknesses in the case of the slab geometries) were varied. The diameter for which k-effective equals 1.00 for a given concentration could then be calculated and graphed. These graphs are included in this report. The pages that follow describe the regions modeled, formulas for calculating the various parameters, a list of cross-sections used in the calculations, a description of the automation routine and data, and finally the data output. The data of most interest are the critical dimensions of the various systems analyzed. This is presented graphically, and in table format, in Appendix B. Appendix C provides a text listing of the same data in a format that is compatible with the CritView code. Appendices D and E provide listing of example Template files and MCNP input files (these are discussed further in Section 4). Appendix F is a complete listing of all of the output data (i.e., all of the analyzed dimensions and the
Pozzi, S.A.
2002-01-17
The present simulations performed with the Monte Carlo code MCNP-POLIMI [1] have the scope of evaluating the associated-particle sealed tube neutron generator (APSTNG) for use as an interrogation source in the source-driven noise analysis method for the assay of nuclear materials. In the Nuclear Materials Identification System (NMIS) developed at the Oak Ridge National Laboratory, the time dependent cross-correlation of the timed neutron source and detector responses is one of the signatures acquired. Previous studies and measurements have demonstrated the sensitivity of this and other related signatures to fissile mass [2-3]. In a recent report [4], we outlined the advantages of the APSTNG interrogation source for use with NMIS when compared with the Cf-252 source. In particular, we showed that when the distance between the source and the sample and the sample and the detectors is large, the APSTNG source outperforms the Cf-252 in sensitivity to fissile mass. This is the case when performing measurements of items that are placed inside containers. The purpose of this report is to investigate the advantages of using the APSTNG source in reducing the effect of floor reflections in the signatures acquired. To this end, a large number of MCNP-POLIMI Monte Carlo simulations were performed to obtain source-detector covariance functions.
MCNP: a general Monte Carlo code for neutron and photon transport
Forster, R.A.; Godfrey, T.N.K.
1985-01-01
MCNP is a very general Monte Carlo neutron photon transport code system with approximately 250 person years of Group X-6 code development invested. It is extremely portable, user-oriented, and a true production code as it is used about 60 Cray hours per month by about 150 Los Alamos users. It has as its data base the best cross-section evaluations available. MCNP contains state-of-the-art traditional and adaptive Monte Carlo techniques to be applied to the solution of an ever-increasing number of problems. Excellent user-oriented documentation is available for all facets of the MCNP code system. Many useful and important variants of MCNP exist for special applications. The Radiation Shielding Information Center (RSIC) in Oak Ridge, Tennessee is the contact point for worldwide MCNP code and documentation distribution. A much improved MCNP Version 3A will be available in the fall of 1985, along with new and improved documentation. Future directions in MCNP development will change the meaning of MCNP to Monte Carlo N Particle where N particle varieties will be transported.
Standard Neutron, Photon, and Electron Data Libraries for MCNP4B.
Energy Science and Technology Software Center (ESTSC)
1997-04-01
Version 00 US DOE 10CFR810 Jurisdiction. DLC-189/MCNPXS is for use with Version 4B and later of the MCNP transport code. This data library provides a comprehensive set of cross sections for a wide range of radiation transport applications using the Monte Carlo code package CCC-660/MCNP4B.
The MCNP-DSP code for calculations of time and frequency analysis parameters for subcritical systems
Valentine, T.E.; Mihalczo, J.T.
1995-12-31
This paper describes a modified version of the MCNP code, the MCNP-DSP. Variance reduction features were disabled to have strictly analog particle tracking in order to follow fluctuating processes more accurately. Some of the neutron and photon physics routines were modified to better represent the production of particles. Other modifications are discussed.
Electron photon verification calculations using MCNP4B
Gierga, D.P.; Adams, K.J.
1998-07-01
MCNP4B 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. Three sets of bremsstrahlung experiments were simulated. The first verification calculations for bremsstrahlung production used the experimental results in Faddegon for 15 MeV electrons incident on lead, aluminum, and beryllium targets. The calculated integrated bremsstrahlung yields, the bremsstrahlung energy spectra, and the mean energy of the bremsstrahlung beam were compared with experiment. The impact of several MCNP tally options and physics parameters was explored in detail. The second was the experiment of O`Dell which measured the bremsstrahlung spectra from 10 and 20.9 MeV electrons incident on a gold/tungsten target. The final set was a comparison of relative experimental spectra with calculated results for 9.66 MeV electrons incident on tungsten based on the experiment of Starfelt and Koch. The transmission experiments of Ebert were also studied, including comparisons of transmission coefficients for 10.2 MeV electrons incident on carbon, silver, and uranium foils. The agreement between experiment and simulation was usually within two standard deviations of the experimental and calculational errors.
MCNP/MCNPX model of the annular core research reactor.
DePriest, Kendall Russell; Cooper, Philip J.; Parma, Edward J., Jr.
2006-10-01
Many experimenters at the Annular Core Research Reactor (ACRR) have a need to predict the neutron/gamma environment prior to testing. In some cases, the neutron/gamma environment is needed to understand the test results after the completion of an experiment. In an effort to satisfy the needs of experimenters, a model of the ACRR was developed for use with the Monte Carlo N-Particle transport codes MCNP [Br03] and MCNPX [Wa02]. The model contains adjustable safety, transient, and control rods, several of the available spectrum-modifying cavity inserts, and placeholders for experiment packages. The ACRR model was constructed such that experiment package models can be easily placed in the reactor after being developed as stand-alone units. An addition to the 'standard' model allows the FREC-II cavity to be included in the calculations. This report presents the MCNP/MCNPX model of the ACRR. Comparisons are made between the model and the reactor for various configurations. Reactivity worth curves for the various reactor configurations are presented. Examples of reactivity worth calculations for a few experiment packages are presented along with the measured reactivity worth from the reactor test of the experiment packages. Finally, calculated neutron/gamma spectra are presented.
Valentine, T.E.; Mihalczo, J.T.
1995-12-31
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.
NASA Astrophysics Data System (ADS)
Sharma, Manish K.; Alajo, Ayodeji Babatunde; Liu, Xin
2014-12-01
The shielding of a neutron generator producing fast neutrons should be sufficient to limit the dose rates to the prescribed values. A deuterium-deuterium neutron generator has been installed in the Nuclear Engineering Department at Missouri University of Science and Technology (Missouri S&T). The generator produces fast neutrons with an approximate energy of 2.5 MeV. The generator is currently shielded with different materials like lead, high-density polyethylene, and borated polyethylene. An MCNP transport simulation has been performed to estimate the dose rates at various places in and around the facility. The simulations incorporated the geometric and composition information of these shielding materials to determine neutron and photon dose rates at three central planes passing through the neutron source. Neutron and photon dose rate contour plots at these planes were provided using a MATLAB program. Furthermore, the maximum dose rates in the vicinity of the facility were used to estimate the annual limit for the generator's hours of operation. A successful operation of this generator will provide a convenient neutron source for basic and applied research at the Nuclear Engineering Department of Missouri S&T.
MCNP simulation of a Theratron 780 radiotherapy unit.
Miró, R; Soler, J; Gallardo, S; Campayo, J M; Díez, S; Verdú, G
2005-01-01
A Theratron 780 (MDS Nordion) 60Co radiotherapy unit has been simulated with the Monte Carlo code MCNP. The unit has been realistically modelled: the cylindrical source capsule and its housing, the rectangular collimator system, both the primary and secondary jaws and the air gaps between the components. Different collimator openings, ranging from 5 x 5 cm2 to 20 x 20 cm2 (narrow and broad beams) at a source-surface distance equal to 80 cm have been used during the study. In the present work, we have calculated spectra as a function of field size. A study of the variation of the electron contamination of the 60Co beam has also been performed. PMID:16604598
Fuel Element Transfer Cask Modelling Using MCNP Technique
NASA Astrophysics Data System (ADS)
Darmawan, Rosli; Topah, Budiman Naim
2010-01-01
After operating for more than 25 years, some of the Reaktor TRIGA Puspati (RTP) fuel elements would have been depleted. A few addition and fuel reconfiguration exercises have to be conducted in order to maintain RTP capacity. Presently, RTP spent fuels are stored at the storage area inside RTP tank. The need to transfer the fuel element outside of RTP tank may be prevalence in the near future. The preparation shall be started from now. A fuel element transfer cask has been designed according to the recommendation by the fuel manufacturer and experience of other countries. A modelling using MCNP code has been conducted to analyse the design. The result shows that the design of transfer cask fuel element is safe for handling outside the RTP tank according to recent regulatory requirement.
Fuel Element Transfer Cask Modelling Using MCNP Technique
Darmawan, Rosli; Topah, Budiman Naim
2010-01-05
After operating for more than 25 years, some of the Reaktor TRIGA Puspati (RTP) fuel elements would have been depleted. A few addition and fuel reconfiguration exercises have to be conducted in order to maintain RTP capacity. Presently, RTP spent fuels are stored at the storage area inside RTP tank. The need to transfer the fuel element outside of RTP tank may be prevalence in the near future. The preparation shall be started from now. A fuel element transfer cask has been designed according to the recommendation by the fuel manufacturer and experience of other countries. A modelling using MCNP code has been conducted to analyse the design. The result shows that the design of transfer cask fuel element is safe for handling outside the RTP tank according to recent regulatory requirement.
Treating electron transport in MCNP{sup trademark}
Hughes, H.G.
1996-12-31
The transport of electrons and other charged particles is fundamentally different from that of neutrons and photons. A neutron, in aluminum slowing down from 0.5 MeV to 0.0625 MeV will have about 30 collisions; a photon will have fewer than ten. An electron with the same energy loss will undergo 10{sup 5} individual interactions. This great increase in computational complexity makes a single- collision Monte Carlo approach to electron transport unfeasible for many situations of practical interest. Considerable theoretical work has been done to develop a variety of analytic and semi-analytic multiple-scattering theories for the transport of charged particles. The theories used in the algorithms in MCNP are the Goudsmit-Saunderson theory for angular deflections, the Landau an theory of energy-loss fluctuations, and the Blunck-Leisegang enhancements of the Landau theory. In order to follow an electron through a significant energy loss, it is necessary to break the electron`s path into many steps. These steps are chosen to be long enough to encompass many collisions (so that multiple-scattering theories are valid) but short enough that the mean energy loss in any one step is small (for the approximations in the multiple-scattering theories). The energy loss and angular deflection of the electron during each step can then be sampled from probability distributions based on the appropriate multiple- scattering theories. This subsumption of the effects of many individual collisions into single steps that are sampled probabilistically constitutes the ``condensed history`` Monte Carlo method. This method is exemplified in the ETRAN series of electron/photon transport codes. The ETRAN codes are also the basis for the Integrated TIGER Series, a system of general-purpose, application-oriented electron/photon transport codes. The electron physics in MCNP is similar to that of the Integrated TIGER Series.
Comparisons between MCNP, EGS4 and experiment for clinical electron beams.
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. PMID:10211804
An enhanced geometry-independent mesh weight window generator for MCNP
Evans, T.M.; Hendricks, J.S.
1997-12-31
A new, enhanced, weight window generator suite has been developed for MCNP{trademark}. The new generator correctly estimates importances in either an user-specified, geometry-independent orthogonal grid or in MCNP geometric cells. The geometry-independent option alleviates the need to subdivide the MCNP cell geometry for variance reduction purposes. In addition, the new suite corrects several pathologies in the existing MCNP weight window generator. To verify the correctness of the new implementation, comparisons are performed with the analytical solution for the cell importance. Using the new generator, differences between Monte Carlo generated and analytical importances are less than 0.1%. Also, assumptions implicit in the original MCNP generator are shown to be poor in problems with high scattering media. The new generator is fully compatible with MCNP`s AVATAR{trademark} automatic variance reduction method. The new generator applications, together with AVATAR, gives MCNP an enhanced suite of variance reduction methods. The flexibility and efficacy of this suite is demonstrated in a neutron porosity tool well-logging problem.
Impact of MCNP unresolved resonance probability-table treatment on {sup 233}U benchmarks
Mosteller, R.D.
1999-06-01
Previous versions of the MCNP Monte Carlo code, up through and including MCNP4B, have not accurately modeled neutron self-shielding effects in the unresolved resonance energy region. Recently, a probability-table treatment has been incorporated into an intermediate version called MCNP4XS, and compatible continuous-energy cross-section libraries have been developed for 27 different isotopes. Preliminary results for a variety of uranium and plutonium benchmarks have been presented previously, and this paper extends those results to include several {sup 233}U benchmarks. The objective of the current study is to assess the reactivity impact of the probability-table treatment on {sup 233}U systems.
Impact of MCNP Unresolved Resonance Probability-Table Treatment on Uranium and Plutonium Benchmarks
Mosteller, R.D.; Little, R.C.
1999-09-20
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 {sup 233}U systems with intermediate spectra. More importantly, it can produce substantial reactivity increases for systems with large amounts of {sup 238}U and intermediate spectra.
a New Method for Neutron Capture Therapy (nct) and Related Simulation by MCNP4C Code
NASA Astrophysics Data System (ADS)
Shirazi, Mousavi; Alireza, Seyed; Ali, Taheri
2010-01-01
Neutron capture therapy (NCT) is enumerated as one of the most important methods for treatment of some strong maladies among cancers in medical science thus is unavoidable controlling and protecting instances in use of this science. Among of treatment instances of this maladies with use of nuclear medical science is use of neutron therapy that is one of the most important and effective methods in treatment of cancers. But whereas fast neutrons have too destroyer effects and also sake of protection against additional absorbed energy (absorbed dose) by tissue during neutron therapy and also naught damaging to rest of healthy tissues, should be measured absorbed energy by tissue accurately, because destroyer effects of fast neutrons is almost quintuple more than gamma photons. In this article for neutron therapy act of male's liver has been simulated a system by the Monte Carlo method (MCNP4C code) and also with use of analytical method, thus absorbed dose by this tissue has been obtained for sources with different energies accurately and has been compared results of this two methods together.
Current status of MCNP6 as a simulation tool useful for space and accelerator applications
Mashnik, Stepan G; Bull, Jeffrey S; Hughes, H. Grady; Prael, Richard E; Sierk, Arnold J
2012-07-20
For the past several years, a major effort has been undertaken at Los Alamos National Laboratory (LANL) to develop the transport code MCNP6, the latest LANL Monte-Carlo transport code representing a merger and improvement of MCNP5 and MCNPX. We emphasize a description of the latest developments of MCNP6 at higher energies to improve its reliability in calculating rare-isotope production, high-energy cumulative particle production, and a gamut of reactions important for space-radiation shielding, cosmic-ray propagation, and accelerator applications. We present several examples of validation and verification of MCNP6 compared to a wide variety of intermediate- and high-energy experimental data on reactions induced by photons, mesons, nucleons, and nuclei at energies from tens of MeV to about 1 TeV/nucleon, and compare to results from other modern simulation tools.
A Patch to MCNP5 for Multiplication Inference: Description and User Guide
Solomon, Jr., Clell J.
2014-05-05
A patch to MCNP5 has been written to allow generation of multiple neutrons from a spontaneous-fission event and generate list-mode output. This report documents the implementation and usage of this patch.
Comparisons of TORT and MCNP dose calculations for BNCT treatment planning
Ingersol, D.T.; Slater, C.O.; Williams, L.R.; Redmond, E.L., II; Zamenhof, R.G.
1996-12-31
The relative merit of using a deterministic code to calculate dose distributions for BNCT applications were examined. The TORT discrete deterministic ordinated code was used in comparison to MCNP4A to calculate dose distributions for BNCT applications
Validation of MCNP6.1 for Criticality Safety of Pu-Metal, -Solution, and -Oxide Systems
Kiedrowski, Brian C.; Conlin, Jeremy Lloyd; Favorite, Jeffrey A.; Kahler, III, Albert C.; Kersting, Alyssa R.; Parsons, Donald K.; Walker, Jessie L.
2014-05-13
Guidance is offered to the Los Alamos National Laboratory Nuclear Criticality Safety division towards developing an Upper Subcritical Limit (USL) for MCNP6.1 calculations with ENDF/B-VII.1 nuclear data for three classes of problems: Pu-metal, -solution, and -oxide systems. A benchmark suite containing 1,086 benchmarks is prepared, and a sensitivity/uncertainty (S/U) method with a generalized linear least squares (GLLS) data adjustment is used to reject outliers, bringing the total to 959 usable benchmarks. For each class of problem, S/U methods are used to select relevant experimental benchmarks, and the calculational margin is computed using extreme value theory. A portion of the margin of sub criticality is defined considering both a detection limit for errors in codes and data and uncertainty/variability in the nuclear data library. The latter employs S/U methods with a GLLS data adjustment to find representative nuclear data covariances constrained by integral experiments, which are then used to compute uncertainties in k_{eff} from nuclear data. The USLs for the classes of problems are as follows: Pu metal, 0.980; Pu solutions, 0.973; dry Pu oxides, 0.978; dilute Pu oxide-water mixes, 0.970; and intermediate-spectrum Pu oxide-water mixes, 0.953.
Analysis of JSI TRIGA MARK II reactor physical parameters calculated with TRIPOLI and MCNP.
Henry, R; Tiselj, I; Snoj, L
2015-03-01
New computational model of the JSI TRIGA Mark II research reactor was built for TRIPOLI computer code and compared with existing MCNP code model. The same modelling assumptions were used in order to check the differences of the mathematical models of both Monte Carlo codes. Differences between the TRIPOLI and MCNP predictions of keff were up to 100pcm. Further validation was performed with analyses of the normalized reaction rates and computations of kinetic parameters for various core configurations. PMID:25576735
MCNP-based computational model for the Leksell Gamma Knife
Trnka, Jiri; Novotny, Josef Jr.; Kluson, Jaroslav
2007-01-15
We have focused on the usage of MCNP code for calculation of Gamma Knife radiation field parameters with a homogenous polystyrene phantom. We have investigated several parameters of the Leksell Gamma Knife radiation field and compared the results with other studies based on EGS4 and PENELOPE code as well as the Leksell Gamma Knife treatment planning system Leksell GammaPlan (LGP). The current model describes all 201 radiation beams together and simulates all the sources in the same time. Within each beam, it considers the technical construction of the source, the source holder, collimator system, the spherical phantom, and surrounding material. We have calculated output factors for various sizes of scoring volumes, relative dose distributions along basic planes including linear dose profiles, integral doses in various volumes, and differential dose volume histograms. All the parameters have been calculated for each collimator size and for the isocentric configuration of the phantom. We have found the calculated output factors to be in agreement with other authors' works except the case of 4 mm collimator size, where averaging over the scoring volume and statistical uncertainties strongly influences the calculated results. In general, all the results are dependent on the choice of the scoring volume. The calculated linear dose profiles and relative dose distributions also match independent studies and the Leksell GammaPlan, but care must be taken about the fluctuations within the plateau, which can influence the normalization, and accuracy in determining the isocenter position, which is important for comparing different dose profiles. The calculated differential dose volume histograms and integral doses have been compared with data provided by the Leksell GammaPlan. The dose volume histograms are in good agreement as well as integral doses calculated in small calculation matrix volumes. However, deviations in integral doses up to 50% can be observed for large
MCNP-based computational model for the Leksell gamma knife.
Trnka, Jiri; Novotny, Josef; Kluson, Jaroslav
2007-01-01
We have focused on the usage of MCNP code for calculation of Gamma Knife radiation field parameters with a homogenous polystyrene phantom. We have investigated several parameters of the Leksell Gamma Knife radiation field and compared the results with other studies based on EGS4 and PENELOPE code as well as the Leksell Gamma Knife treatment planning system Leksell GammaPlan (LGP). The current model describes all 201 radiation beams together and simulates all the sources in the same time. Within each beam, it considers the technical construction of the source, the source holder, collimator system, the spherical phantom, and surrounding material. We have calculated output factors for various sizes of scoring volumes, relative dose distributions along basic planes including linear dose profiles, integral doses in various volumes, and differential dose volume histograms. All the parameters have been calculated for each collimator size and for the isocentric configuration of the phantom. We have found the calculated output factors to be in agreement with other authors' works except the case of 4 mm collimator size, where averaging over the scoring volume and statistical uncertainties strongly influences the calculated results. In general, all the results are dependent on the choice of the scoring volume. The calculated linear dose profiles and relative dose distributions also match independent studies and the Leksell GammaPlan, but care must be taken about the fluctuations within the plateau, which can influence the normalization, and accuracy in determining the isocenter position, which is important for comparing different dose profiles. The calculated differential dose volume histograms and integral doses have been compared with data provided by the Leksell GammaPlan. The dose volume histograms are in good agreement as well as integral doses calculated in small calculation matrix volumes. However, deviations in integral doses up to 50% can be observed for large
Validation and verification of MCNP6 as a new simulation tool useful for medical applications
Mashnik, Stepan G
2011-01-06
MCNP6, the latest and most advanced LANL transport code, representing a merger of MCNP5 and MCNPX has been Validated and Verified (V&V) against different experimental data and results by other codes relevant to medical applications. In the present work, we V&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 well data of interest for medical applications measured on both thin and thick targets and agrees very well with similar results obtained with other codes; MCNP6 may be a very useful tool for medical applications We plan to make MCNP6 available to the public via RSICC at Oak Ridge in the middle of 2011 but we are allowed to provide it to friendly US Beta-users outside LANL already now.
Verification of Unstructured Mesh Capabilities in MCNP6 for Reactor Physics Problems
Burke, Timothy P.; Martz, Roger L.; Kiedrowski, Brian C.; Martin, William R.
2012-08-22
New unstructured mesh capabilities in MCNP6 (developmental version during summer 2012) show potential for conducting multi-physics analyses by coupling MCNP to a finite element solver such as Abaqus/CAE[2]. Before these new capabilities can be utilized, the ability of MCNP to accurately estimate eigenvalues and pin powers using an unstructured mesh must first be verified. Previous work to verify the unstructured mesh capabilities in MCNP was accomplished using the Godiva sphere [1], and this work attempts to build on that. To accomplish this, a criticality benchmark and a fuel assembly benchmark were used for calculations in MCNP using both the Constructive Solid Geometry (CSG) native to MCNP and the unstructured mesh geometry generated using Abaqus/CAE. The Big Ten criticality benchmark [3] was modeled due to its geometry being similar to that of a reactor fuel pin. The C5G7 3-D Mixed Oxide (MOX) Fuel Assembly Benchmark [4] was modeled to test the unstructured mesh capabilities on a reactor-type problem.
MCNP6 Simulation of Light and Medium Nuclei Fragmentation at Intermediate Energies
Mashnik, Stepan Georgievich; Kerby, Leslie Marie
2015-05-22
MCNP6, the latest and most advanced LANL Monte Carlo transport code, representing a merger of MCNP5 and MCNPX, is actually much more than the sum of those two computer codes; MCNP6 is available to the public via RSICC at Oak Ridge, TN, USA. In the present work, MCNP6 was validated and verified (V&V) against different experimental data on intermediate-energy fragmentation reactions, and results by several other codes, using mainly the latest modifications of the Cascade-Exciton Model (CEM) and of the Los Alamos version of the Quark-Gluon String Model (LAQGSM) event generators CEM03.03 and LAQGSM03.03. It was found that MCNP6 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.
IER-163 Post-Experiment MCNP Calculations (U)
Favorite, Jeffrey A.
2012-06-04
IER-163 has been modeled with high fidelity in MCNP6. The model k{sub eff} was high, as in other similar calculations. The fission ratio {sup 238}U(n,f)/{sup 235}U(n,f) was 12.6% too small compared with measurements; the ratio {sup 239}Pu(n,f)/{sup 235}U(n,f) was 11.5% too small compared with measurements; the iridium ratio {sup 193}Ir(n,n{prime})/{sup 191}Ir(n,{gamma}) was 16.4% too large; and the gold ratios {sup 197}Au(n,2n)/{sup 197}Au(n,{gamma}), {sup 197}Au(n,2n)/{sup 235}U(n,f), and {sup 197}Au(n,{gamma})/{sup 235}U(n,f) were within one standard deviation of the measured values. It is suggested that the calculated {sup 235}U fission rate is too large and the calculated {sup 238}U fission rate is too small.
Fission matrix capability for MCNP, Part I - Theory
Brown, F. B.; Carney, S. E.; Kiedrowski, B. C.; Martin, W. R.
2013-07-01
The theory underlying the fission matrix method is derived using a rigorous Green's function approach. The method is then used to investigate fundamental properties of the transport equation for a continuous-energy physics treatment. We provide evidence that an infinite set of discrete, real eigenvalues and eigenfunctions exist for the continuous-energy problem, and that the eigenvalue spectrum converges smoothly as the spatial mesh for the fission matrix is refined. We also derive equations for the adjoint solution. We show that if the mesh is sufficiently refined so that both forward and adjoint solutions are valid, then the adjoint fission matrix is identical to the transpose of the forward matrix. While the energy-dependent transport equation is strictly bi-orthogonal, we provide surprising results that the forward modes are very nearly self-adjoint for a variety of continuous-energy problems. A companion paper (Part II - Applications) describes the initial experience and results from implementing this fission matrix capability into the MCNP Monte Carlo code. (authors)
MCNP-REN - A Monte Carlo Tool for Neutron Detector Design Without Using the Point Model
Abhold, M.E.; Baker, M.C.
1999-07-25
The development of neutron detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model fails to accurately predict detector response in common applications. For this reason, the general Monte Carlo N-Particle code (MCNP) was modified to simulate the pulse streams that would be generated by a neutron detector and normally analyzed by a shift register. This modified code, MCNP - Random Exponentially Distributed Neutron Source (MCNP-REN), along with the Time Analysis Program (TAP) predict neutron detector response without using the 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 MOX fresh fuel made using the Underwater Coincidence Counter (UWCC) as well as measurements of HEU reactor fuel using the active neutron Research Reactor Fuel Counter (RRFC) are compared with calculations. The method used in MCNP-REN is demonstrated to be fundamentally sound and shown to eliminate the need to use the point model for detector performance predictions.
Testing actinide fission yield treatment in CINDER90 for use in MCNP6 burnup calculations
Fensin, Michael Lorne; Umbel, Marissa
2015-09-18
Most of the development of the MCNPX/6 burnup capability focused on features that were applied to the Boltzman transport or used to prepare coefficients for use in CINDER90, with little change to CINDER90 or the CINDER90 data. Though a scheme exists for best solving the coupled Boltzman and Bateman equations, the most significant approximation is that the employed nuclear data are correct and complete. Thus, the CINDER90 library file contains 60 different actinide fission yields encompassing 36 fissionable actinides (thermal, fast, high energy and spontaneous fission). Fission reaction data exists for more than 60 actinides and as a result, fissionmore » yield data must be approximated for actinides that do not possess fission yield information. Several types of approximations are used for estimating fission yields for actinides which do not possess explicit fission yield data. The objective of this study is to test whether or not certain approximations of fission yield selection have any impact on predictability of major actinides and fission products. Further we assess which other fission products, available in MCNP6 Tier 3, result in the largest difference in production. Because the CINDER90 library file is in ASCII format and therefore easily amendable, we assess reasons for choosing, as well as compare actinide and major fission product prediction for the H. B. Robinson benchmark for, three separate fission yield selection methods: (1) the current CINDER90 library file method (Base); (2) the element method (Element); and (3) the isobar method (Isobar). Results show that the three methods tested result in similar prediction of major actinides, Tc-99 and Cs-137; however, certain fission products resulted in significantly different production depending on the method of choice.« less
Testing actinide fission yield treatment in CINDER90 for use in MCNP6 burnup calculations
Fensin, Michael Lorne; Umbel, Marissa
2015-09-18
Most of the development of the MCNPX/6 burnup capability focused on features that were applied to the Boltzman transport or used to prepare coefficients for use in CINDER90, with little change to CINDER90 or the CINDER90 data. Though a scheme exists for best solving the coupled Boltzman and Bateman equations, the most significant approximation is that the employed nuclear data are correct and complete. Thus, the CINDER90 library file contains 60 different actinide fission yields encompassing 36 fissionable actinides (thermal, fast, high energy and spontaneous fission). Fission reaction data exists for more than 60 actinides and as a result, fission yield data must be approximated for actinides that do not possess fission yield information. Several types of approximations are used for estimating fission yields for actinides which do not possess explicit fission yield data. The objective of this study is to test whether or not certain approximations of fission yield selection have any impact on predictability of major actinides and fission products. Further we assess which other fission products, available in MCNP6 Tier 3, result in the largest difference in production. Because the CINDER90 library file is in ASCII format and therefore easily amendable, we assess reasons for choosing, as well as compare actinide and major fission product prediction for the H. B. Robinson benchmark for, three separate fission yield selection methods: (1) the current CINDER90 library file method (Base); (2) the element method (Element); and (3) the isobar method (Isobar). Results show that the three methods tested result in similar prediction of major actinides, Tc-99 and Cs-137; however, certain fission products resulted in significantly different production depending on the method of choice.
Hiroshima Air-Over-Ground Analysis: Comparison of DORT and MCNP Calculations
Santoro, RT
2001-09-04
Monte Carlo (MCNP4B) and Discrete Ordinates (DORT) calculations were carried out to estimate {sup 60}Co and {sup 152}Eu activation as a function of ground range due to neutrons emitted from the Hiroshima A-bomb. Results of ORNL DORT and MCNP calculations using RZ cylindrical air-over-ground models are compared with LANL MCNP results obtained with an XYZ air-over-ground model. All of the calculations were carried out using ENDF/B-VI cross-section data and detailed angle and energy resolved neutron emission spectra from the weapon. Favorable agreement was achieved for the {sup 60}Co and {sup 152}Eu activation for ground ranges out to 1000m from the three calculations.
Simulation of Photon energy Spectra Using MISC, SOURCES, MCNP and GADRAS
Tucker, Lucas P.; Shores, Erik F.; Myers, Steven C.; Felsher, Paul D.; Garner, Scott E.; Solomon, Clell J. Jr.
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.
Correction to the MCNP{trademark} perturbation feature for cross-section dependent tallies
Densmore, J.D.; McKinney, G.W.; Hendricks, J.S.
1997-10-01
The differential operator perturbation technique is a new feature of the Monte Carlo N-Particle Transport Code MCNP version 4B that will allow users to calculate the effects of cross-section data perturbations on tallies. The implementation of the differential operator perturbation technique in MCNP assumes that the tally is independent of any perturbed cross-section data, an assumption that may not be valid for some tallies. The authors provide derivations of both the first- and second-order corrected perturbations. In addition, the appropriate perturbation corrections are demonstrated so users may accurately calculate perturbation effects for any cross-section dependent tally. Finally, corrected perturbations from six example problems are compared to actual MCNP results.
Verification of MCNP simulation of neutron flux parameters at TRIGA MK II reactor of Malaysia.
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. PMID:22885391
Monte Carlo importance sampling for the MCNP{trademark} general source
Lichtenstein, H.
1996-01-09
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.
MCNP Modeling Results for Location of Buried TRU Waste Drums
NASA Astrophysics Data System (ADS)
Steinman, D. K.; Schweitzer, J. S.
2006-05-01
In the 1960's, fifty-five gallon drums of TRU waste were buried in shallow pits on remote U.S. Government facilities such as the Idaho National Engineering Laboratory (now split into the Idaho National Laboratory and the Idaho Completion Project [ICP]). Subsequently, it was decided to remove the drums and the material that was in them from the burial pits and send the material to the Waste Isolation Pilot Plant in New Mexico. Several technologies have been tried to locate the drums non-intrusively with enough precision to minimize the chance for material to be spread into the environment. One of these technologies is the placement of steel probe holes in the pits into which wireline logging probes can be lowered to measure properties and concentrations of material surrounding the probe holes for evidence of TRU material. There is also a concern that large quantities of volatile organic compounds (VOC) are also present that would contaminate the environment during removal. In 2001, the Idaho National Engineering and Environmental Laboratory (INEEL) built two pulsed neutron wireline logging tools to measure TRU and VOC around the probe holes. The tools are the Prompt Fission Neutron (PFN) and the Pulsed Neutron Gamma (PNG), respectively. They were tested experimentally in surrogate test holes in 2003. The work reported here estimates the performance of the tools using Monte-Carlo modelling prior to field deployment. A MCNP model was constructed by INEEL personnel. It was modified by the authors to assess the ability of the tools to predict quantitatively the position and concentration of TRU and VOC materials disposed around the probe holes. The model was used to simulate the tools scanning the probe holes vertically in five centimetre increments. A drum was included in the model that could be placed near the probe hole and at other locations out to forty-five centimetres from the probe-hole in five centimetre increments. Scans were performed with no chlorine in the
Morgan C. White
2000-07-01
The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V&V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second, the ability to
Radiation Transport Analysis in Chalcogenide-Based Devices and a Neutron Howitzer Using MCNP
NASA Astrophysics Data System (ADS)
Bowler, Herbert
As photons, electrons, and neutrons traverse a medium, they impart their energy in ways that are analytically difficult to describe. Monte Carlo methods provide valuable insight into understanding this behavior, especially when the radiation source or environment is too complex to simplify. This research investigates simulating various radiation sources using the Monte Carlo N-Particle (MCNP) transport code, characterizing their impact on various materials, and comparing the simulation results to general theory and measurements. A total of five sources were of interest: two photon sources of different incident particle energies (3.83 eV and 1.25 MeV), two electron sources also of different energies (30 keV and 100 keV), and a californium-252 (Cf-252) spontaneous fission neutron source. Lateral and vertical programmable metallization cells (PMCs) were developed by other researchers for exposure to these photon and electron sources, so simplified PMC models were implemented in MCNP to estimate the doses and fluences. Dose rates measured around the neutron source and the predicted maximum activity of activation foils exposed to the neutrons were determined using MCNP and compared to experimental results obtained from gamma-ray spectroscopy. The analytical fluence calculations for the photon and electron cases agreed with MCNP results, and differences are due to MCNP considering particle movements that hand calculations do not. Doses for the photon cases agreed between the analytical and simulated results, while the electron cases differed by a factor of up to 4.8. Physical dose rate measurements taken from the neutron source agreed with MCNP within the 10% tolerance of the measurement device. The activity results had a percent error of up to 50%, which suggests a need to further evaluate the spectroscopy setup.
Verification of MCNP and DANT/sys With the Analytic Benchmark Test Set
Parsons, D.K.; Sood, A.; Forster, R.A.; Little, R.C.
1999-09-20
The recently published analytic benchmark test set has been used to verify the multigroup option of MCNP and also the deterministic DANT/sys series of codes for criticality calculations. All seventy-five problems of the test set give values for K{sub eff} accurate to at least five significant digits. Flux ratios and flux shapes are also available for many of the problems. All seventy-five problems have been run by both the MCNP and DANT/sys codes and comparisons to K{sub eff} and flux shapes have been made. Results from this verification exercise are given below.
Preliminary results of 3D dose calculations with MCNP-4B code from a SPECT image.
Rodríguez Gual, M; Lima, F F; Sospedra Alfonso, R; González González, J; Calderón Marín, C
2004-01-01
Interface software was developed to generate the input file to run Monte Carlo MCNP-4B code from medical image in Interfile format version 3.3. The software was tested using a spherical phantom of tomography slides with known cumulated activity distribution in Interfile format generated with IMAGAMMA medical image processing system. The 3D dose calculation obtained with Monte Carlo MCNP-4B code was compared with the voxel S factor method. The results show a relative error between both methods less than 1 %. PMID:15625058
Grant, C.; Mollerach, R.; Leszczynski, F.; Serra, O.; Marconi, J.; Fink, J.
2006-07-01
In 2005 the Argentine Government took the decision to complete the construction of the Atucha-II nuclear power plant, which has been progressing slowly during the last ten years. Atucha-II is a 745 MWe nuclear station moderated and cooled with heavy water, of German (Siemens) design located in Argentina. It has a pressure vessel design with 451 vertical coolant channels and the fuel assemblies (FA) are clusters of 37 natural UO{sub 2} rods with an active length of 530 cm. For the reactor physics area, a revision and update of reactor physics calculation methods and models was recently carried out covering cell, supercell (control rod) and core calculations. This paper presents benchmark comparisons of core parameters of a slightly idealized model of the Atucha-I core obtained with the PUMA reactor code with MCNP5. The Atucha-I core was selected because it is smaller, similar from a neutronic point of view, more symmetric than Atucha-II, and has some experimental data available. To validate the new models benchmark comparisons of k-effective, channel power and axial power distributions obtained with PUMA and MCNP5 have been performed. In addition, a simple cell heterogeneity correction recently introduced in PUMA is presented, which improves significantly the agreement of calculated channel powers with MCNP5. To complete the validation, the calculation of some of the critical configurations of the Atucha-I reactor measured during the experiments performed at first criticality is also presented. (authors)
NASA Astrophysics Data System (ADS)
Volmert, Ben; Pantelias, Manuel; Mutnuru, R. K.; Neukaeter, Erwin; Bitterli, Beat
2016-02-01
In this paper, an overview of the Swiss Nuclear Power Plant (NPP) activation methodology is presented and the work towards its validation by in-situ NPP foil irradiation campaigns is outlined. Nuclear Research and consultancy Group (NRG) in The Netherlands has been given the task of performing the corresponding neutron metrology. For this purpose, small Aluminium boxes containing a set of circular-shaped neutron activation foils have been prepared. After being irradiated for one complete reactor cycle, the sets have been successfully retrieved, followed by gamma-spectrometric measurements of the individual foils at NRG. Along with the individual activities of the foils, the reaction rates and thermal, intermediate and fast neutron fluence rates at the foil locations have been determined. These determinations include appropriate corrections for gamma self-absorption and neutron self-shielding as well as corresponding measurement uncertainties. The comparison of the NPP Monte Carlo calculations with the results of the foil measurements is done by using an individual generic MCNP model functioning as an interface and allowing the simulation of individual foil activation by predetermined neutron spectra. To summarize, the comparison between calculation and measurement serve as a sound validation of the Swiss NPP activation methodology by demonstrating a satisfying agreement between measurement and calculation. Finally, the validation offers a chance for further improvements of the existing NPP models by ensuing calibration and/or modelling optimizations for key components and structures.
TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.
Kurosawa, Masahiko
2005-01-01
For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data. PMID:16604689
Certification of MCNP Version 4A for WHC computer platforms. Revision 7
Carter, L.L.
1995-05-03
MCNP is a general-purpose Monte Carlo code that can 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).
TALYS-Based Cross Section Library for Use with MCNP(X).
Energy Science and Technology Software Center (ESTSC)
2009-11-11
Version 00 The TENDL-2008 library has been checked with the CHECKR, FIZCON and PSYCHE checking programs and successfully processed with NJOY-99.161 into ACE format to create this library for use in MCNP5 and MCNPX calculations. ACE files are provided for neutrons, protons, deuterons, tritons, helions and alpha particles.
NASA Astrophysics Data System (ADS)
Tiyapun, K.; Chimtin, M.; Munsorn, S.; Somchit, S.
2015-05-01
The objective of this work is to demonstrate the method for validating the predication of the calculation methods for neutron flux distribution in the irradiation tubes of TRIGA research reactor (TRR-1/M1) using the MCNP computer code model. The reaction rate using in the experiment includes 27Al(n, α)24Na and 197Au(n, γ)198Au reactions. Aluminium (99.9 wt%) and gold (0.1 wt%) foils and the gold foils covered with cadmium were irradiated in 9 locations in the core referred to as CT, C8, C12, F3, F12, F22, F29, G5, and G33. The experimental results were compared to the calculations performed using MCNP which consisted of the detailed geometrical model of the reactor core. The results from the experimental and calculated normalized reaction rates in the reactor core are in good agreement for both reactions showing that the material and geometrical properties of the reactor core are modelled very well. The results indicated that the difference between the experimental measurements and the calculation of the reactor core using the MCNP geometrical model was below 10%. In conclusion the MCNP computational model which was used to calculate the neutron flux and reaction rate distribution in the reactor core can be used for others reactor core parameters including neutron spectra calculation, dose rate calculation, power peaking factors calculation and optimization of research reactor utilization in the future with the confidence in the accuracy and reliability of the calculation.
Comparison of scientific computing platforms for MCNP4A Monte Carlo calculations
Hendricks, J.S.; Brockhoff, R.C. . Applied Theoretical Physics Division)
1994-04-01
The performance of seven computer platforms is evaluated with the widely used and internationally available MCNP4A Monte Carlo radiation transport code. All results are reproducible and are presented in such a way as to enable comparison with computer platforms not in the study. The authors observed that the HP/9000-735 workstation runs MCNP 50% faster than the Cray YMP 8/64. Compared with the Cray YMP 8/64, the IBM RS/6000-560 is 68% as fast, the Sun Sparc10 is 66% as fast, the Silicon Graphics ONYX is 90% as fast, the Gateway 2000 model 4DX2-66V personal computer is 27% as fast, and the Sun Sparc2 is 24% as fast. In addition to comparing the timing performance of the seven platforms, the authors observe that changes in compilers and software over the past 2 yr have resulted in only modest performance improvements, hardware improvements have enhanced performance by less than a factor of [approximately]3, timing studies are very problem dependent, MCNP4Q runs about as fast as MCNP4.
Application of MCNP for neutronic calculations at VR-1 training reactor
NASA Astrophysics Data System (ADS)
Huml, Ondřej; Rataj, Jan; Bílý, Tomáš
2014-06-01
The paper presents utilization of Monte Carlo MCNP transport code for neutronic calculations of training reactor VR-1. Results of calculations are compared with results of measurements realized during last few critical experiments with various reactor core configurations. Very good agreement between calculations and measurements is observed.
MCNP5 CRITICALITY VALIDATION AND BIAS FOR INTERMEDIATE ENRICHED URANIUM SYSTEMS
FINFROCK SH
2009-12-10
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_sign} 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{sub 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{sub 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{sub 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{sub 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{sub eff} limit for calculations of the intermediate enriched uranium type systems.
MCNP calculations for Russian criticality-safety benchmarks
Capell, B.M.; Mosteller, R.D.; Pelowitz, D.B.
1996-12-31
The current edition of the International Handbook of Evaluated Criticality Safety Benchmark Experiments contains evaluations of 20 critical experiments performed and evaluated by the Institute for Experimental Physics of the Russian Federal Nuclear Center (VNIIEF) at Arzamas-16 and 16 critical experiments performed and evaluated by the Institute for Technical Physics of the Russian Federal Nuclear Center (VNIITF) at Chelyabinsk-70. These fast-spectrum experiments are of particular interest for data testing of ENDF/B-VI because they contain uranium metal systems of intermediate enrichment as well as uranium and plutonium metal systems with reflectors such as graphite, stainless steel, polyethylene, beryllium, and beryllium oxide. This paper presents the first published results for such systems using cross-section libraries based on ENDF/B-VI.
Mueller, Don E.; Marshall, William J.; Wagner, John C.; Bowen, Douglas G.
2015-09-01
The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (k_{eff}) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the bias due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of k_{eff} calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.
RADBALLTECHNOLOGY TESTING AND MCNP MODELING OF THE TUNGSTEN COLLIMATOR
Farfan, E.
2010-07-08
The United Kingdom's National Nuclear Laboratory (NNL) has developed a remote, non-electrical, radiation-mapping device known as RadBall{trademark}, which can locate and quantify radioactive hazards within contaminated areas of the nuclear industry. RadBall{trademark} consists of a colander-like outer shell that houses a radiation-sensitive polymer sphere. The outer shell works to collimate radiation sources and those areas of the polymer sphere that are exposed react, becoming increasingly more opaque, in proportion to the absorbed dose. The polymer sphere is imaged in an optical-CT scanner, which produces a high resolution 3D map of optical attenuation coefficients. Subsequent analysis of the optical attenuation matrix provides information on the spatial distribution of sources in a given area forming a 3D characterization of the area of interest. RadBall{trademark} has no power requirements and can be positioned in tight or hard-to reach locations. The RadBall{trademark} technology has been deployed in a number of technology trials in nuclear waste reprocessing plants at Sellafield in the United Kingdom and facilities of the Savannah River National Laboratory (SRNL). This study focuses on the RadBall{trademark} testing and modeling accomplished at SRNL.
Perkasa, Y. S.; Waris, A. Kurniadi, R. Su'ud, Z.
2014-09-30
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.
Valentine, T.E.; Rugama, Y. Munoz-Cobos, J.; Perez, R.
2000-10-23
The design of reactivity monitoring systems for accelerator-driven systems must be investigated to ensure that such systems remain subcritical during operation. The Monte Carlo codes LAHET and MCNP-DSP were combined together to facilitate the design of reactivity monitoring systems. The coupling of LAHET and MCNP-DSP provides a tool that can be used to simulate a variety of subcritical measurements such as the pulsed neutron, Rossi-{alpha}, or noise analysis measurements.
Neutron Cross Section Library Based on JEFF3.1 for Use with MCNP.
Energy Science and Technology Software Center (ESTSC)
2007-03-20
Version 00 This continuous energy cross-section data library in ACE format is for shielding and criticality applications done with MCNP. In addition to the description of the NJOY processing procedure used to create the library, the included report NEA/NSC/DOC(2006)18 contains results from the benchmarking activity aimed at testing the quality of the data for criticality and shielding applications. The library at 300K has been verified: visually (no discontinuities, correct processing in all range) and withmore » comparisons with other libraries available for the same purposes (ENDF/B-VI.8, JEF2.2, JENDL3.3, ) A set of experiments using MCNP4c are used in order to validate the processed library.« less
Gamma irradiator dose mapping simulation using the MCNP code and benchmarking with dosimetry.
Sohrabpour, M; Hassanzadeh, M; Shahriari, M; Sharifzadeh, M
2002-10-01
The Monte Carlo transport code, MCNP, has been applied in simulating dose rate distribution in the IR-136 gamma irradiator system. Isodose curves, cumulative dose values, and system design data such as throughputs, over-dose-ratios, and efficiencies have been simulated as functions of product density. Simulated isodose curves, and cumulative dose values were compared with dosimetry values obtained using polymethyle-methacrylate, Fricke, ethanol-chlorobenzene, and potassium dichromate dosimeters. The produced system design data were also found to agree quite favorably with those of the system manufacturer's data. MCNP has thus been found to be an effective transport code for handling of various dose mapping excercises for gamma irradiators. PMID:12361333
MCNP6 simulation of light and medium nuclei fragmentation at intermediate energies
NASA Astrophysics Data System (ADS)
Mashnik, Stepan G.; Kerby, Leslie M.
2016-05-01
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 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.
Total reaction cross sections in CEM and MCNP6 at intermediate energies
Kerby, Leslie M.; Mashnik, Stepan G.
2015-05-14
Accurate total reaction cross section models are important to achieving reliable predictions from spallation and transport codes. The latest version of the Cascade Exciton Model (CEM) as incorporated in the code CEM03.03, and the Monte Carlo N-Particle transport code (MCNP6), both developed at Los Alamos National Laboratory (LANL), each use such cross sections. Having accurate total reaction cross section models in the intermediate energy region (50 MeV to 5 GeV) is very important for different applications, including analysis of space environments, use in medical physics, and accelerator design, to name just a few. The current inverse cross sections used in the preequilibrium and evaporation stages of CEM are based on the Dostrovsky et al. model, published in 1959. Better cross section models are now available. Implementing better cross section models in CEM and MCNP6 should yield improved predictions for particle spectra and total production cross sections, among other results.
MCNP6 Simulation of Light and Medium Nuclei Fragmentation at Intermediate Energies
Mashnik, Stepan Georgievich; Kerby, Leslie Marie
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.
Lahodova, Z.; Flibor, S.; Klupak, V.; Kucera, J.; Marek, M.; Viererbl, L.
2006-07-01
The main purpose of this work was to evaluate the neutron energy distribution in a channel of the LVR-15 reactor used mostly for short-time neutron activation analysis. Twenty types of activation monitors were irradiated in this channel equipped with a pneumatic facility with a transport time of 3.5 s. The activities measured and the corresponding reaction rates were used to determinate the neutron spectrum. The reaction rates were compared with MCNP calculations to confirm the results. The second purpose of this work was to verify our nuclear data library used for the reaction rate calculations. The experiment results were also incorporated into our database system of neutron energy distribution at the reactor core. (authors)
Calculation of self-shielding factor for neutron activation experiments using GEANT4 and MCNP
NASA Astrophysics Data System (ADS)
Romero-Barrientos, Jaime; Molina, F.; Aguilera, Pablo; Arellano, H. F.
2016-07-01
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 1.10-5eV to 2.107eV 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.
SABRINA: an interactive three-dimensional geometry-mnodeling program for MCNP
West, J.T. III
1986-10-01
SABRINA is a fully interactive three-dimensional geometry-modeling program for MCNP, a Los Alamos Monte Carlo code for neutron and photon transport. In SABRINA, a user constructs either body geometry or surface geometry models and debugs spatial descriptions for the resulting objects. This enhanced capability significantly reduces effort in constructing and debugging complicated three-dimensional geometry models for Monte Carlo analysis. 2 refs., 33 figs.
Conversion of Input Data between KENO and MCNP File Formats for Computer Criticality Assessments
Schwarz, Randolph A.; Carter, Leland L.; Schwarz Alysia L.
2006-11-30
KENO is a Monte Carlo criticality code that is maintained by Oak Ridge National Laboratory (ORNL). KENO is included in the SCALE (Standardized Computer Analysis for Licensing Evaluation) package. KENO is often used because it was specifically designed for criticality calculations. Because KENO has convenient geometry input, including the treatment of lattice arrays of materials, it is frequently used for production calculations. Monte Carlo N-Particle (MCNP) is a Monte Carlo transport code maintained by Los Alamos National Laboratory (LANL). MCNP has a powerful 3D geometry package and an extensive cross section database. It is a general-purpose code and may be used for calculations involving shielding or medical facilities, for example, but can also be used for criticality calculations. MCNP is becoming increasingly more popular for performing production criticality calculations. Both codes have their own specific advantages. After a criticality calculation has been performed with one of the codes, it is often desirable (or may be a safety requirement) to repeat the calculation with the other code to compare the important parameters using a different geometry treatment and cross section database. This manual conversion of input files between the two codes is labor intensive. The industry needs the capability of converting geometry models between MCNP and KENO without a large investment in manpower. The proposed conversion package will aid the user in converting between the codes. It is not intended to be used as a “black box”. The resulting input file will need to be carefully inspected by criticality safety personnel to verify the intent of the calculation is preserved in the conversion. The purpose of this package is to help the criticality specialist in the conversion process by converting the geometry, materials, and pertinent data cards.
UW MCNP source patch for the EPFL Haefely source. EPFL (Swiss) fusion-fission hybrid experiment
McKinney, G; Woodruff, G L
1986-06-01
The development of a source patch which describes the Haefely neutron source for use in the MCNP Monte Carlo code has been described in progress reports of the EPFL (Swiss) Fusion Blanket Project at the University of Washington. The most recent of these reports dealing with the source patch was Progress Report No. 14. This report reviews some of the physical description included in the report, and also includes additional details of the patch as well as a listing of the patch itself.
The Use of MCNP in Flash Radiographic Applications at AWE
NASA Astrophysics Data System (ADS)
Quillin, S.; Crotch, I.; McAlpin, S.; O'Malley, J.
AWE performs experiments to investigate the hydrodynamic behavior of explosive metal systems in order to underwrite the UK nuclear deterrent. The experiments involve the manufacture of a device to mock up some aspect of the weapon. Inert simulant materials replace fissile weapon components. The device is then detonated under remote control within specially designed explosive containment buildings called firing chambers. During the experiment a very brief, intense, collimated flash of high energy x-rays are used to take a snapshot of the implosion (see Fig. 1). Prom the resulting image measurements of the dynamic configuration and density distribution of the components in the device are inferred. These are then used to compare with calculations of the hydrodynamic operation of the weapon and understand how the device would perform under various conditions. This type of experiment is known as a core punch experiment.
Mashnik, Stepan G
2010-11-22
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 andV 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.
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. PMID:16532938
Enger, Shirin A.; Munck af Rosenschoeld, Per; Rezaei, Arash; Lundqvist, Hans
2006-02-15
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.
Calculated organ doses for Mayak production association central hall using ICRP and MCNP.
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. PMID:12645766
An analysis of MCNP cross-sections and tally methods for low-energy photon emitters
NASA Astrophysics Data System (ADS)
DeMarco, John J.; Wallace, Robert E.; Boedeker, Kirsten
2002-04-01
Monte Carlo calculations are frequently used to analyse a variety of radiological science applications using low-energy (10-1000 keV) photon sources. This study seeks to create a low-energy benchmark for the MCNP Monte Carlo code by simulating the absolute dose rate in water and the air-kerma rate for monoenergetic point sources with energies between 10 keV and 1 MeV. The analysis compares four cross-section datasets as well as the tally method for collision kerma versus absorbed dose. The total photon attenuation coefficient cross-section for low atomic number elements has changed significantly as cross-section data have changed between 1967 and 1989. Differences of up to 10% are observed in the photoelectric cross-section for water at 30 keV between the standard MCNP cross-section dataset (DLC-200) and the most recent XCOM/NIST tabulation. At 30 keV, the absolute dose rate in water at 1.0 cm from the source increases by 7.8% after replacing the DLC-200 photoelectric cross-sections for water with those from the XCOM/NIST tabulation. The differences in the absolute dose rate are analysed when calculated with either the MCNP absorbed dose tally or the collision kerma tally. Significant differences between the collision kerma tally and the absorbed dose tally can occur when using the DLC-200 attenuation coefficients in conjunction with a modern tabulation of mass energy-absorption coefficients.
MCNP6 Simulation of Reactions of Interest to FRIB, Medical, and Space Applications
NASA Astrophysics Data System (ADS)
Mashnik, Stepan G.
The latest production-version of the Los Alamos Monte Carlo N-Particle transport code MCNP6 has been used to simulate a variety of particle-nucleus and nucleus-nucleus reactions of academic and applied interest to research subjects at the Facility for Rare Isotope Beams (FRIB), medical isotope production, space-radiation shielding, cosmic-ray propagation, and accelerator applications, including several reactions induced by radioactive isotopes, analyzing production of both stable and radioactive residual nuclei. Here, we discuss examples of validation and verification of MCNP6 by comparing with recent neutron spectra measured at the Heavy Ion Medical Accelerator in Chiba, Japan; spectra of light fragments from several reactions measured recently at GANIL, France; INFN Laboratori Nazionali del Sud, Catania, Italy; COSY of the Jülich Research Center, Germany; and cross sections of products from several reactions measured lately at GSI, Darmstadt, Germany; ITEP, Moscow, Russia; and, LANSCE, LANL, Los Alamos, U.S.A. As a rule, MCNP6 provides quite good predictions for most of the reactions we analyzed so far, allowing us to conclude that it can be used as a reliable and useful simulation tool for various applications for FRIB, medical, and space applications involving stable and radioactive isotopes.
An investigation of voxel geometries for MCNP-based radiation dose calculations.
Zhang, Juying; Bednarz, Bryan; Xu, X George
2006-11-01
Voxelized geometry such as those obtained from medical images is increasingly used in Monte Carlo calculations of absorbed doses. One useful application of calculated absorbed dose is the determination of fluence-to-dose conversion factors for different organs. However, confusion still exists about how such a geometry is defined and how the energy deposition is best computed, especially involving a popular code, MCNP5. This study investigated two different types of geometries in the MCNP5 code, cell and lattice definitions. A 10 cm x 10 cm x 10 cm test phantom, which contained an embedded 2 cm x 2 cm x 2 cm target at its center, was considered. A planar source emitting parallel photons was also considered in the study. The results revealed that MCNP5 does not calculate total target volume for multi-voxel geometries. Therefore, tallies which involve total target volume must be divided by the user by the total number of voxels to obtain a correct dose result. Also, using planar source areas greater than the phantom size results in the same fluence-to-dose conversion factor. PMID:17023800
Voxel2MCNP: software for handling voxel models for Monte Carlo radiation transport calculations.
Hegenbart, Lars; Pölz, Stefan; Benzler, Andreas; Urban, Manfred
2012-02-01
Voxel2MCNP is a program that sets up radiation protection scenarios with voxel models and generates corresponding input files for the Monte Carlo code MCNPX. Its technology is based on object-oriented programming, and the development is platform-independent. It has a user-friendly graphical interface including a two- and three-dimensional viewer. A row of equipment models is implemented in the program. Various voxel model file formats are supported. Applications include calculation of counting efficiency of in vivo measurement scenarios and calculation of dose coefficients for internal and external radiation scenarios. Moreover, anthropometric parameters of voxel models, for instance chest wall thickness, can be determined. Voxel2MCNP offers several methods for voxel model manipulations including image registration techniques. The authors demonstrate the validity of the program results and provide references for previous successful implementations. The authors illustrate the reliability of calculated dose conversion factors and specific absorbed fractions. Voxel2MCNP is used on a regular basis to generate virtual radiation protection scenarios at Karlsruhe Institute of Technology while further improvements and developments are ongoing. PMID:22217596
Comparison of KENO-VI and MCNP5 Criticality Analyses for a Lunar Regolith Clustered-Reactor System
NASA Astrophysics Data System (ADS)
Bess, John Darrell
2008-01-01
The Lunar Regolith Clustered-Reactor System design has been presented as an alternative method for providing surface power to a lunar facility using a fast-fission, heatpipe-cooled nuclear reactor. The reactor system is divided into subcritical units that can be safely launched into orbit without risk of inadvertent criticality in the event of a launch accident. The reactor subunits are emplaced into the lunar surface to form a clustered-reactor system, utilizing the regolith as both radiation shielding and neutron-reflector material. Coordinated placement of multiple subunits can provision a critical reactor system proportional to localized lunar surface power demand. Reactor units assembled using proven and tested materials in radiation environments such as UO2 fuel, stainless-steel cladding and support, and compatible liquid-metal heatpipes promote safety and reliability, with ease of manufacture and testing. Reactor power levels of approximately 100 kWth per subunit significantly reduces the negative effects of elevated temperature and radiation environments associated with single nuclear power reactors operated at higher power levels. The analysis of subunit criticality in various accident scenarios differs by up to 4% (~$6 in reactivity) between results generated using conventional criticality analysis codes, MCNP5 and KENO-VI. A demonstrated trend exists between results of the two criticality codes as accident conditions approach a multiplication factor of one. Code comparison of a tri-cluster system on the lunar surface provides comparable results with calculated system reactivity within 0.5%. Iron concentration is confirmed as the dominant element in the lunar regolith influencing system reactivity.
Comparison of KENO-VI and MCNP5 Criticality Analyses for a Lunar Regolith Clustered-Reactor System
Bess, John Darrell
2008-01-21
The Lunar Regolith Clustered-Reactor System design has been presented as an alternative method for providing surface power to a lunar facility using a fast-fission, heatpipe-cooled nuclear reactor. The reactor system is divided into subcritical units that can be safely launched into orbit without risk of inadvertent criticality in the event of a launch accident. The reactor subunits are emplaced into the lunar surface to form a clustered-reactor system, utilizing the regolith as both radiation shielding and neutron-reflector material. Coordinated placement of multiple subunits can provision a critical reactor system proportional to localized lunar surface power demand. Reactor units assembled using proven and tested materials in radiation environments such as UO{sub 2} fuel, stainless-steel cladding and support, and compatible liquid-metal heatpipes promote safety and reliability, with ease of manufacture and testing. Reactor power levels of approximately 100 kWth per subunit significantly reduces the negative effects of elevated temperature and radiation environments associated with single nuclear power reactors operated at higher power levels. The analysis of subunit criticality in various accident scenarios differs by up to 4% ({approx}$6 in reactivity) between results generated using conventional criticality analysis codes, MCNP5 and KENO-VI. A demonstrated trend exists between results of the two criticality codes as accident conditions approach a multiplication factor of one. Code comparison of a tri-cluster system on the lunar surface provides comparable results with calculated system reactivity within 0.5%. Iron concentration is confirmed as the dominant element in the lunar regolith influencing system reactivity.
Simulations of neutron multiplicity measurements with MCNP-PoliMi.
Mattingly, John K.; Pozzi, Sara A.; Clarke, Shaun D.; Dennis, Ben D.; Miller, Eric C.
2010-09-01
The heightened focus on nuclear safeguards and accountability has increased the need to develop and verify simulation tools for modeling these applications. The ability to accurately simulate safeguards techniques, such as neutron multiplicity counting, aids in the design and development of future systems. This work focuses on validating the ability of the Monte Carlo code MCNPX-PoliMi to reproduce measured neutron multiplicity results for a highly multiplicative sample. The benchmark experiment for this validation consists of a 4.5-kg sphere of plutonium metal that was moderated by various thicknesses of polyethylene. The detector system was the nPod, which contains a bank of 15 3He detectors. Simulations of the experiments were compared to the actual measurements and several sources of potential bias in the simulation were evaluated. The analysis included the effects of detector dead time, source-detector distance, density, and adjustments made to the value of {nu}-bar in the data libraries. Based on this analysis it was observed that a 1.14% decrease in the evaluated value of {nu}-bar for 239Pu in the ENDF-VII library substantially improved the accuracy of the simulation.
NASA Astrophysics Data System (ADS)
Poškus, Andrius
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 that the
NASA Astrophysics Data System (ADS)
Adjei, Christian Amevi
The main objective of this thesis is twofold. The starting objective was to develop a model for meaningful benchmarking of different versions of GEANT4 against an experimental set-up and MCNP5 pertaining to photon transport and interactions. The following objective was to develop a preliminary design of a Fast Neutron Pencil Beam (FNPB) Facility to be applicable for the University of Utah research reactor (UUTR) using MCNP5 and GEANT4. The three various GEANT4 code versions, GEANT4.9.4, GEANT4.9.3, and GEANT4.9.2, were compared to MCNP5 and the experimental measurements of gamma attenuation in air. The average gamma dose rate was measured in the laboratory experiment at various distances from a shielded cesium source using a Ludlum model 19 portable NaI detector. As it was expected, the gamma dose rate decreased with distance. All three GEANT4 code versions agreed well with both the experimental data and the MCNP5 simulation. Additionally, a simple GEANT4 and MCNP5 model was developed to compare the code agreements for neutron interactions in various materials. Preliminary FNPB design was developed using MCNP5; a semi-accurate model was developed using GEANT4 (because GEANT4 does not support the reactor physics modeling, the reactor was represented as a surface neutron source, thus a semi-accurate model). Based on the MCNP5 model, the fast neutron flux in a sample holder of the FNPB is obtained to be 6.52×107 n/cm2s, which is one order of magnitude lower than gigantic fast neutron pencil beam facilities existing elsewhere. The MCNP5 model-based neutron spectrum indicates that the maximum expected fast neutron flux is at a neutron energy of ~1 MeV. In addition, the MCNP5 model provided information on gamma flux to be expected in this preliminary FNPB design; specifically, in the sample holder, the gamma flux is to be expected to be around 108 γ/cm 2s, delivering a gamma dose of 4.54×103 rem/hr. This value is one to two orders of magnitudes below the gamma
V&V of MCNP 6.1.1 Beta Against Intermediate and High-Energy Experimental Data
Mashnik, Stepan G
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.
Koivunoro, Hanna; Siiskonen, Teemu; Kotiluoto, Petri; Auterinen, Iiro; Hippelaeinen, Eero; Savolainen, Sauli
2012-03-15
Purpose: In this work, accuracy of the mcnp5 code in the electron transport calculations and its suitability for ionization chamber (IC) response simulations in photon beams are studied in comparison to egsnrc and penelope codes. Methods: The electron transport is studied by comparing the depth dose distributions in a water phantom subdivided into thin layers using incident energies (0.05, 0.1, 1, and 10 MeV) for the broad parallel electron beams. The IC response simulations are studied in water phantom in three dosimetric gas materials (air, argon, and methane based tissue equivalent gas) for photon beams ({sup 60}Co source, 6 MV linear medical accelerator, and mono-energetic 2 MeV photon source). Two optional electron transport models of mcnp5 are evaluated: the ITS-based electron energy indexing (mcnp5{sub ITS}) and the new detailed electron energy-loss straggling logic (mcnp5{sub new}). The electron substep length (ESTEP parameter) dependency in mcnp5 is investigated as well. Results: For the electron beam studies, large discrepancies (>3%) are observed between the mcnp5 dose distributions and the reference codes at 1 MeV and lower energies. The discrepancy is especially notable for 0.1 and 0.05 MeV electron beams. The boundary crossing artifacts, which are well known for the mcnp5{sub ITS}, are observed for the mcnp5{sub new} only at 0.1 and 0.05 MeV beam energies. If the excessive boundary crossing is eliminated by using single scoring cells, the mcnp5{sub ITS} provides dose distributions that agree better with the reference codes than mcnp5{sub new}. The mcnp5 dose estimates for the gas cavity agree within 1% with the reference codes, if the mcnp5{sub ITS} is applied or electron substep length is set adequately for the gas in the cavity using the mcnp5{sub new}. The mcnp5{sub new} results are found highly dependent on the chosen electron substep length and might lead up to 15% underestimation of the absorbed dose. Conclusions: Since the mcnp5 electron
Nuclear criticality research at the University of New Mexico
Busch, R.D.
1997-06-01
Two projects at the University of New Mexico are briefly described. The university`s Chemical and Nuclear Engineering Department has completed the final draft of a primer for MCNP4A, which it plans to publish soon. The primer was written to help an analyst who has little experience with the MCNP code to perform criticality safety analyses. In addition, the department has carried out a series of approach-to-critical experiments on the SHEBA-II, a UO{sub 2}F{sub 2} solution critical assembly at Los Alamos National Laboratory. The results obtained differed slightly from what was predicted by the TWODANT code.
PWR Facility Dose Modeling Using MCNP5 and the CADIS/ADVANTG Variance-Reduction Methodology
Blakeman, Edward D; Peplow, Douglas E.; Wagner, John C; Murphy, Brian D; Mueller, Don
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.
NASA Astrophysics Data System (ADS)
Mauerhofer, E.; Havenith, A.; Carasco, C.; Payan, E.; Kettler, J.; Ma, J. L.; Perot, B.
2013-04-01
The Forschungszentrum Jülich GmbH (FZJ), together with the Aachen University Rheinisch-Westfaelische Technische Hochschule (RWTH) and the French Alternative Energies and Atomic Energy Commission (CEA Cadarache) are involved in a cooperation aiming at characterizing toxic and reactive elements in radioactive waste packages by means of Prompt Gamma Neutron Activation Analysis (PGNAA) [1]. The French and German waste management agencies have indeed defined acceptability limits concerning these elements in view of their projected geological repositories. A first measurement campaign was performed in the new Prompt Gamma Neutron Activation Analysis (PGNAA) facility called MEDINA, at FZJ, to assess the capture gamma-ray signatures of some elements of interest in large samples up to waste drums with a volume of 200 liter. MEDINA is the acronym for Multi Element Detection based on Instrumental Neutron Activation. This paper presents MCNP calculations of the MEDINA facility and quantitative comparison between measurement and simulation. Passive gamma-ray spectra acquired with a high purity germanium detector and calibration sources are used to qualify the numerical model of the crystal. Active PGNAA spectra of a sodium chloride sample measured with MEDINA then allow for qualifying the global numerical model of the measurement cell. Chlorine indeed constitutes a usual reference with reliable capture gamma-ray production data. The goal is to characterize the entire simulation protocol (geometrical model, nuclear data, and postprocessing tools) which will be used for current measurement interpretation, extrapolation of the performances to other types of waste packages or other applications, as well as for the study of future PGNAA facilities.
Mauerhofer, E.; Havenith, A.; Kettler, J.; Carasco, C.; Payan, E.; Ma, J. L.; Perot, B.
2013-04-19
The Forschungszentrum Juelich GmbH (FZJ), together with the Aachen University Rheinisch-Westfaelische Technische Hochschule (RWTH) and the French Alternative Energies and Atomic Energy Commission (CEA Cadarache) are involved in a cooperation aiming at characterizing toxic and reactive elements in radioactive waste packages by means of Prompt Gamma Neutron Activation Analysis (PGNAA). The French and German waste management agencies have indeed defined acceptability limits concerning these elements in view of their projected geological repositories. A first measurement campaign was performed in the new Prompt Gamma Neutron Activation Analysis (PGNAA) facility called MEDINA, at FZJ, to assess the capture gamma-ray signatures of some elements of interest in large samples up to waste drums with a volume of 200 liter. MEDINA is the acronym for Multi Element Detection based on Instrumental Neutron Activation. This paper presents MCNP calculations of the MEDINA facility and quantitative comparison between measurement and simulation. Passive gamma-ray spectra acquired with a high purity germanium detector and calibration sources are used to qualify the numerical model of the crystal. Active PGNAA spectra of a sodium chloride sample measured with MEDINA then allow for qualifying the global numerical model of the measurement cell. Chlorine indeed constitutes a usual reference with reliable capture gamma-ray production data. The goal is to characterize the entire simulation protocol (geometrical model, nuclear data, and postprocessing tools) which will be used for current measurement interpretation, extrapolation of the performances to other types of waste packages or other applications, as well as for the study of future PGNAA facilities.
Calculation of the effective dose from natural radioactivity in soil using MCNP code.
Krstic, D; Nikezic, D
2010-01-01
Effective dose delivered by photon emitted from natural radioactivity in soil was calculated in this work. Calculations have been done for the most common natural radionuclides in soil (238)U, (232)Th series and (40)K. A ORNL human phantoms and the Monte Carlo transport code MCNP-4B were employed to calculate the energy deposited in all organs. The effective dose was calculated according to ICRP 74 recommendations. Conversion factors of effective dose per air kerma were determined. Results obtained here were compared with other authors. PMID:20045343
SABRINA - An interactive geometry modeler for MCNP (Monte Carlo Neutron Photon)
West, J.T.; Murphy, J.
1988-01-01
SABRINA is an interactive three-dimensional geometry modeler developed to produce complicated models for the Los Alamos Monte Carlo Neutron Photon program MCNP. SABRINA produces line drawings and color-shaded drawings for a wide variety of interactive graphics terminals. It is used as a geometry preprocessor in model development and as a Monte Carlo particle-track postprocessor in the visualization of complicated particle transport problem. SABRINA is written in Fortran 77 and is based on the Los Alamos Common Graphics System, CGS. 5 refs., 2 figs.
Efficiency of whole-body counter for various body size calculated by MCNP5 software.
Krstic, D; Nikezic, D
2012-11-01
The efficiency of a whole-body counter for (137)Cs and (40)K was calculated using the MCNP5 code. The ORNL phantoms of a human body of different body sizes were applied in a sitting position in front of a detector. The aim was to investigate the dependence of efficiency on the body size (age) and the detector position with respect to the body and to estimate the accuracy of real measurements. The calculation work presented here is related to the NaI detector, which is available in the Serbian Whole-body Counter facility in Vinca Institute. PMID:22923253
Application of Numerical Phantoms and MCNP Calculation for In Vivo Calibration
NASA Astrophysics Data System (ADS)
Franck, D.; Borisov, N. M.; Laval, L.
The paper reports on development of numeric phantoms for Monte Carlo calculations for in vivo measurements of radionuclides deposited in tissues. The individual properties of each person require rather precise geometric representations. It is particularly important for low energy gamma ray emitting sources as thorium, uranium, plutonium and other actinides. The new utility which allows automatic creation of MCNP initial file from individual scanning information, was developed. It includes segmentation of voxel matrix, obtained with computer tomography, for distinguishing tissues by level of brightness, association colors with certain tissues, source and detector specification and, finally, voxel coupling to reduce the consumed memory and increase speed of calculations.
Input files with ORNL—mathematical phantoms of the human body for MCNP-4B
NASA Astrophysics Data System (ADS)
Krstić, D.; Nikezić, D.
2007-01-01
Protection against ionizing radiation requires information on the absorbed doses in organs of the human body. Implantation of many dosimeters in the human body is undesirable (or impossible), so the doses in organs are not measurable and some kind of dose calculation has to be applied. Calculation of doses in organs requests: (a) an exact description of the geometry of organs, (b) the chemical constitution of tissues, and (c) appropriate computer programs. The first two items, (a) and (b), make a so-called "phantom". In another words, the "phantom of a human body" is a mathematical representation of the human body including all other relevant information. All organs are represented with geometrical bodies (like cylinders, ellipsoids, tori, cones etc.), which are described with suitable mathematical equations. A corresponding chemical constitution for various types of organ tissues is also defined. MCNP-4B ( Monte Carlo N- Particle) is often used as transport code. Users of this software prepare an "input file" providing all necessary information for program execution. This information includes: (a) source definition—type of ionizing radiation, energy spectrum, and geometry of the source; (b) target definition—material constitution, geometry, location in respect to the source etc.; (c) characterization of absorbing media between the source and target; (d) output tally, etc. This paper presents input files with "human phantoms" for the MCNP-4B code. The input files with "phantoms" were prepared based on publications issued by the Oak Ridge National Laboratory (ORNL). Seven input files relating to different age groups (newborn, 1, 5, 10, 15 years, as well as, male and female adults) are presented here. A test example and comparison with other data found in literature are also given. Program summaryTitle of program: INPUT FILES, AMALE, AFEMALE, AGE15, AGE10, AGE5, AGE01, NEWB Catalogue identifier:ADYF_v1_0 Program summary URL
NASA Astrophysics Data System (ADS)
Singh, Vishwanath P.; Medhat, M. E.; Shirmardi, S. P.
2015-01-01
The mass attenuation coefficients, μ/ρ and effective atomic numbers, Zeff of some carbon steel and stainless steel alloys have been calculated by using Geant4, MCNP simulation codes for different gamma ray energies, 279.1 keV, 661.6 keV, 662 keV, 1115.5 keV, 1173 keV and 1332 keV. The simulation results of Zeff using Geant4 and MCNP codes have been compared with possible available experimental results and theoretical WinXcom, and good agreement has been observed. The simulated μ/ρ and Zeff values using Geant4 and MCNP code signifies that both the simulation process can be followed to determine the gamma ray interaction properties of the alloys for energies wherever analogous experimental results may not be available. This kind of studies can be used for various applications such as for radiation dosimetry, medical and radiation shielding.
Holly R. Trellue
1998-12-01
Monteburns is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code 0RIGEN2. Monteburns produces many criticality and burnup computational parameters based on material feed/removal specifications, power(s), and time intervals. This code processes input from the user indicating the system geometry, initial material compositions, feed/removal, and other code-specific parameters. Results from MCNP, 0RIGEN2, and other calculations are then output successively as the code runs. The principle function of monteburns is to first transfer one-group cross sections and fluxes from MCNP to 0RIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from 0RIGEN2 back to MCNP in a repeated, cyclic fashion. The main requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with 0RIGEN2 and other calculations are performed by monteburns. This report presents the results obtained from the benchmarking of monteburns to measured and previously obtained data from traditional Light Water Reactor systems. The majority of the differences seen between the two were less than five percent. These were primarily a result of variances in cross sections between MCNP, cross section libraries used by other codes, and observed values. With this understanding, this code can now be used with confidence for burnup calculations in three-dimensional systems. It was designed for use in the Accelerator Transmutation of Waste project at Los Alamos National Laboratory but is also being applied to the analysis of isotopic production/destruction of transuranic actinides in a reactor system. The code has now been shown to sufficiently support these calculations.
Zaker, Neda; Zehtabian, Mehdi; Sina, Sedigheh; Koontz, Craig; Meigooni, 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. PMID:27074460
Chen, A Y; Liu, Y-W H; Sheu, R J
2008-01-01
This study investigates the radiation shielding design of the treatment room for boron neutron capture therapy at Tsing Hua Open-pool Reactor using "TORT-coupled MCNP" method. With this method, the computational efficiency is improved significantly by two to three orders of magnitude compared to the analog Monte Carlo MCNP calculation. This makes the calculation feasible using a single CPU in less than 1 day. Further optimization of the photon weight windows leads to additional 50-75% improvement in the overall computational efficiency. PMID:17825572