MCNP capabilities for nuclear well logging calculations
International Nuclear Information System (INIS)
Forster, R.A.; Little, R.C.; Briesmeister, J.F.; Hendricks, J.S.
1990-01-01
The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. This paper discusses how the general-purpose continuous-energy Monte Carlo code MCNP (Monte Carlo neutron photon), part of the LARTCS, provides a computational predictive capability for many applications of interest to the nuclear well logging community. The generalized three-dimensional geometry of MCNP is well suited for borehole-tool models. SABRINA, another component of the LARTCS, is a graphics code that can be used to interactively create a complex MCNP geometry. Users can define many source and tally characteristics with standard MCNP features. The time-dependent capability of the code is essential when modeling pulsed sources. Problems with neutrons, photons, and electrons as either single particle or coupled particles can be calculated with MCNP. The physics of neutron and photon transport and interactions is modeled in detail using the latest available cross-section data
Nuclear densimeter of soil simulated in MCNP-4C code
International Nuclear Information System (INIS)
Braga, Mario R.M.S.S.; Penna, Rodrigo; Vasconcelos, Danilo C.; Pereira, Claubia; Guerra, Bruno T.; Silva, Clemente J.G.C.
2009-01-01
The Monte Carlo code (MCNPX) was used to simulate a nuclear densimeter for measuring soil density. An Americium source (E = 60 keV) and a NaI (Tl) detector were placed on soil surface. Results from MCNP shown that scattered photon fluxes may be used to determining soil density. Linear regressions between scattered photons fluxes and soil density were calculated and shown correlation coefficients near unity. (author)
Verification of MCNP6.2 for Nuclear Criticality Safety Applications
Energy Technology Data Exchange (ETDEWEB)
Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2017-05-10
Several suites of verification/validation benchmark problems were run in early 2017 to verify that the new production release of MCNP6.2 performs correctly for nuclear criticality safety applications (NCS). MCNP6.2 results for several NCS validation suites were compared to the results from MCNP6.1 [1] and MCNP6.1.1 [2]. MCNP6.1 is the production version of MCNP® released in 2013, and MCNP6.1.1 is the update released in 2014. MCNP6.2 includes all of the standard features for NCS calculations that have been available for the past 15 years, along with new features for sensitivity-uncertainty based methods for NCS validation [3]. Results from the benchmark suites were compared with results from previous verification testing [4-8]. Criticality safety analysts should consider testing MCNP6.2 on their particular problems and validation suites. No further development of MCNP5 is planned. MCNP6.1 is now 4 years old, and MCNP6.1.1 is now 3 years old. In general, released versions of MCNP are supported only for about 5 years, due to resource limitations. All future MCNP improvements, bug fixes, user support, and new capabilities are targeted only to MCNP6.2 and beyond.
Using NJOY to Create MCNP ACE Files and Visualize Nuclear Data
Energy Technology Data Exchange (ETDEWEB)
Kahler, Albert Comstock [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2016-10-14
We provide lecture materials that describe the input requirements to create various MCNP ACE files (Fast, Thermal, Dosimetry, Photo-nuclear and Photo-atomic) with the NJOY Nuclear Data Processing code system. Input instructions to visualize nuclear data with NJOY are also provided.
Energy Technology Data Exchange (ETDEWEB)
Cadenas Mendicoa, A. M.
2016-08-01
Due to the lack of graphical representation capability of same nuclear codes like MCNP of GOTHIC, widely used in the industry, the following article describes the development of an interface to use a graphical representation open source (Paraview) with the outputs generated by the nuclear codes. Moreover, this article aims at describing the advantage of this type of visualization programs for the modeling and decision making in the calculation. (Author)
Implementation of a tree algorithm in MCNP code for nuclear well logging applications.
Li, Fusheng; Han, Xiaogang
2012-07-01
The goal of this paper is to develop some modeling capabilities that are missing in the current MCNP code. Those missing capabilities can greatly help for some certain nuclear tools designs, such as a nuclear lithology/mineralogy spectroscopy tool. The new capabilities to be developed in this paper include the following: zone tally, neutron interaction tally, gamma rays index tally and enhanced pulse-height tally. The patched MCNP code also can be used to compute neutron slowing-down length and thermal neutron diffusion length. Copyright © 2011 Elsevier Ltd. All rights reserved.
MCNP benchmark analyses of critical experiments for the Space Nuclear Thermal Propulsion program
International Nuclear Information System (INIS)
Selcow, E.C.; Cerbone, R.J.; Ludewig, H.; Mughabghab, S.F.; Schmidt, E.; Todosow, M.; Parma, E.J.; Ball, R.M.; Hoovler, G.S.
1993-01-01
Benchmark analyses have been performed of Particle Bed Reactor (PBR) critical experiments (CX) using the MCNP radiation transport code. The experiments have been conducted at the Sandia National Laboratory reactor facility in support of the Space Nuclear Thermal Propulsion (SNTP) program. The test reactor is a nineteen element water moderated and reflected thermal system. A series of integral experiments have been carried out to test the capabilities of the radiation transport codes to predict the performance of PBR systems. MCNP was selected as the preferred radiation analysis tool for the benchmark experiments. Comparison between experimental and calculational results indicate close agreement. This paper describes the analyses of benchmark experiments designed to quantify the accuracy of the MCNP radiation transport code for predicting the performance characteristics of PBR reactors
MCNP benchmark analyses of critical experiments for the Space Nuclear Thermal Propulsion program
Selcow, Elizabeth C.; Cerbone, Ralph J.; Ludewig, Hans; Mughabghab, Said F.; Schmidt, Eldon; Todosow, Michael; Parma, Edward J.; Ball, Russell M.; Hoovler, Gary S.
1993-01-01
Benchmark analyses have been performed of Particle Bed Reactor (PBR) critical experiments (CX) using the MCNP radiation transport code. The experiments have been conducted at the Sandia National Laboratory reactor facility in support of the Space Nuclear Thermal Propulsion (SNTP) program. The test reactor is a nineteen element water moderated and reflected thermal system. A series of integral experiments have been carried out to test the capabilities of the radiation transport codes to predict the performance of PBR systems. MCNP was selected as the preferred radiation analysis tool for the benchmark experiments. Comparison between experimental and calculational results indicate close agreement. This paper describes the analyses of benchmark experiments designed to quantify the accuracy of the MCNP radiation transport code for predicting the performance characteristics of PBR reactors.
MCNP and other nuclear codes output graphical representation using python scripts
International Nuclear Information System (INIS)
Cadenas Mendicoa, A. M.
2016-01-01
Due to the lack of graphical representation capability of same nuclear codes like MCNP of GOTHIC, widely used in the industry, the following article describes the development of an interface to use a graphical representation open source (Paraview) with the outputs generated by the nuclear codes. Moreover, this article aims at describing the advantage of this type of visualization programs for the modeling and decision making in the calculation. (Author)
Implementation of a tree algorithm in MCNP code for nuclear well logging applications
Energy Technology Data Exchange (ETDEWEB)
Li Fusheng, E-mail: fusheng.li@bakerhughes.com [Baker Hughes Incorporated, 2001 Rankin Rd. Houston, TX 77073-5101 (United States); Han Xiaogang [Baker Hughes Incorporated, 2001 Rankin Rd. Houston, TX 77073-5101 (United States)
2012-07-15
The goal of this paper is to develop some modeling capabilities that are missing in the current MCNP code. Those missing capabilities can greatly help for some certain nuclear tools designs, such as a nuclear lithology/mineralogy spectroscopy tool. The new capabilities to be developed in this paper include the following: zone tally, neutron interaction tally, gamma rays index tally and enhanced pulse-height tally. The patched MCNP code also can be used to compute neutron slowing-down length and thermal neutron diffusion length. - Highlights: Black-Right-Pointing-Pointer Tree structure programming is suitable for Monte-Carlo based particle tracking. Black-Right-Pointing-Pointer Enhanced pulse height tally is developed for oilwell logging tool simulation. Black-Right-Pointing-Pointer Neutron interaction tally and gamma ray index tally for geochemical logging.
Use experiences of MCNP in nuclear energy study. 2. Review of variance reduction techniques
Energy Technology Data Exchange (ETDEWEB)
Sakurai, Kiyoshi; Yamamoto, Toshihiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; eds.
1998-03-01
`MCNP Use Experience` Working Group was established in 1996 under the Special Committee on Nuclear Code Evaluation. This year`s main activity of the working group has been focused on the review of variance reduction techniques of Monte Carlo calculations. This working group dealt with the variance reduction techniques of (1) neutron and gamma ray transport calculation of fusion reactor system, (2) concept design of nuclear transmutation system using accelerator, (3) JMTR core calculation, (4) calculation of prompt neutron decay constant, (5) neutron and gamma ray transport calculation for exposure evaluation, (6) neutron and gamma ray transport calculation of shielding system, etc. Furthermore, this working group started an activity to compile `Guideline of Monte Carlo Calculation` which will be a standard in the future. The appendices of this report include this `Guideline`, the use experience of MCNP 4B and examples of Monte Carlo calculations of high energy charged particles. The 11 papers are indexed individually. (J.P.N.)
Use experiences of MCNP in nuclear energy study. 2. Review of variance reduction techniques
International Nuclear Information System (INIS)
Sakurai, Kiyoshi; Yamamoto, Toshihiro
1998-03-01
''MCNP Use Experience'' Working Group was established in 1996 under the Special Committee on Nuclear Code Evaluation. This year''s main activity of the working group has been focused on the review of variance reduction techniques of Monte Carlo calculations. This working group dealt with the variance reduction techniques of (1) neutron and gamma ray transport calculation of fusion reactor system, (2) concept design of nuclear transmutation system using accelerator, (3) JMTR core calculation, (4) calculation of prompt neutron decay constant, (5) neutron and gamma ray transport calculation for exposure evaluation, (6) neutron and gamma ray transport calculation of shielding system, etc. Furthermore, this working group started an activity to compile ''Guideline of Monte Carlo Calculation'' which will be a standard in the future. The appendices of this report include this ''Guideline'', the use experience of MCNP 4B and examples of Monte Carlo calculations of high energy charged particles. The 11 papers are indexed individually. (J.P.N.)
About the application of MCNP4 code in nuclear reactor core design calculations
International Nuclear Information System (INIS)
Svarny, J.
2000-01-01
This paper provides short review about application of MCNP code for reactor physics calculations performed in SKODA JS. Problems of criticality safety analysis of spent fuel systems for storage and transport of spent fuel are discussed and relevant applications are presented. Application of standard Monte Carlo code for accelerator driven system for LWR waste destruction is shown and conclusions are reviewed. Specific heterogeneous effects in neutron balance of WWER nuclear cores are solved for adjusting standard design codes. (Authors)
Current status of ACE format libraries for MCNP at nuclear date center of KAERI
Energy Technology Data Exchange (ETDEWEB)
Kim, Do Heon; Gil, Choong Sup; Lee, Young Ouk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2016-09-15
The current status of ACE format MCNP/MCNPX libraries by NDC of KAERI is presented with a short description of each library. Validation calculations with recent nuclear data evaluations ENDF/B-VII.0, ENDF/B-VII.1, JEFF-3.2, and JENDL-4.0 have been carried out by the MCNP5 code for 119 criticality benchmark problems taken from the expanded criticality validation suite supplied by LANL. The overall performances of the ACE format KN-libraries have been analyzed in comparison with the results calculated with the ENDF/B-VII.0-based ENDF70 library of LANL. It was confirmed that the ENDF/B-VII.1-based KNE71 library showed better performances than the others by comparing the RMS errors and χ2 values for five benchmark categories as well as whole benchmark problems. ENDF/B-VII.1 and JEFF-3.2 have a tendency to yield more reliable MCNP calculation results within certain confidence intervals regarding the total uncertainties for the keff values. It is found that the adoption of the latest evaluated nuclear data might ensure better outcomes in various research and development areas.
Comparative Analysis of the Dalat Nuclear Research Reactor with HEU Fuel Using SRAC and MCNP5
Directory of Open Access Journals (Sweden)
Giang Phan
2017-01-01
Full Text Available Neutronics analysis has been performed for the 500 kW Dalat Nuclear Research Reactor loaded with highly enriched uranium fuel using the SRAC code system. The effective multiplication factors, keff, were analyzed for the core at criticality conditions and in two cases corresponding to the complete withdrawal and the full insertion of control rods. MCNP5 calculations were also conducted and compared to that obtained with the SRAC code. The results show that the difference of the keff values between the codes is within 55 pcm. Compared to the criticality conditions established in the experiments, the maximum differences of the keff values obtained from the SRAC and MCNP5 calculations are 119 pcm and 64 pcm, respectively. The radial and axial power peaking factors are 1.334 and 1.710, respectively, in the case of no control rod insertion. At the criticality condition these values become 1.445 and 1.832 when the control rods are partially inserted. Compared to MCNP5 calculations, the deviation of the relative power densities is less than 4% at the fuel bundles in the middle of the core, while the maximum deviation is about 7% appearing at some peripheral bundles. This agreement indicates the verification of the analysis models.
Nuclear reactor multi-physics simulations with coupled MCNP5 and STAR-CCM+
International Nuclear Information System (INIS)
Cardoni, Jeffrey Neil; Rizwan-uddin
2011-01-01
The MCNP5 Monte Carlo particle transport code has been coupled to the computational fluid dynamics code, STAR-CCM+, to provide a high fidelity multi-physics simulation tool for pressurized water nuclear reactors. The codes are executed separately and coupled externally through a Perl script. The Perl script automates the exchange of temperature, density, and volumetric heating information between the codes using ASCII text data files. Fortran90 and Java utility programs assist job automation with data post-processing and file management. The MCNP5 utility code, MAKXSF, pre-generates temperature dependent cross section libraries for the thermal feedback calculations. The MCNP5–STAR-CCM+ coupled simulation tool, dubbed MULTINUKE, was applied to a steady state, PWR cell model to demonstrate its usage and capabilities. The demonstration calculation showed reasonable results that agree with PWR values typically reported in literature. Temperature and fission reaction rate distributions were realistic and intuitive. Reactivity coefficients were also deemed reasonable in comparison to historically reported data. The demonstration problem consisted of 9,984 CFD cells and 7,489 neutronic cells. MCNP5 tallied fission energy deposition over 3,328 UO_2 cells. The coupled solution converged within eight hours and in three MULTINUKE iterations. The simulation was carried out on a 64 bit, quad core, Intel 2.8 GHz microprocessor with 1 GB RAM. The simulations on a quad core machine indicated that a massively parallelized implementation of MULTINUKE can be used to assess larger multi-million cell models. (author)
International Nuclear Information System (INIS)
Hendricks, J.S.
1994-01-01
The MCNP code development program is a relatively large and rapidly changing project in the small and highly-specialized field of radiation transport, specifically radiation protection and shielding. A number of major new MCNP initiatives are described in the subsequent papers in this session. The focus of this paper is the important new developments not described elsewhere and a number of recent developments that have been available since MCNP4A but have gone unnoticed. In particular, we report for the first time a new MCNP quality assurance initiative providing 97% test coverage, a new MCNP feature enabling plotting of nuclear data, and the other new features developed so far for MCNP4B. Finally, an attempt is made to articulate how all these fit together into the overall MCNP development program
International Nuclear Information System (INIS)
Goorley, T.; James, M.; Booth, T.; Brown, F.; Bull, J.; Cox, L.J.; Durkee, J.; Elson, J.; Fensin, M.; Forster, R.A.; Hendricks, J.; Hughes, H.G.; Johns, R.; Kiedrowski, B.; Martz, R.; Mashnik, S.; McKinney, G.; Pelowitz, D.; Prael, R.; Sweezy, J.
2016-01-01
Highlights: • MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, but it is much more than the sum of these two computer codes. • MCNP6 is the result of six years of effort by the MCNP5 and MCNPX code development teams. • These groups of people, residing in Los Alamos National Laboratory’s X Computational Physics Division, Monte Carlo Codes Group (XCP-3) and Nuclear Engineering and Nonproliferation Division, Radiation Transport Modeling Team (NEN-5) respectively, have combined their code development efforts to produce the next evolution of MCNP. • While maintenance and major bug fixes will continue for MCNP5 1.60 and MCNPX 2.7.0 for upcoming years, new code development capabilities only will be developed and released in MCNP6. • In fact, the initial release of MCNP6 contains numerous new features not previously found in either code. • These new features are summarized in this document. • Packaged with MCNP6 is also the new production release of the ENDF/B-VII.1 nuclear data files usable by MCNP. • The high quality of the overall merged code, usefulness of these new features, along with the desire in the user community to start using the merged code, have led us to make the first MCNP6 production release: MCNP6 version 1. • High confidence in the MCNP6 code is based on its performance with the verification and validation test suites, comparisons to its predecessor codes, our automated nightly software debugger tests, the underlying high quality nuclear and atomic databases, and significant testing by many beta testers. - Abstract: MCNP6 can be described as the merger of MCNP5 and MCNPX capabilities, but it is much more than the sum of these two computer codes. MCNP6 is the result of six years of effort by the MCNP5 and MCNPX code development teams. These groups of people, residing in Los Alamos National Laboratory’s X Computational Physics Division, Monte Carlo Codes Group (XCP-3) and Nuclear Engineering and
Modeling of a planning system in radiotherapy and Nuclear Medicine using the MCNP6 code
International Nuclear Information System (INIS)
Massicano, Felipe
2015-01-01
Cancer therapy has many branches and one of them is the use of radiation sources as treatment leading method. Radiotherapy and nuclear medicine are examples of these treatment types. For using the ionization radiation as main tool for the therapy, there is the need of crafting many treatment simulation in order to maximum the tumoral tissue dose without surpass the dose limit in health tissue surrounding. Treatment planning systems (TPS) are systems which have the purpose of simulating these therapy types. Nuclear medicine and radiotherapy have many distinct features linked to the therapy mode and consequently they have different TPS destined for each. The radiotherapy TPS is more developed than the nuclear medicine TPS and by that reason the development of a TPS that was similar to the radiotherapy TPS, but enough generic for include other therapy types, it will contribute with significant advances in nuclear medicine and in others therapy types with radiation. Based on this, the goal of work was to model a TPS that utilizes the Monte Carlo N-Particle Transport code (MCNP6) in order to simulate radiotherapy therapy, nuclear medicine therapy and with potential for simulating other therapy types too. The result of this work was the creation of a Framework in Java language, object oriented, named IBMC which will assist in the development of new TPS with MCNP6 code. The IBMC allowed to develop rapidly and easily TPS for radiotherapy and nuclear medicine and the results were validated with systems already consolidated. The IBMC showed high potential for developing TPS by new therapy types. (author)
Benchmarking of MCNP for calculating dose rates at an interim storage facility for nuclear waste.
Heuel-Fabianek, Burkhard; Hille, Ralf
2005-01-01
During the operation of research facilities at Research Centre Jülich, Germany, nuclear waste is stored in drums and other vessels in an interim storage building on-site, which has a concrete shielding at the side walls. Owing to the lack of a well-defined source, measured gamma spectra were unfolded to determine the photon flux on the surface of the containers. The dose rate simulation, including the effects of skyshine, using the Monte Carlo transport code MCNP is compared with the measured dosimetric data at some locations in the vicinity of the interim storage building. The MCNP data for direct radiation confirm the data calculated using a point-kernel method. However, a comparison of the modelled dose rates for direct radiation and skyshine with the measured data demonstrate the need for a more precise definition of the source. Both the measured and the modelled dose rates verified the fact that the legal limits (<1 mSv a(-1)) are met in the area outside the perimeter fence of the storage building to which members of the public have access. Using container surface data (gamma spectra) to define the source may be a useful tool for practical calculations and additionally for benchmarking of computer codes if the discussed critical aspects with respect to the source can be addressed adequately.
International Nuclear Information System (INIS)
Naito, Yoshitaka
2001-01-01
To assist succeeding reports which will be presented in this research meeting, following items on the computer code MCNP developed in USA are presented: (1) history of development of MCNP, (2) meaning of the development, (3) progress of study on Monte Carlo codes in the nuclear code committee and (4) expectation to Monte Carlo codes. (author)
MCNP Progress & Performance Improvements
Energy Technology Data Exchange (ETDEWEB)
Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bull, Jeffrey S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2015-04-14
Twenty-eight slides give information about the work of the US DOE/NNSA Nuclear Criticality Safety Program on MCNP6 under the following headings: MCNP6.1.1 Release, with ENDF/B-VII.1; Verification/Validation; User Support & Training; Performance Improvements; and Work in Progress. Whisper methodology will be incorporated into the code, and run speed should be increased.
Hartling, K.; Ciungu, B.; Li, G.; Bentoumi, G.; Sur, B.
2018-05-01
Monte Carlo codes such as MCNP and Geant4 rely on a combination of physics models and evaluated nuclear data files (ENDF) to simulate the transport of neutrons through various materials and geometries. The grid representation used to represent the final-state scattering energies and angles associated with neutron scattering interactions can significantly affect the predictions of these codes. In particular, the default thermal scattering libraries used by MCNP6.1 and Geant4.10.3 do not accurately reproduce the ENDF/B-VII.1 model in simulations of the double-differential cross section for thermal neutrons interacting with hydrogen nuclei in a thin layer of water. However, agreement between model and simulation can be achieved within the statistical error by re-processing ENDF/B-VII.I thermal scattering libraries with the NJOY code. The structure of the thermal scattering libraries and sampling algorithms in MCNP and Geant4 are also reviewed.
International Nuclear Information System (INIS)
Campolina, Daniel de Almeida Magalhaes
2009-01-01
In Many situations of nuclear system study, it is necessary to know the detailed particle flux in a geometry. Deterministic 1-D and 2-D methods aren't suitable to represent some strong 3-D behavior configurations, for example in cores where the neutron flux varies considerably in the space and Monte Carlo analysis are necessary. The majority of Monte Carlo transport calculation codes, performs time static simulations, in terms of fuel isotopic composition. This work is a initial project to incorporate depletion capability to the MCNP code, by means of a connection with ORIGEN2.1 burnup code. The method to develop the program proposed followed the methodology of other programs used to the same purpose. Essentially, MCNP data library are used to generate one group microscopic cross sections that override default ORIGEN libraries. To verify the actual implemented part, comparisons which MCNPX (version 2.6.0) results were made. The neutron flux and criticality value of core agree. The neutron flux and criticality value of the core agree, especially in beginning of burnup when the influence of fission products are not very considerable. The small difference encountered was probably caused by the difference in the number of isotopes considered in the transport models (89 MCNPX x 25 GB). Next step of this work is to adapt MCNP version 4C to work with a memory higher than its standard value (4MB), in order to allow a greater number of isotopes in the transport model. (author)
International Nuclear Information System (INIS)
Valentine, T.E.
1997-01-01
The Monte Carlo code MCNP-DSP was developed from the Los Alamos MCNP4a code to calculate the time and frequency response statistics obtained from the 252 Cf-source-driven frequency analysis measurements. This code can be used to validate calculational methods and cross section data sets from subcritical experiments. This code provides a more general model for interpretation and planning of experiments for nuclear criticality safety, nuclear safeguards, and nuclear weapons identification and replaces the use of point kinetics models for interpreting the measurements. The use of MCNP-DSP extends the usefulness of this measurement method to systems with much lower neutron multiplication factors
MCNP Perturbation Capability for Monte Carlo Criticality Calculations
International Nuclear Information System (INIS)
Hendricks, J.S.; Carter, L.L.; McKinney, G.W.
1999-01-01
The differential operator perturbation capability in MCNP4B has been extended to automatically calculate perturbation estimates for the track length estimate of k eff in MCNP4B. The additional corrections required in certain cases for MCNP4B are no longer needed. Calculating the effect of small design changes on the criticality of nuclear systems with MCNP is now straightforward
International Nuclear Information System (INIS)
Romdhani, Ibtissem
2014-01-01
As part of developing its nuclear infrastructure base, the National Science and Technology Center Nuclear (CNSTN) examines the technical feasibility of setting up a new installation of subcritical assembly. Our study focuses on determining the neutron parameters of a nuclear zero power reactor based on Monte Carlo simulation MCNP. The objective of the simulation is to model the installation, determine the effective multiplication factor, and spatial distribution of neutron flux.
International Nuclear Information System (INIS)
Katalenich, Jeff; Flaska, Marek; Pozzi, Sara A.; Hartman, Michael R.
2011-01-01
Fast and robust methods for interrogation of special nuclear material (SNM) are of interest to many agencies and institutions in the United States. It is well known that passive interrogation methods are typically sufficient for plutonium identification because of a relatively high neutron production rate from 240 Pu . On the other hand, identification of shielded uranium requires active methods using neutron or photon sources . Deuterium-deuterium (2.45 MeV) and deuterium-tritium (14.1 MeV) neutron-generator sources have been previously tested and proven to be relatively reliable instruments for active interrogation of nuclear materials . In addition, the newest generators of this type are small enough for applications requiring portable interrogation systems. Active interrogation techniques using high-energy neutrons are being investigated as a method to detect hidden SNM in shielded containers . Due to the thickness of some containers, penetrating radiation such as high-energy neutrons can provide a potential means of probing shielded SNM. In an effort to develop the capability to assess the signal seen from various forms of shielded nuclear materials, University of Michigan Neutron Science Laboratory's D-T neutron generator and its shielding were accurately modeled in MCNP. The generator, while operating at nominal power, produces approximately 1x10 10 neutrons/s, a source intensity which requires a large amount of shielding to minimize the dose rates around the generator. For this reason, the existing shielding completely encompasses the generator and does not include beam ports. Therefore, several MCNP simulations were performed to estimate the yield of uncollided 14.1-MeV neutrons from the generator for active interrogation experiments. Beam port diameters of 5, 10, 15, 20, and 25 cm were modeled to assess the resulting neutron fluxes. The neutron flux outside the beam ports was estimated to be approximately 2x10 4 n/cm 2 s.
International Nuclear Information System (INIS)
Love, E.F.; Pauley, K.A.; Reid, B.D.
1995-09-01
This study describes the use of the Monte Carlo Neutron-Photon (MCNP) code for determining activation levels of irradiated reactor vessel internals hardware. The purpose of the analysis is to produce data for the Department of Energy's Greater-Than-Class C Low-Level Radioactive Waste Program. An MCNP model was developed to analyze the Yankee Rowe reactor facility. The model incorporates reactor geometry, material compositions, and operating history data acquired from Yankee Atomic Electric Company. In addition to the base activation analysis, parametric studies were performed to determine the sensitivity of activation to specific parameters. A component sampling plan was also developed to validate the model results, although the plan was not implemented. The calculations for the Yankee Rowe reactor predict that only the core baffle and the core support plates will be activated to levels above the Class C limits. The parametric calculations show, however, that the large uncertainties in the material compositions could cause errors in the estimates that could also increase the estimated activation level of the core barrel to above the Class C limits. Extrapolation of the results to other reactor facilities indicates that in addition to the baffle and support plates, core barrels may also be activated to above Class C limits; however the classification will depend on the specific operating conditions of the reactor and the specific material compositions of the metal, as well as the use of allowable concentration averaging practices in packaging and classifying the waste
The comparison of MCNP perturbation technique with MCNP difference method in critical calculation
International Nuclear Information System (INIS)
Liu Bin; Lv Xuefeng; Zhao Wei; Wang Kai; Tu Jing; Ouyang Xiaoping
2010-01-01
For a nuclear fission system, we calculated Δk eff , which arise from system material composition changes, by two different approaches, the MCNP perturbation technique and the MCNP difference method. For every material composition change, we made four different runs, each run with different cycles or each cycle generating different neutrons, then we compared the two Δk eff that are obtained by two different approaches. As a material composition change in any particular cell of the nuclear fission system is small compared to the material compositions in the whole nuclear fission system, in other words, this composition change can be treated as a small perturbation, the Δk eff results obtained from the MCNP perturbation technique are much quicker, much more efficient and reliable than the results from the MCNP difference method. When a material composition change in any particular cell of the nuclear fission system is significant compared to the material compositions in the whole nuclear fission system, both the MCNP perturbation technique and the MCNP difference method can give satisfactory results. But for the run with the same cycles and each cycle generating the same neutrons, the results obtained from the MCNP perturbation technique are systemically less than the results obtained from the MCNP difference method. To further confirm our calculation results from the MCNP4C, we run the exact same MCNP4C input file in MCNP5, the calculation results from MCNP5 are the same as the calculation results from MCNP4C. We need caution when using the MCNP perturbation technique to calculate the Δk eff as the material composition change is large compared to the material compositions in the whole nuclear fission system, even though the material composition changes of any particular cell of the fission system still meet the criteria of MCNP perturbation technique.
International Nuclear Information System (INIS)
Goorley, John T.
2012-01-01
We, the development teams for MCNP, NJOY, and parts of ENDF, would like to invite you to a proposed 3 day workshop October 30, 31 and November 1 2012, to be held at Los Alamos National Laboratory. At this workshop, we will review new and developing missions that MCNP6 and the underlying nuclear data are being asked to address. LANL will also present its internal plans to address these missions and recent advances in these three capabilities and we will be interested to hear your input on these topics. Additionally we are interested in hearing from you additional technical advances, missions, concerns, and other issues that we should be considering for both short term (1-3 years) and long term (4-6 years)? What are the additional existing capabilities and methods that we should be investigating? The goal of the workshop is to refine priorities for mcnp6 transport methods, algorithms, physics, data and processing as they relate to the intersection of MCNP, NJOY and ENDF.
International Nuclear Information System (INIS)
Braga, Mario R.M.S.S.; Oliveira, Arno H.; Lima, Claubia P.B.
2013-01-01
The aim of this work is to evaluate the behavior of the variation the elements: Mg, Ca, Fe in the soils composition on a nuclear probe to measure the density of porous materials nondestructive in testing based on coherent Compton Effect, the effect Rayleigh. To study the effect of composition in soil was used nuclear code MCNP4X where was simulated two sources, a source 14mCi americium-241 and other source 4mCi cesium-137, lead shielding and volume scintillator. To avoid problems with geometries were simulated spheres with 1.00 meters of diameter filled with soil to be evaluated. Data analysis allowed establishing correction parameters for nuclear probe. (author)
International Nuclear Information System (INIS)
Cramer, S.N.
1984-01-01
The MCNP code is the major Monte Carlo coupled neutron-photon transport research tool at the Los Alamos National Laboratory, and it represents the most extensive Monte Carlo development program in the United States which is available in the public domain. The present code is the direct descendent of the original Monte Carlo work of Fermi, von Neumaum, and Ulam at Los Alamos in the 1940s. Development has continued uninterrupted since that time, and the current version of MCNP (or its predecessors) has always included state-of-the-art methods in the Monte Carlo simulation of radiation transport, basic cross section data, geometry capability, variance reduction, and estimation procedures. The authors of the present code have oriented its development toward general user application. The documentation, though extensive, is presented in a clear and simple manner with many examples, illustrations, and sample problems. In addition to providing the desired results, the output listings give a a wealth of detailed information (some optional) concerning each state of the calculation. The code system is continually updated to take advantage of advances in computer hardware and software, including interactive modes of operation, diagnostic interrupts and restarts, and a variety of graphical and video aids
ZZ-FSXLIB-JD99, MCNP nuclear data library based on JENDL Dosimetry File 99
International Nuclear Information System (INIS)
Shibata, Keiichi
2007-01-01
Description: JENDL Dosimetry File 99 processed into ACE for Monte Carlo calculations. JENDL/D-99 based MCNP library. Format: ACE. Number of groups: Continuous energy cross section library. Nuclides: 47 Nuclides and 67 reactions: Li-6 (n, triton) alpha; Li-6 alpha-production; Li-7 triton- production; B-10 (n, alpha) Li-7; B-10 alpha-production; F-19 (n, 2n) F-18; Na-23 (n, 2n) Na-22; Na-23 (n, gamma) Na-24; Mg-24 (n, p) Na-24; Al-27 (n, p) Mg-27; Al-27 (n, alpha) Na-24; P-31 (n, p) Si-31; S-32 (n, p) P-32; Sc-45 (n, gamma) Sc-46; Ti-nat (n, x) Sc-46; Ti-nat (n, x) Sc-47; Ti-nat (n, x) Sc-48; Ti-46 (n, 2n) Ti-45; Ti-46 (n, p) Sc-46; Ti-47 (n, np) Sc-46; Ti-47 (n, p) Sc-47; Ti-48 (n, np) Sc-47; Ti-48 (n, p) Sc-48; Ti-49 (n, np) Sc-48; Cr-50 (n, gamma) Cr-51; Cr-52 (n, 2n) Cr-51; Mn-55 (n, 2n) Mn-54; Mn-55 (n, gamma) Mn-56; Fe-54 (n, p) Mn-54; Fe-56 (n, p) Mn-56; Fe-57 (n, np) Mn-56; Fe-58 (n, gamma) Fe-59; Co-59 (n, 2n) Co-58; Co-59 (n, gamma) Co-60; Co-59 (n, alpha) Mn-56; Ni-58 (n, 2n) Ni-57; Ni-58 (n, p) Co-58; Ni-60 (n, p) Co-60; Cu-63 (n, 2n) Cu-62; Cu-63 (n, gamma) Cu-64; Cu-63 (n, alpha) Co-60; Cu-65 (n, 2n) Cu-64; Zn-64 (n, p) Cu-64; Y-89 (n, 2n) Y-88; Zr-90 (n, 2n) Zr-89; Nb-93 (n, n') Nb-93m; Nb-93 (n, 2n) Nb-92m; Rh-103 (n, n') Rh-103m; Ag-109 (n, gamma) Ag-110m; In-115 (n, n') In-115m; In-115 (n, gamma) In-116m; I-127 (n,2n) I-126; Eu-151 (n, gamma) Eu-152; Tm-169 (n,2n) Tm-168; Ta-181 (n, gamma) Ta-182; W-186 (n, gamma) W-187; Au-197 (n, 2n) Au-196; Au-197 (n, gamma) Au-198; Hg-199 (n, n') Hg-199m; Th-232 - fission; Th-232 (n, gamma) Th-233; U-235 - fission; U-238 - fission; U-238 (n, gamma) U-239; Np-237 - fission; Pu-239 - fission; Am-241 - fission. The data were produced on the 31 of March, 2006
Energy Technology Data Exchange (ETDEWEB)
Mosteller, R. D. (Russell D.)
2004-01-01
The MCNP Criticality Validation Suite is a collection of 31 benchmarks taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments. MCNP5 calculations clearly demonstrate that, overall, nuclear data for a preliminary version of ENDFB-VII produce better agreement with the benchmarks in the suite than do corresponding data from ENDF/B-VI. Additional calculations identify areas where improvements in the data still are needed. Based on results for the MCNP Criticality Validation Suite, the Pre-ENDF/B-VII nuclear data produce substantially better overall results than do their ENDF/B-VI counterparts. The calculated values for k{sub eff} for bare metal spheres and for an IEU cylinder reflected by normal uranium are in much better agreement with the benchmark values. In addition, the values of k{sub eff} for the bare metal spheres are much more consistent with those for corresponding metal spheres reflected by normal uranium or water. In addition, a long-standing controversy about the need for an ad hoc adjustment to the {sup 238}U resonance integral for thermal systems may finally be resolved. On the other hand, improvements still are needed in a number of areas. Those areas include intermediate-energy cross sections for {sup 235}U, angular distributions for elastic scattering in deuterium, and fast cross sections for {sup 237}Np.
Calculation of neutron activation of control rods of a nuclear reactor, using MCNP5
International Nuclear Information System (INIS)
Pena V, J.D.
2016-01-01
The control rods of a nuclear reactor are activated by neutron irradiation. The generated activity produces a dose around the rod which is irrelevant inside the reactor, but significant when the rod is withdrawn and placed in a storage pool, because this dose is a potential risk to the surrounding personnel. On the other hand, most of the activation occurs in the stainless steel components of the rod. The Monte Carlo model can reliably determine the activation produced in a stainless steel part exposed to a neutron flux in a reactor and the dose measurement around this part. This thesis presents the Monte Carlo models developed for the activation of the control rods of the TRIGA Mark III reactor of Instituto Nacional de Investigaciones Nucleares (ININ) when only standard fuel was available. Therefore, the validations of the Monte Carlo models are reliable. (Author)
International Nuclear Information System (INIS)
Hendricks, J.S.; Frankle, S.C.; Court, J.D.
1994-01-01
We report here for the first time the availability of an official set of ENDF/B-VI neutron data for MCNP(trademark). The LANL Radiation Transport group engaged the Nuclear Theory and Applications Group to construct a complete library based on ENDF/B-VI Release in the Spring of 1994. A new and thorough set of quality assurance tests was established and data passing those tests were subject only to a limited set of benchmarking tests. All nuclides were subjected to infinite medium calculations. The fissionable materials were benchmarked against critical assemblies, and 28 nuclides were benchmarked against the LLNL pulsed sphere experiments
MCNP benchmark analyses of critical experiments for space nuclear thermal propulsion
International Nuclear Information System (INIS)
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
MCNP variance reduction overview
International Nuclear Information System (INIS)
Hendricks, J.S.; Booth, T.E.
1985-01-01
The MCNP code is rich in variance reduction features. Standard variance reduction methods found in most Monte Carlo codes are available as well as a number of methods unique to MCNP. We discuss the variance reduction features presently in MCNP as well as new ones under study for possible inclusion in future versions of the code
Status of electron transport in MCNP trademark
International Nuclear Information System (INIS)
Hughes, H.G.
1997-01-01
The latest version of MCNP, the Los Alamos Monte Carlo transport code, has now been officially released. MCNP4B has been sent to the Radiation Safety Information Computational Center (RSICC), in Oak Ridge, Tennessee, which is responsible for the further distribution of the code within the US. International distribution of MCNP is done by the Nuclear Energy Agency (ECD/NEA), in Paris, France. Readers with access to the World-Wide-Web should consult the MCNP distribution site http://www-xdiv.lanl.gov/XTM/mcnp/about.html for specific information about contacting RSICC and OECD/NEA. A variety of new features are available in MCNP4B. Among these are differential operator perturbations, cross-section plotting capabilities, enhanced diagnostics for transport in repeated structures and lattices, improved efficiency in distributed-memory multiprocessing, corrected particle lifetime and lifespan estimators, and expanded software quality assurance procedures and testing, including testing of the multigroup Boltzmann-Fokker-Planck capability. New and improved cross section sets in the form of ENDF/B-VI evaluations have also been recently released and can be used in MCNP4B. Perhaps most significant for the interests of this special session, the electron transport algorithm has been improved, especially in the collisional energy-loss straggling and the angular-deflection treatments. In this paper, the author concentrates on a fairly complete documentation of the current status of the electron transport methods in MCNP
Criticality calculations with MCNP trademark: A primer
International Nuclear Information System (INIS)
Harmon, C.D. II; Busch, R.D.; Briesmeister, J.F.; Forster, R.A.
1994-01-01
With the closure of many experimental facilities, the nuclear criticality safety analyst increasingly is required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, in many cases, the analyst has little experience with the specific codes available at his/her facility. This primer will help you, the analyst, understand and use the MCNP Monte Carlo code for nuclear criticality safety analyses. It assumes that you have a college education in a technical field. There is no assumption of familiarity with Monte Carlo codes in general or with MCNP in particular. Appendix A gives an introduction to Monte Carlo techniques. The primer is designed to teach by example, with each example illustrating two or three features of MCNP that are useful in criticality analyses. Beginning with a Quickstart chapter, the primer gives an overview of the basic requirements for MCNP input and allows you to run a simple criticality problem with MCNP. This chapter is not designed to explain either the input or the MCNP options in detail; but rather it introduces basic concepts that are further explained in following chapters. Each chapter begins with a list of basic objectives that identify the goal of the chapter, and a list of the individual MCNP features that are covered in detail in the unique chapter example problems. It is expected that on completion of the primer you will be comfortable using MCNP in criticality calculations and will be capable of handling 80 to 90 percent of the situations that normally arise in a facility. The primer provides a set of basic input files that you can selectively modify to fit the particular problem at hand
International Nuclear Information System (INIS)
Pauzi, A M
2013-01-01
The neutron transport code, Monte Carlo N-Particle (MCNP) which was wellkown as the gold standard in predicting nuclear reaction was used to model the small nuclear reactor core called U -battery TM, which was develop by the University of Manchester and Delft Institute of Technology. The paper introduces on the concept of modeling the small reactor core, a high temperature reactor (HTR) type with small coated TRISO fuel particle in graphite matrix using the MCNPv4C software. The criticality of the core were calculated using the software and analysed by changing key parameters such coolant type, fuel type and enrichment levels, cladding materials, and control rod type. The criticality results from the simulation were validated using the SCALE 5.1 software by [1] M Ding and J L Kloosterman, 2010. The data produced from these analyses would be used as part of the process of proposing initial core layout and a provisional list of materials for newly design reactor core. In the future, the criticality study would be continued with different core configurations and geometries.
MCNP: Photon benchmark problems
International Nuclear Information System (INIS)
Whalen, D.J.; Hollowell, D.E.; Hendricks, J.S.
1991-09-01
The recent widespread, markedly increased use of radiation transport codes has produced greater user and institutional demand for assurance that such codes give correct results. Responding to these pressing requirements for code validation, the general purpose Monte Carlo transport code MCNP has been tested on six different photon problem families. MCNP was used to simulate these six sets numerically. Results for each were compared to the set's analytical or experimental data. MCNP successfully predicted the analytical or experimental results of all six families within the statistical uncertainty inherent in the Monte Carlo method. From this we conclude that MCNP can accurately model a broad spectrum of photon transport problems. 8 refs., 30 figs., 5 tabs
Directory of Open Access Journals (Sweden)
Kabach Ouadie
2017-12-01
Full Text Available To validate the new Evaluated Nuclear Data File (ENDF/B-VIII.0β4 library, 31 different critical cores were selected and used for a benchmark test of the important parameter keff. The four utilized libraries are processed using Nuclear Data Processing Code (NJOY2016. The results obtained with the ENDF/B-VIII.0β4 library were compared against those calculated with ENDF/B-VI.8, ENDF/B-VII.0, and ENDF/B-VII.1 libraries using the Monte Carlo N-Particle (MCNP(X code. All the MCNP(X calculations of keff values with these four libraries were compared with the experimentally measured results, which are available in the International Critically Safety Benchmark Evaluation Project. The obtained results are discussed and analyzed in this paper.
Development of automatic editing system for MCNP library 'autonj'
International Nuclear Information System (INIS)
Maekawa, Fujio; Sakurai, Kiyoshi; Kume, Etsuo; Nomura, Yasushi; Kosako, Kazuaki; Kawasaki, Nobuo; Naito, Yoshitaka
1999-12-01
As an activity of the MCNP High-Temperature Library Production Working Group under the Nuclear Code Evaluation Special Committee of Nuclear Code Committee, the automatic editing system for MCNP library 'autonj' was developed. The autonj includes the NJOY-97 code as its main body, and is a system that enables us to easily produce cross section libraries for MCNP from evaluated nuclear data files such as JENDL-3.2. A temperature dependent library at six temperature points based on JENDL-3.2 was produced by using autonj. The autonj system and the temperature dependent library were installed on the JAERI AP3000 computer. (author)
Adjoint-Based Uncertainty Quantification with MCNP
Energy Technology Data Exchange (ETDEWEB)
Seifried, Jeffrey E. [Univ. of California, Berkeley, CA (United States)
2011-09-01
This work serves to quantify the instantaneous uncertainties in neutron transport simulations born from nuclear data and statistical counting uncertainties. Perturbation and adjoint theories are used to derive implicit sensitivity expressions. These expressions are transformed into forms that are convenient for construction with MCNP6, creating the ability to perform adjoint-based uncertainty quantification with MCNP6. These new tools are exercised on the depleted-uranium hybrid LIFE blanket, quantifying its sensitivities and uncertainties to important figures of merit. Overall, these uncertainty estimates are small (< 2%). Having quantified the sensitivities and uncertainties, physical understanding of the system is gained and some confidence in the simulation is acquired.
Possible Improvements to MCNP6 and its CEM/LAQGSM Event-Generators
Energy Technology Data Exchange (ETDEWEB)
Mashnik, Stepan Georgievich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2015-08-04
This report is intended to the MCNP6 developers and sponsors of MCNP6. It presents a set of suggested possible future improvements to MCNP6 and to its CEM03.03 and LAQGSM03.03 event-generators. A few suggested modifications of MCNP6 are quite simple, aimed at avoiding possible problems with running MCNP6 on various computers, i.e., these changes are not expected to change or improve any results, but should make the use of MCNP6 easier; such changes are expected to require limited man-power resources. On the other hand, several other suggested improvements require a serious further development of nuclear reaction models, are expected to improve significantly the predictive power of MCNP6 for a number of nuclear reactions; but, such developments require several years of work by real experts on nuclear reactions.
CTEx Beowulf cluster for MCNP performance
International Nuclear Information System (INIS)
Gonzaga, Roberto N.; Amorim, Aneuri S. de; Balthar, Mario Cesar V.
2011-01-01
This work is an introduction to the CTEx Nuclear Defense Department's Beowulf Cluster. Building a Beowulf Cluster is a complex learning process that greatly depends upon your hardware and software requirements. The feasibility and efficiency of performing MCNP5 calculations with a small, heterogeneous computing cluster built in Red Hat's Fedora Linux operating system personal computers (PC) are explored. The performance increases that may be expected with such clusters are estimated for cases that typify general radiation transport calculations. Our results show that the speed increase from additional slave PCs is nearly linear up to 10 processors. The pre compiled parallel binary version of MCNP uses the Message-Passing Interface (MPI) protocol. The use of this pre compiled parallel version of MCNP5 with the MPI protocol on a small, heterogeneous computing cluster built from Red Hat's Fedora Linux operating system PCs is the subject of this work. (author)
Utilization of MCNP code in the research and design for China advanced research reactor
International Nuclear Information System (INIS)
Shen Feng
2006-01-01
MCNP, which is the internationalized neutronics code, is used for nuclear research and design in China Advanced Research Reactor (CARR). MCNP is an important neutronics code in the research and design for CARR since many calculation tasks could be undertaken by it. Many nuclear parameters on reactor core, the design and optimization research for many reactor utilizations, much verification for other nuclear calculation code and so on are conducted with help of MCNP. (author)
Energy Technology Data Exchange (ETDEWEB)
Pena V, J.D.
2016-07-01
The control rods of a nuclear reactor are activated by neutron irradiation. The generated activity produces a dose around the rod which is irrelevant inside the reactor, but significant when the rod is withdrawn and placed in a storage pool, because this dose is a potential risk to the surrounding personnel. On the other hand, most of the activation occurs in the stainless steel components of the rod. The Monte Carlo model can reliably determine the activation produced in a stainless steel part exposed to a neutron flux in a reactor and the dose measurement around this part. This thesis presents the Monte Carlo models developed for the activation of the control rods of the TRIGA Mark III reactor of Instituto Nacional de Investigaciones Nucleares (ININ) when only standard fuel was available. Therefore, the validations of the Monte Carlo models are reliable. (Author)
Estimation and interpretation of keff confidence intervals in MCNP
International Nuclear Information System (INIS)
Urbatsch, T.J.
1995-01-01
MCNP has three different, but correlated, estimators for Calculating k eff in nuclear criticality calculations: collision, absorption, and track length estimators. The combination of these three estimators, the three-combined k eff estimator, is shown to be the best k eff estimator available in MCNP for estimating k eff confidence intervals. Theoretically, the Gauss-Markov Theorem provides a solid foundation for MCNP's three-combined estimator. Analytically, a statistical study, where the estimates are drawn using a known covariance matrix, shows that the three-combined estimator is superior to the individual estimator with the smallest variance. The importance of MCNP's batch statistics is demonstrated by an investigation of the effects of individual estimator variance bias on the combination of estimators, both heuristically with the analytical study and emprically with MCNP
Estimation and interpretation of keff confidence intervals in MCNP
International Nuclear Information System (INIS)
Urbatsch, T.J.
1995-01-01
The Monte Carlo code MCNP has three different, but correlated, estimators for calculating k eff in nuclear criticality calculations: collision, absorption, and track length estimators. The combination of these three estimators, the three-combined k eff estimator, is shown to be the best k eff estimator available in MCNP for estimating k eff confidence intervals. Theoretically, the Gauss-Markov theorem provides a solid foundation for MCNP's three-combined estimator. Analytically, a statistical study, where the estimates are drawn using a known covariance matrix, shows that the three-combined estimator is superior to the estimator with the smallest variance. Empirically, MCNP examples for several physical systems demonstrate the three-combined estimator's superiority over each of the three individual estimators and its correct coverage rates. Additionally, the importance of MCNP's statistical checks is demonstrated
International Nuclear Information System (INIS)
Hendricks, J.S.; Briesmeister, J.F.
1991-01-01
MCNP is a widely used and actively developed Monte Carlo radiation transport code. Many important features have recently been added and more are under development. Benchmark studies not only indicate that MCNP is accurate but also that modern computer codes can give answers basically as accurate as the physics data that goes in them. Even deep penetration problems can be correct to within a factor of two after 10 to 25 mean free paths of penetration. And finally, Monte Carlo calculations, once thought to be too expensive to run routinely, can now be run effectively on desktop computers which compete with the supercomputers of yesteryear. 21 refs., 3 tabs
International Nuclear Information System (INIS)
Hendricks, J.S.; Whalen, D.J.; Cardon, D.A.; Uhle, J.L.
1991-01-01
Over 50 neutron benchmark calculations have recently been completed as part of an ongoing program to validate the MCNP Monte Carlo radiation transport code. The new and significant aspects of this work are as follows: These calculations are the first attempt at a validation program for MCNP and the first official benchmarking of version 4 of the code. We believe the chosen set of benchmarks is a comprehensive set that may be useful for benchmarking other radiation transport codes and data libraries. These calculations provide insight into how well neutron transport calculations can be expected to model a wide variety of problems
Using Machine Learning to Predict MCNP Bias
Energy Technology Data Exchange (ETDEWEB)
Grechanuk, Pavel Aleksandrovi [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2018-01-09
For many real-world applications in radiation transport where simulations are compared to experimental measurements, like in nuclear criticality safety, the bias (simulated - experimental k_{eff}) in the calculation is an extremely important quantity used for code validation. The objective of this project is to accurately predict the bias of MCNP6 [1] criticality calculations using machine learning (ML) algorithms, with the intention of creating a tool that can complement the current nuclear criticality safety methods. In the latest release of MCNP6, the Whisper tool is available for criticality safety analysts and includes a large catalogue of experimental benchmarks, sensitivity profiles, and nuclear data covariance matrices. This data, coming from 1100+ benchmark cases, is used in this study of ML algorithms for criticality safety bias predictions.
A Microsoft Windows version of the MCNP visual editor
International Nuclear Information System (INIS)
Schwarz, R.A.; Carter, L.L.; Pfohl, J.
1999-01-01
Work has started on a Microsoft Windows version of the MCNP visual editor. The MCNP visual editor provides a graphical user interface for displaying and creating MCNP geometries. The visual editor is currently available from the Radiation Safety Information Computational Center (RSICC) and the Nuclear Energy Agency (NEA) as software package PSR-358. It currently runs on the major UNIX platforms (IBM, SGI, HP, SUN) and Linux. Work has started on converting the visual editor to work in a Microsoft Windows environment. This initial work focuses on converting the display capabilities of the visual editor; the geometry creation capability of the visual editor may be included in future upgrades
International Nuclear Information System (INIS)
Sood, Avnet; Forster, R. Arthur; Parsons, D. Kent
2001-01-01
Monte Carlo simulations of nuclear criticality eigenvalue problems are often performed by general purpose radiation transport codes such as MCNP. MCNP performs detailed statistical analysis of the criticality calculation and provides feedback to the user with warning messages, tables, and graphs. The purpose of the analysis is to provide the user with sufficient information to assess spatial convergence of the eigenfunction and thus the validity of the criticality calculation. As a test of this statistical analysis package in MCNP, analytic criticality verification benchmark problems have been used for the first time to assess the performance of the criticality convergence tests in MCNP. The MCNP statistical analysis capability has been recently assessed using the 75 multigroup criticality verification analytic problem test set. MCNP was verified with these problems at the 10 -4 to 10 -5 statistical error level using 40 000 histories per cycle and 2000 active cycles. In all cases, the final boxed combined k eff answer was given with the standard deviation and three confidence intervals that contained the analytic k eff . To test the effectiveness of the statistical analysis checks in identifying poor eigenfunction convergence, ten problems from the test set were deliberately run incorrectly using 1000 histories per cycle, 200 active cycles, and 10 inactive cycles. Six problems with large dominance ratios were chosen from the test set because they do not achieve the normal spatial mode in the beginning of the calculation. To further stress the convergence tests, these problems were also started with an initial fission source point 1 cm from the boundary thus increasing the likelihood of a poorly converged initial fission source distribution. The final combined k eff confidence intervals for these deliberately ill-posed problems did not include the analytic k eff value. In no case did a bad confidence interval go undetected. Warning messages were given signaling that
International Nuclear Information System (INIS)
Mosteller, Russell D.
2002-01-01
Two validation suites, one for criticality and another for radiation shielding, have been defined and tested for the MCNP Monte Carlo code. All of the cases in the validation suites are based on experiments so that calculated and measured results can be compared in a meaningful way. The cases in the validation suites are described, and results from those cases are discussed. For several years, the distribution package for the MCNP Monte Carlo code1 has included an installation test suite to verify that MCNP has been installed correctly. However, the cases in that suite have been constructed primarily to test options within the code and to execute quickly. Consequently, they do not produce well-converged answers, and many of them are physically unrealistic. To remedy these deficiencies, sets of validation suites are being defined and tested for specific types of applications. All of the cases in the validation suites are based on benchmark experiments. Consequently, the results from the measurements are reliable and quantifiable, and calculated results can be compared with them in a meaningful way. Currently, validation suites exist for criticality and radiation-shielding applications.
MCNP and OMEGA criticality calculations
International Nuclear Information System (INIS)
Seifert, E.
1998-04-01
The reliability of OMEGA criticality calculations is shown by a comparison with calculations by the validated and widely used Monte Carlo code MCNP. The criticality of 16 assemblies with uranium as fissionable is calculated with the codes MCNP (Version 4A, ENDF/B-V cross sections), MCNP (Version 4B, ENDF/B-VI cross sections), and OMEGA. Identical calculation models are used for the three codes. The results are compared mutually and with the experimental criticality of the assemblies. (orig.)
Lecture Notes on Criticality Safety Validation Using MCNP & Whisper
Energy Technology Data Exchange (ETDEWEB)
Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2016-03-11
Training classes for nuclear criticality safety, MCNP documentation. The need for, and problems surrounding, validation of computer codes and data area considered first. Then some background for MCNP & Whisper is given--best practices for Monte Carlo criticality calculations, neutron spectra, S(α,β) thermal neutron scattering data, nuclear data sensitivities, covariance data, and correlation coefficients. Whisper is computational software designed to assist the nuclear criticality safety analyst with validation studies with the Monte Carlo radiation transport package MCNP. Whisper's methodology (benchmark selection – C_{k}'s, weights; extreme value theory – bias, bias uncertainty; MOS for nuclear data uncertainty – GLLS) and usage are discussed.
Criticality Calculations with MCNP6 - Practical Lectures
Energy Technology Data Exchange (ETDEWEB)
Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3)
2016-11-29
These slides are used to teach MCNP (Monte Carlo N-Particle) usage to nuclear criticality safety analysts. The following are the lecture topics: course information, introduction, MCNP basics, criticality calculations, advanced geometry, tallies, adjoint-weighted tallies and sensitivities, physics and nuclear data, parameter studies, NCS validation I, NCS validation II, NCS validation III, case study 1 - solution tanks, case study 2 - fuel vault, case study 3 - B&W core, case study 4 - simple TRIGA, case study 5 - fissile mat. vault, criticality accident alarm systems. After completion of this course, you should be able to: Develop an input model for MCNP; Describe how cross section data impact Monte Carlo and deterministic codes; Describe the importance of validation of computer codes and how it is accomplished; Describe the methodology supporting Monte Carlo codes and deterministic codes; Describe pitfalls of Monte Carlo calculations; Discuss the strengths and weaknesses of Monte Carlo and Discrete Ordinants codes; The diffusion theory model is not strictly valid for treating fissile systems in which neutron absorption, voids, and/or material boundaries are present. In the context of these limitations, identify a fissile system for which a diffusion theory solution would be adequate.
Criticality Calculations with MCNP6 - Practical Lectures
International Nuclear Information System (INIS)
Brown, Forrest B.; Rising, Michael Evan; Alwin, Jennifer Louise
2016-01-01
These slides are used to teach MCNP (Monte Carlo N-Particle) usage to nuclear criticality safety analysts. The following are the lecture topics: course information, introduction, MCNP basics, criticality calculations, advanced geometry, tallies, adjoint-weighted tallies and sensitivities, physics and nuclear data, parameter studies, NCS validation I, NCS validation II, NCS validation III, case study 1 - solution tanks, case study 2 - fuel vault, case study 3 - B&W core, case study 4 - simple TRIGA, case study 5 - fissile mat. vault, criticality accident alarm systems. After completion of this course, you should be able to: Develop an input model for MCNP; Describe how cross section data impact Monte Carlo and deterministic codes; Describe the importance of validation of computer codes and how it is accomplished; Describe the methodology supporting Monte Carlo codes and deterministic codes; Describe pitfalls of Monte Carlo calculations; Discuss the strengths and weaknesses of Monte Carlo and Discrete Ordinants codes; The diffusion theory model is not strictly valid for treating fissile systems in which neutron absorption, voids, and/or material boundaries are present. In the context of these limitations, identify a fissile system for which a diffusion theory solution would be adequate.
MCNP trademark Monte Carlo: A precis of MCNP
International Nuclear Information System (INIS)
Adams, K.J.
1996-01-01
MCNP trademark is a general purpose three-dimensional time-dependent neutron, photon, and electron transport code. It is highly portable and user-oriented, and backed by stringent software quality assurance practices and extensive experimental benchmarks. The cross section database is based upon the best evaluations available. MCNP incorporates state-of-the-art analog and adaptive Monte Carlo techniques. The code is documented in a 600 page manual which is augmented by numerous Los Alamos technical reports which detail various aspects of the code. MCNP represents over a megahour of development and refinement over the past 50 years and an ongoing commitment to excellence
Simulación con el código MCNP del reactor nuclear RP-10 en su configuración #14, BOC
Lázaro, Gerardo; Parreño, Fernando
2001-01-01
Se presenta los resultados de exceso de reactividad del núcleo del reactor RP-10 en su configuración 14. Este exceso de reactividad ha sido calculado con MCNP4B con un modelo que describe en detalle las características de los elementos combustibles normales y de control, así como de cada elemento que constituye la configuración de trabajo #14. Este modelo fue previamente utilizado en el reactor RP-0 y ha sido aplicado en la configuración de arranque para el cálculo del exceso de reactividad y...
Improved photon production data for MCNP trademark
International Nuclear Information System (INIS)
Adams, A.A.; Frankle, S.C.; Little, R.C.
1998-04-01
Computer simulations with MCNP are often used to obtain information from measurements of neutron induced gamma-ray spectra. For such simulations to be useful, the complicated spectra produced by a wide variety of nuclides must be reproduced, requiring high quality nuclear data. A previous assessment of the neutron induced photon production data in the MCNP data libraries indicated a need for improvement. The photon production data were often based on outdated experiments and binned in such wide energy groups as to be of limited value for some applications. This paper describes the work that is underway at Los Alamos National Laboratory to improve the photon production data for thermal neutron capture reactions. To date, high quality photon production data for each stable isotope of chlorine, chromium, iron, copper, and nickel have been obtained. The improved spectra have been incorporated into ENDF formatted evaluations and processed into corresponding MCNP data files. Similar improvements for aluminum, manganese, silicon, calcium, and vanadium are also planned. The methodology used to produce the spectra is discussed, and sample results for chlorine are presented
International Nuclear Information System (INIS)
Brockhoff, R.C.; Hendricks, J.S.
1994-09-01
The MCNP test set is used to test the MCNP code after installation on various computer platforms. For MCNP4 and MCNP4A this test set included 25 test problems designed to test as many features of the MCNP code as possible. A new and better test set has been devised to increase coverage of the code from 85% to 97% with 28 problems. The new test set is as fast as and shorter than the MCNP4A test set. The authors describe the methodology for devising the new test set, the features that were not covered in the MCNP4A test set, and the changes in the MCNP4A test set that have been made for MCNP4B and its developmental versions. Finally, new bugs uncovered by the new test set and a compilation of all known MCNP4A bugs are presented
Development of automatic cross section compilation system for MCNP
International Nuclear Information System (INIS)
Maekawa, Fujio; Sakurai, Kiyoshi
1999-01-01
A development of a code system to automatically convert cross-sections for MCNP is in progress. The NJOY code is, in general, used to convert the data compiled in the ENDF format (Evaluated Nuclear Data Files by BNL) into the cross-section libraries required by various reactor physics codes. While the cross-section library: FSXLIB-J3R2 was already converted from the JENDL-3.2 version of Japanese Evaluated Nuclear Data Library for a continuous energy Monte Carlo code MCNP, the library keeps only the cross-sections at room temperature (300 K). According to the users requirements which want to have cross-sections at higher temperature, say 600 K or 900 K, a code system named 'autonj' is under development to provide a set of cross-section library of arbitrary temperature for the MCNP code. This system can accept any of data formats adopted JENDL that may not be treated by NJOY code. The input preparation that is repeatedly required at every nuclide on NJOY execution is greatly reduced by permitting the conversion process of as many nuclides as the user wants in one execution. A few MCNP runs were achieved for verification purpose by using two libraries FSXLIB-J3R2 and the output of autonj'. The almost identical MCNP results within the statistical errors show the 'autonj' output library is correct. In FY 1998, the system will be completed, and in FY 1999, the user's manual will be published. (K. Tsuchihashi)
Potential MCNP enhancements for NCT
International Nuclear Information System (INIS)
Estes, G.P.; Taylor, W.M.
1992-01-01
MCNP a Monte Carlo radiation transport code, is currently widely used in the medical community for a variety of purposes including treatment planning, diagnostics, beam design, tomographic studies, and radiation protection. This is particularly true in the Neutron Capture Therapy (NCT) community. The current widespread medical use of MCNP after its general public distribution in about 1980 attests to the code's general versatility and usefulness, particularly since its development to date has not been influenced by medical applications. This paper discusses enhancements to MCNP that could be implemented at Los Alamos for the benefit of the NCT community. These enhancements generally fall into two categories, namely those that have already been developed to some extent but are not yet publicly available, and those that seem both needed based on our current understanding of NCT goals, and achievable based on our working knowledge of the MCNP code. MCNP is a general, coupled neutron/photon/electron Monte Carlo code developed and maintained by the Radiation Transport Group at Los Alamos. It has been used extensively for radiation shielding studies, reactor analysis, detector design, physics experiment interpretation, oil and gas well logging, radiation protection studies, accelerator design, etc. over the years. MCNP is a three-dimensional geometry, continuous energy physics code capable of modeling complex geometries, specifying material regions such as organs by the intersections of analytical surfaces
Suitability study of MCNP Monte Carlo program for use in medical physics
International Nuclear Information System (INIS)
Jeraj, R.
1998-01-01
MCNP is widely used Monte Carlo program in reactor and nuclear physics. However, an option of simulating electrons was added into the code a few years ago. With this extension MCNP became a code, potentially applicable for applications in medical physics. In 1997, a new version of the code, named MCNP4B was released, which contains several improvements in electron transport modeling. To test suitability of the code, several important issues were considered and examined. Default sampling in MCNP electron transport was found to be inappropriate, because it gives wrong depth dose curves for electron energies of interest in radiotherapy (Me V range). The problem can be solved if ITS-style energy sampling is used instead. One of the most difficult problems in electron transport is simulation of electron backscattering, which MCNP predicts well for all, low and high Z materials. One of the potential drawbacks, if somebody wanted to use MCNP for dosimetry on real patient geometries is that MCNP lattice calculation (e.g. when calculating dose distributions) becomes very slow for large number of scoring voxels. However, if just one scoring voxel is used, the number of geometry voxels only slightly affects the speed. In the study it was found that MCNP could be reliability used for many applications in medical physics. However, the established limitations should be taken into account when MCNP is used for a particular application.(author)
International Nuclear Information System (INIS)
Klasky, Marc Louis; Myers, Steven Charles; James, Michael R.; Mayo, Douglas R.
2016-01-01
To facilitate the timely execution of System Threat Reviews (STRs) for DNDO, and also to develop a methodology for performing STRs, LANL performed comparisons of several radiation transport codes (MCNP, GADRAS, and Gamma-Designer) that have been previously utilized to compute radiation signatures. While each of these codes has strengths, it is of paramount interest to determine the limitations of each of the respective codes and also to identify the most time efficient means by which to produce computational results, given the large number of parametric cases that are anticipated in performing STR's. These comparisons serve to identify regions of applicability for each code and provide estimates of uncertainty that may be anticipated. Furthermore, while performing these comparisons, examination of the sensitivity of the results to modeling assumptions was also examined. These investigations serve to enable the creation of the LANL methodology for performing STRs. Given the wide variety of radiation test sources, scenarios, and detectors, LANL calculated comparisons of the following parameters: decay data, multiplicity, device (n,γ) leakages, and radiation transport through representative scenes and shielding. This investigation was performed to understand potential limitations utilizing specific codes for different aspects of the STR challenges.
Simulations for the neutron detector TETRA with MCNP
International Nuclear Information System (INIS)
Testov, D.; Kuznetsova, E.; Wilson, Jh.
2013-01-01
To study the nuclear structure of β-delayed neutron precursors at ALTO ISOL-facility at IPN (Orsay), the high efficiency 4π neutron detector TETRA with 3 He filled counters built at JINR (Dubna) was modified. The MCNP simulations to optimize the future configuration were necessary. The details of the calculations and the major results obtained are discussed
Benchmark analysis of MCNP trademark ENDF/B-VI iron
International Nuclear Information System (INIS)
Court, J.D.; Hendricks, J.S.
1994-12-01
The MCNP ENDF/B-VI iron cross-section data was subjected to four benchmark studies as part of the Hiroshima/Nagasaki dose re-evaluation for the National Academy of Science and the Defense Nuclear Agency. The four benchmark studies were: (1) the iron sphere benchmarks from the Lawrence Livermore Pulsed Spheres; (2) the Oak Ridge National Laboratory Fusion Reactor Shielding Benchmark; (3) a 76-cm diameter iron sphere benchmark done at the University of Illinois; (4) the Oak Ridge National Laboratory Benchmark for Neutron Transport through Iron. MCNP4A was used to model each benchmark and computational results from the ENDF/B-VI iron evaluations were compared to ENDF/B-IV, ENDF/B-V, the MCNP Recommended Data Set (which includes Los Alamos National Laboratory Group T-2 evaluations), and experimental data. The results show that the ENDF/B-VI iron evaluations are as good as, or better than, previous data sets
International Nuclear Information System (INIS)
Cox, Lawrence J.; Barrett, Richard F.; Booth, Thomas Edward; Briesmeister, Judith F.; Brown, Forrest B.; Bull, Jeffrey S.; Giesler, Gregg Carl; Goorley, John T.; Mosteller, Russell D.; Forster, R. Arthur; Post, Susan E.; Prael, Richard E.; Selcow, Elizabeth Carol; Sood, Avneet
2002-01-01
The Monte Carlo transport workhorse, MCNP, is undergoing a massive renovation at Los Alamos National Laboratory (LANL) in support of the Eolus Project of the Advanced Simulation and Computing (ASCI) Program. MCNP Version 5 (V5) (expected to be released to RSICC in Spring, 2002) will consist of a major restructuring from FORTRAN-77 (with extensions) to ANSI-standard FORTRAN-90 with support for all of the features available in the present release (MCNP-4C2/4C3). To most users, the look-and-feel of MCNP will not change much except for the improvements (improved graphics, easier installation, better online documentation). For example, even with the major format change, full support for incremental patching will still be provided. In addition to the language and style updates, MCNP V5 will have various new user features. These include improved photon physics, neutral particle radiography, enhancements and additions to variance reduction methods, new source options, and improved parallelism support (PVM, MPI, OpenMP).
Hot Cell Window Shielding Analysis Using MCNP
International Nuclear Information System (INIS)
Pope, Chad L.; Scates, Wade W.; Taylor, J. Todd
2009-01-01
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.
International Nuclear Information System (INIS)
Pierre, J.R.M.
1996-01-01
Following the commissioning of the Low Enrichment Uranium (LEU) Fuelled SLOWPOKE-2 research reactor at the Royal Military College-College Militaire Royal (RMC-CMR), excess reactivity measurements were conducted over a range of temperature and power. The results showed a maximum excess reactivity of 3.37 mk at 33 o C. Several deterministic models using computer codes like WIMS-CRNL, CITATION, TRIVAC and DRAGON have been used to try to reproduce the excess reactivity and temperature trend of both the LEU and HEU SLOWPOKE-2 reactors. The best simulations had been obtained at Ecole Polytechnique de Montreal. They were able to reproduce the temperature trend of their HEU-fuelled reactor using TRIVAC calculations, but this model over-estimated the absolute value of the excess reactivity by 119 mk. Although calculations using DRAGON did not reproduce the temperature trend as well as TRIVAC, these calculations represented a significant improvement on the absolute value at 20 o C reducing the discrepancy to 13 mk. Given the advance in computer technology, a probabilistic approach was tried in this work, using the Monte-Carlo N-Particle Transport Code System MCNP 4A, to model the RMC-CMR SLOWPOKE-2 reactor.
SUPERIMPOSED MESH PLOTTING IN MCNP
Energy Technology Data Exchange (ETDEWEB)
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.
Preliminary evaluation of pin power distribution for fuel assemblies of SMART by MCNP
International Nuclear Information System (INIS)
Kim, Kyo Youn
1998-08-01
Monte Carlo transport code MCNP can describe an object sophisticately by use of three-dimensional modelling and can adopt a continuous energy cross-section library. Therefore MCNP has been widely utilized in the field of radiation physics to estimate fluxes and dose rates for nuclear facilities and to review results from conventional methods such a as discrete ordinates method and point kernel method. The Monte Carlo method has recently been introduced to estimated the neutron multiplication factor and pin power distribution in the fuel assembly of a reactor core. The operating thermal power of SMART core is 330 MWt and there are 57 fuel assemblies in the core. In this study it was assumed that the core has 4 types of fuel assemblies. In this study, MCNP4a was used to perform to estimate criticality and normalized pin power distribution in a fuel assembly of SMART core. The results from MCNP4a calculations are able to be used review those from nuclear design/analysis code. It is very complicated to pick up interested data from MCNP output list and to normalize pin power distribution in a fuel assembly because MCNP is not only a nuclear design/analysis code. In this study a program FAPIN was developed to generated a generate a normalized pin power distribution from the MCNP output list. (author). 11 refs
International Nuclear Information System (INIS)
Randolph Schwarz; Leland L. Carter; Alysia Schwarz
2005-01-01
Monte Carlo N-Particle Transport Code (MCNP) is the code of choice for doing complex neutron/photon/electron transport calculations for the nuclear industry and research institutions. The Visual Editor for Monte Carlo N-Particle is internationally recognized as the best code for visually creating and graphically displaying input files for MCNP. The work performed in this grant was used to enhance the capabilities of the MCNP Visual Editor to allow it to read in both 2D and 3D Computer Aided Design (CAD) files, allowing the user to electronically generate a valid MCNP input geometry
Whole core burnup calculations using 'MCNP'
International Nuclear Information System (INIS)
Haran, O.; Shaham, Y.
1996-01-01
Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors)
Whole core burnup calculations using `MCNP`
Energy Technology Data Exchange (ETDEWEB)
Haran, O; Shaham, Y [Israel Atomic Energy Commission, Beersheba (Israel). Nuclear Research Center-Negev
1996-12-01
Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors).
Comparison of MCNP5 and experimental results on neutron shielding effects for materials
Energy Technology Data Exchange (ETDEWEB)
Torres, D. A. (Daniel A.); Mosteller, R. D. (Russell D.); Sweezy, J. E. (Jeremy E.)
2004-01-01
The MCNP Radiation-Shielding Validation Suite was created to assess the impact on dose rates and attenuation factors of future improvements in the MCNP Monte Carlo code or its nuclear data libraries. However, it does not currently contain any deep-penetration cases. For this reason, a set of deep-penetration benchmarks has been investigated for possible inclusion in the Suite. Overall, the MCNP5 results match the measured values quite well. Furthermore, with the exception of Resin-F, there is no systematic trend in the ratio of calculated to measured results.
MCNP(TM) Release 6.1.1 beta: Creating and Testing the Code Distribution
Energy Technology Data Exchange (ETDEWEB)
Cox, Lawrence J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Casswell, Laura [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2014-06-12
This report documents the preparations for and testing of the production release of MCNP6™1.1 beta through RSICC at ORNL. It addresses tests on supported operating systems (Linux, MacOSX, Windows) with the supported compilers (Intel, Portland Group and gfortran). Verification and Validation test results are documented elsewhere. This report does not address in detail the overall packaging of the distribution. Specifically, it does not address the nuclear and atomic data collection, the other included software packages (MCNP5, MCNPX and MCNP6) and the collection of reference documents.
A review of radiation dosimetry applications using the MCNP Monte Carlo code
Energy Technology Data Exchange (ETDEWEB)
Solberg, T.D.; DeMarco, J.J.; Chetty, I.J.; Mesa, A.V.; Cagnon, C.H.; Li, A.N.; Mather, K.K.; Medin, P.M.; Arellano, A.R.; Smathers, J.B. [California Univ., Los Angeles, CA (United States). Dept. of Radiation Oncology
2001-07-01
The Monte Carlo code MCNP (Monte Carlo N-Particle) has a significant history dating to the early years of the Manhattan Project. More recently, MCNP has been used successfully to solve many problems in the field of medical physics. In radiotherapy applications MCNP has been used successfully to calculate the bremsstrahlung spectra from medical linear accelerators, for modeling the dose distributions around high dose rate brachytherapy sources, and for evaluating the dosimetric properties of new radioactive sources used in intravascular irradiation for prevention of restenosis following angioplasty. MCNP has also been used for radioimmunotherapy and boron neutron capture therapy applications. It has been used to predict fast neutron activation of shielding and biological materials. One area that holds tremendous clinical promise is that of radiotherapy treatment planning. In diagnostic applications, MCNP has been used to model X-ray computed tomography and positron emission tomography scanners, to compute the dose delivered from CT procedures, and to determine detector characteristics of nuclear medicine devices. MCNP has been used to determine particle fluxes around radiotherapy treatment devices and to perform shielding calculations in radiotherapy treatment rooms. This manuscript is intended to provide to the reader a comprehensive summary of medical physics applications of the MCNP code. (orig.)
A review of radiation dosimetry applications using the MCNP Monte Carlo code
International Nuclear Information System (INIS)
Solberg, T.D.; DeMarco, J.J.; Chetty, I.J.; Mesa, A.V.; Cagnon, C.H.; Li, A.N.; Mather, K.K.; Medin, P.M.; Arellano, A.R.; Smathers, J.B.
2002-01-01
The Monte Carlo code MCNP (Monte Carlo N-Particle) has a significant history dating to the early years of the Manhattan Project. More recently, MCNP has been used successfully to solve many problems in the field of medical physics. In radiotherapy applications MCNP has been used successfully to calculate the bremsstrahlung spectra from medical linear accelerators, for modeling the dose distributions around high dose rate brachytherapy sources, and for evaluating the dosimetric properties of new radioactive sources used in intravascular irradiation for prevention of restenosis following angioplasty. MCNP has also been used for radioimmunotherapy and boron neutron capture therapy applications. It has been used to predict fast neutron activation of shielding and biological materials. One area that holds tremendous clinical promise is that of radiotherapy treatment planning. In diagnostic applications, MCNP has been used to model X-ray computed tomography and positron emission tomography scanners, to compute the dose delivered from CT procedures, and to determine detector characteristics of nuclear medicine devices. MCNP has been used to determine particle fluxes around radiotherapy treatment devices and to perform shielding calculations in radiotherapy treatment rooms. This manuscript is intended to provide to the reader a comprehensive summary of medical physics applications of the MCNP code. (author)
Criticality safety validation of MCNP5 using continuous energy libraries
International Nuclear Information System (INIS)
Salome, Jean A.D.; Pereira, Claubia; Assuncao, Jonathan B.A.; Veloso, Maria Auxiliadora F.; Costa, Antonella L.; Silva, Clarysson A.M. da
2013-01-01
The study of subcritical systems is very important in the design, installation and operation of various devices, mainly nuclear reactors and power plants. The information generated by these systems guide the decisions to be taken in the executive project, the economic viability and the safety measures to be employed in a nuclear facility. Simulating some experiments from the International Handbook of Evaluated Criticality Safety Benchmark Experiments, the code MCNP5 was validated to nuclear criticality analysis. Its continuous libraries were used. The average values and standard deviation (SD) were evaluated. The results obtained with the code are very similar to the values obtained by the benchmark experiments. (author)
New developments enhancing MCNP for criticality safety
International Nuclear Information System (INIS)
Hendricks, J.S.; McKinney, G.W.; Forster, R.A.
1993-01-01
Since the early 80's MCNP has had three estimates of k eff : collision, absorption, and track length. MCNP has also had collision and absorption estimators of removal lifetime. These are calculated for every cycle and are averaged over the cycles as simple averages and covariance weighted averages. Correlation coefficients between estimators are also calculated. These criticality estimators are all in addition to the extensive summary information and tally edits used in shielding and other problems. A number of significant new developments have been made to enhance the MCNP Monte Carlo radiation transport code for criticality safety applications. These are available in the newly released MCNP4A version of the code
International Nuclear Information System (INIS)
Jo, Y. S.; Kim, J. D.; Kil, C. S.; Jang, J. H.
1999-01-01
The scattering laws and MCNP thermal libraries for liquid hydrogen and deuterium are comparatively calculated on HP715 (32-bit computer) and SGI IP27 (64-bit computer) using NJOY97. The results are also compared with the experimental data. In addition, MCNP calculations for the nuclear design of a cold neutron source at HANARO are performed with the newly generated MCNP thermal libraries from two different computers and the results are compared
Radiation shielding calculation using MCNP
International Nuclear Information System (INIS)
Masukawa, Fumihiro
2001-01-01
To verify the Monte Carlo code MCNP4A as a tool to generate the reference data in the shielding designs and the safety evaluations, various shielding benchmark experiments were analyzed using this code. These experiments were categorized in three types of the shielding subjects; bulk shielding, streaming, and skyshine. For the variance reduction technique, which is indispensable to get meaningful results with the Monte Carlo shielding calculation, we mainly used the weight window, the energy dependent Russian roulette and spitting. As a whole, our analyses performed enough small statistical errors and showed good agreements with these experiments. (author)
MCNP Version 6.2 Release Notes
Energy Technology Data Exchange (ETDEWEB)
Werner, Christopher John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bull, Jeffrey S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Solomon, C. J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McKinney, Gregg Walter [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dixon, David A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martz, Roger Lee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hughes, Henry G. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Cox, Lawrence James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Zukaitis, Anthony J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Armstrong, J. C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Forster, Robert Arthur [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Casswell, Laura [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2018-02-05
Monte Carlo N-Particle or MCNP^{®} is a general-purpose Monte Carlo radiation-transport code designed to track many particle types over broad ranges of energies. This MCNP Version 6.2 follows the MCNP6.1.1 beta version and has been released in order to provide the radiation transport community with the latest feature developments and bug fixes for MCNP. Since the last release of MCNP major work has been conducted to improve the code base, add features, and provide tools to facilitate ease of use of MCNP version 6.2 as well as the analysis of results. These release notes serve as a general guide for the new/improved physics, source, data, tallies, unstructured mesh, code enhancements and tools. For more detailed information on each of the topics, please refer to the appropriate references or the user manual which can be found at http://mcnp.lanl.gov. This release of MCNP version 6.2 contains 39 new features in addition to 172 bug fixes and code enhancements. There are still some 33 known issues the user should familiarize themselves with (see Appendix).
MCNP trademark Software Quality Assurance plan
International Nuclear Information System (INIS)
Abhold, H.M.; Hendricks, J.S.
1996-04-01
MCNP is a computer code that models the interaction of radiation with matter. MCNP is developed and maintained by the Transport Methods Group (XTM) of the Los Alamos National Laboratory (LANL). This plan describes the Software Quality Assurance (SQA) program applied to the code. The SQA program is consistent with the requirements of IEEE-730.1 and the guiding principles of ISO 900
Energy Technology Data Exchange (ETDEWEB)
Bull, Jeffrey S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2017-11-13
This presentation describes how to build MCNP 6.2. MCNP®* 6.2 can be compiled on Macs, PCs, and most Linux systems. It can also be built for parallel execution using both OpenMP and Messing Passing Interface (MPI) methods. MCNP6 requires Fortran, C, and C++ compilers to build the code.
Status Report on the MCNP 2020 Initiative
Energy Technology Data Exchange (ETDEWEB)
Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2017-10-02
The discussion below provides a status report on the MCNP 2020 initiative. It includes discussion of the history of MCNP 2020, accomplishments during 2013-17, priorities for near-term development, other related efforts, a brief summary, and a list of references for the plans and work accomplished.
The new MCNP6 depletion capability
International Nuclear Information System (INIS)
Fensin, M. L.; James, M. R.; Hendricks, J. S.; Goorley, J. T.
2012-01-01
The first MCNP based in-line Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. Both the MCNP5 and MCNPX codes have historically provided a successful combinatorial geometry based, continuous energy, Monte Carlo radiation transport solution for advanced reactor modeling and simulation. However, due to separate development pathways, useful simulation capabilities were dispersed between both codes and not unified in a single technology. MCNP6, the next evolution in the MCNP suite of codes, now combines the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. We describe here the new capabilities of the MCNP6 depletion code dating from the official RSICC release MCNPX 2.6.0, reported previously, to the now current state of MCNP6. NEA/OECD benchmark results are also reported. The MCNP6 depletion capability enhancements beyond MCNPX 2.6.0 reported here include: (1) new performance enhancing parallel architecture that implements both shared and distributed memory constructs; (2) enhanced memory management that maximizes calculation fidelity; and (3) improved burnup physics for better nuclide prediction. MCNP6 depletion enables complete, relatively easy-to-use depletion calculations in a single Monte Carlo code. The enhancements described here help provide a powerful capability as well as dictate a path forward for future development to improve the usefulness of the technology. (authors)
The New MCNP6 Depletion Capability
International Nuclear Information System (INIS)
Fensin, Michael Lorne; James, Michael R.; Hendricks, John S.; Goorley, John T.
2012-01-01
The first MCNP based inline Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. Both the MCNP5 and MCNPX codes have historically provided a successful combinatorial geometry based, continuous energy, Monte Carlo radiation transport solution for advanced reactor modeling and simulation. However, due to separate development pathways, useful simulation capabilities were dispersed between both codes and not unified in a single technology. MCNP6, the next evolution in the MCNP suite of codes, now combines the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. We describe here the new capabilities of the MCNP6 depletion code dating from the official RSICC release MCNPX 2.6.0, reported previously, to the now current state of MCNP6. NEA/OECD benchmark results are also reported. The MCNP6 depletion capability enhancements beyond MCNPX 2.6.0 reported here include: (1) new performance enhancing parallel architecture that implements both shared and distributed memory constructs; (2) enhanced memory management that maximizes calculation fidelity; and (3) improved burnup physics for better nuclide prediction. MCNP6 depletion enables complete, relatively easy-to-use depletion calculations in a single Monte Carlo code. The enhancements described here help provide a powerful capability as well as dictate a path forward for future development to improve the usefulness of the technology.
Development of MCNP interface code in HFETR
International Nuclear Information System (INIS)
Qiu Liqing; Fu Rong; Deng Caiyu
2007-01-01
In order to describe the HFETR core with MCNP method, the interface code MCNPIP for HFETR and MCNP code is developed. This paper introduces the core DXSY and flowchart of MCNPIP code, and the handling of compositions of fuel elements and requirements on hardware and software. Finally, MCNPIP code is validated against the practical application. (authors)
MCNP4A: Features and philosophy
International Nuclear Information System (INIS)
Hendricks, J.S.
1993-01-01
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 ENSDF based radionuclide source for MCNP
International Nuclear Information System (INIS)
Berlizov, A.N.; Tryshyn, V.V.
2003-01-01
A utility for generating source code of the Source subroutine of MCNP (a general Monte Carlo NxParticle transport code) on the basis of ENSDF (Evaluated Nuclear Structure Data File) is described. The generated code performs statistical simulation of processes, accompanying radioactive decay of a chosen radionuclide through a specified decay branch, providing characteristics of emitted correlated particles on its output. At modeling the following processes are taken into account: emission of continuum energy electrons at beta - -decay to different exited levels of a daughter nucleus; annihilation photon emission accompanying beta + -decay; gamma-ray emission; emission of discrete energy electrons resulted from internal conversion process on atomic K- and L I,II,III -shells; K and LX-ray emission at single and double fluorescence, accompanying electron capture and internal conversion processes. Number of emitted particles, their types, energies and emission times are sampled according to characteristics of a decay scheme of a particular radionuclide as well as characteristics of atomic shells of mother and daughter nuclei. Angular correlations, calculated for a particular combination of nuclear level spins, mixing ratios and gamma-ray multipolarities, are taken into account at sampling of directional cosines of emitted gamma-rays. The paper contains examples of spectrometry system response simulation at measurements with real radionuclide sources. (authors)
MCNP capabilities at the dawn of the 21st century: Neutron-gamma applications
International Nuclear Information System (INIS)
Selcow, E.C.; McKinney, G.W.
2000-01-01
The Los Alamos National Laboratory Monte Carlo N-Particle radiation transport code, MCNP, has become an international standard for a wide spectrum of neutron-gamma radiation transport applications. These include nuclear criticality safety, radiation shielding, nuclear safeguards, nuclear well-logging, fission and fusion reactor design, accelerator target design, detector design and analysis, health physics, medical radiation therapy and imaging, radiography, decontamination and decommissioning, and waste storage and disposal. The latest version of the code, MCNP4C, was released to the Radiation Safety Information Computational Center (RSICC) in February 2000.This paper described the new features and capabilities of the code, and discusses the specific applicability to neutron-gamma problems. We will also discuss the future directions for MCNP code development, including rewriting the code in Fortran 90
Analysis of parallel computing performance of the code MCNP
International Nuclear Information System (INIS)
Wang Lei; Wang Kan; Yu Ganglin
2006-01-01
Parallel computing can reduce the running time of the code MCNP effectively. With the MPI message transmitting software, MCNP5 can achieve its parallel computing on PC cluster with Windows operating system. Parallel computing performance of MCNP is influenced by factors such as the type, the complexity level and the parameter configuration of the computing problem. This paper analyzes the parallel computing performance of MCNP regarding with these factors and gives measures to improve the MCNP parallel computing performance. (authors)
MatMCNP: A Code for Producing Material Cards for MCNP
Energy Technology Data Exchange (ETDEWEB)
DePriest, Kendall Russell [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Saavedra, Karen C. [American Structurepoint, Inc., Indianapolis, IN (United States)
2014-09-01
A code for generating MCNP material cards (MatMCNP) has been written and verified for naturally occurring, stable isotopes. The program allows for material specification as either atomic or weight percent (fractions). MatMCNP also permits the specification of enriched lithium, boron, and/or uranium. In addition to producing the material cards for MCNP, the code calculates the atomic (or number) density in atoms/barn-cm as well as the multiplier that should be used to convert neutron and gamma fluences into dose in the material specified.
International Nuclear Information System (INIS)
Bourauel, Peter; Nabbi, Rahim; Biel, Wolfgang; Forrest, Robin
2009-01-01
The MCNP 3D Monte Carlo computer code is used not only for criticality calculations of nuclear systems but also to simulate transports of radiation and particles. The findings so obtained about neutron flux distribution and the associated spectra allow information about materials activation, nuclear heating, and radiation damage to be obtained by means of activation codes such as FISPACT. The stochastic character of particle and radiation transport processes normally links findings to the materials cells making up the geometry model of MCNP. Where high spatial resolution is required for the activation calculations with FISPACT, fine segmentation of the MCNP geometry becomes compulsory, which implies considerable expense for the modeling process. For this reason, an alternative simulation technique has been developed in an effort to automate and optimize data transfer between MCNP and FISPACT. (orig.)
MCNP5 development, verification, and performance
International Nuclear Information System (INIS)
Forrest B, Brown
2003-01-01
MCNP is a well-known and widely used Monte Carlo code for neutron, photon, and electron transport simulations. During the past 18 months, MCNP was completely reworked to provide MCNP5, a modernized version with many new features, including plotting enhancements, photon Doppler broadening, radiography image tallies, enhancements to source definitions, improved variance reduction, improved random number generator, tallies on a superimposed mesh, and edits of criticality safety parameters. Significant improvements in software engineering and adherence to standards have been made. Over 100 verification problems have been used to ensure that MCNP5 produces the same results as before and that all capabilities have been preserved. Testing on large parallel systems shows excellent parallel scaling. (author)
Installation and validation of MCNP-4A
International Nuclear Information System (INIS)
Marks, N.A.
1997-01-01
MCNP-4A is a multi-purpose Monte Carlo program suitable for the modelling of neutron, photon, and electron transport problems. It is a particularly useful technique when studying systems containing irregular shapes. MCNP has been developed over the last 25 years by Los Alamos, and is distributed internationally via RSIC at Oak Ridge. This document describes the installation of MCNP-4A (henceforth referred to as MCNP) on the Silicon Graphics workstation (bluey.ansto.gov.au). A limited number of benchmarks pertaining to fast and thermal systems were performed to check the installation and validate the code. The results are compared to deterministic calculations performed using the AUS neutronics code system developed at ANSTO. (author)
MCNP5 development, verification, and performance
Energy Technology Data Exchange (ETDEWEB)
Forrest B, Brown [Los Alamos National Laboratory (United States)
2003-07-01
MCNP is a well-known and widely used Monte Carlo code for neutron, photon, and electron transport simulations. During the past 18 months, MCNP was completely reworked to provide MCNP5, a modernized version with many new features, including plotting enhancements, photon Doppler broadening, radiography image tallies, enhancements to source definitions, improved variance reduction, improved random number generator, tallies on a superimposed mesh, and edits of criticality safety parameters. Significant improvements in software engineering and adherence to standards have been made. Over 100 verification problems have been used to ensure that MCNP5 produces the same results as before and that all capabilities have been preserved. Testing on large parallel systems shows excellent parallel scaling. (author)
MCNP application for the 21 century
International Nuclear Information System (INIS)
McKinney, G.W.
2000-01-01
The Los Alamos National Laboratory (LANL) Monte Carlo N-Particle radiation transport code, MCNP, has become an international standard for a wide spectrum of neutron, photon, and electron radiation transport applications. The latest version of the code, MCNP 4C, was released to the Radiation Safety Information Computational Center (RSICC) in February 2000. This paper describes the code development philosophy, new features and capabilities, applicability to various problems, and future directions
Neutron-induced photon production in MCNP
International Nuclear Information System (INIS)
Little, R.C.; Seamon, R.E.
1983-01-01
An improved method of neutron-induced photon production has been incorporated into the Monte Carlo transport code MCNP. The new method makes use of all partial photon-production reaction data provided by ENDF/B evaluators including photon-production cross sections as well as energy and angular distributions of secondary photons. This faithful utilization of sophisticated ENDF/B evaluations allows more precise MCNP calculations for several classes of coupled neutron-photon problems
Visualizing MCNP Tally Segment Geometry and Coupling Results with ABAQUS
International Nuclear Information System (INIS)
J. R. Parry; J. A. Galbraith
2007-01-01
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
International Nuclear Information System (INIS)
Little, R.C.; Frankle, S.C.; Hughes, H.G. III; Prael, R.E.
1997-01-01
MCNP trademark and LAHET trademark are two of the codes included in the LARAMIE (Los Alamos Radiation Modeling Interactive Environment) code system. Both MCNP and LAHET are three-dimensional continuous-energy Monte Carlo radiation transport codes. The capabilities of MCNP and LAHET are currently being merged into one code for the Accelerator Production of Tritium (APT) program at Los Alamos National Laboratory. Concurrently, a significant effort is underway to improve the accuracy of the physics in the merged code. In particular, full nuclear-data evaluations (in ENDF6 format) for many materials of importance to APT are being produced for incident neutrons and protons up to an energy of 150-MeV. After processing, cross-section tables based on these new evaluations will be available for use fin the merged code. In order to utilize these new cross-section tables, significant enhancements are required for the merged code. Neutron cross-section tables for MCNP currently specify emission data for neutrons and photons only; the new evaluations also include complete neutron-induced data for protons, deuterons, tritons, and alphas. In addition, no provision in either MCNP or LAHET currently exists for the use of incident charged-particle tables other than for electrons. To accommodate the new neutron-induced data, it was first necessary to expand the format definition of an MCNP neutron cross-section table. The authors have prepared a 150-MeV neutron cross-section library in this expanded format for 15 nuclides. Modifications to MCNP have been implemented so that this expanded neutron library can be utilized
TET_2MCNP: A conversion program to implement tetrahearal-mesh models in MCNP
International Nuclear Information System (INIS)
Han, Min Cheol; Yeom, Yeon Soo; Nguyen, Thng Tat; Choi, Chan Soo; Lee, Hyun Su; Kim, Chan Hyeong
2016-01-01
Tetrahedral-mesh geometries can be used in the MCNP code, but the MCNP code accepts only the geometry in the Abaqus input file format; hence, the existing tetrahedral-mesh models first need to be converted to the Abacus input file format to be used in the MCNP code. In the present study, we developed a simple but useful computer program, TET_2MCNP, for converting TetGen-generated tetrahedral-mesh models to the Abacus input file format. TET_2MCNP is written in C++ and contains two components: one for converting a TetGen output file to the Abacus input file and the other for the reverse conversion process. The TET_2MCP program also produces an MCNP input file. Further, the program provides some MCNP-specific functions: the maximum number of elements (i.e., tetrahedrons) per part can be limited, and the material density of each element can be transferred to the MCNP input file. To test the developed program, two tetrahedral-mesh models were generated using TetGen and converted to the Abaqus input file format using TET_2MCNP. Subsequently, the converted files were used in the MCNP code to calculate the object- and organ-averaged absorbed dose in the sphere and phantom, respectively. The results show that the converted models provide, within statistical uncertainties, identical dose values to those obtained using the PHITS code, which uses the original tetrahedral-mesh models produced by the TetGen program. The results show that the developed program can successfully convert TetGen tetrahedral-mesh models to Abacus input files. In the present study, we have developed a computer program, TET_2MCNP, which can be used to convert TetGen-generated tetrahedral-mesh models to the Abaqus input file format for use in the MCNP code. We believe this program will be used by many MCNP users for implementing complex tetrahedral-mesh models, including computational human phantoms, in the MCNP code
TET{sub 2}MCNP: A conversion program to implement tetrahearal-mesh models in MCNP
Energy Technology Data Exchange (ETDEWEB)
Han, Min Cheol; Yeom, Yeon Soo; Nguyen, Thng Tat; Choi, Chan Soo; Lee, Hyun Su; Kim, Chan Hyeong [Dept. of Nuclear Engineering, Hanyang University, Seoul (Korea, Republic of)
2016-12-15
Tetrahedral-mesh geometries can be used in the MCNP code, but the MCNP code accepts only the geometry in the Abaqus input file format; hence, the existing tetrahedral-mesh models first need to be converted to the Abacus input file format to be used in the MCNP code. In the present study, we developed a simple but useful computer program, TET{sub 2}MCNP, for converting TetGen-generated tetrahedral-mesh models to the Abacus input file format. TET{sub 2}MCNP is written in C++ and contains two components: one for converting a TetGen output file to the Abacus input file and the other for the reverse conversion process. The TET{sub 2}MCP program also produces an MCNP input file. Further, the program provides some MCNP-specific functions: the maximum number of elements (i.e., tetrahedrons) per part can be limited, and the material density of each element can be transferred to the MCNP input file. To test the developed program, two tetrahedral-mesh models were generated using TetGen and converted to the Abaqus input file format using TET{sub 2}MCNP. Subsequently, the converted files were used in the MCNP code to calculate the object- and organ-averaged absorbed dose in the sphere and phantom, respectively. The results show that the converted models provide, within statistical uncertainties, identical dose values to those obtained using the PHITS code, which uses the original tetrahedral-mesh models produced by the TetGen program. The results show that the developed program can successfully convert TetGen tetrahedral-mesh models to Abacus input files. In the present study, we have developed a computer program, TET{sub 2}MCNP, which can be used to convert TetGen-generated tetrahedral-mesh models to the Abaqus input file format for use in the MCNP code. We believe this program will be used by many MCNP users for implementing complex tetrahedral-mesh models, including computational human phantoms, in the MCNP code.
Determination of neutron flux distribution in an Am-Be irradiator using the MCNP.
Shtejer-Diaz, K; Zamboni, C B; Zahn, G S; Zevallos-Chávez, J Y
2003-10-01
A neutron irradiator has been assembled at IPEN facilities to perform qualitative-quantitative analysis of many materials using thermal and fast neutrons outside the nuclear reactor premises. To establish the prototype specifications, the neutron flux distribution and the absorbed dose rates were calculated using the MCNP computer code. These theoretical predictions then allow one to discuss the optimum irradiator design and its performance.
MCNP and visualization of neutron flux and power distributions
International Nuclear Information System (INIS)
Snoj, L.; Lengar, I.; Zerovnik, G.; Ravnik, M.
2009-01-01
The visualization of neutron flux and power distributions in two nuclear reactors (TRIG A type research reactor and typical PWR) and one thermonuclear reactor (tokamak type) are treated in the paper. The distributions are calculated with MCNP computer code and presented using Amira and Voxler software. The results in the form of figures are presented in the paper together with comments qualitatively explaining the figures. The remembrance of most of the people is better, if they visualize a process. Therefore a representation of the reactor and neutron transport parameters is a convenient modern educational tool for (nuclear power plant) operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for core and irradiation planning. (authors)
International Nuclear Information System (INIS)
Schwarz, Randy A.; Carter, Leeland L.
2004-01-01
Monte Carlo N-Particle Transport Code (MCNP) (Reference 1) is the code of choice for doing complex neutron/photon/electron transport calculations for the nuclear industry and research institutions. The Visual Editor for Monte Carlo N-Particle (References 2 to 11) is recognized internationally as the best code for visually creating and graphically displaying input files for MCNP. The work performed in this grant enhanced the capabilities of the MCNP Visual Editor to allow it to read in a 2D Computer Aided Design (CAD) file, allowing the user to modify and view the 2D CAD file and then electronically generate a valid MCNP input geometry with a user specified axial extent
MCNP modelling of a combined neutron/gamma counter
Bourva, L C A; Ottmar, H; Weaver, D R
1999-01-01
A series of Monte Carlo neutron calculations for a combined gamma/passive neutron coincidence counter has been performed. This type of device, part of a suite of non-destructive assay instruments utilised for the enforcement of the Euratom nuclear safeguards within the European Union, is to be used for high accuracy measurements of the plutonium content of small samples of nuclear materials. The multi-purpose Monte Carlo N-particle (MCNP) code version 4B has been used to model in detail the neutron coincidence detector and to investigate the leakage self-multiplication of PuO sub 2 and mixed U-Pu oxide (MOX) reference samples used to calibrate the instrument. The MCNP calculations have been used together with a neutron coincidence counting interpretative model to determine characteristic parameters of the detector. A comparative study to both experimental and previous numerical results has been performed. Sensitivity curves of the variation of the detector's efficiency, epsilon, to, alpha, the ratio of (alpha...
New calculations for critical assemblies using MCNP4B
International Nuclear Information System (INIS)
Adams, A.A.; Frankle, S.C.; Little, R.C.
1997-07-01
A suite of 41 criticality benchmarks has been modeled using MCNP trademark (version 4B). Most of the assembly specifications were obtained from the Cross Section Evaluation Working Group (CSEWG) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) compendiums of experimental benchmarks. A few assembly specifications were obtained from experimental papers. The suite contains thermal and fast assemblies, bare and reflected assemblies, and emphasizes 233 U, 235 U, 238 U, and 239 Pu. The values of k eff for each assembly in the suite were calculated using MCNP libraries derived primarily from release 2 of ENDF/B-V and release 2 of ENDF/B-VI. The results show that the new ENDF/B-VI.2 evaluations for H, O, N, B, 235 U, 238 U, and 239 Pu can have a significant impact on the values of k eff . In addition to the integral quantity k eff , several additional experimental measurements were performed and documented. These experimental measurements include central fission and reaction-rate ratios for various isotopes, and neutron leakage and flux spectra. They provide more detailed information about the accuracy of the nuclear data than can k eff . Comparison calculations were performed using both ENDF/B-V.2 and ENDF/B-VI.2-based data libraries. The purpose of this paper is to compare the results of these additional calculations with experimental data, and to use these results to assess the quality of the nuclear data
Biasing secondary particle interaction physics and production in MCNP6
International Nuclear Information System (INIS)
Fensin, M.L.; James, M.R.
2016-01-01
Highlights: • Biasing secondary production and interactions of charged particles in the tabular energy regime. • Examining lower weight window bounds for rare events when using Russian roulette. • The new biasing strategy can speedup calculations by a factor of 1 million or more. - Abstract: Though MCNP6 will transport elementary charged particles and light ions to low energies (i.e. less than 20 MeV), MCNP6 has historically relied on model physics with suggested minimum energies of ∼20 to 200 MeV. Use of library data for the low energy regime was developed for MCNP6 1.1.Beta to read and use light ion libraries. Thick target yields of neutron production for alphas on fluoride result in 1 production event per roughly million sampled alphas depending on the energy of the alpha (for other isotopes the yield can be even rarer). Calculation times to achieve statistically significant and converged thick target yields are quite laborious, needing over one hundred processor hours. The MUCEND code possess a biasing technique for improving the sampling of secondary particle production by forcing a nuclear interaction to occur per each alpha transported. We present here a different biasing strategy for secondary particle production from charged particles. During each substep, as the charged particle slows down, we bias both a nuclear collision event to occur at each substep and the production of secondary particles at the collision event, while still continuing to progress the charged particle until reaching a region of zero importance or an energy/time cutoff. This biasing strategy is capable of speeding up calculations by a factor of a million or more as compared to the unbiased calculation. Further presented here are both proof that the biasing strategy is capable of producing the same results as the unbiased calculation and the limitations to consider in order to achieve accurate results of secondary particle production. Though this strategy was developed for MCNP
Energy Technology Data Exchange (ETDEWEB)
Chapman, Bryan Scott [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); MacQuigg, Michael Robert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wysong, Andrew Russell [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2016-04-21
In this document, the code MCNP is validated with ENDF/B-VII.1 cross section data under the purview of ANSI/ANS-8.24-2007, for use with uranium systems. MCNP is a computer code based on Monte Carlo transport methods. While MCNP has wide reading capability in nuclear transport simulation, this validation is limited to the functionality related to neutron transport and calculation of criticality parameters such as k_{eff}.
International Nuclear Information System (INIS)
Chapman, Bryan Scott; MacQuigg, Michael Robert; Wysong, Andrew Russell
2016-01-01
In this document, the code MCNP is validated with ENDF/B-VII.1 cross section data under the purview of ANSI/ANS-8.24-2007, for use with uranium systems. MCNP is a computer code based on Monte Carlo transport methods. While MCNP has wide reading capability in nuclear transport simulation, this validation is limited to the functionality related to neutron transport and calculation of criticality parameters such as k eff .
International Nuclear Information System (INIS)
Karriem, Z.; Zamonsky, O.M.
2014-01-01
The South African Nuclear Energy Corporation SOC Ltd (Necsa) is a state owned nuclear facility which owns and operates SAFARI-1, a 20 MW material testing reactor. SAFARI-1 is a multi-purpose reactor and is used for the production of radioisotopes through in-core sample irradiation. The Radiation and Reactor Theory (RRT) Section of Necsa supports SAFARI-1 operations with nuclear engineering analyses which include core-reload design, core-follow and radiation transport analyses. The primary computer codes that are used for the analyses are the OSCAR-4 nodal diffusion core simulator and the Monte Carlo transport code MCNP. RRT has developed a calculation methodology based on OSCAR-4 and MCNP to simulate the diverse in-core irradiation conditions in SAFARI-1, for the purpose of radioisotope production. In this paper we present the OSCAR-4/MCNP calculation methodology and the software tools that were developed for rapid and reliable construction of MCNP analysis models. The paper will present the application and accuracy of the methodology for the production of yttrium-90 ( 90 Y) and will include comparisons between calculation results and experimental measurements. The paper will also present sensitivity analyses that were performed to determine the effects of control rod bank position, representation of core depletion state and sample loading configuration, on the calculated 90 Y sample activity. (author)
Depleted Reactor Analysis With MCNP-4B
International Nuclear Information System (INIS)
Caner, M.; Silverman, L.; Bettan, M.
2004-01-01
Monte Carlo neutronics calculations are mostly done for fresh reactor cores. There is today an ongoing activity in the development of Monte Carlo plus burnup code systems made possible by the fast gains in computer processor speeds. In this work we investigate the use of MCNP-4B for the calculation of a depleted core of the Soreq reactor (IRR-1). The number densities as function of burnup were taken from the WIMS-D/4 cell code calculations. This particular code coupling has been implemented before. The Monte Carlo code MCNP-4B calculates the coupled transport of neutrons and photons for complicated geometries. We have done neutronics calculations of the IRR-1 core with the WIMS and CITATION codes in the past Also, we have developed an MCNP model of the IRR-1 standard fuel for a criticality safety calculation of a spent fuel storage pool
SABRINA, Geometry Plot Program for MCNP
International Nuclear Information System (INIS)
SEIDL, Marcus
2003-01-01
1 - Description of program or function: SABRINA is an interactive, three-dimensional, geometry-modeling code system, primarily for use with CCC-200/MCNP. SABRINA's capabilities include creation, visualization, and verification of three-dimensional geometries specified by either surface- or body-base combinatorial geometry; display of particle tracks are calculated by MCNP; and volume fraction generation. 2 - Method of solution: Rendering is performed by ray tracing or an edge and intersection algorithm. Volume fraction calculations are made by ray tracing. 3 - Restrictions on the complexity of the problem: A graphics display with X Window capability is required
Modeling the PUSPATI TRIGA Reactor using MCNP code
International Nuclear Information System (INIS)
Mohamad Hairie Rabir; Mark Dennis Usang; Naim Syauqi Hamzah; Julia Abdul Karim; Mohd Amin Sharifuldin Salleh
2012-01-01
The 1 MW TRIGA MARK II research reactor at Malaysian Nuclear Agency achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutron flux distribution and depletion study of TRIGA fuel. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core and shielding with literally no physical approximation. (author)
MOCUP, MCNP/ORIGEN Coupling Utility Programs
International Nuclear Information System (INIS)
SEIDL, Marcus
2003-01-01
1 - Description of program or function: MOCUP is a series of utility and data manipulation programs to solve time and space-dependent coupled neutronics/isotopics problems. 2 - Methods: The neutronics calculation is performed by the Los Alamos National Laboratory code system, version 4a or later (CCC-200 or CCC-660),and the depletion and isotopics calculation is performed by CCC-371/ORIGEN2.1 developed at Oak Ridge National Laboratory. MCNP and ORIGEN2.1 are NOT included in this package. MOCUP consists of three utility programs (mcnpPRO, origenPRO, compPRO) to, respectively, search the MCNP output and tally files for relevant cell and tally parameters, prepare ORIGEN2.1 input files and execute the ORIGEN2.1 runs, and search ORIGEN2.1 punch files for relevant isotope concentrations and produce new MCNP input files. A graphical user interface is provided for execution convenience. 3 - Restrictions on the complexity of the problem: At present, no mechanism exists for automatic serial execution of the program modules. The user must interface with the GUI to run each of the modules
Investigation of the applicability of MCNP code to complicated geometries
International Nuclear Information System (INIS)
Higuchi, Kenji; Yamaguchi, Yukichi
1994-03-01
Applicability of MCNP code, which is a general purpose Monte Carlo code for particle transport problems, to complicated geometries, has been investigated as a study in Human Acts Simulation Program (HASP), in which basic studies for intelligent robot for patrol and inspection of nuclear facilities are being performed. In HASP, basic software systems simulating the behavior of intelligent robot of human shape working in Japan Research Reactor No.3 are being developed. The aim of Dose Evaluation system in HASP is to establish the methodology to evaluate irradiation damage of the LSI/VLSI circuits embedded within a robot body and to give design criteria of intelligent robot. Monte Carlo method is used to solve particle transport problem in a complicated geometry such as robot body. Preliminary evaluation to establish the methodology has been conducted using continuous energy Monte Carlo code, MCNP with the anthropomorphic phantom. The phantom has the same degree of geometric complexity as robot body and is widely used for the calculation of the effective dose equivalent for radiological protection. It allowed us to verify the validity of the methodology by comparison of calculation results with the data in ICRP Pub. 51. In this report, the method used in the calculation of effective dose equivalent, visualization system supporting visualization of input data for complicated geometry and the results in the evaluation of validity of the method by the comparison of the calculated results with the data in the ICRP publication are described. (author)
Calculation of power density with MCNP in TRIGA reactor
International Nuclear Information System (INIS)
Snoj, L.; Ravnik, M.
2006-01-01
Modern Monte Carlo codes (e.g. MCNP) allow calculation of power density distribution in 3-D geometry assuming detailed geometry without unit-cell homogenization. To normalize MCNP calculation by the steady-state thermal power of a reactor, one must use appropriate scaling factors. The description of the scaling factors is not adequately described in the MCNP manual and requires detailed knowledge of the code model. As the application of MCNP for power density calculation in TRIGA reactors has not been reported in open literature, the procedure of calculating power density with MCNP and its normalization to the power level of a reactor is described in the paper. (author)
Methodology for converting CT medical images to MCNP input using the Scan2MCNP system
International Nuclear Information System (INIS)
Boia, L.S.; Silva, A.X.; Cardoso, S.C.; Castro, R.C.
2009-01-01
This paper develops a methodology for the application software Scan2MCNP, which converts medical images DICOM (Digital Imaging and Communications in Medicine) for MCNP input file. The Scan2MCNP handles, processes and executes the medical images generated by CT equipment, allowing the user to perform the selection and parameterization of the study area in question (tissues and organs). The details of these worked in medical imaging software, therefore, will be converted to equity to the process of language analysis of MCNP radiation transport, through the generation of a code input file. With this file, it is possible to simulate any situation/problem of the type and level of radiation to the proposed treatment chosen by the medical staff responsible for the patient. Within a computational process oriented, the Scan2MCNP can contribute along with other software that has been used recently in the area of medical physics, to improve the levels of quality and precision of radiotherapy treatments. In this work, medical images DICOM of the Anthropomorphic Rando Phantom were used in the process of analysis and development of computer software Scan2MCNP. However, it emphasized that the software is successful in certain situations, depending upon a number of auxiliary procedures and software that can help in the solution of certain problems in the natural radiation treatment or express agility by the team of medical physics. (author)
Energy Technology Data Exchange (ETDEWEB)
Monk, S.D., E-mail: s.monk@lancaster.ac.uk [Department of Engineering, Lancaster University, Lancaster LA1 4YW (United Kingdom); Shippen, B.A. [Department of Engineering, Lancaster University, Lancaster LA1 4YW (United Kingdom); Colling, B.R. [Department of Engineering, Lancaster University, Lancaster LA1 4YW (United Kingdom); Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Cheneler, D.; Al Hamrashdi, H.; Alton, T. [Department of Engineering, Lancaster University, Lancaster LA1 4YW (United Kingdom)
2017-05-15
Highlights: • Comparison of the use of MCNP6 and GEANT4 Monte Carlo software when large distances and thicknesses are considered. • The Thermal Neutron Facility (TNF) at TRIUMF used as an example real life example location. • The effects of water, aluminium, iron and lead considered over various thicknesses up to 3 m. - Abstract: A comparison of the Monte Carlo based simulation codes MCNP6-1.0 and GEANT4-10.1 as used for modelling large scale structures is presented here. The high-energy neutron field at the Tri Universities Meson Facility (TRIUMF) in Vancouver, British Columbia is the structure modelled in this work. Work with the emphasis on the modelling of the facility and comparing with experimental results has been published previously, whereas this work is focussed on comparing the performance of the codes over relatively high depths of material rather than the accuracy of the results themselves in comparison to experimental data. Comparisons of three different locations within the neutron facility are modelled and presented using both codes as well as analysis of the transport of typical neutrons fields through large blocks of iron, water, lead and aluminium in order to determine where any deviations are likely to have occurred. Results indicate that over short distances, results from the two codes are in broad agreement – although over greater distances and within more complex geometries, deviation increases dramatically. The conclusions reached are that it is likely the deviations between the codes is caused by both the compounding effect of slight differences between the cross section files used by the two codes to determine the neutron transport through iron, and differences in the processes used by both codes.
International Nuclear Information System (INIS)
Yamamoto, Toshihiro; Sakurai, Kiyoshi; Tonoike, Kotaro; Miyoshi, Yoshinori
2001-01-01
Reactor noise analysis methods using Monte Carlo technique have been proposed and developed in the field of nuclear criticality safety. The Monte Carlo simulation for noise analysis can be made by simulating physical phenomena in the course of neutron transport in a nuclear fuel as practically as possible. MCNP-DSP was developed by T. Valentine of ORNL for this purpose and it is a modified version of MCNP-4A. The authors applied this code to frequency analysis measurements performed in light-water critical assembly TCA. Prompt neutron generation times for critical and subcritical cores were measured by doing the frequency analysis of detector signals. The Monte Carlo simulations for these experiments were carried out using MCNP-DSP, and prompt neutron generation times were calculated. (author)
Evaluation of Geometric Progression (GP Buildup Factors using MCNP Codes (MCNP6.1 and MCNP5-1.60
Directory of Open Access Journals (Sweden)
Kim Kyung-O
2016-01-01
Full Text Available The gamma-ray buildup factors of three-dimensional point kernel code (QAD-CGGP are re-evaluated by using MCNP codes (MCNP6.1 and MCNPX5-1.60 and ENDF/B-VI.8 photoatomic data, which cover an energy range of 0.015–15 MeV and an iron thickness of 0.5–40 Mean Free Path (MFP. These new data are fitted to the Geometric Progression (GP fitting function and are then compared with ANS standard data equipped with QAD-CGGP. In addition, a simple benchmark calculation was performed to compare the QAD-CGGP results applied with new and existing buildup factors based on the MCNP codes. In the case of the buildup factors of low-energy gamma-rays, new data are evaluated to be about 5% higher than the existing data. In other cases, these new data present a similar trend based on the specific penetration depth, while existing data continuously increase beyond that depth. In a simple benchmark, the calculations using the existing data were slightly underestimated compared to the reference data at a deep penetration depth. On the other hand, the calculations with new data were stabilized with an increasing penetration depth, despite a slight overestimation at a shallow penetration depth.
Experimental validation of lead cross sections for scale and MCNP
International Nuclear Information System (INIS)
Henrikson, D.J.
1995-01-01
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 eff which require further study include diameters of styrofoam inserts and homogenization
Potential of the MCNP computer code
International Nuclear Information System (INIS)
Kyncl, J.
1995-01-01
The MCNP code is designed for numerical solution of neutron, photon, and electron transport problems by the Monte Carlo method. The code is based on the linear transport theory of behavior of the differential flux of the particles. The code directly uses data from the cross section point data library for input. Experience is outlined, gained in the application of the code to the calculation of the effective parameters of fuel assemblies and of the entire reactor core, to the determination of the effective parameters of the elementary fuel cell, and to the numerical solution of neutron diffusion and/or transport problems of the fuel assembly. The agreement between the calculated and observed data gives evidence that the MCNP code can be used with advantage for calculations involving WWER type fuel assemblies. (J.B.). 4 figs., 6 refs
International Nuclear Information System (INIS)
Cashwell, E.D.; Schrandt, R.G.
1980-01-01
The current state of the art of calculating flux at a point with MCNP is discussed. Various techniques are touched upon, but the main emphasis is on the fast improved version of the once-more-collided flux estimator, which has been modified to treat neutrons thermalized by the free gas model. The method is tested on several problems on interest and the results are presented
Data analysis and visualization in MCNP trademark
International Nuclear Information System (INIS)
Waters, L.S.
1994-01-01
There are many situations where the user may wish to go beyond current MCNP capabilities. For example, data produced by the code may need formatting for input into an external graphics package. Limitations on disk space may hinder writing out large PTRAK files. Specialized data analysis routines may be needed to model complex experimental results. One may wish to produce particle histories in a format not currently available in the code. To address these and other similar concerns a new capability in MCNP is being tested. A number of real, integer, logical and character variables describing the current and past characteristics of a particle are made available online to the user in three subroutines. The type of data passed can be controlled by cards in the INP file. The subroutines otherwise are empty, and the user may code in any desired analysis. A new MCNP executable is produced by compiling these subroutines and linking to a library which contains the object files for the rest of the code
MCNP and MATXS cross section libraries based on JENDL-3.3
International Nuclear Information System (INIS)
Kosako, Kazuaki; Konno, Chikara; Fukahori, Tokio; Shibata, Keiichi
2003-01-01
The continuous energy cross section library for the Monte Carlo transport code MCNP-4C, FSXLIB-J33, has been generated from the latest version of JENDL-3.3. The multigroup cross section library with the MATXS format, MATXS-J33, has been generated also from JENDL-3.3. Both libraries contain all nuclides in JENDL-3.3 and are processed at 300 K by the nuclear data processing system NJOY99. (author)
Analysis of Topaz-II reactor performance using MCNP and TFEHX
International Nuclear Information System (INIS)
Lee, H.H.; Klein, A.C.
1993-01-01
Data reported by Russian scientist and engineers for the TOPAZ-II Space Nuclear Power is compared with analytical results calculated using the Monte Carlo Neutron and Photon (MCNP) and TFEHX computer codes. The results of these comparisons show good agreement with the TOPAZ-II neutronics, thermionic and thermal hydraulics performance. A detailed description of the TOPAZ-II reactor and of the TFE should enhance the performance of the both codes in modeling the reactor and TFE performances
Elaborate SMART MCNP Modelling Using ANSYS and Its Applications
Song, Jaehoon; Surh, Han-bum; Kim, Seung-jin; Koo, Bonsueng
2017-09-01
An MCNP 3-dimensional model can be widely used to evaluate various design parameters such as a core design or shielding design. Conventionally, a simplified 3-dimensional MCNP model is applied to calculate these parameters because of the cumbersomeness of modelling by hand. ANSYS has a function for converting the CAD `stp' format into an MCNP input in the geometry part. Using ANSYS and a 3- dimensional CAD file, a very detailed and sophisticated MCNP 3-dimensional model can be generated. The MCNP model is applied to evaluate the assembly weighting factor at the ex-core detector of SMART, and the result is compared with a simplified MCNP SMART model and assembly weighting factor calculated by DORT, which is a deterministic Sn code.
Application of MCNP in the criticality calculation for reactors
International Nuclear Information System (INIS)
Zhong Zhaopeng; Shi Gong; Hu Yongming
2003-01-01
The criticality calculation is carried out with 3-D Monte Carlo code (MCNP). The author focuses on the introduction of modelling of the core and reflector. The core description is simplified by using repetition structure function of MCNP. k eff in different control rods positions are calculated for the case of JRR3, and the results is consistent with that of the reference. This work shows that MCNP is applicable for reactor criticality calculation
Development of visual platform of MCNP4B
International Nuclear Information System (INIS)
Fan Jiajin; Wang Yi; Cheng Jianping
2002-01-01
For convenience of using MCNP, the authors successfully developed a new code named McnpClient. With friend man-machine interface, the users can create input files very easily. If any error occurs during running process, McnpClient will give detailed fatal error or bad trouble messages. When the running is done, all the data can be obtained and in the mean time the curves associated with the data can be displayed
Monte Carlo parameter studies and uncertainty analyses with MCNP5
International Nuclear Information System (INIS)
Brown, F. B.; Sweezy, J. E.; Hayes, R.
2004-01-01
A software tool called mcnp p study has been developed to automate the setup, execution, and collection of results from a series of MCNP5 Monte Carlo calculations. This tool provides a convenient means of performing parameter studies, total uncertainty analyses, parallel job execution on clusters, stochastic geometry modeling, and other types of calculations where a series of MCNP5 jobs must be performed with varying problem input specifications. (authors)
Benchmark of WIMS-IST against MCNP for CANDU pressure tube fast fluxes
International Nuclear Information System (INIS)
Donders, R.E.; Douglas, S.R.
2002-01-01
Pressure tube fast-flux data in CANDU are currently calculated using the multi-group neutron transport code WIMS-IST. In this study, the WIMS-IST fast flux calculations are benchmarked against MCNP calculations (a Monte Carlo particle transport code), over the range of fuel burnup and coolant density in CANDU. The comparison shows good agreement between WIMS and MCNP, with WIMS fast fluxes being 1.5% to 4% lower than the MCNP values. The difference is smallest for fresh fuel, and increases with burnup. The fast flux gradient across the pressure tube (factor of 1.23 from inner edge to outer edge) is accurately calculated by WIMS. When reporting fast fluxes in pressure tubes, these are generally given as >1.000 MeV fluxes. For WIMS, this requires an extra conversion step, since the WIMS ENDF/B libraries do not have a group boundary at 1 MeV. The conversion step is based on a fictitious isotope ONEMEV in the WIMS nuclear data library. The conversion factor in WIMS was found to be about one percent too high. When providing >1 MeV fluxes from WIMS, this partially compensates for the slight under prediction of the fast flux. Pressure tube >1 MeV fluxes from WIMS are therefore 0.5% to 3% lower than MCNP values. To obtain accurate fast flux data, neutron transport calculations must be performed on a critical cell. For this study, all calculations were performed with radial albedo boundary conditions giving a critical cell. This required the use of an albedo version of MCNP, developed at AECL. (author)
Use of MCNP + GADRAS in Generating More Realistic Gamma-Ray Spectra for Plutonium and HEU Objects
International Nuclear Information System (INIS)
Rawool-Sullivan, Mohini; Mattingly, John; Mitchell, Dean
2012-01-01
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 output data analysis with ROOT (MODAR)
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
International Nuclear Information System (INIS)
Karriem, Z.; Ivanov, K.; Zamonsky, O.
2011-01-01
This paper presents work that has been performed to develop an integrated Monte Carlo- Deterministic transport methodology in which the two methods make use of exactly the same general geometry and multigroup nuclear data. The envisioned application of this methodology is in reactor lattice physics methods development and shielding calculations. The methodology will be based on the Method of Long Characteristics (MOC) and the Monte Carlo N-Particle Transport code MCNP5. Important initial developments pertaining to ray tracing and the development of an MOC flux solver for the proposed methodology are described. Results showing the viability of the methodology are presented for two 2-D general geometry transport problems. The essential developments presented is the use of MCNP as geometry construction and ray tracing tool for the MOC, verification of the ray tracing indexing scheme that was developed to represent the MCNP geometry in the MOC and the verification of the prototype 2-D MOC flux solver. (author)
Development of interface between MCNP-FISPACT-MCNP (IPR-MFM) based on rigorous two step method
International Nuclear Information System (INIS)
Shaw, A.K.; Swami, H.L.; Danani, C.
2015-01-01
In this work we present the development of interface tool between MCNP-FISPACT-MCNP (MFM) based on Rigorous Two Step method for the shutdown dose rate (SDDR) calculation. The MFM links MCNP radiation transport and the FISPACT inventory code through a suitable coupling scheme. MFM coupling scheme has three steps. In first step it picks neutron spectrum and total flux from MCNP output file to use as input parameter for FISPACT. It prepares the FISPACT input files by using irradiation history, neutron flux and neutron spectrum and then execute the FISPACT input file in the second step. Third step of MFM coupling scheme extracts the decay gammas from the FISPACT output file and prepares MCNP input file for decay gamma transport followed by execution of MCNP input file and estimation of SDDR. Here detailing of MFM methodology and flow scheme has been described. The programming language PYTHON has been chosen for this development of the coupling scheme. A complete loop of MCNP-FISPACT-MCNP has been developed to handle the simplified geometrical problems. For validation of MFM interface a manual cross-check has been performed which shows good agreements. The MFM interface also has been validated with exiting MCNP-D1S method for a simple geometry with 14 MeV cylindrical neutron source. (author)
Verification of the AZNHEX code v.1.4 with MCNP6 for different reference cases
International Nuclear Information System (INIS)
Galicia A, J.; Francois L, J. L.; Bastida O, G. E.; Del Valle G, E.
2017-09-01
The codes that make up the AZTLAN platform (AZTHECA, AZTRAN, AZKIND and AZNHEX) are currently in the testing phase simulating a variety of nuclear reactor assemblies and cores to compare and validate the results obtained for a particular case, with codes globally used in the nuclear area such as CASMO, Serpent and MCNP. The objective of this work is to continue improving the future versions of the codes of the AZTLAN platform so that accurate and reliable results can be obtained for the user. To test the current version of the AZNHEX code, 3 cases were taken into account, the first being the simulation of a VVER-440 reactor assembly; for the second case, the assembly of a fast reactor cooled with helium was simulated and for the third case it was decided to take up the case of the core of a fast reactor cooled with sodium, this because the previous versions of AZNHEX did not show adequate results and, in addition, they presented a considerable amount of limitations. The comparison and validation of the results (neutron multiplication factor, radial power, radial flow, axial power) for these three cases were made using the code MCNP6. The results obtained show that this version of AZNHEX produces values of the neutron multiplication factor and the neutron and power flow distributions very close to those of MCNP6. (Author)
Burnup calculation of a CANDU6 reactor using the Serpent and MCNP6 codes
Energy Technology Data Exchange (ETDEWEB)
Hussein, M.S.; Bonin, H.W., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, ON (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, ON (Canada)
2014-07-01
A study of fuel burnup for the CANDU6 reactor is carried out to validate the most recent versions of the probabilistic transport code (MCNP6) and the continuous energy burnup calculation code (Serpent). These two codes allow for 3-D geometry calculation accounting for a detailed analysis without unit-cell homogenization. On the other hand, the WIMS-AECL computer program is used to model neutron transport in nuclear-reactor lattices for design, safety analysis, and operation. It works with two-dimensional regions and can perform collision probability calculations for a periodic structure of the lattice cell. In the present work, the multiplication factor, the total flux and fuel burnup could be calculated for a CANDU6 nuclear reactor based on the GENTILLY-2 core design. The MCNP6 and Serpent codes provide a calculation of the track length estimated flux per neutron source. This estimated flux is then scaled with normalization to the reactor power in order to provide a flux in unit of n/cm{sup 2}s. Good agreement is observed between the actual total flux calculated by MCNP6, Serpent and WIMS-AECL. The effective multiplication factors of the whole core CANDU6 reactor are further calculated as a function of burnup and further compared to those calculated by WIMS-AECL where excellent agreement is also obtained. (author)
Burnup calculation of a CANDU6 reactor using the Serpent and MCNP6 codes
International Nuclear Information System (INIS)
Hussein, M.S.; Bonin, H.W.; Lewis, B.J.
2014-01-01
A study of fuel burnup for the CANDU6 reactor is carried out to validate the most recent versions of the probabilistic transport code (MCNP6) and the continuous energy burnup calculation code (Serpent). These two codes allow for 3-D geometry calculation accounting for a detailed analysis without unit-cell homogenization. On the other hand, the WIMS-AECL computer program is used to model neutron transport in nuclear-reactor lattices for design, safety analysis, and operation. It works with two-dimensional regions and can perform collision probability calculations for a periodic structure of the lattice cell. In the present work, the multiplication factor, the total flux and fuel burnup could be calculated for a CANDU6 nuclear reactor based on the GENTILLY-2 core design. The MCNP6 and Serpent codes provide a calculation of the track length estimated flux per neutron source. This estimated flux is then scaled with normalization to the reactor power in order to provide a flux in unit of n/cm 2 s. Good agreement is observed between the actual total flux calculated by MCNP6, Serpent and WIMS-AECL. The effective multiplication factors of the whole core CANDU6 reactor are further calculated as a function of burnup and further compared to those calculated by WIMS-AECL where excellent agreement is also obtained. (author)
MCNPX{trademark} -- The LAHET{trademark}/MCNP{trademark} code merger
Energy Technology Data Exchange (ETDEWEB)
Hughes, H.G.; Adams, K.J.; Chadwick, M.B. [and others
1997-08-01
The MCNP code is written and maintained by Group X-TM at Los Alamos National Laboratory. In response to the demands of the accelerator community, the authors have undertaken a major effort to expand the capabilities of MCNP to increase the set of transportable particles; to make use of newly evaluated high-energy nuclear data tables for neutrons, protons, and potentially other particles; and to incorporate physics models for use where tabular data are unavailable. A preliminary version of the expanded code, called MCNPX, has now been issued for testing. The new code includes all existing LAHET physics modules, and has the ability to utilize the 150-MeV data libraries that have recently been released by LANL Group T-2.
An MCNP model of glove boxes in a plutonium processing facility
International Nuclear Information System (INIS)
Dooley, D.E.; Kornreich, D.E.
1998-01-01
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
Validation of MCNP: SPERT-D and BORAX-V fuel
International Nuclear Information System (INIS)
Crawford, C.; Palmer, B.
1992-11-01
This report discusses critical experiments involving SPERT-D 1,2 fuel elements and BORAX-V 3-8 fuel which have been modeled and calculations performed with MCNP. MCNP is a Monte Carlo based transport code. For this study continuous-energy nuclear data from the ENDF/B-V cross section library was used. The SPERT-D experiments consisted of various arrays of fuel elements moderated and reflected with either water or a uranyl nitrate solution. Some SPERT-D experiments used cadmium as a fixed neutron poison, while others were poisoned with various concentrations of boron in the moderating/reflecting solution. ne BORAX-V experiments were arrays of either boiling fuel rod assemblies or superheater assemblies, both types of arrays were moderated and reflected with water. In one boiling fuel experiment, two fuel rods were replaced with borated stainless steel poison rods
Validation of MCNP4A for repository scattered radiation analysis
International Nuclear Information System (INIS)
Haas, M.N.; Su, S.
1998-02-01
Comparison is made between experimentally determined albedo (scattered) radiation and MCNP4A predictions in order to provide independent validation for repository shielding analysis. Both neutron and gamma scattered radiation fields from concrete ducts are compared in this paper. Satisfactory agreement is found between actual and calculated results with conservative values calculated by the MCNP4A code for all conditions
Development and improvement for MCNP-3B interactive plotter
International Nuclear Information System (INIS)
Gao Yanfeng
1996-01-01
The author briefly explains the development and improvement for the MCNP-3B interactive plotter. It describes the functions of geometry visualization and tally result plot, and introduces the progresses in user interface, process display and surface matching. The construction of MCNP-3B/PC is given
An assessment of the MCNP4C weight window
International Nuclear Information System (INIS)
Culbertson, Christopher N.; Hendricks, John S.
1999-01-01
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
Semi-Analytical Benchmarks for MCNP6
Energy Technology Data Exchange (ETDEWEB)
Grechanuk, Pavel Aleksandrovi [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2016-11-07
Code verification is an extremely important process that involves proving or disproving the validity of code algorithms by comparing them against analytical results of the underlying physics or mathematical theory on which the code is based. Monte Carlo codes such as MCNP6 must undergo verification and testing upon every release to ensure that the codes are properly simulating nature. Specifically, MCNP6 has multiple sets of problems with known analytic solutions that are used for code verification. Monte Carlo codes primarily specify either current boundary sources or a volumetric fixed source, either of which can be very complicated functions of space, energy, direction and time. Thus, most of the challenges with modeling analytic benchmark problems in Monte Carlo codes come from identifying the correct source definition to properly simulate the correct boundary conditions. The problems included in this suite all deal with mono-energetic neutron transport without energy loss, in a homogeneous material. The variables that differ between the problems are source type (isotropic/beam), medium dimensionality (infinite/semi-infinite), etc.
International Nuclear Information System (INIS)
Lara, Rafael G.; Maiorino, Jose R.
2013-01-01
This work aimed at the implementation and qualification of MCNP code in a supercomputer of the Universidade Federal do ABC, so that may be available a next-generation simulation tool for precise calculations of nuclear reactors and systems subject to radiation. The implementation of this tool will have multidisciplinary applications, covering various areas of engineering (nuclear, aerospace, biomedical), radiation physics and others
Compilation of MCNP data library based on JENDL-3T and test through analysis of benchmark experiment
International Nuclear Information System (INIS)
Sakurai, K.; Sasamoto, N.; Kosako, K.; Ishikawa, T.; Sato, O.; Oyama, Y.; Narita, H.; Maekawa, H.; Ueki, K.
1989-01-01
Based on an evaluated nuclear data library JENDL-3T, a temporary version of JENDL-3, a pointwise neutron cross section library for MCNP code is compiled which involves 39 nuclides from H-1 to Am-241 which are important for shielding calculations. Compilation is performed with the code system which consists of the nuclear data processing code NJOY-83 and library compilation code MACROS. Validity of the code system and reliability of the library are certified by analysing benchmark experiments. (author)
ENDF/B-VI data for MCNP trademark
International Nuclear Information System (INIS)
Hendricks, J.S.; Frankle, S.C.; Court, J.D.
1994-12-01
Nuclear and atomic data are the foundation upon which the radiation transport codes are built. For neutron transport the international standard is the Evaluated Nuclear Data File from Brookhaven National Laboratory. The latest version, ENDF/B-VI release 2, has recently become available for use in the Monte Carlo N-Particle (MCNP) radiation transport code. These neutron cross-section data are designated by ZAID identifiers ending in .60c and are referred to as the ENDF60 library. The ENDF60 data library was processed from the ENDF/B-VI evaluations using the NJOY code. Fifty-two percent of the data evaluations are translations from ENDF/B-V. The remaining 48% are new evaluations which have sometimes changed significantly. The RSIC release package contains the ENDF60 neutron library, a new photon library MCPLIB02, the electron library EL1, and an updated XSDIR file. The authors report here the work done by the LANL Radiation Transport Group (X-6) in testing and validating the ENDF60 data library and in developing the necessary new sampling and detector schemes. When the ENDF60 library should be used in preference to the previous libraries, is also considered. The development of the new photon library MCPLIB02 is also discussed
New Neutron, Proton, and S(α,β) MCNP Data Libraries Based on ENDF/B-VII
International Nuclear Information System (INIS)
Little, Robert C.; Trellue, Holly R.; MacFarlane, Robert E.; Kahler, A.C.; Lee, Mary Beth; White, Morgan C.
2008-01-01
The general-purpose Evaluated Nuclear Data File ENDF/B-VII.0 was released in December 2006. A number of sub-libraries were included in ENDF/B-VII.0 such that data were provided for incident neutrons, photons, and charged particles. This paper describes the creation of MCNP data libraries at Los Alamos National Laboratory based on three ENDF/B-VII.0 sub-libraries: neutrons, protons, and thermal scattering. An ACE-formatted continuous-energy neutron data library called ENDF70 for MCNP has been produced. This library provides data for 390 materials at five temperatures: 293.6, 600, 900, 1200, and 2500 K. The library was processed primarily with Version 248 of NJOY99. Extensive checking and quality-assurance tests were applied to the data. Improvements to the processing code were made and certain evaluations were modified as a result of these tests. ENDF/B-VII.0 included proton evaluations for 48 target materials. Forty-seven proton evaluations (all except for 13 C) were processed at room temperature and combined into the MCNP library ENDF70PROT. Neutron thermal S(α,β) scattering data exist for twenty different materials in ENDF/B-VII.0. All twenty of these evaluations were processed at all applicable temperatures (these vary for each evaluation), and combined into the MCNP library ENDF70SAB. All of these ENDF/B-VII.0 based MCNP libraries (ENDF70, ENDF70PROT, and ENDF70SAB) are available as part of the MCNP5 1.50 release. (authors)
Characteristics of Multihole Collimator Gamma Camera Simulation Modeled Using MCNP5
International Nuclear Information System (INIS)
Saripan, M. I.; Mashohor, S.; Adnan, W. A. Wan; Marhaban, M. H.; Hashim, S.
2008-01-01
This paper describes the characteristics of the multihole collimator gamma camera that is simulated using the combination of the Monte Carlo N-Particles Code (MCNP) version 5 and in-house software. The model is constructed based on the GCA-7100A Toshiba Gamma Camera at the Royal Surrey County Hospital, Guildford, Surrey, UK. The characteristics are analyzed based on the spatial resolution of the images detected by the Sodium Iodide (NaI) detector. The result is recorded in a list-mode file referred to as a PTRAC file within MCNP5. All pertinent nuclear reaction mechanisms, such as Compton and Rayleigh scattering and photoelectric absorption are undertaken by MCNP5 for all materials encountered by each photon. The experiments were conducted on Tl-201, Co-57, Tc-99 m and Cr-51 radio nuclides. The comparison of full width half maximum value of each datasets obtained from experimental work, simulation and literature are also reported in this paper. The relationship of the simulated data is in agreement with the experimental results and data obtained in the literature. A careful inspection at each of the data points of the spatial resolution of Tc-99 m shows a slight discrepancy between these sets. However, the difference is very insignificant, i.e. less than 3 mm only, which corresponds to a size of less than 1 pixel only (of the segmented detector)
MCNP perturbation technique for criticality analysis
International Nuclear Information System (INIS)
McKinney, G.W.; Iverson, J.L.
1995-01-01
The differential operator perturbation technique has been incorporated into the Monte Carlo N-Particle transport code MCNP and will become a standard feature of future releases. This feature includes first and/or second order terms of the Taylor Series expansion for response perturbations related to cross-section data (i.e., density, composition, etc.). Criticality analyses can benefit from this technique in that predicted changes in the track-length tally estimator of K eff may be obtained for multiple perturbations in a single run. A key advantage of this method is that a precise estimate of a small change in response (i.e., < 1%) is easily obtained. This technique can also offer acceptable accuracy, to within a few percent, for up to 20-30% changes in a response
Monte Carlo MCNP-4B-based absorbed dose distribution estimates for patient-specific dosimetry.
Yoriyaz, H; Stabin, M G; dos Santos, A
2001-04-01
This study was intended to verify the capability of the Monte Carlo MCNP-4B code to evaluate spatial dose distribution based on information gathered from CT or SPECT. A new three-dimensional (3D) dose calculation approach for internal emitter use in radioimmunotherapy (RIT) was developed using the Monte Carlo MCNP-4B code as the photon and electron transport engine. It was shown that the MCNP-4B computer code can be used with voxel-based anatomic and physiologic data to provide 3D dose distributions. This study showed that the MCNP-4B code can be used to develop a treatment planning system that will provide such information in a time manner, if dose reporting is suitably optimized. If each organ is divided into small regions where the average energy deposition is calculated with a typical volume of 0.4 cm(3), regional dose distributions can be provided with reasonable central processing unit times (on the order of 12-24 h on a 200-MHz personal computer or modest workstation). Further efforts to provide semiautomated region identification (segmentation) and improvement of marrow dose calculations are needed to supply a complete system for RIT. It is envisioned that all such efforts will continue to develop and that internal dose calculations may soon be brought to a similar level of accuracy, detail, and robustness as is commonly expected in external dose treatment planning. For this study we developed a code with a user-friendly interface that works on several nuclear medicine imaging platforms and provides timely patient-specific dose information to the physician and medical physicist. Future therapy with internal emitters should use a 3D dose calculation approach, which represents a significant advance over dose information provided by the standard geometric phantoms used for more than 20 y (which permit reporting of only average organ doses for certain standardized individuals)
MCNP HPGe detector benchmark with previously validated Cyltran model.
Hau, I D; Russ, W R; Bronson, F
2009-05-01
An exact copy of the detector model generated for Cyltran was reproduced as an MCNP input file and the detection efficiency was calculated similarly with the methodology used in previous experimental measurements and simulation of a 280 cm(3) HPGe detector. Below 1000 keV the MCNP data correlated to the Cyltran results within 0.5% while above this energy the difference between MCNP and Cyltran increased to about 6% at 4800 keV, depending on the electron cut-off energy.
Energy Technology Data Exchange (ETDEWEB)
Lara, Rafael G.; Maiorino, Jose R., E-mail: rafael.lara@aluno.ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais Aplicadas
2013-07-01
This work aimed at the implementation and qualification of MCNP code in a supercomputer of the Universidade Federal do ABC, so that may be available a next-generation simulation tool for precise calculations of nuclear reactors and systems subject to radiation. The implementation of this tool will have multidisciplinary applications, covering various areas of engineering (nuclear, aerospace, biomedical), radiation physics and others.
MOCUP: MCNP-ORIGEN2 coupled utility program
International Nuclear Information System (INIS)
Moore, R.L.; Schnitzler, B.G.; Wemple, C.A.
1995-01-01
MOCUP is a system of external processors that allow for a limited treatment of the temporal composition of the user-selected MCNP cells in a time-dependent flux environment. The ORIGEN2 code computes the time-dependent compositions of these individually selected MCNP cells. All data communication between the two codes is accomplished through the MCNP and ORIGEN2 input/output files, the MOCUP Processor Output files, and two user supplied tables. MOCUP is either command line or interactively driven. The interactive interface is based on the portable XII window environment and the Motif tool kit. MOCUP was constructed so that no modifications to either MCNP or ORIGEN2 were necessary. Section 4 of the writeup contains the input instructions needed to set up the MOCUP run. MOCUP is extremely useful for analysts who perform isotope production, material transformation, and depletion and isotope analyses on complex, non-lattice geometries, and uniform and non-uniform lattices
E language based on MCNP modeling software for autonomous
International Nuclear Information System (INIS)
Li Fei; Ge Liangquan; Zhang Qingxian
2010-01-01
MCNP (Monte Carlo N-Particle Code) is based on the Monte Carlo method for computing neutron, photon and other particles as the object of the movement simulation computer program. Because of its powerful computing simulation, flexible and universal features in many fields has been widely used, but due to a software professional in the operating area has been greatly restricted, so that in later development has been greatly hindered. E-language was used in order to develop the autonomy of MCNP modeling software, used to address users not familiar with MCNP and can not create object model, get rid of dull red tape 'notebook' type of program type and built a new MCNP modeling system. (authors)
Particle Track Visualization using the MCNP Visual Editor
International Nuclear Information System (INIS)
Schwarz, Randolph A.; Carter, Lee; Brown, Wendi A.
2001-01-01
The Monte Carlo N-Particle (MCNP) visual editor1,2,3 is used throughout the world for displaying and creating complex MCNP geometries. The visual editor combines the Los Alamos MCNP Fortran code with a C front end to provide a visual interface. A big advantage of this approach is that the particle transport routines for MCNP are available to the visual front end. The latest release of the visual editor by Pacific Northwest National Laboratory enables the user to plot transport data points on top of a two-dimensional geometry plot. The user can plot source points, collisions points, surface crossings, and tally contributions. This capability can be used to show where particle collisions are occurring, verify the effectiveness of the particle biasing, or show which collisions contribute to a tally. For a KCODE (criticality source) calculation, the visual editor can be used to plot the source points for specific cycles
GB - a preliminary linking code between MCNP4C and Origen2.1 - DEN/UFMG version
International Nuclear Information System (INIS)
Campolina, Daniel; Pereira, Claubia; Veloso, Maria Auxiliadora Fortini; Cavatoni, Andre
2009-01-01
Nowadays it is possible to perform burnup simulation in a detailed 3D geometry and a continuous energy description by the Monte Carlo method. This paper describes an initial project to create and verify a connection code to link Origen2.1 (Oak Ridge National Laboratory) and MCNP4C (Los Alamos National Laboratory). Essentially the code includes point depletion capability to the MCNP code. The incorporation of point depletion capability is explicit and can be summarized by three steps: 1-Monte Carlo determines reaction rates, 2-the reaction rates are used to determine microscopic cross sections for depletion equations, 3-solution of depletion equations (given by Origen2.1) determines number densities for next MCNP step. To evaluate the initial version of the program, we focused on comparing the results with one of the major Monte Carlo burnup codes: MCNPX version 2.6.0. The input files for all codes share the same MCNP geometry, nuclear data library and core thermal power. While simulating 75 time steps at 800 kw of a Heat Pipe Power System model, we have found that the codes generate very similar results. The neutron flux and criticality value of the core agree, especially in the begin of burnup when the influence of fission products are not very considerable. The small difference encountered was probably caused by the difference in the number of isotopes considered in the transport models (89 MCNPX x 25 GB (author)
Lecture note on neutron and photon transport calculation with MCNP
International Nuclear Information System (INIS)
Sakurai, Kiyoshi
2003-01-01
This paper is a lecture note on the continuous energy Monte Carlo method. The contents are as follows; history of the Monte Carlo study, continuous energy Monte Carlo codes, libraries, evaluation method for calculation results, integral emergent particle density equation, pseudorandom number, random walk, variance reduction techniques, MCNP weight window method, MCNP weight window generator, exponential transform, estimators, criticality problem and research subjects. This paper is a textbook for beginners on the Monte Carlo calculation. (author)
Spectral measurements in critical assemblies: MCNP specifications and calculated results
Energy Technology Data Exchange (ETDEWEB)
Stephanie C. Frankle; Judith F. Briesmeister
1999-12-01
Recently, a suite of 86 criticality benchmarks for the Monte Carlo N-Particle (MCNP) transport code was developed, and the results of testing the ENDF/B-V and ENDF/B-VI data (through Release 2) were published. In addition to the standard k{sub eff} measurements, other experimental measurements were performed on a number of these benchmark assemblies. In particular, the Cross Section Evaluation Working Group (CSEWG) specifications contain experimental data for neutron leakage and central-flux measurements, central-fission ratio measurements, and activation ratio measurements. Additionally, there exists another set of fission reaction-rate measurements performed at the National Institute of Standards and Technology (NIST) utilizing a {sup 252}Cf source. This report will describe the leakage and central-flux measurements and show a comparison of experimental data to MCNP simulations performed using the ENDF/B-V and B-VI (Release 2) data libraries. Central-fission and activation reaction-rate measurements will be described, and the comparison of experimental data to MCNP simulations using available data libraries for each reaction of interest will be presented. Finally, the NIST fission reaction-rate measurements will be described. A comparison of MCNP results published previously with the current MCNP simulations will be presented for the NIST measurements, and a comparison of the current MCNP simulations to the experimental measurements will be presented.
Spectral measurements in critical assemblies: MCNP specifications and calculated results
International Nuclear Information System (INIS)
Frankle, Stephanie C.; Briesmeister, Judith F.
1999-01-01
Recently, a suite of 86 criticality benchmarks for the Monte Carlo N-Particle (MCNP) transport code was developed, and the results of testing the ENDF/B-V and ENDF/B-VI data (through Release 2) were published. In addition to the standard k eff measurements, other experimental measurements were performed on a number of these benchmark assemblies. In particular, the Cross Section Evaluation Working Group (CSEWG) specifications contain experimental data for neutron leakage and central-flux measurements, central-fission ratio measurements, and activation ratio measurements. Additionally, there exists another set of fission reaction-rate measurements performed at the National Institute of Standards and Technology (NIST) utilizing a 252 Cf source. This report will describe the leakage and central-flux measurements and show a comparison of experimental data to MCNP simulations performed using the ENDF/B-V and B-VI (Release 2) data libraries. Central-fission and activation reaction-rate measurements will be described, and the comparison of experimental data to MCNP simulations using available data libraries for each reaction of interest will be presented. Finally, the NIST fission reaction-rate measurements will be described. A comparison of MCNP results published previously with the current MCNP simulations will be presented for the NIST measurements, and a comparison of the current MCNP simulations to the experimental measurements will be presented
Analysis of radiation field distribution in Yonggwang unit 3 with MCNP code
International Nuclear Information System (INIS)
Lee, Cheol Woo; Ha, Wi Ho; Shin, Chang Ho; Kim, Soon Young; Kim, Jong Kyung
2004-01-01
Radiation field analysis is performed at the inside of the containment building of nuclear power plant(NPP) using the well-known MCNP code. The target NPP in this study is Yonggwang Unit 3 Cycle 8. In this work, whole transport calculations were done using MCNPX 2.4.0 due to the functional benefits, such as Mesh Tally, that the code provides. The neutron spectra released from the operating reactor core were firstly evaluated as a radiation source term, and then dose distributions in the work areas of the NPP were calculated
Criticality safety analysis of spent fuel storage for NPP Mochovce using MCNP5
International Nuclear Information System (INIS)
Farkas, G.; Hascik, J.; Lueley, J.; Vrban, B.; Petriska, M.; Slugen, V.; Urban, P.
2011-01-01
The paper presents results of nuclear criticality safety analysis of spent fuel storage for the first and second unit of NPP Mochovce. The spent fuel storage pool (compact and reserve grid) was modeled using the Monte Carlo code MCNP5. Conservative approach was applied and calculation of k eff values was performed for normal and various postulated emergency conditions in order to evaluate the final maximal k eff values. The requirement of current safety regulations to ensure 5% subcriticality was met except one especially conservative case. (Authors)
International Nuclear Information System (INIS)
Singh, Tej; Kumar, Jainendra; Sharma, Archana; Singh, Kanchhi; Raina, V.K.; Srinivasan, P.
2009-01-01
At present Dhruva and Cirus reactors provide majority of research reactor based experimental/irradiation facilities to cater to various needs of the vast pool of researchers in the field of sciences research and development work for nuclear power plants and production of radioisotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 30 MWt Multi Purpose Research Reactor is proposed to be constructed. This paper describes some of the physics design features of this reactor using MCNP code to validate the deterministic methods. The criticality calculations for 100 material testing reactor (JHR) of France and 610 MW SAVANNAH thermal reactor were performed using MCNP computer codes to boost the confidence level in designing the physics design of reactor core. (author)
Energy Technology Data Exchange (ETDEWEB)
Park, Jong Sung, E-mail: jspark@nfri.re.kr; Kwon, Sungjin; Im, Kihak
2016-11-01
Highlights: • A comparison study of main parameter calculations: neutron wall loading (NWL), tritium breeding ratio (TBR), and nuclear heating, on a Korean fusion demonstration reactor (K-DEMO) neutronic analysis model using MCNP and ATTILA was performed to investigate the feasibility of using ATTILA. • The calculation results of this study indicates that ATTILA showed close agreement with MCNP within ranges (3.3–28%). • Partly high discrepancy (17–28%) results between two codes existed to the nuclear heating calculation in high attenuating materials and radially thick structure regions. • The rest of the results showed small differences of NWL calculation (3.3%) and TBR distribution (3.9%). • ATTILA could be acceptable for K-DEMO neutronic analysis considering discrepancy (3.3–28%). - Abstract: A comparison study of main parameter calculations: neutron wall loading (NWL), tritium breeding ratio (TBR), and nuclear heating, on a Korean fusion demonstration reactor (K-DEMO) neutronic analysis model using MCNP and ATTILA was performed to investigate the feasibility of using ATTILA for the main parameter calculations. The model was created by commercial CAD program (Pro-Engineer™) as a 22.5° sector of tokamak consisting of major components such as blankets, shields, divertors, vacuum vessels (VV), toroidal field (TF) coils, and others, which was directly imported into ATTILA by Parasolid file. The discretizing in space, angle, and energy variables were refined for application of the K-DEMO neutronic analysis model through an iterative process since these variables greatly impact on accuracy, solution times, and memory consumptions in ATTILA. The main parameter calculations using ATTILA and the result of comparison studies indicate that the NWL distributions by two codes were almost agreed within discrepancy of 3.3%; the TBR distribution using ATTILA was slightly bigger than MCNP with a difference 3.9%; the nuclear heating values on TF coils and VV
KENO2MCNP, Version 5L, Conversion of Input Data between KENOV.a and MCNP File Formats
International Nuclear Information System (INIS)
2008-01-01
1 - Description of program or function: The KENO2MCNP program was written to convert KENO V.a input files to MCNP Format. This program currently only works with KENO Va geometries and will not work with geometries that contain more than a single array. A C++ graphical user interface was created that was linked to Fortran routines from KENO V.a that read the material library and Fortran routines from the MCNP Visual Editor that generate the MCNP input file. Either SCALE 5.0 or SCALE 5.1 cross section files will work with this release. 2 - Methods: The C++ binary executable reads the KENO V.a input file, the KENO V.a material library and SCALE data libraries. When an input file is read in, the input is stored in memory. The converter goes through and loads different sections of the input file into memory including parameters, composition, geometry information, array information and starting information. Many of the KENO V.a materials represent compositions that must be read from the KENO V.a material library. KENO2MCNP includes the KENO V.a FORTRAN routines used to read this material file for creating the MCNP materials. Once the file has been read in, the user must select 'Convert' to convert the file from KENO V.a to MCNP. This will generate the MCNP input file along with an output window that lists the KENO V.a composition information for the materials contained in the KENO V.a input file. The program can be run interactively by clicking on the executable or in batch mode from the command prompt. 3 - Restrictions on the complexity of the problem: Not all KENO V.a input files are supported. Only one array is allowed in the input file. Some of the more complex material descriptions also may not be converted
Application of the NJOY code for unresolved resonance treatment in the MCNP utility code
International Nuclear Information System (INIS)
Milosevic, M.; Greenspan, E.; Vujic, J. . E-mail addresses of corresponding authors: mmilos@vin.bg.ac.yu , vujic@nuc.berkeley.edu ,; Milosevic, M.; Vujic, J.)
2005-01-01
There are numerous uncertainties in the prediction of neutronic characteristics of reactor cores, particularly in the case of innovative reactor designs, arising from approximations used in the solution of the transport equation, and in nuclear data processing and cross section libraries generation. This paper describes the problems encountered in the analysis of the Encapsulated Nuclear Heat Source (ENHS) benchmark core and the new procedures and cross section libraries developed to overcome these problems. The ENHS is a new lead-bismuth or lead cooled novel reactor concept that is fuelled with metallic alloy of Pu, U and Zr, and it is designed to operate for 20 effective full power years without refuelling and with very small burnup reactivity swing. The computational tools benchmarked include: MOCUP - a coupled MCNP-4C and ORIGEN2.1 utility codes with MCNP data libraries based on the ENDF/B-VI evaluations; and KWO2 - a coupled KENO-V.a and ORIGEN2.1 code with ENDFB-V.2 based 238 group library. Calculations made for the ENHS benchmark have shown that the differences between the results obtained using different code systems and cross section libraries are significant and should be taken into account in assessing the quality of nuclear data libraries. (author)
Analysis of Gamma Dose Rate for RTP 2 MW Core Configuration Using MCNP
International Nuclear Information System (INIS)
Mohamad Hairie Rabir; Mohd Amin Sharifuldin Salleh; Julia Abdul Karim
2011-01-01
The Malaysian 1 MW TRIGA MARK II research reactor at Malaysian Nuclear Agency achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes for their use in agriculture, industry, and medicine. This study deals with the calculation of gamma dose rate at water pool surface and concrete shielding surface of the proposed 2-MW core configuration of PUSPATI TRIGA Reactor. The 3-D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA core with pool water and concrete shielding and validation of the input by comparisons with the measured and available safety analysis report (SAR) of the reactor. The model represents in detailed all components of the reactor with literally no physical approximation. Continuous energy cross section data from the more recent nuclear data as well as S(α, β) thermal neutron scattering functions distributed with the MCNP code were used. Results of calculations are analyzed and discussed. (author)
Comparisons of the MCNP criticality benchmark suite with ENDF/B-VI.8, JENDL-3.3, and JEFF-3.0
International Nuclear Information System (INIS)
Kim, Do Heon; Gil, Choong-Sup; Kim, Jung-Do; Chang, Jonghwa
2003-01-01
A comparative study has been performed with the latest evaluated nuclear data libraries ENDF/B-VI.8, JENDL-3.3, and JEFF-3.0. The study has been conducted through the benchmark calculations for 91 criticality problems with the libraries processed for MCNP4C. The calculation results have been compared with those of the ENDF60 library. The self-shielding effects of the unresolved-resonance (UR) probability tables have also been estimated for each library. The χ 2 differences between the MCNP results and experimental data were calculated for the libraries. (author)
UNR. A code for processing unresolved resonance data for MCNP
International Nuclear Information System (INIS)
Hogenbirk, A.
1994-09-01
In neutron transport problems the correct treatment of self-shielding is important for those nuclei present in large concentrations. Monte Carlo calculations using continuous-energy cross section data, such as calculations with the code MCNP, offer the advantage that neutron transport is calculated in a very accurate way. Self-shielding in the resolved resonance region is taken into account exactly in MCNP. However, self-shielding in the unresolved resonance region can not be taken into account by MCNP, although the effect of it may be important in many applications. In this report a description is given of the computer code UNR. With this code problem-dependent cross section libraries can be produced for MCNP. In these libraries self-shielded cross section data in the unresolved resonance range are given, which are produced by NJOY-module UNRESR. It is noted, that the treatment for resonance self-shielding presented in this report is approximate. However, the current version of MCNP does not allow the use of probability tables, which would be a general solution. (orig.)
Benchmarking comparison and validation of MCNP photon interaction data
Directory of Open Access Journals (Sweden)
Colling Bethany
2017-01-01
Full Text Available The objective of the research was to test available photoatomic data libraries for fusion relevant applications, comparing against experimental and computational neutronics benchmarks. Photon flux and heating was compared using the photon interaction data libraries (mcplib 04p, 05t, 84p and 12p. Suitable benchmark experiments (iron and water were selected from the SINBAD database and analysed to compare experimental values with MCNP calculations using mcplib 04p, 84p and 12p. In both the computational and experimental comparisons, the majority of results with the 04p, 84p and 12p photon data libraries were within 1σ of the mean MCNP statistical uncertainty. Larger differences were observed when comparing computational results with the 05t test photon library. The Doppler broadening sampling bug in MCNP-5 is shown to be corrected for fusion relevant problems through use of the 84p photon data library. The recommended libraries for fusion neutronics are 84p (or 04p with MCNP6 and 84p if using MCNP-5.
Benchmarking comparison and validation of MCNP photon interaction data
Colling, Bethany; Kodeli, I.; Lilley, S.; Packer, L. W.
2017-09-01
The objective of the research was to test available photoatomic data libraries for fusion relevant applications, comparing against experimental and computational neutronics benchmarks. Photon flux and heating was compared using the photon interaction data libraries (mcplib 04p, 05t, 84p and 12p). Suitable benchmark experiments (iron and water) were selected from the SINBAD database and analysed to compare experimental values with MCNP calculations using mcplib 04p, 84p and 12p. In both the computational and experimental comparisons, the majority of results with the 04p, 84p and 12p photon data libraries were within 1σ of the mean MCNP statistical uncertainty. Larger differences were observed when comparing computational results with the 05t test photon library. The Doppler broadening sampling bug in MCNP-5 is shown to be corrected for fusion relevant problems through use of the 84p photon data library. The recommended libraries for fusion neutronics are 84p (or 04p) with MCNP6 and 84p if using MCNP-5.
Measurements by activation foils and comparative computations by MCNP code
International Nuclear Information System (INIS)
Kyncl, J.
2008-01-01
Systematic study of the radioactive waste minimisation problem is subject of the SPHINX project. Its idea is that burning or transmutation of the waste inventory problematic part will be realized in a nuclear reactor the fuel of which is in the form of liquid fluorides. In frame of the project, several experiments have been performed with so-called inserted experimental channel. The channel was filled up by the fluorides mixture, surrounded by six fuel assemblies with moderator and placed into LR-0 reactor vessel. This formation was brought to critical state and measurement with activation foil detectors were carried out at selected positions of the inserted channel. Main aim of the measurements was to determine reaction rates for the detectors mentioned. For experiment evaluation, comparative computations were accomplished by code MCNP4a. The results obtained show that very often, computed values of reaction rates differ substantially from the values that were obtained from the experiment. This contribution deals with analysis of the reasons of these differences from the point of view of computations by Monte Carlo method. The analysis of concrete cases shows that the inaccuracy of reaction rate computed is caused mostly by three circumstances:-space region that is occupied by detector is relatively very small;- microscopic effective cross-section R(E) of the reaction changes strongly with energy just in the energy interval that gives the greatest contribution to the reaction; - in the energy interval that gives the greatest contribution to reaction rate, the error of the computed neutron flux is great. These circumstances evoke that the computation of reaction rate with casual accuracy submits extreme demands on computing time. (Author)
Energy Technology Data Exchange (ETDEWEB)
Galicia A, J.; Francois L, J. L.; Bastida O, G. E. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, Circuito Exterior s/n, 04510 Ciudad de Mexico (Mexico); Del Valle G, E., E-mail: jgaliciaa87@gmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, 07738 Ciudad de Mexico (Mexico)
2017-09-15
The codes that make up the AZTLAN platform (AZTHECA, AZTRAN, AZKIND and AZNHEX) are currently in the testing phase simulating a variety of nuclear reactor assemblies and cores to compare and validate the results obtained for a particular case, with codes globally used in the nuclear area such as CASMO, Serpent and MCNP. The objective of this work is to continue improving the future versions of the codes of the AZTLAN platform so that accurate and reliable results can be obtained for the user. To test the current version of the AZNHEX code, 3 cases were taken into account, the first being the simulation of a VVER-440 reactor assembly; for the second case, the assembly of a fast reactor cooled with helium was simulated and for the third case it was decided to take up the case of the core of a fast reactor cooled with sodium, this because the previous versions of AZNHEX did not show adequate results and, in addition, they presented a considerable amount of limitations. The comparison and validation of the results (neutron multiplication factor, radial power, radial flow, axial power) for these three cases were made using the code MCNP6. The results obtained show that this version of AZNHEX produces values of the neutron multiplication factor and the neutron and power flow distributions very close to those of MCNP6. (Author)
MCNP load balancing and fault tolerance with PVM
International Nuclear Information System (INIS)
McKinney, G.W.
1995-01-01
Version 4A of the Monte Carlo neutron, photon, and electron transport code MCNP, developed by LANL (Los Alamos National Laboratory), supports distributed-memory multiprocessing through the software package PVM (Parallel Virtual Machine, version 3.1.4). Using PVM for interprocessor communication, MCNP can simultaneously execute a single problem on a cluster of UNIX-based workstations. This capability provided system efficiencies that exceeded 80% on dedicated workstation clusters, however, on heterogeneous or multiuser systems, the performance was limited by the slowest processor (i.e., equal work was assigned to each processor). The next public release of MCNP will provide multiprocessing enhancements that include load balancing and fault tolerance which are shown to dramatically increase multiuser system efficiency and reliability
Monte Carlo importance sampling for the MCNP trademark general source
International Nuclear Information System (INIS)
Lichtenstein, H.
1996-01-01
Research was performed to develop an importance sampling procedure for a radiation source. The procedure was developed for the MCNP radiation transport code, but the approach itself is general and can be adapted to other Monte Carlo codes. The procedure, as adapted to MCNP, relies entirely on existing MCNP capabilities. It has been tested for very complex descriptions of a general source, in the context of the design of spent-reactor-fuel storage casks. Dramatic improvements in calculation efficiency have been observed in some test cases. In addition, the procedure has been found to provide an acceleration to acceptable convergence, as well as the benefit of quickly identifying user specified variance-reduction in the transport that effects unstable convergence
Generating and verification of ACE-multigroup library for MCNP
International Nuclear Information System (INIS)
Chen Chaobin; Hu Zehua; Chen Yixue; Wu Jun; Yang Shouhai
2012-01-01
The Monte Carlo code MCNP can handle multigroup calculations and a sample multigroup set based on ENDF/B-V, MGXSNP, is available for MCNP for coupled neutron-photon transport. However, this library is not suit- able for all problems, and there is a need for users to be able to generate multigroup libraries tailored to their specific applications. For these purposes CSPT (cross section processing tool) is created to generate multigroup library for MCNP from deterministic multigroup cross sections (GENDF or ANISN format at present). Several ACE-multigroup libraries based on ENDF/B-VII.0 converted and verified in this work, we drawn the conclusion that the CSPT code works correctly and the libraries produced are credible. (authors)
S values at voxels level for 188Re and 90Y calculated with the MCNP-4C code
International Nuclear Information System (INIS)
Coca, M.A.; Torres, L.A.; Cornejo, N.; Martin, G.
2008-01-01
Full text: MIRD formalism at voxel level has been suggested as an optional methodology to perform internal radiation dosimetry calculation during internal radiation therapy in Nuclear Medicine. Voxel S values for Y 90 , 131 I, 32 P, 99m Tc and 89 Sr have been published to different sizes. Currently, 188 Re has been proposed as a promising radionuclide for therapy due to its physical features and availability from generators. The main objective of this work was to estimate the voxel S values for 188 Re at cubical geometry using the MCNP-4C code for the simulations of radiation transport and energy deposition. Mean absorbed dose to target voxels per radioactive decay in a source voxel were estimated and reported for 188 Re and Y 90 . A comparison of voxel S values computed with the MCNP code and the data reported in MIRD Pamphlet 17 for 90 Y was performed in order to evaluate our results. (author)
MCNP speed advances for boron neutron capture therapy
International Nuclear Information System (INIS)
Goorley, J.T.; McKinney, G.; Adams, K.; Estes, G.
1998-04-01
The Boron Neutron Capture Therapy (BNCT) treatment planning process of the Beth Israel Deaconess Medical Center-M.I.T team relies on MCNP to determine dose rates in the subject's head for various beam orientations. In this time consuming computational process, four or five potential beams are investigated. Of these, one or two final beams are selected and thoroughly evaluated. Recent advances greatly decreased the time needed to do these MCNP calculations. Two modifications to the new MCNP4B source code, lattice tally and tracking enhancements, reduced the wall-clock run times of a typical one million source neutrons run to one hour twenty five minutes on a 200 MHz Pentium Pro computer running Linux and using the GNU FORTRAN compiler. Previously these jobs used a special version of MCNP4AB created by Everett Redmond, which completed in two hours two minutes. In addition to this 30% speedup, the MCNP4B version was adapted for use with Parallel Virtual Machine (PVM) on personal computers running the Linux operating system. MCNP, using PVM, can be run on multiple computers simultaneously, offering a factor of speedup roughly the same as the number of computers used. With two 200 MHz Pentium Pro machines, the run time was reduced to forty five minutes, a 1.9 factor of improvement over the single Linux computer. While the time of a single run was greatly reduced, the advantages associated with PVM derive from using computational power not already used. Four possible beams, currently requiring four separate runs, could be run faster when each is individually run on a single machine under Windows NT, rather than using Linux and PVM to run one after another with each multiprocessed across four computers. It would be advantageous, however, to use PVM to distribute the final two beam orientations over four computers
Reactor physics verification of the MCNP6 unstructured mesh capability
International Nuclear Information System (INIS)
Burke, T. P.; Kiedrowski, B. C.; Martz, R. L.; Martin, W. R.
2013-01-01
The Monte Carlo software package MCNP6 has the ability to transport particles on unstructured meshes generated from the Computed-Aided Engineering software Abaqus. Verification is performed using benchmarks with features relevant to reactor physics - Big Ten and the C5G7 computational benchmark. Various meshing strategies are tested and results are compared to reference solutions. Computational performance results are also given. The conclusions show MCNP6 is capable of producing accurate calculations for reactor physics geometries and the computational requirements for small lattice benchmarks are reasonable on modern computing platforms. (authors)
An Electron/Photon/Relaxation Data Library for MCNP6
Energy Technology Data Exchange (ETDEWEB)
Hughes, III, H. Grady [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2015-08-07
The capabilities of the MCNP6 Monte Carlo code in simulation of electron transport, photon transport, and atomic relaxation have recently been significantly expanded. The enhancements include not only the extension of existing data and methods to lower energies, but also the introduction of new categories of data and methods. Support of these new capabilities has required major additions to and redesign of the associated data tables. In this paper we present the first complete documentation of the contents and format of the new electron-photon-relaxation data library now available with the initial production release of MCNP6.
Accelerating Pseudo-Random Number Generator for MCNP on GPU
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.
Simplification of an MCNP model designed for dose rate estimation
Laptev, Alexander; Perry, Robert
2017-09-01
A study was made to investigate the methods of building a simplified MCNP model for radiological dose estimation. The research was done using an example of a complicated glovebox with extra shielding. The paper presents several different calculations for neutron and photon dose evaluations where glovebox elements were consecutively excluded from the MCNP model. The analysis indicated that to obtain a fast and reasonable estimation of dose, the model should be realistic in details that are close to the tally. Other details may be omitted.
Simplification of an MCNP model designed for dose rate estimation
Directory of Open Access Journals (Sweden)
Laptev Alexander
2017-01-01
Full Text Available A study was made to investigate the methods of building a simplified MCNP model for radiological dose estimation. The research was done using an example of a complicated glovebox with extra shielding. The paper presents several different calculations for neutron and photon dose evaluations where glovebox elements were consecutively excluded from the MCNP model. The analysis indicated that to obtain a fast and reasonable estimation of dose, the model should be realistic in details that are close to the tally. Other details may be omitted.
Reactor physics verification of the MCNP6 unstructured mesh capability
Energy Technology Data Exchange (ETDEWEB)
Burke, T. P. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Boulevard, Ann Arbor, MI 48109 (United States); Kiedrowski, B. C.; Martz, R. L. [X-Computational Physics Division, Monte Carlo Codes Group, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Martin, W. R. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Boulevard, Ann Arbor, MI 48109 (United States)
2013-07-01
The Monte Carlo software package MCNP6 has the ability to transport particles on unstructured meshes generated from the Computed-Aided Engineering software Abaqus. Verification is performed using benchmarks with features relevant to reactor physics - Big Ten and the C5G7 computational benchmark. Various meshing strategies are tested and results are compared to reference solutions. Computational performance results are also given. The conclusions show MCNP6 is capable of producing accurate calculations for reactor physics geometries and the computational requirements for small lattice benchmarks are reasonable on modern computing platforms. (authors)
Estimation and interpretation of keff confidence intervals in MCNP
International Nuclear Information System (INIS)
Urbatsch, T.J.
1995-11-01
MCNP's criticality methodology and some basic statistics are reviewed. Confidence intervals are discussed, as well as how to build them and their importance in the presentation of a Monte Carlo result. The combination of MCNP's three k eff estimators is shown, theoretically and empirically, by statistical studies and examples, to be the best k eff estimator. The method of combining estimators is based on a solid theoretical foundation, namely, the Gauss-Markov Theorem in regard to the least squares method. The confidence intervals of the combined estimator are also shown to have correct coverage rates for the examples considered
Impact of MCNP unresolved resonance probability-table treatment on uranium and plutonium benchmarks
International Nuclear Information System (INIS)
Mosteller, R.D.; Little, R.C.
1998-01-01
Versions of MCNP up through and including 4B have not accurately modeled neutron self-shielding effects in the unresolved resonance energy region. Recently, a probability-table treatment has been incorporated into a developmental version of MCNP. This paper presents MCNP results for a variety of uranium and plutonium critical benchmarks, calculated with and without the probability-table treatment
Gas Core Reactor Numerical Simulation Using a Coupled MHD-MCNP Model
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.
International Nuclear Information System (INIS)
Wang Xinghua; Zhou Sichun; Zhang Qingxian; Zhao Feng; Liu Jun; Zhu Jian
2013-01-01
Taking Qinshan nuclear power plant as an example, in this paper, Monte Carlo method was used in the MCNP procedures for the establishment of nuclear power station simulation model, construct the reactor pressure vessel and vessel core component composition and arrangement, KCODE card was used to calculate the effect of the number and the location of burnable poison control rod factor K eff by the boron acid. The calculation results show that, with the increasing in the number of burnable poison control rod value-added factor K eff shown a downward trend, and with the burnable poison control rod from the dense to sparse, which K eff will be decreasing slowly. This condition is consistent with the theoretical. (authors)
Parallel MCNP Monte Carlo transport calculations with MPI
International Nuclear Information System (INIS)
Wagner, J.C.; Haghighat, A.
1996-01-01
The steady increase in computational performance has made Monte Carlo calculations for large/complex systems possible. However, in order to make these calculations practical, order of magnitude increases in performance are necessary. The Monte Carlo method is inherently parallel (particles are simulated independently) and thus has the potential for near-linear speedup with respect to the number of processors. Further, the ever-increasing accessibility of parallel computers, such as workstation clusters, facilitates the practical use of parallel Monte Carlo. Recognizing the nature of the Monte Carlo method and the trends in available computing, the code developers at Los Alamos National Laboratory implemented the message-passing general-purpose Monte Carlo radiation transport code MCNP (version 4A). The PVM package was chosen by the MCNP code developers because it supports a variety of communication networks, several UNIX platforms, and heterogeneous computer systems. This PVM version of MCNP has been shown to produce speedups that approach the number of processors and thus, is a very useful tool for transport analysis. Due to software incompatibilities on the local IBM SP2, PVM has not been available, and thus it is not possible to take advantage of this useful tool. Hence, it became necessary to implement an alternative message-passing library package into MCNP. Because the message-passing interface (MPI) is supported on the local system, takes advantage of the high-speed communication switches in the SP2, and is considered to be the emerging standard, it was selected
Duplicating MC-15 Output with Python and MCNP
Energy Technology Data Exchange (ETDEWEB)
McSpaden, Alexander Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2017-08-23
Two Python scripts have been written that process the output files of MCNP6 into a format that mimics the list-mode output of Los Alamos National Laboratory’s MC-15 and NPOD neutron detection systems. This report details the methods implemented in these scripts and instructions on their use.
Performance of scientific computing platforms with MCNP4B
International Nuclear Information System (INIS)
McLaughlin, H.E.; Hendricks, J.S.
1998-01-01
Several computing platforms were evaluated with the MCNP4B Monte Carlo radiation transport code. The DEC AlphaStation 500/500 was the fastest to run MCNP4B. Compared to the HP 9000-735, the fastest platform 4 yr ago, the AlphaStation is 335% faster, the HP C180 is 133% faster, the SGI Origin 2000 is 82% faster, the Cray T94/4128 is 1% faster, the IBM RS/6000-590 is 93% as fast, the DEC 3000/600 is 81% as fast, the Sun Sparc20 is 57% as fast, the Cray YMP 8/8128 is 57% as fast, the sun Sparc5 is 33% as fast, and the Sun Sparc2 is 13% as fast. All results presented are reproducible and allow for comparison to computer platforms not included in this study. Timing studies are seen to be very problem dependent. The performance gains resulting from advances in software were also investigated. Various compilers and operating systems were seen to have a modest impact on performance, whereas hardware improvements have resulted in a factor of 4 improvement. MCNP4B also ran approximately as fast as MCNP4A
Thermal lattice benchmarks for testing basic evaluated data files, developed with MCNP4B
International Nuclear Information System (INIS)
Maucec, M.; Glumac, B.
1996-01-01
The development of unit cell and full reactor core models of DIMPLE S01A and TRX-1 and TRX-2 benchmark experiments, using Monte Carlo computer code MCNP4B is presented. Nuclear data from ENDF/B-V and VI version of cross-section library were used in the calculations. In addition, a comparison to results obtained with the similar models and cross-section data from the EJ2-MCNPlib library (which is based upon the JEF-2.2 evaluation) developed in IRC Petten, Netherlands is presented. The results of the criticality calculation with ENDF/B-VI data library, and a comparison to results obtained using JEF-2.2 evaluation, confirm the MCNP4B full core model of a DIMPLE reactor as a good benchmark for testing basic evaluated data files. On the other hand, the criticality calculations results obtained using the TRX full core models show less agreement with experiment. It is obvious that without additional data about the TRX geometry, our TRX models are not suitable as Monte Carlo benchmarks. (author)
Physics and Algorithm Enhancements for a Validated MCNP/X Monte Carlo Simulation Tool, Phase VII
International Nuclear Information System (INIS)
McKinney, Gregg W.
2012-01-01
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.
Use of McCad for the conversion of ITER CAD data to MCNP geometry
International Nuclear Information System (INIS)
Tsige-Tamirat, H.; Fischer, U.; Serikov, A.; Stickel, S.
2008-01-01
The program McCad provides a CAD interface for the Monte Carlo transport code MCNP. It is able to convert CAD data into MCNP input geometry description and provides GUI components for modeling, visualization, and data exchange. It performs sequences of tests on CAD data to check its validity and neutronics appropriateness including completion of the final MCNP model by void geometries. McCad has been used to convert a 40 deg. ITER torus sector CAD model to a suitable MCNP geometry model. Results of MCNP calculations performed to validate the converted geometry are presented
International Nuclear Information System (INIS)
Cintra, Felipe Belonsi de
2010-01-01
This study made a comparison between some of the major transport codes that employ the Monte Carlo stochastic approach in dosimetric calculations in nuclear medicine. We analyzed in detail the various physical and numerical models used by MCNP5 code in relation with codes like EGS and Penelope. The identification of its potential and limitations for solving microdosimetry problems were highlighted. The condensed history methodology used by MCNP resulted in lower values for energy deposition calculation. This showed a known feature of the condensed stories: its underestimates both the number of collisions along the trajectory of the electron and the number of secondary particles created. The use of transport codes like MCNP and Penelope for micrometer scales received special attention in this work. Class I and class II codes were studied and their main resources were exploited in order to transport electrons, which have particular importance in dosimetry. It is expected that the evaluation of available methodologies mentioned here contribute to a better understanding of the behavior of these codes, especially for this class of problems, common in microdosimetry. (author)
Criticality Benchmark Results Using Various MCNP Data Libraries
International Nuclear Information System (INIS)
Frankle, Stephanie C.
1999-01-01
A suite of 86 criticality benchmarks has been recently implemented in MCNPtrademark 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, 235,238 U, 237 Np, and 239,240 Pu. When examining the results of these calculations for the five manor categories of 233 U, intermediate-enriched 235 U (IEU), highly enriched 235 U (HEU), 239 Pu, and mixed metal assembles, we find the following: (1) The new evaluations for 9 Be, 12 C, and 14 N show no net effect on k eff ; (2) There is a consistent decrease in k eff for all of the solution assemblies for ENDF/B-VI due to 1 H and 16 O, moving k eff further from the benchmark value for uranium solutions and closer to the benchmark value for plutonium solutions; (3) k eff decreased for the ENDF/B-VI Fe isotopic data, moving the calculated k eff further from the benchmark value; (4) k eff decreased for the ENDF/B-VI Ni isotopic data, moving the calculated k 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 235 U tends to decrease k eff while the 238 U data tends to increase k 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 235,238 U evaluations tend to increase k eff . For the mixed graphite and normal uranium-reflected assembly, a large increase in k eff due to changes in the 238 U evaluation moved the calculated k eff much closer to the benchmark value. (8) There is little change in k eff for the uranium solutions due to the new 235,238 U evaluations; and (9) There is little change in k eff
Benchmarking ENDF/B-VII.1, JENDL-4.0 and JEFF-3.1.1 with MCNP6
International Nuclear Information System (INIS)
Marck, Steven C. van der
2012-01-01
Recent releases of three major world nuclear reaction data libraries, ENDF/B-VII.1, JENDL-4.0, and JEFF-3.1.1, have been tested extensively using benchmark calculations. The calculations were performed with the latest release of the continuous energy Monte Carlo neutronics code MCNP, i.e. MCNP6. Three types of benchmarks were used, viz. criticality safety benchmarks, (fusion) shielding benchmarks, and reference systems for which the effective delayed neutron fraction is reported. For criticality safety, more than 2000 benchmarks from the International Handbook of Criticality Safety Benchmark Experiments were used. Benchmarks from all categories were used, ranging from low-enriched uranium, compound fuel, thermal spectrum ones (LEU-COMP-THERM), to mixed uranium-plutonium, metallic fuel, fast spectrum ones (MIX-MET-FAST). For fusion shielding many benchmarks were based on IAEA specifications for the Oktavian experiments (for Al, Co, Cr, Cu, LiF, Mn, Mo, Si, Ti, W, Zr), Fusion Neutronics Source in Japan (for Be, C, N, O, Fe, Pb), and Pulsed Sphere experiments at Lawrence Livermore National Laboratory (for 6 Li, 7 Li, Be, C, N, O, Mg, Al, Ti, Fe, Pb, D2O, H2O, concrete, polyethylene and teflon). The new functionality in MCNP6 to calculate the effective delayed neutron fraction was tested by comparison with more than thirty measurements in widely varying systems. Among these were measurements in the Tank Critical Assembly (TCA in Japan) and IPEN/MB-01 (Brazil), both with a thermal spectrum, two cores in Masurca (France) and three cores in the Fast Critical Assembly (FCA, Japan), all with fast spectra. The performance of the three libraries, in combination with MCNP6, is shown to be good. The results for the LEU-COMP-THERM category are on average very close to the benchmark value. Also for most other categories the results are satisfactory. Deviations from the benchmark values do occur in certain benchmark series, or in isolated cases within benchmark series. Such
An MCNP parametric study of George C. Laurence's subcritical pile experiment
Energy Technology Data Exchange (ETDEWEB)
Dranga, R.; Blomeley, L., E-mail: ruxandra.dranga@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carrington, R. [McGill Univ., Dept. of Mathematics and Statistics, Montreal, Quebec (Canada)
2014-12-01
In the early 1940s at the National Research Council (NRC) Laboratories in Ottawa, Canada, Dr. George Laurence conducted several experiments to determine if a sustained nuclear fission chain reaction in a carbon-uranium arrangement (or 'pile') was possible. Although Dr. Laurence did not achieve criticality, these pioneering experiments marked a significant historical event in nuclear science, and they provided a valuable reference for subsequent experiments that led to the design of Canada's first heavy-water reactors at the Chalk River Nuclear Laboratories. This paper summarizes the results of a recent collaborative project between Atomic Energy of Canada Limited and the Deep River Science Academy undertaken to numerically explore the experiments carried out at the NRC Laboratories by Dr. Laurence, while teaching high school students about nuclear science and technology. In this study, a modern Monte Carlo reactor physics code, MCNP6, was utilized to identify and study the key parameters impacting the subcritical pile's neutron multiplication factor (e.g., moderation, geometry, material impurities) and quantify their effect on the extent of subcriticality. The findings presented constitute the first endeavour to model, using a current computational reactor physics tool, the seminal experiment that provided the foundation of Canada's nuclear science and technology program. (author)
An MCNP parametric study of George C. Laurence's subcritical pile experiment
International Nuclear Information System (INIS)
Dranga, R.; Blomeley, L.; Carrington, R.
2014-01-01
In the early 1940s at the National Research Council (NRC) Laboratories in Ottawa, Canada, Dr. George Laurence conducted several experiments to determine if a sustained nuclear fission chain reaction in a carbon-uranium arrangement (or 'pile') was possible. Although Dr. Laurence did not achieve criticality, these pioneering experiments marked a significant historical event in nuclear science, and they provided a valuable reference for subsequent experiments that led to the design of Canada's first heavy-water reactors at the Chalk River Nuclear Laboratories. This paper summarizes the results of a recent collaborative project between Atomic Energy of Canada Limited and the Deep River Science Academy undertaken to numerically explore the experiments carried out at the NRC Laboratories by Dr. Laurence, while teaching high school students about nuclear science and technology. In this study, a modern Monte Carlo reactor physics code, MCNP6, was utilized to identify and study the key parameters impacting the subcritical pile's neutron multiplication factor (e.g., moderation, geometry, material impurities) and quantify their effect on the extent of subcriticality. The findings presented constitute the first endeavour to model, using a current computational reactor physics tool, the seminal experiment that provided the foundation of Canada's nuclear science and technology program. (author)
Monte Carlo simulation on nuclear energy study. Annual report of Nuclear Code Evaluation Committee
International Nuclear Information System (INIS)
Sakurai, Kiyoshi; Yamamoto, Toshihiro
1999-03-01
In this report, research results discussed in 1998 fiscal year at Nuclear Code Evaluation Special Committee of Nuclear Code Committee were summarised. Present status of Monte Carlo calculation in high energy region investigated / discussed at Monte Carlo simulation working-group and automatic compilation system for MCNP cross sections developed at MCNP high temperature library compilation working-group were described. The 6 papers are indexed individually. (J.P.N.)
Introduction to the simulation with MCNP Monte Carlo code and its applications in Medical Physics
International Nuclear Information System (INIS)
Parreno Z, F.; Paucar J, R.; Picon C, C.
1998-01-01
The simulation by Monte Carlo is tool which Medical Physics counts with it for the development of its research, the interest by this tool is growing, as we may observe in the main scientific journals for the years 1995-1997 where more than 27 % of the papers treat over Monte Carlo and/or its applications in the radiation transport.In the Peruvian Institute of Nuclear Energy we are implementing and making use of the MCNP4 and EGS4 codes. In this work are presented the general features of the Monte Carlo method and its more useful applications in Medical Physics. Likewise, it is made a simulation of the calculation of isodose curves in an interstitial treatment with Ir-192 wires in a mammary gland carcinoma. (Author)
Multi-canister overpack project - verification and validation, MCNP 4A
International Nuclear Information System (INIS)
Goldmann, L.H.
1997-01-01
This supporting document contains the software verification and validation (V and V) package used for Phase 2 design of the Spent Nuclear Fuel Multi-Canister Overpack. V and V packages for both ANSYS and MCNP are included. Description of Verification Run(s): This software requires that it be compiled specifically for the machine it is to be used on. Therefore to facilitate ease in the verification process the software automatically runs 25 sample problems to ensure proper installation and compilation. Once the runs are completed the software checks for verification by performing a file comparison on the new output file and the old output file. Any differences between any of the files will cause a verification error. Due to the manner in which the verification is completed a verification error does not necessarily indicate a problem. This indicates that a closer look at the output files is needed to determine the cause of the error
International Nuclear Information System (INIS)
Bakkari, B El; Bardouni, T El.; Erradi, L.; Chakir, E.; Meroun, O.; Azahra, M.; Boukhal, H.; Khoukhi, T El.; Htet, A.
2007-01-01
Full text: New releases of nuclear data files made available during the few recent years. The reference MCNP5 code (1) for Monte Carlo calculations is usually distributed with only one standard nuclear data library for neutron interactions based on ENDF/B-VI. The main goal of this work is to process new neutron cross sections libraries in ACE continuous format for MCNP code based on the most recent data files recently made available for the scientific community : ENDF/B-VII.b2, ENDF/B-VI (release 8), JEFF3.0, JEFF-3.1, JENDL-3.3 and JEF2.2. In our data treatment, we used the modular NJOY system (release 99.9) (2) in conjunction with its most recent upadates. Assessment of the processed point wise cross sections libraries performances was made by means of some criticality prediction and analysis of other integral parameters for a set of reactor benchmarks. Almost all the analyzed benchmarks were taken from the international handbook of Evaluated criticality safety benchmarks experiments from OECD (3). Some revised benchmarks were taken from references (4,5). These benchmarks use Pu-239 or U-235 as the main fissionable materiel in different forms, different enrichments and cover various geometries. Monte Carlo calculations were performed in 3D with maximum details of benchmark description and the S(α,β) cross section treatment was adopted in all thermal cases. The resulting one standard deviation confidence interval for the eigenvalue is typically +/-13% to +/-20 pcm [fr
Electron/Photon Verification Calculations Using MCNP4B
Energy Technology Data Exchange (ETDEWEB)
D. P. Gierga; K. J. Adams
1999-04-01
MCNP4BW was released in February 1997 with significant enhancements to electron/photon transport methods. These enhancements have been verified against a wide range of published electron/photon experiments, spanning high energy bremsstrahlung production to electron transmission and reflection. The impact of several MCNP tally options and physics parameters was explored in detail. The agreement between experiment and simulation was usually within two standard deviations of the experimental and calculational errors. Furthermore, sub-step artifacts for bremsstrahlung production were shown to be mitigated. A detailed suite of electron depth dose calculations in water is also presented. Areas for future code development have also been explored and include the dependence of cell and detector tallies on different bremsstrahlung angular models and alternative variance reduction splitting schemes for bremsstrahlung production.
Convergence testing for MCNP5 Monte Carlo eigenvalue calculations
International Nuclear Information System (INIS)
Brown, F.; Nease, B.; Cheatham, J.
2007-01-01
Determining convergence of Monte Carlo criticality problems is complicated by the statistical noise inherent in the random, walks of the neutrons in each generation. The latest version of MCNP5 incorporates an important new tool for assessing convergence: the Shannon entropy of the fission source distribution, H src . Shannon entropy is a well-known concept from information theory and provides a single number for each iteration to help characterize convergence trends for the fission source distribution. MCNP5 computes H src for each iteration, and these values may be plotted to examine convergence trends. Convergence testing should include both k eff and H src , since the fission distribution will converge more slowly than k eff , especially when the dominance ratio is close to 1.0. (authors)
A DRAGON-MCNP comparison of void reactivity calculations
Energy Technology Data Exchange (ETDEWEB)
Marleau, G [Ecole Polytechnique, Montreal, PQ (Canada). Inst. de Genie Nucleaire; Milgram, M S [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)
1996-12-31
The determination of the reactivity coefficients associated with coolant voiding in a CANDU reactor is a subject which has attracted a large amount of interest in the last few years both from the theoretical and experimental point of view. One expects that deterministic codes such as DRAGON and WIMS-AECL or the MCNP4 Monte Carlo code should be able to adequately simulate the cell behaviour upon coolant voiding. However, the absence of an experimental database at equilibrium and discharge burnups has not permitted the full validation of any of these lattice codes, although a partial validation through comparison of two different computer codes has been considered. Here we present a comparison between DRAGON and MCNP4 of the void reactivity evaluation for fresh fuel. (author). 16 refs., 5 tabs.
A Monte Carlo burnup code linking MCNP and REBUS
International Nuclear Information System (INIS)
Hanan, N.A.; Olson, A.P.; Pond, R.B.; Matos, J.E.
1998-01-01
The REBUS-3 burnup code, used in the anl RERTR Program, is a very general code that uses diffusion theory (DIF3D) to obtain the fluxes required for reactor burnup analyses. Diffusion theory works well for most reactors. However, to include the effects of exact geometry and strong absorbers that are difficult to model using diffusion theory, a Monte Carlo method is required. MCNP, a general-purpose, generalized-geometry, time-dependent, Monte Carlo transport code, is the most widely used Monte Carlo code. This paper presents a linking of the MCNP code and the REBUS burnup code to perform these difficult analyses. The linked code will permit the use of the full capabilities of REBUS which include non-equilibrium and equilibrium burnup analyses. Results of burnup analyses using this new linked code are also presented. (author)
A DRAGON-MCNP comparison of void reactivity calculations
International Nuclear Information System (INIS)
Marleau, G.
1995-01-01
The determination of the reactivity coefficients associated with coolant voiding in a CANDU reactor is a subject which has attracted a large amount of interest in the last few years both from the theoretical and experimental point of view. One expects that deterministic codes such as DRAGON and WIMS-AECL or the MCNP4 Monte Carlo code should be able to adequately simulate the cell behaviour upon coolant voiding. However, the absence of an experimental database at equilibrium and discharge burnups has not permitted the full validation of any of these lattice codes, although a partial validation through comparison of two different computer codes has been considered. Here we present a comparison between DRAGON and MCNP4 of the void reactivity evaluation for fresh fuel. (author). 16 refs., 5 tabs
A Monte Carlo burnup code linking MCNP and REBUS
International Nuclear Information System (INIS)
Hanan, N. A.
1998-01-01
The REBUS-3 burnup code, used in the ANL RERTR Program, is a very general code that uses diffusion theory (DIF3D) to obtain the fluxes required for reactor burnup analyses. Diffusion theory works well for most reactors. However, to include the effects of exact geometry and strong absorbers that are difficult to model using diffusion theory, a Monte Carlo method is required. MCNP, a general-purpose, generalized-geometry, time-dependent, Monte Carlo transport code, is the most widely used Monte Carlo code. This paper presents a linking of the MCNP code and the REBUS burnup code to perform these difficult burnup analyses. The linked code will permit the use of the full capabilities of REBUS which include non-equilibrium and equilibrium burnup analyses. Results of burnup analyses using this new linked code are also presented
A fast, automated, semideterministic weight windows generator for MCNP
International Nuclear Information System (INIS)
Mickael, M.W.
1995-01-01
A fast automated method is developed to estimate particle importance in the Los Alamos Carlo code MCNP. It provides an automated and efficient way of predicting and setting up an important map for the weight windows technique. A short analog simulation is first performed to obtain effective group parameters based on the input description of the problem. A solution of the multigroup time-dependent adjoint diffusion equation is then used to estimate particle importance. At any point in space, time, and energy, the particle importance is determined, based on the calculated parameters, and used as the lower limit of the weight window. The method has been tested for neutron, photon, and coupled neutron-photon problems. Significant improvement in the simulation efficiency is obtained using this technique at no additional computer time and with no prior knowledge of the nature of the problem. Moreover, time and angular importance that are not available yet in MCNP are easily implemented in this method
MCNP/X TRANSPORT IN THE TABULAR REGIME
Energy Technology Data Exchange (ETDEWEB)
HUGHES, H. GRADY [Los Alamos National Laboratory
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.
A photoneutron production option for MCNP4A
International Nuclear Information System (INIS)
Gallmeier, F.X.
1996-01-01
A photoneutron production option was implemented in the MCNP4A code, mainly to supply a tool for reactor shielding calculations in beryllium and heavy water environments of complicated three dimensional geometries. Subroutines were developed to calculate the probability of the photoneutron production at the photon collision sites and the energy and flight direction of the created photoneutrons with the help of user supplied data. These subroutines are accessed through subroutine colidp which processes the photon collisions
Characteristics of multiprocessing MCNP5 on small personal computer clusters
International Nuclear Information System (INIS)
Robinson, S M; Mc Conn, R J Jr; Pagh, R T; Schweppe, J E; Siciliano, E R
2006-01-01
The feasibility and efficiency of performing MCNP5 calculations with a small, heterogeneous computing cluster built from Microsoft ( R) Windows TM personal computers (PC) are explored. The performance increases that may be expected with such clusters are estimated for cases that typify general radiation-shielding calculations. Our results show that the speed increase from additional slave PCs is nearly linear up to 10 processors. Guidance is given as to the specific advantages of changing various parameters present in the system. Implementing load balancing, and reducing the overhead from the MCNP rendezvous mechanism add to heterogeneous cluster efficiency. Hyper-threading technology and matching the total number of slave processes to the total number of logical processors also yield modest speed increases in the range below 7 processors. Because of the ease of acquisition of heterogeneous desktop computers, and the peak in efficiency at the level of a few physical processors, a strong case is made for the use of small clusters as a tool for producing MCNP5 calculations rapidly, and detailed instructions for constructing such clusters are provided
MCNP6 fragmentation of light nuclei at intermediate energies
Energy Technology Data Exchange (ETDEWEB)
Mashnik, Stepan G., E-mail: mashnik@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Kerby, Leslie M. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); University of Idaho, Moscow, ID 83844 (United States)
2014-11-11
Fragmentation reactions induced on light target nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below are studied with the latest Los Alamos Monte Carlo transport code MCNP6 and with its cascade-exciton model (CEM) and Los Alamos version of the quark-gluon string model (LAQGSM) event generators, version 03.03, used as stand-alone codes. Such reactions are involved in different applications, like cosmic-ray-induced single event upsets (SEU's), radiation protection, and cancer therapy with proton and ion beams, among others; therefore, it is important that MCNP6 simulates them as well as possible. CEM and LAQGSM assume that intermediate-energy fragmentation reactions on light nuclei occur generally in two stages. The first stage is the intranuclear cascade (INC), followed by the second, Fermi breakup disintegration of light excited residual nuclei produced after the INC. Both CEM and LAQGSM account also for coalescence of light fragments (complex particles) up to {sup 4}He from energetic nucleons emitted during INC. We investigate the validity and performance of MCNP6, CEM, and LAQGSM in simulating fragmentation reactions at intermediate energies and discuss possible ways of further improving these codes.
Parallelization of MCNP4 code by using simple FORTRAN algorithms
International Nuclear Information System (INIS)
Yazid, P.I.; Takano, Makoto; Masukawa, Fumihiro; Naito, Yoshitaka.
1993-12-01
Simple FORTRAN algorithms, that rely only on open, close, read and write statements, together with disk files and some UNIX commands have been applied to parallelization of MCNP4. The code, named MCNPNFS, maintains almost all capabilities of MCNP4 in solving shielding problems. It is able to perform parallel computing on a set of any UNIX workstations connected by a network, regardless of the heterogeneity in hardware system, provided that all processors produce a binary file in the same format. Further, it is confirmed that MCNPNFS can be executed also on Monte-4 vector-parallel computer. MCNPNFS has been tested intensively by executing 5 photon-neutron benchmark problems, a spent fuel cask problem and 17 sample problems included in the original code package of MCNP4. Three different workstations, connected by a network, have been used to execute MCNPNFS in parallel. By measuring CPU time, the parallel efficiency is determined to be 58% to 99% and 86% in average. On Monte-4, MCNPNFS has been executed using 4 processors concurrently and has achieved the parallel efficiency of 79% in average. (author)
LEU-fueled SLOWPOKE-2 modelling with MCNP4A
International Nuclear Information System (INIS)
Pierre, J.R.M.; Bonin, H.W.J.
1996-01-01
Following the commissioning of the Low Enrichment Uranium (LEU) Fueled SLOWPOKE-2 research reactor at Royal Military College,excess reactivity measurements were conducted over a range of temperature and power. Given the advance in computer technology, the use of Monte Carlo N-Particle Transport Code System MCNP 4A appeared possible for the simulation of the LEU-fueled SLOWPOKE-2 reactor core, and this work demonstrates that this is indeed the case. MCNP 4A is a full three dimensional program allowing the user to enter a large amount of complexity. The limit on the geometry complexity is the computing time required to achieve a reasonable standard deviation. To this point several models of the SLOWPOKE-2 have been developed giving some insight on the sensitivity of the code. MCNP4A can use various cross section libraries. The aim of this work is to calculate accurately the reactivity of the core and reproduce The temperature trend of the reactivity. The model preserved as much as possible the details of the core and facility in order to allow further study in the flux mapping
International Nuclear Information System (INIS)
Youssef, M.Z.; Feder, R.; Davis, I.
2007-01-01
The ITER IT has adopted the newly developed FEM, 3-D, and CAD-based Discrete Ordinates code, ATTILA for the neutronics studies contingent on its success in predicting key neutronics parameters and nuclear field according to the stringent QA requirements set forth by the Management and Quality Program (MQP). ATTILA has the advantage of providing a full flux and response functions mapping everywhere in one run where components subjected to excessive radiation level and strong streaming paths can be identified. The ITER neutronics community had agreed to use a standard CAD model of ITER (40 degree sector, denoted ''Benchmark CAD Model'') to compare results for several responses selected for calculation benchmarking purposes to test the efficiency and accuracy of the CAD-MCNP approach developed by each party. Since ATTILA seems to lend itself as a powerful design tool with minimal turnaround time, it was decided to benchmark this model with ATTILA as well and compare the results to those obtained with the CAD MCNP calculations. In this paper we report such comparison for five responses, namely: (1) Neutron wall load on the surface of the 18 shield blanket module (SBM), (2) Neutron flux and nuclear heating rate in the divertor cassette, (3) nuclear heating rate in the winding pack of the inner leg of the TF coil, (4) Radial flux profile across dummy port plug and shield plug placed in the equatorial port, and (5) Flux at seven point locations situated behind the equatorial port plug. (orig.)
Importance sampling techniques and treatment of electron transport in MCNP 4A
International Nuclear Information System (INIS)
Ueki, K.
1994-01-01
The continuous energy Monte Carlo code MCNP was developed by the Radiation Transport Group at Los Alamos National Laboratory and the MCNP 4A version is available, now. The MCNP 4A is able to do the coupled neutron-secondary gamma-ray-electron-bremsstrahlung calculation. The calculated results, such as energy spectra, tally fluctuation chart, and geometrical input data can be displayed by using a work station. The document of the MCNP 4A code has no description on the subroutines, except few ones of 'SOURCE', 'TALLYX'. However, when we want to improve the MCNP Monte Carlo sampling techniques to get more accuracy or efficiency results for some problems, some subroutines are required or needed to revised. Three subroutines have been revised and built in the MCNP 4A code. (author)
Development and application of MCNP auto-modeling tool: Mcam 3.0
International Nuclear Information System (INIS)
Liu Xiaoping; Luo Yuetong; Tong Lili
2005-01-01
Mcam is abbreviation of 'MCNP Automatic Modeling', which is a CAD interface program of MCNP geometry model based on CAD technology. Making use of existing CAD technology is Mcam's major characteristic. In rough, CAD technology is utilized in the following two ways: (1) Mcam makes it possible to create MCNP geometry model in some CAD software; (2) accelerate creation of MCNP geometry model by inheriting some existing 3D CAD model. The paper gives an introduction of Mcam's major ability: (1) ability to convert CAD model into MCNP geometry model; (2) ability to convert MCNP geometry model into CAD model; (3) ability to construct CAD model. At the end of the paper, several models are given to demonstrate Mcam's different ability respectively
Systems guide to MCNP (Monte Carlo Neutron and Photon Transport Code)
International Nuclear Information System (INIS)
Kirk, B.L.; West, J.T.
1984-06-01
The subject of this report is the implementation of the Los Alamos National Laboratory Monte Carlo Neutron and Photon Transport Code - Version 3 (MCNP) on the different types of computer systems, especially the IBM MVS system. The report supplements the documentation of the RSIC computer code package CCC-200/MCNP. Details of the procedure to follow in executing MCNP on the IBM computers, either in batch mode or interactive mode, are provided
MCNP6 Fission Cross Section Calculations at Intermediate and High Energies
Mashnik, Stepan G.; Sierk, Arnold J.; Prael, Richard E.
2013-01-01
MCNP6 has been Validated and Verified (V&V) against intermediate- and high-energy fission cross-section experimental data. An error in the calculation of fission cross sections of 181Ta and a few nearby target nuclei by the CEM03.03 event generator in MCNP6 and a "bug: in the calculation of fission cross sections with the GENXS option of MCNP6 while using the LAQGSM03.03 event generator were detected during our V&V work. After fixing both problems, we find that MCNP6 using CEM03.03 and LAQGSM...
International Nuclear Information System (INIS)
Mirzakhanian, L; Enger, S; Giusti, V
2015-01-01
Purpose: A major concern in proton therapy is the production of secondary neutrons causing secondary cancers, especially in young adults and children. Most utilized Monte Carlo codes in proton therapy are Geant4 and MCNP. However, the default versions of Geant4 and MCNP6 do not have suitable cross sections or physical models to properly handle secondary particle production in proton energy ranges used for therapy. In this study, default versions of Geant4 and MCNP6 were modified to better handle production of secondaries by adding the TENDL-2012 cross-section library. Methods: In-water proton depth-dose was measured at the “The Svedberg Laboratory” in Uppsala (Sweden). The proton beam was mono-energetic with mean energy of 178.25±0.2 MeV. The measurement set-up was simulated by Geant4 version 10.00 (default and modified version) and MCNP6. Proton depth-dose, primary and secondary particle fluence and neutron equivalent dose were calculated. In case of Geant4, the secondary particle fluence was filtered by all the physics processes to identify the main process responsible for the difference between the default and modified version. Results: The proton depth-dose curves and primary proton fluence show a good agreement between both Geant4 versions and MCNP6. With respect to the modified version, default Geant4 underestimates the production of secondary neutrons while overestimates that of gammas. The “ProtonInElastic” process was identified as the main responsible process for the difference between the two versions. MCNP6 shows higher neutron production and lower gamma production than both Geant4 versions. Conclusion: Despite the good agreement on the proton depth dose curve and primary proton fluence, there is a significant discrepancy on secondary neutron production between MCNP6 and both versions of Geant4. Further studies are thus in order to find the possible cause of this discrepancy or more accurate cross-sections/models to handle the nuclear
Reactor Simulations for Safeguards with the MCNP Utility for Reactor Evolution Code
International Nuclear Information System (INIS)
Shiba, T.; Fallot, M.
2015-01-01
To tackle nuclear material proliferation, we conducted several proliferation scenarios using the MURE (MCNP Utility for Reactor Evolution) code. The MURE code, developed by CNRS laboratories, is a precision, open-source code written in C++ that automates the preparation and computation of successive MCNP (Monte Carlo N-Particle) calculations and solves the Bateman equations in between, for burnup or thermal-hydraulics purposes. In addition, MURE has been completed recently with a module for the CHaracterization of Radioactive Sources, called CHARS, which computes the emitted gamma, beta and alpha rays associated to any fuel composition. Reactor simulations could allow knowing how plutonium or other material generation evolves inside reactors in terms of time and amount. The MURE code is appropriate for this purpose and can also provide knowledge on associated particle emissions. Using MURE, we have both developed a cell simulation of a typical CANDU reactor and a detailed model of light water PWR core, which could be used to analyze the composition of fuel assemblies as a function of time or burnup. MURE is also able to provide, thanks to its extension MURE-CHARTS, the emitted gamma rays from fuel assemblies unloaded from the core at any burnup. Diversion cases of Generation IV reactors have been also developed; a design of Very High Temperature Reactor (a Pebble Bed Reactor (PBR), loaded with UOx, PuOx and ThUOx fuels), and a Na-cooled Fast Breeder Reactor (FBR) (with depleted Uranium or Minor Actinides in the blanket). The loading of Protected Plutonium Production (P3) in the FBR was simulated. The simulations of various reactor designs taking into account reactor physics constraints may bring valuable information to inspectors. At this symposium, we propose to show the results of these reactor simulations as examples of the potentiality of reactor simulations for safeguards. (author)
MCNP-DSP, Monte Carlo Neutron-Particle Transport Code with Digital Signal Processing
International Nuclear Information System (INIS)
2002-01-01
1 - Description of program or function: MCNP-DSP is recommended only for experienced MCNP users working with subcritical measurements. It is a modification of the Los Alamos National Laboratory's Monte Carlo code MCNP4a that is used to simulate a variety of subcritical measurements. The DSP version was developed to simulate frequency analysis measurements, correlation (Rossi-) measurements, pulsed neutron measurements, Feynman variance measurements, and multiplicity measurements. CCC-700/MCNP4C is recommended for general purpose calculations. 2 - Methods:MCNP-DSP performs calculations very similarly to MCNP and uses the same generalized geometry capabilities of MCNP. MCNP-DSP can only be used with the continuous-energy cross-section data. A variety of source and detector options are available. However, unlike standard MCNP, the source and detector options are limited to those described in the manual because these options are specified in the MCNP-DSP extra data file. MCNP-DSP is used to obtain the time-dependent response of detectors that are modeled in the simulation geometry. The detectors represent actual detectors used in measurements. These time-dependent detector responses are used to compute a variety of quantities such as frequency analysis signatures, correlation signatures, multiplicity signatures, etc., between detectors or sources and detectors. Energy ranges are 0-60 MeV for neutrons (data generally only available up to 20 MeV) and 1 keV - 1 GeV for photons and electrons. 3 - Restrictions on the complexity of the problem: None noted
Nuclear Data Libraries for Hydrogen in Light Water Ice
International Nuclear Information System (INIS)
Torres, L; Gillette, V.H
2000-01-01
Nuclear data libraries were produced for hydrogen (H) in light water ice at different temperatures, 20, 30, 50, 77, 112, 180, 230 K.These libraries were produced using the NJOY nuclear data processing system.With this code we produce pointwise cross sections and related quantities, in the ENDF format, and in the ACE format for MCNP.Experimental neutron spectra at such temperatures were compared with MCNP4B simulations, based on the locally produced libraries, leading to satisfactory results
Radiation calculations using LAHET/MCNP/CINDER90
International Nuclear Information System (INIS)
Waters, L.
1994-01-01
The LAHET monte carlo code system has recently been expanded to include high energy hadronic interactions via the FLUKA code, while retaining the original Los Alamos versions of HETC and ISABEL at lower energies. Electrons and photons are transported with EGS4 or ITS, while the MCNP coupled neutron/photon monte carlo code provides analysis of neutrons with kinetic energies less than 20 MeV. An interface with the CINDER activation code is now in common use. Various other changes have been made to facilitate analysis of high energy accelerator radiation environments and experimental physics apparatus, such as those found at SSC and RHIC. Current code developments and applications are reviewed
Using MCNP code for neutron and photon skyshine analysis
Energy Technology Data Exchange (ETDEWEB)
Zharkov, V.P.; Dikareva, O.F.; Kartashev, I.A.; Kiselev, A.N.; Netecha, M.E. [Research and Development Inst. of Power Engineering, Moscow (Russian Federation); Nomura, Y.; Tsubosaka, A. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)
2000-03-01
The MCNP Monte-Carlo code was used for the investigation of the sensitivity of neutron and neutron-induced secondary photon dose rate, total and thermal neutron fluxes and space-energy distributions to energy and angular distribution of radiation source, to thickness and composition of the ground, air density (including it changing with height), humidities of air and ground, thermalization effects, detector's dimension and its disposal above the ground level. The calculations were performed with the assumption that the source or released radiation into the atmosphere can be treated as a point source and the source containment structure has a negligible perturbation on the skyshine radiation field. (author)
MCNP calculations for the HCPB submodules in-pile test
Energy Technology Data Exchange (ETDEWEB)
Pijlgroms, B.J. [Section Nuclear and Reactor Physics, ECN Nuclear Research, Petten (Netherlands)
1998-11-01
This report describes the MCNP calculations that have been performed for the Helium Cooled Pebble Bed (HCPB) Submodules In-pile Test that has been planned for irradiation in the materials testing High Flux Reactor (HFR) at Petten. In this test, four HSM-8 submodules will be placed at core position H4. The report presents the neutron flux and power density profiles to be expected in the submodules. For the gamma induced heating only a rough estimation could be made. In the HCPB submodules the total specific heating does not exceed (36.7 {+-} 2.9)[W/cc]. 8 refs.
Processing methods for temperature-dependent MCNP libraries
International Nuclear Information System (INIS)
Li Songyang; Wang Kan; Yu Ganglin
2008-01-01
In this paper,the processing method of NJOY which transfers ENDF files to ACE (A Compact ENDF) files (point-wise cross-Section file used for MCNP program) is discussed. Temperatures that cover the range for reactor design and operation are considered. Three benchmarks are used for testing the method: Jezebel Benchmark, 28 cm-thick Slab Core Benchmark and LWR Benchmark with Burnable Absorbers. The calculation results showed the precision of the neutron cross-section library and verified the correct processing methods in usage of NJOY. (authors)
Development of temperature related thermal neutron scattering database for MCNP
International Nuclear Information System (INIS)
Mei Longwei; Cai Xiangzhou; Jiang Dazhen; Chen Jingen; Guo Wei
2013-01-01
Based on ENDF/B-Ⅶ neutron library, the thermal neutron scattering library S(α, β) for molten salt reactor moderators was developed. The temperatures of this library were chose as the characteristic temperature of the molten salt reactor. The cross section of the thermal neutron scattering of ACE format was investigated, and this library was also validated by the benchmarks of ICSBEP. The uncertainties shown in the validation were in reasonable range when compared with the thermal neutron scattering library tmccs which included in the MCNP data library. It was proved that the thermal neutron scattering library processed in this study could be used in the molten salt reactor design. (authors)
Wielandt acceleration for MCNP5 Monte Carlo eigenvalue calculations
International Nuclear Information System (INIS)
Brown, F.
2007-01-01
Monte Carlo criticality calculations use the power iteration method to determine the eigenvalue (k eff ) and eigenfunction (fission source distribution) of the fundamental mode. A recently proposed method for accelerating convergence of the Monte Carlo power iteration using Wielandt's method has been implemented in a test version of MCNP5. The method is shown to provide dramatic improvements in convergence rates and to greatly reduce the possibility of false convergence assessment. The method is effective and efficient, improving the Monte Carlo figure-of-merit for many problems. In addition, the method should eliminate most of the underprediction bias in confidence intervals for Monte Carlo criticality calculations. (authors)
Effect of the MCNP model definition on the computation time
International Nuclear Information System (INIS)
Šunka, Michal
2017-01-01
The presented work studies the influence of the method of defining the geometry in the MCNP transport code and its impact on the computational time, including the difficulty of preparing an input file describing the given geometry. Cases using different geometric definitions including the use of basic 2-dimensional and 3-dimensional objects and theirs combinations were studied. The results indicate that an inappropriate definition can increase the computational time by up to 59% (a more realistic case indicates 37%) for the same results and the same statistical uncertainty. (orig.)
MCNP Techniques for Modeling Sodium Iodide Spectra of Kiwi Surveys
International Nuclear Information System (INIS)
Robert B Hayes
2007-01-01
This work demonstrates how MCNP can be used to predict the response of mobile search and survey equipment from base principles. The instrumentation evaluated comes from the U.S. Department of Energy's Aerial Measurement Systems. Through reconstructing detector responses to various point-source measurements, detector responses to distributed sources can be estimated through superposition. Use of this methodology for currently deployed systems allows predictive determinations of activity levels and distributions for common configurations of interest. This work helps determine the quality and efficacy of certain surveys in fully characterizing an effected site following a radiological event of national interest
International Nuclear Information System (INIS)
Hussein, M.S.; Bonin, H.W.; Lewis, B.J.
2014-01-01
The most recent versions of the Monte Carlo-based probabilistic transport code MCNP6 and the continuous energy reactor physics burnup calculation code Serpent allow for a 3-D geometry calculation accounting for the detailed geometry without unit-cell homogenization. These two codes are used to calculate the axial and radial flux and power distributions for a CANDU6 GENTILLY-2 nuclear reactor core with 37-element fuel bundles. The multiplication factor, actual flux distribution and power density distribution were calculated by using a tally combination for MCNP6 and detector analysis for Serpent. Excellent agreement was found in the calculated flux and power distribution. The Serpent code is most efficient in terms of the computational time. (author)
Energy Technology Data Exchange (ETDEWEB)
Hussein, M.S.; Bonin, H.W., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, ON (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, ON (Canada)
2014-07-01
The most recent versions of the Monte Carlo-based probabilistic transport code MCNP6 and the continuous energy reactor physics burnup calculation code Serpent allow for a 3-D geometry calculation accounting for the detailed geometry without unit-cell homogenization. These two codes are used to calculate the axial and radial flux and power distributions for a CANDU6 GENTILLY-2 nuclear reactor core with 37-element fuel bundles. The multiplication factor, actual flux distribution and power density distribution were calculated by using a tally combination for MCNP6 and detector analysis for Serpent. Excellent agreement was found in the calculated flux and power distribution. The Serpent code is most efficient in terms of the computational time. (author)
International Nuclear Information System (INIS)
Luneville, L.; Chiron, M.; Toubon, H.; Dogny, S.; Huver, M.; Berger, L.
2001-01-01
The research performed in common these last 3 years by the French Atomic Commission CEA, COGEMA and Eurisys Mesures had for main subject the realization of a complete tool of modelization for the largest range of realistic cases, the Pascalys modelization software. The main purpose of the modelization was to calculate the global measurement efficiency, which delivers the most accurate relationship between the photons emitted by the nuclear source in volume, punctual or deposited form and the germanium hyper pure detector, which detects and analyzes the received photons. It has been stated since long time that experimental global measurement efficiency becomes more and more difficult to address especially for complex scene as we can find in decommissioning and dismantling or in case of high activities for which the use of high activity reference sources become difficult to use for both health physics point of view and regulations. The choice of a calculation code is fundamental if accurate modelization is searched. MCNP represents the reference code but its use is long time calculation consuming and then not practicable in line on the field. Direct line-of-sight point kernel code as the French Atomic Commission 3-D analysis Mercure code can represent the practicable compromise between the most accurate MCNP reference code and the realistic performances needed in modelization. The comparison between the results of Pascalys-Mercure and MCNP code taking in account the last improvements of Mercure in the low energy range where the most important errors can occur, is presented in this paper, Mercure code being supported in line by the recent Pascalys 3-D modelization scene software. The incidence of the intrinsic efficiency of the Germanium detector is also approached for the total efficiency of measurement. (authors)
Using MCNP for in-core instrument calibration in CANDU
Energy Technology Data Exchange (ETDEWEB)
Taylor, D.C. [Point Lepreau Generating Station, NB Power, Lepreau, New Brunswick (Canada); Anghel, V.N.P.; Sur, B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)
2002-07-01
The calibration of in-core instruments is important for safe and economical CANDU operation. However, in-core detectors are not normally suited to bench calibration procedures. This paper describes the use and validation of detailed neutron transport calculations for the purpose of calibrating the response of in-core neutron flux detectors. The Monte-Carlo transport code, MCNP, was used to model the thermal neutron flux distribution in the region around self-powered in-core flux detectors (ICFDs), and in the vicinity of the calandria edge. The ICFD model was used to evaluate the reduction in signal of a given detector (the 'detector shading factor') due to neutron absorption in surrounding materials, detectors, and lead-cables. The calandria edge model was used to infer the accuracy of the calandria edge position from flux scans performed by AECL's traveling flux detector (TFD) system. The MCNP results were checked against experimental results on ICFDs, and also against shading factors computed by other means. The use of improved in-core detector calibration factors obtained by this new methodology will improve the accuracy of spatial flux control performance in CANDU-6 reactors. The accurate determination of TFD based calandria edge position is useful in the quantitative measurement of changes in in-core component dimensions and position due to aging, such as pressure tube sag. (author)
Visualization of geometry and tally data using MCNP and Justine
International Nuclear Information System (INIS)
Cox, L.J.; Favorite, J.A.
1999-01-01
The Monte Carlo N-Particle (MCNP) transport code is a general-purpose code that can be used for neutron, photon, electron, or coupled neutron/photon/electron transport, including the capability to calculate eigenvalues for neutron-multiplying systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and fourth-degree elliptical tori. Justine is the graphical user interface and problem setup tool for the Los Alamos Radiation Modeling Interactive Environment (LARAMIE). Its purpose is to serve as a convenient and very general interface for setting up physics calculations and linking together the disparate radiation transport codes under a single front-end. Currently, the LARAMIE system includes MCNP and the deterministic transport code suit DANTSYS (ONEDANT, TWODANT, and THREEDANT, for one-, two-, and three-dimensional geometries, respectively). Justine is currently available through the Radiation Safety Information Computational Center to members of the criticality safety community for evaluation and use. The authors will demonstrate the capabilities of both codes for visualization of geometries and results from a variety of criticality problems
MCNP analysis of the nine-cell LWR gadolinium benchmark
International Nuclear Information System (INIS)
Arkuszewski, J.J.
1988-01-01
The Monte Carlo results for a 9-cell fragment of the light water reactor square lattice with a central gadolinium-loaded pin are presented. The calculations are performed with the code MCNP-3A and the ENDF-B/5 library and compared with the results obtained from the BOXER code system and the JEF-1 library. The objective of this exercise is to study the feasibility of BOXER for the analysis of a Gd-loaded LWR lattice in the broader framework of GAP International Benchmark Analysis. A comparison of results indicates that, apart from unavoidable discrepancies originating from different data evaluations, the BOXER code overestimates the multiplication factor by 1.4 % and underestimates the power release in a Gd cell by 4.66 %. It is hoped that further similar studies with use of the JEF-1 library for both BOXER and MCNP will help to isolate and explain these discrepancies in a cleaner way. (author) 4 refs., 9 figs., 10 tabs
International Nuclear Information System (INIS)
Valentine, T.E.; Mihalczo, J.T.
1995-01-01
This paper describes calculations performed to validate the modified version of the MCNP code, the MCNP-DSP, used for: the neutron and photon spectra of the spontaneous fission of californium 252; the representation of the detection processes for scattering detectors; the timing of the detection process; and the calculation of the frequency analysis parameters for the MCNP-DSP code
Continuous energy cross section library for MCNP/MCNPX based on JENDL high energy file 2007. FXJH7
International Nuclear Information System (INIS)
Sasa, Toshinobu; Sugawara, Takanori; Fukahori, Tokio; Kosako, Kazuaki
2008-11-01
The latest JENDL High Energy File (JENDL/HE) was released in 2007 to respond the requirements of reaction data in high energy range up to several GeV to design accelerator facilities such as accelerator-driven systems and research complex like J-PARC. To apply the JENDL/HE-2007 file to the design study, the cross section library of FXJH7 series was constructed from the JENDL/HE file for the calculation using MCNP and MCNPX codes which are widely used in the field of nuclear reactors, fusion reactors, accelerator facilities, medical applications, and so on. In this report, the outline of the JENDL/HE-2007 file, modification of nuclear data processing code NJOY99, construction of FXJH7 library and test calculations for shielding and eigenvalue analyses are summarized. (author)
Comparison of MCNP and Experimental Measurements for an HPGe-based Spectroscopy Portal Monitor
International Nuclear Information System (INIS)
Keyser, Ronald M.; Hensley, Walter K.; Twomey, Timothy R.; UPP, Daniel L.
2008-01-01
The necessity to monitor international commercial transportation for illicit nuclear materials resulted in the installation of many nuclear radiation detection systems in Portal Monitors. These were mainly gross counters which alarmed at any indication of high radioactivity in the shipment, the vehicle or even the driver. The innocent alarm rate, due to legal shipments of sources and NORM, or medical isotopes in patients, caused interruptions and delays in commerce while the legality of the shipment was verified. To overcome this difficulty, Department of Homeland Security (DHS) supported the writing of the ANSI N42.38 standard (Performance Criteria for Spectroscopy-Based Portal Monitors used for Homeland Security) to define the performance of a Portal Monitor with nuclide identification capabilities, called a Spectroscopy Portal Monitor. This standard defines detection levels and response characteristics for the system for energies from 25 keV to 3 MeV. To accomplish the necessary performance, several different HPGe detector configurations were modeled using MCNP for the horizontal field of view (FOV) and vertical linearity of response over the detection zone of 5 meters by 4.5 meters for 661 keV as representative of the expected nuclides of interest. The configuration with the best result was built and tested. The results for the FOV as a function of energy and the linearity show good agreement with the model and performance exceeding the requirements of N42.38
International Nuclear Information System (INIS)
He Tao; Su Bingjing
2011-01-01
Highlights: → The performance of the MCNP differential operator perturbation technique is compared with that of the MCNP correlated sampling method for three types of fixed-source problems. → In terms of precision, the MCNP perturbation technique outperforms correlated sampling for one type of problem but performs comparably with or even under-performs correlated sampling for the other two types of problems. → In terms of accuracy, the MCNP perturbation calculations may predict inaccurate results for some of the test problems. However, the accuracy can be improved if the midpoint correction technique is used. - Abstract: Correlated sampling and the differential operator perturbation technique are two methods that enable MCNP (Monte Carlo N-Particle) to simulate small response change between an original system and a perturbed system. In this work the performance of the MCNP differential operator perturbation technique is compared with that of the MCNP correlated sampling method for three types of fixed-source problems. In terms of precision of predicted response changes, the MCNP perturbation technique outperforms correlated sampling for the problem involving variation of nuclide concentrations in the same direction but performs comparably with or even underperforms correlated sampling for the other two types of problems that involve void or variation of nuclide concentrations in opposite directions. In terms of accuracy, the MCNP differential operator perturbation calculations may predict inaccurate results that deviate from the benchmarks well beyond their uncertainty ranges for some of the test problems. However, the accuracy of the MCNP differential operator perturbation can be improved if the midpoint correction technique is used.
International Nuclear Information System (INIS)
Youssef, M. Z.
2007-01-01
Attila is a newly developed finite element code based on Sn neutron, gamma, and charged particle transport in 3-D geometry in which unstructured tetrahedral meshes are generated to describe complex geometry that is based on CAD input (Solid Works, Pro/Engineer, etc). In the present work we benchmark its calculation accuracy by comparing its prediction to the measured data inside two experimental mock-ups bombarded with 14 MeV neutrons. The results are also compared to those based on MCNP calculations. The experimental mock-ups simulate parts of the International Thermonuclear Experimental Reactor (ITER) in-vessel components, namely: (1) the Tungsten mockup configuration (54.3 cm x 46.8 cm x 45 cm), and (2) the ITER shielding blanket followed by the SCM region (simulated by alternating layers of SS316 and copper). In the latter configuration, a high aspect ratio rectangular streaming channel was introduced (to simulate steaming paths between ITER blanket modules) which ends with a rectangular cavity. The experiments on these two fusion-oriented integral experiments were performed at the Fusion Neutron Generator (FNG) facility, Frascati, Italy. In addition, the nuclear performance of the ITER MCNP 'Benchmark' CAD model has been performed with Attila to compare its results to those obtained with CAD-based MCNP approach developed by several ITER participants. The objective of this paper is to compare results based on two distinctive 3-D calculation tools using the same nuclear data, FENDL2.1, and the same response functions of several reaction rates measured in ITER mock-ups and to enhance confidence from the international neutronics community in the Attila code and how it can precisely quantify the nuclear field in large and complex systems, such as ITER. Attila has the advantage of providing a full flux mapping visualization everywhere in one run where components subjected to excessive radiation level and strong streaming paths can be identified. In addition, the
International Nuclear Information System (INIS)
Richard, Joshua; Galloway, Jack; Fensin, Michael; Trellue, Holly
2015-01-01
Highlights: • A modular mapping methodogy for neutronic-thermal hydraulic nuclear reactor multiphysics, SMITHERS, has been developed. • Written in Python, SMITHERS takes a novel object-oriented approach for facilitating data transitions between solvers. This approach enables near-instant compatibility with existing MCNP/MONTEBURNS input decks. • It also allows for coupling with thermal-hydraulic solvers of various levels of fidelity. • Two BWR and PWR test problems are presented for verifying correct functionality of the SMITHERS code routines. - Abstract: A novel object-oriented modular mapping methodology for externally coupled neutronics–thermal hydraulics multiphysics simulations was developed. The Simulator using MCNP with Integrated Thermal-Hydraulics for Exploratory Reactor Studies (SMITHERS) code performs on-the-fly mapping of material-wise power distribution tallies implemented by MCNP-based neutron transport/depletion solvers for use in estimating coolant temperature and density distributions with a separate thermal-hydraulic solver. The key development of SMITHERS is that it reconstructs the hierarchical geometry structure of the material-wise power generation tallies from the depletion solver automatically, with only a modicum of additional information required from the user. Additionally, it performs the basis mapping from the combinatorial geometry of the depletion solver to the required geometry of the thermal-hydraulic solver in a generalizable manner, such that it can transparently accommodate varying levels of thermal-hydraulic solver geometric fidelity, from the nodal geometry of multi-channel analysis solvers to the pin-cell level of discretization for sub-channel analysis solvers. The mapping methodology was specifically developed to be flexible enough such that it could successfully integrate preexisting depletion solver case files with different thermal-hydraulic solvers. This approach allows the user to tailor the selection of a
International Nuclear Information System (INIS)
Pölz, Stefan; Laubersheimer, Sven; Eberhardt, Jakob S; Harrendorf, Marco A; Keck, Thomas; Benzler, Andreas; Breustedt, Bastian
2013-01-01
The basic idea of Voxel2MCNP is to provide a framework supporting users in modeling radiation transport scenarios using voxel phantoms and other geometric models, generating corresponding input for the Monte Carlo code MCNPX, and evaluating simulation output. Applications at Karlsruhe Institute of Technology are primarily whole and partial body counter calibration and calculation of dose conversion coefficients. A new generic data model describing data related to radiation transport, including phantom and detector geometries and their properties, sources, tallies and materials, has been developed. It is modular and generally independent of the targeted Monte Carlo code. The data model has been implemented as an XML-based file format to facilitate data exchange, and integrated with Voxel2MCNP to provide a common interface for modeling, visualization, and evaluation of data. Also, extensions to allow compatibility with several file formats, such as ENSDF for nuclear structure properties and radioactive decay data, SimpleGeo for solid geometry modeling, ImageJ for voxel lattices, and MCNPX’s MCTAL for simulation results have been added. The framework is presented and discussed in this paper and example workflows for body counter calibration and calculation of dose conversion coefficients is given to illustrate its application. (paper)
Comparison calculations of WWER-1000 fuel assemblies by using the MCNP 4.2 a KASSETA codes
International Nuclear Information System (INIS)
Trgina, M.
1993-12-01
The power multiplication and distribution factors are compared for various geometries and material configurations of WWER-1000 fuel assemblies. The calculations were performed in 2 ways: (i) using nuclear data, employing older and current data collections, and (ii) using the author's own model based on the KASSETA code. The comparison code MCNP 4.2 is described, intended for computerized simulation of the transport of neutrons, photons and electrons. This code uses its own cross section library. The methodology is outlined and a specification of the Monte Carlo method employed is given. The use of the refined data library gave rise to appreciable deviations of the multiplication factors in all variants. The use of the older data library led to identical criticality results for the variant with water holes. For inserted absorbers the discrepancies in criticality and in power distribution data are appreciable. The marked disagreement between the results of application of the MCNP 4.2 and KASSETA codes for the variants with inserted control elements is indicative of inappropriateness of the approximation procedure in the latter code. (J.B.). 2 tabs., 11 figs., 11 refs
Energy Technology Data Exchange (ETDEWEB)
Xolocostli M, J. V.; Vargas E, S.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Reyes F, M. del C.; Del Valle G, E., E-mail: vicente.xolocostli@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, UP - Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico)
2014-10-15
In this paper a comparison is made in analyzing the suite of Benchmark problems for reactors type BWR between CASMO-4, MCNP6 and Serpent code. The Benchmark problem consists of two different geometries: a fuel cell of a pin and assembly type BWR. To facilitate the study of reactors physics in the fuel pin their nuclear characteristics are provided to detail, such as burnt dependence, the reactivity of selected nuclide, etc. With respect to the fuel assembly, the presented results are regarding to infinite multiplication factor for burning different steps and different vacuum conditions. Making the analysis of this set of Benchmark problems provides comprehensive test problems for the next fuels generation of BWR reactors with high extended burned. It is important to note that when making this comparison the purpose is to validate the methodologies used in modeling for different operating conditions, if the case is of other BWR assembly. The results will be within a range with some uncertainty, considering that does not depend on code that is used. Escuela Superior de Fisica y Matematicas of Instituto Politecnico Nacional (IPN (Mexico) has accumulated some experience in using Serpent, due to the potential of this code over other commercial codes such as CASMO and MCNP. The obtained results for the infinite multiplication factor are encouraging and motivate the studies to continue with the generation of the X S of a core to a next step a respective nuclear data library is constructed and this can be used by codes developed as part of the development project of the Mexican Analysis Platform of Nuclear Reactors AZTLAN. (Author)
Improvement of Monte Carlo code A3MCNP for large-scale shielding problems
International Nuclear Information System (INIS)
Miyake, Y.; Ohmura, M.; Hasegawa, T.; Ueki, K.; Sato, O.; Haghighat, A.; Sjoden, G.E.
2004-01-01
A 3 MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, that automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic 'importance' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A 3 MCNP uses the 3-D Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A 3 MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A 3 MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A 3 MCNP (referred to as A 3 MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A 3 MCNPV for a concrete cask streaming problem and a PWR dosimetry problem. (author)
Performance of the improved version of Monte Carlo Code A3MCNP for cask shielding design
International Nuclear Information System (INIS)
Hasegawa, T.; Ueki, K.; Sato, O.; Sjoden, G.E.; Miyake, Y.; Ohmura, M.; Haghighat, A.
2004-01-01
A 3 MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, that automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic ''importance'' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A 3 MCNP uses the 3-D Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A3MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A 3 MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A 3 MCNP (referred to as A 3 MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A 3 MCNPV for cask neutron and gamma-ray shielding problem
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.
Comparisons between MCNP, EGS4 and experiment for clinical electron beams
International Nuclear Information System (INIS)
Jeraj, R.; Keall, P.J.; Ostwald, P.M.
1999-01-01
Understanding the limitations of Monte Carlo codes is essential in order to avoid systematic errors in simulations, and to suggest further improvement of the codes. MCNP and EGS4, Monte Carlo codes commonly used in medical physics, were compared and evaluated against electron depth dose data and experimental backscatter results obtained using clinical radiotherapy beams. Different physical models and algorithms used in the codes give significantly different depth dose curves and electron backscattering factors. The default version of MCNP calculates electron depth dose curves which are too penetrating. The MCNP results agree better with experiment if the ITS-style energy-indexing algorithm is used. EGS4 underpredicts electron backscattering for high- Z materials. The results slightly improve if optimal PRESTA-I parameters are used. MCNP simulates backscattering well even for high- Z materials. To conclude the comparison, a timing study was performed. EGS4 is generally faster than MCNP and use of a large number of scoring voxels dramatically slows down the MCNP calculation. However, use of a large number of geometry voxels in MCNP only slightly affects the speed of the calculation. (author)
An enhanced geometry-independent mesh weight window generator for MCNP
International Nuclear Information System (INIS)
Evans, T.M.; Hendricks, J.S.
1997-01-01
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
Optimal space-energy splitting in MCNP with the DSA
International Nuclear Information System (INIS)
Dubi, A.; Gurvitz, N.
1990-01-01
The Direct Statistical Approach (DSA) particle transport theory is based on the possibility of obtaining exact explicit expressions for the dependence of the second moment and calculation time on the splitting parameters. This allows the automatic optimization of the splitting parameters by ''learning'' the bulk parameters from which the problem dependent coefficients of the quality function (second moment time) are constructed. The above procedure was exploited to implement an automatic optimization of the splitting parameters in the Monte Carlo Neutron Photon (MCNP) code. This was done in a number of steps. In the first instance, only spatial surface splitting was considered. In this step, the major obstacle has been the truncation of an infinite series of ''products'' of ''surface path's'' leading from the source to the detector. Encouraging results from the first phase led to the inclusion of full space/energy phase space splitting. (author)
Problem and solution of tally segment card in MCNP code
International Nuclear Information System (INIS)
Xie Jiachun; Zhao Shouzhi; Sun Zheng; Jia Baoshan
2010-01-01
Wrong results may be given when FS card (tally segment card) was used for tally with other tally cards in Monte Carlo code MCNP. According to the comparison of segment tally results which were obtained by FS card of three different models of the same geometry, the tally results of fuel regions were found to be wrong in fill pattern. The reason is that the fuel cells were described by Universe card and FILL card, and the filled cells were always considered at Universe card definition place. A proposed solution was that the segment tally for filled cells was done at Universe card definition place. Radial flux distribution of one example was calculated in this way. The results show that the fault of segment tally with FS card in fill pattern could be solved by this method. (authors)
BWR Fuel Assemblies Physics Analysis Utilizing 3D MCNP Modeling
International Nuclear Information System (INIS)
Chiang, Ren-Tai; Williams, John B.; Folk, Ken S.
2008-01-01
MCNP is used to model a partially controlled BWR fresh fuel four assemblies (2x2) system for better understanding BWR fuel behavior and for benchmarking production codes. The impact of the GE14 plenum regions on axial power distribution is observed by comparing against the GE13 axial power distribution, in which the GE14 relative power is lower than the GE13 relative power at the 15. node and at the 16. node due to presence of the plenum regions in GE14 fuel in these two nodes. The segmented rod power distribution study indicates that the azimuthally dependent power distribution is very significant for the fuel rods next to the water gap in the uncontrolled portion. (authors)
BWR Fuel Assemblies Physics Analysis Utilizing 3D MCNP Modeling
Energy Technology Data Exchange (ETDEWEB)
Chiang, Ren-Tai [University of Florida, Gainesville, Florida 32611 (United States); Williams, John B.; Folk, Ken S. [Southern Nuclear Company, Birmingham, Alabama 35242 (United States)
2008-07-01
MCNP is used to model a partially controlled BWR fresh fuel four assemblies (2x2) system for better understanding BWR fuel behavior and for benchmarking production codes. The impact of the GE14 plenum regions on axial power distribution is observed by comparing against the GE13 axial power distribution, in which the GE14 relative power is lower than the GE13 relative power at the 15. node and at the 16. node due to presence of the plenum regions in GE14 fuel in these two nodes. The segmented rod power distribution study indicates that the azimuthally dependent power distribution is very significant for the fuel rods next to the water gap in the uncontrolled portion. (authors)
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.
MCNP simulation of the TRIGA Mark II benchmark experiment
International Nuclear Information System (INIS)
Jeraj, R.; Glumac, B.; Maucec, M.
1996-01-01
The complete 3D MCNP model of the TRIGA Mark II reactor is presented. It enables precise calculations of some quantities of interest in a steady-state mode of operation. Calculational results are compared to the experimental results gathered during reactor reconstruction in 1992. Since the operating conditions were well defined at that time, the experimental results can be used as a benchmark. It may be noted that this benchmark is one of very few high enrichment benchmarks available. In our simulations experimental conditions were thoroughly simulated: fuel elements and control rods were precisely modeled as well as entire core configuration and the vicinity of the core. ENDF/B-VI and ENDF/B-V libraries were used. Partial results of benchmark calculations are presented. Excellent agreement of core criticality, excess reactivity and control rod worths can be observed. (author)
Fuel element transfer cask modelling using MCNP technique
International Nuclear Information System (INIS)
Rosli Darmawan
2009-01-01
Full text: 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. (author)
Fuel Element Transfer Cask Modelling Using MCNP Technique
International Nuclear Information System (INIS)
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.
Installation of MCNP on 64-bit parallel computers
International Nuclear Information System (INIS)
Meginnis, A.B.; Hendricks, J.S.; McKinney, G.W.
1995-01-01
The Monte Carlo radiation transport code MCNP has been successfully ported to two 64-bit workstations, the SGI and DEC Alpha. We found the biggest problem for installation on these machines to be Fortran and C mismatches in argument passing. Correction of these mismatches enabled, for the first time, dynamic memory allocation on 64-bit workstations. Although the 64-bit hardware is faster because 8-bytes are processed at a time rather than 4-bytes, we found no speed advantage in true 64-bit coding versus implicit double precision when porting an existing code to the 64-bit workstation architecture. We did find that PVM multiasking is very successful and represents a significant performance enhancement for scientific workstations
International Nuclear Information System (INIS)
Xolocostli M, J. V.; Vargas E, S.; Gomez T, A. M.; Reyes F, M. del C.; Del Valle G, E.
2014-10-01
In this paper a comparison is made in analyzing the suite of Benchmark problems for reactors type BWR between CASMO-4, MCNP6 and Serpent code. The Benchmark problem consists of two different geometries: a fuel cell of a pin and assembly type BWR. To facilitate the study of reactors physics in the fuel pin their nuclear characteristics are provided to detail, such as burnt dependence, the reactivity of selected nuclide, etc. With respect to the fuel assembly, the presented results are regarding to infinite multiplication factor for burning different steps and different vacuum conditions. Making the analysis of this set of Benchmark problems provides comprehensive test problems for the next fuels generation of BWR reactors with high extended burned. It is important to note that when making this comparison the purpose is to validate the methodologies used in modeling for different operating conditions, if the case is of other BWR assembly. The results will be within a range with some uncertainty, considering that does not depend on code that is used. Escuela Superior de Fisica y Matematicas of Instituto Politecnico Nacional (IPN (Mexico) has accumulated some experience in using Serpent, due to the potential of this code over other commercial codes such as CASMO and MCNP. The obtained results for the infinite multiplication factor are encouraging and motivate the studies to continue with the generation of the X S of a core to a next step a respective nuclear data library is constructed and this can be used by codes developed as part of the development project of the Mexican Analysis Platform of Nuclear Reactors AZTLAN. (Author)
Impact of MCNP Unresolved Resonance Probability-Table Treatment on Uranium and Plutonium Benchmarks
International Nuclear Information System (INIS)
Mosteller, R.D.; Little, R.C.
1999-01-01
A probability-table treatment recently has been incorporated into an intermediate version of the MCNP Monte Carlo code named MCNP4XS. This paper presents MCNP4XS results for a variety of uranium and plutonium criticality benchmarks, calculated with and without the probability-table treatment. It is shown that the probability-table treatment can produce small but significant reactivity changes for plutonium and 233 U systems with intermediate spectra. More importantly, it can produce substantial reactivity increases for systems with large amounts of 238 U and intermediate spectra
A Validated MCNP(X) Cross Section Library based on JEFF 3.1
International Nuclear Information System (INIS)
Haeck, W.; Verboomen, B.
2006-01-01
ALEPH-LIB is a multi-temperature neutron transport library for standard use by MCNP(X) and ALEPH generated with ALEPH-DLG. This is an auxiliary computer code to ALEPH, the Monte Carlo burn-up code under development at SCK-CEN in collaboration with Ghent university. ALEPH-DLG automates the entire process of generating library files with NJOY and takes care of the first requirement of a validated application library: verify the processing. It produces tailor made NJOY input files using data from the original ENDF file (initial temperature, the fact if the nuclide is fissile or if it has unresolved resonances, etc.) When the library files have been generated, ALEPH-DLG will also process the output from NJOY by extracting all messages and warnings. If ALEPH-DLG finds anything out of the ordinary, it will either warn the user or perform corrective actions. The temperatures included in the ALEPH-LIB library are 300, 600, 900, 1200, 1500 and 1800 K. Library files were produced for the JEF 2.2, JEFF 3.0, JEFF 3.1, JENDL 3.3 and ENDF/B-VI.8 nuclear data libraries. This will be extended with ENDF/B-VII when it becomes available. This report deals with the JEFF 3.1 files included in ALEPH-LIB that are now released by the NEA-OECD.
Development and validation of a model TRIGA Mark III reactor with code MCNP5
International Nuclear Information System (INIS)
Galicia A, J.; Francois L, J. L.; Aguilar H, F.
2015-09-01
The main purpose of this paper is to obtain a model of the reactor core TRIGA Mark III that accurately represents the real operating conditions to 1 M Wth, using the Monte Carlo code MCNP5. To provide a more detailed analysis, different models of the reactor core were realized by simulating the control rods extracted and inserted in conditions in cold (293 K) also including an analysis for shutdown margin, so that satisfied the Operation Technical Specifications. The position they must have the control rods to reach a power equal to 1 M Wth, were obtained from practice entitled Operation in Manual Mode performed at Instituto Nacional de Investigaciones Nucleares (ININ). Later, the behavior of the K eff was analyzed considering different temperatures in the fuel elements, achieving calculate subsequently the values that best represent the actual reactor operation. Finally, the calculations in the developed model for to obtain the distribution of average flow of thermal, epithermal and fast neutrons in the six new experimental facilities are presented. (Author)
Study of salinity in aqueous medium using X-Ray beam with MCNP-X code
Energy Technology Data Exchange (ETDEWEB)
Barbosa, Caroline M.; Braz, Delson [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Salgado, César M., E-mail: cbarbosa@nuclear.ufrj.br, E-mail: delson@nuclear.ufrj.br, E-mail: otero@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)
2017-07-01
In offshore production, it is possible that the produced water presents geochemical characteristics that correspond to the mixture of formation water (connate water) and the sea water (injection water), and the physical-chemical behavior of the injected water allows a considerable variation in the index of salinity altering the water/oil ratio during transportation and/or extraction. Injection water is generally used to raise the reservoir pressure, increasing the percentage of extracted oil. This water has a significant amount of salts that generate some difficulties, such as measuring fractions of volume in multiphase systems. One way to check the effects of salinity would be to regularly measure the amount of salt present in the water. In this way, this work presents a methodology to measure the concentration and the types of salts using nuclear techniques through the MCNP-X computational code. The measurement geometry uses an X-ray beam (40-100 keV) and NaI(Tl) scintillation detector positioned diametrically opposed to the source. The studied samples were the NaCl, KCl and MgCl{sub 2} salts in aqueous solution. The results present the possibility of differentiating the formation and injection waters due to differences in the salt concentrations. (author)
Progress of conversion system from CAD data to MCNP geometry data in Japan
International Nuclear Information System (INIS)
Sato, S.; Nashif, H.; Masuda, F.; Morota, H.; Iida, H.; Konno, C.
2010-01-01
Automatic conversion systems from CAD data to MCNP geometry input data have been developed to convert the CAD data of the fusion reactor with very complicated structure. So far, two conversion systems (GEOMIT-1 and ARCMCP) have been developed and the third system (GEOMIT-2) is under developing. The void data can be created in these systems. GEOMIT-1 was developed in 2007, but a lot of manual shape splitting work for the CAD data was required to convert the complicated geometry. ARCMCP was developed in 2008. The algorithm has been drastically improved on automatic creation of ambiguous surface in ARCMCP, but it still required a little manual shape splitting work. The latest system, GEOMIT-2, does not require additional commercial software packages, though the previous systems require them. It also has functions of the CAD data healing and the automatic shape splitting. Geometrical errors of CAD data can be automatically revised by the healing function, and complicated geometries can be automatically split into simple geometries by the shape splitting function. Any manual works for CAD data are not required in GEOMIT-2. GEOMIT-2 is very useful for nuclear analyses of fusion reactors.
Optimization of Shielding- Collimator Parameters for ING-27 Neutron Generator Using MCNP5
Directory of Open Access Journals (Sweden)
Hegazy Aya Hamdy
2018-01-01
Full Text Available Neutron generators are now used in various fields. They produce only fast neutrons; D-D neutron generator produces 2.45 MeV neutrons and D-T produces 14.1 MeV neutrons. In order to optimize shielding-collimator parameters to achieve higher neutron flux at the investigated sample (The signal with lower neutron and gamma rays flux at the area of the detectors, design iterations are widely used. This work was applied to ROMASHA setup, TANGRA project, FLNP, Joint Institute for Nuclear Research. The studied parameters were; (1 shielding-collimator material, (2 Distance between the shielding-collimator assembly first plate and center of the neutron beam, and (3 thickness of collimator sheets. MCNP5 was used to simulate ROMASHA setup after it was validated on the experimental results of irradiation of Carbon-12 sample for one hour to detect its 4.44 MeV characteristic gamma line. The ratio between the signal and total neutron flux that enters each detector was calculated and plotted, concluding that the optimum shielding-collimator assembly is Tungsten of 5 cm thickness for each plate, and a distance of 2.3 cm. Also, the ratio between the signal and total gamma rays flux was calculated and plotted for each detector, leading to the previous conclusion but the distance was 1 cm.
A Validated MCNP(X) Cross Section Library based on JEFF 3.1
Energy Technology Data Exchange (ETDEWEB)
Haeck, W; Verboomen, B
2006-10-15
ALEPH-LIB is a multi-temperature neutron transport library for standard use by MCNP(X) and ALEPH generated with ALEPH-DLG. This is an auxiliary computer code to ALEPH, the Monte Carlo burn-up code under development at SCK-CEN in collaboration with Ghent university. ALEPH-DLG automates the entire process of generating library files with NJOY and takes care of the first requirement of a validated application library: verify the processing. It produces tailor made NJOY input files using data from the original ENDF file (initial temperature, the fact if the nuclide is fissile or if it has unresolved resonances, etc.) When the library files have been generated, ALEPH-DLG will also process the output from NJOY by extracting all messages and warnings. If ALEPH-DLG finds anything out of the ordinary, it will either warn the user or perform corrective actions. The temperatures included in the ALEPH-LIB library are 300, 600, 900, 1200, 1500 and 1800 K. Library files were produced for the JEF 2.2, JEFF 3.0, JEFF 3.1, JENDL 3.3 and ENDF/B-VI.8 nuclear data libraries. This will be extended with ENDF/B-VII when it becomes available. This report deals with the JEFF 3.1 files included in ALEPH-LIB that are now released by the NEA-OECD.
Optimization of Shielding- Collimator Parameters for ING-27 Neutron Generator Using MCNP5
Hegazy, Aya Hamdy; Skoy, V. R.; Hossny, K.
2018-04-01
Neutron generators are now used in various fields. They produce only fast neutrons; D-D neutron generator produces 2.45 MeV neutrons and D-T produces 14.1 MeV neutrons. In order to optimize shielding-collimator parameters to achieve higher neutron flux at the investigated sample (The signal) with lower neutron and gamma rays flux at the area of the detectors, design iterations are widely used. This work was applied to ROMASHA setup, TANGRA project, FLNP, Joint Institute for Nuclear Research. The studied parameters were; (1) shielding-collimator material, (2) Distance between the shielding-collimator assembly first plate and center of the neutron beam, and (3) thickness of collimator sheets. MCNP5 was used to simulate ROMASHA setup after it was validated on the experimental results of irradiation of Carbon-12 sample for one hour to detect its 4.44 MeV characteristic gamma line. The ratio between the signal and total neutron flux that enters each detector was calculated and plotted, concluding that the optimum shielding-collimator assembly is Tungsten of 5 cm thickness for each plate, and a distance of 2.3 cm. Also, the ratio between the signal and total gamma rays flux was calculated and plotted for each detector, leading to the previous conclusion but the distance was 1 cm.
Energy Technology Data Exchange (ETDEWEB)
Leal, B. [Centro Regional de Estudios Nucleares, A.P. 579C, 98068 Zacatecas (Mexico)
2004-07-01
In the majority of the laboratories, the calibration in efficiency of the detector is carried out by means of the standard sources measurement of gamma photons that have a determined activity, or for matrices that contain a variety of radionuclides that can embrace the energy range of interest. Given the experimental importance that has the determination from the curves of efficiency to the effects of establishing the quantitative results, is appealed to the simulation of the response function of the detector used in the Regional Center of Nuclear Studies inside the energy range of 80 keV to 1400 keV varying the density of the matrix, by means of the application of the Monte Carlo code MCNP-4A. The adjustment obtained shows an acceptance grade in the range of 100 to 600 keV, with a smaller percentage discrepancy to 5%. (Author)
Energy Technology Data Exchange (ETDEWEB)
Parreno Z, F.; Paucar J, R.; Picon C, C. [Instituto Peruano de Energia Nuclear, Av. Canada 1470, San Borja, Lima 41 (Peru)
1998-12-31
The simulation by Monte Carlo is tool which Medical Physics counts with it for the development of its research, the interest by this tool is growing, as we may observe in the main scientific journals for the years 1995-1997 where more than 27 % of the papers treat over Monte Carlo and/or its applications in the radiation transport.In the Peruvian Institute of Nuclear Energy we are implementing and making use of the MCNP4 and EGS4 codes. In this work are presented the general features of the Monte Carlo method and its more useful applications in Medical Physics. Likewise, it is made a simulation of the calculation of isodose curves in an interstitial treatment with Ir-192 wires in a mammary gland carcinoma. (Author)
Energy Technology Data Exchange (ETDEWEB)
Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: samuel.vargas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)
2013-10-15
The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)
Energy Technology Data Exchange (ETDEWEB)
Aredes, Vitor Ottoni; Bitelli, Ulysses d' Utra; Mura, Luiz Ernesto C.; Santos, Diogo Feliciano dos; Lima, Ana Cecilia de Souza, E-mail: ubitelli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)
2015-07-01
This study aims to determine the distribution of thermal neutron flux in the IPEN/MB-01 nuclear reactor core assembled with cylindrical core configuration of minor excess of reactivity with 568 fuel rods (28 fuel rods in diameter). The thermal neutron flux at the positions of irradiation derive from the method of reaction rate using gold foils. The experiment consists in inserting gold activations foils with and without cadmium coverage (cadmium boxes with 0.0502 cm thickness) in several positions throughout the active core. After irradiation, activity induced by nuclear reaction rates over gold foils is assessed by gamma ray spectrometry using a high-purity germanium (HPGe) detector. Experimental results are compared to those derived from calculations performed using a three dimensional CITATION diffusion code and MCNP-5 code and a proper nuclear data library. While calculated neutron flux data shows good agreement with experimental values in regions with little disturbance in the neutron flux, also showing that in the region of the reflectors of neutrons and near the control rods, the diffusion theory is not very precise. The average value of thermal neutron flux obtained experimentally compared to the calculated value by CITATION code and MCNP-5 code respectively show a difference of 1.18% and 0.84% at a nuclear power level of 74.65 ± 3.28 % watts. The average measured value of thermal neutron flux is 4.10 10{sup 8} ± 5.25% n/cm{sup 2}s. (author)
International Nuclear Information System (INIS)
Aredes, Vitor Ottoni; Bitelli, Ulysses d'Utra; Mura, Luiz Ernesto C.; Santos, Diogo Feliciano dos; Lima, Ana Cecilia de Souza
2015-01-01
This study aims to determine the distribution of thermal neutron flux in the IPEN/MB-01 nuclear reactor core assembled with cylindrical core configuration of minor excess of reactivity with 568 fuel rods (28 fuel rods in diameter). The thermal neutron flux at the positions of irradiation derive from the method of reaction rate using gold foils. The experiment consists in inserting gold activations foils with and without cadmium coverage (cadmium boxes with 0.0502 cm thickness) in several positions throughout the active core. After irradiation, activity induced by nuclear reaction rates over gold foils is assessed by gamma ray spectrometry using a high-purity germanium (HPGe) detector. Experimental results are compared to those derived from calculations performed using a three dimensional CITATION diffusion code and MCNP-5 code and a proper nuclear data library. While calculated neutron flux data shows good agreement with experimental values in regions with little disturbance in the neutron flux, also showing that in the region of the reflectors of neutrons and near the control rods, the diffusion theory is not very precise. The average value of thermal neutron flux obtained experimentally compared to the calculated value by CITATION code and MCNP-5 code respectively show a difference of 1.18% and 0.84% at a nuclear power level of 74.65 ± 3.28 % watts. The average measured value of thermal neutron flux is 4.10 10 8 ± 5.25% n/cm 2 s. (author)
Validation of MCNP6.1 for Criticality Safety of Pu-Metal, -Solution, and -Oxide Systems
Energy Technology Data Exchange (ETDEWEB)
Kiedrowski, Brian C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Conlin, Jeremy Lloyd [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Favorite, Jeffrey A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kahler, III, Albert C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kersting, Alyssa R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parsons, Donald K. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Walker, Jessie L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
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.
Development of ITER 3D neutronics model and nuclear analyses
International Nuclear Information System (INIS)
Zeng, Q.; Zheng, S.; Lu, L.; Li, Y.; Ding, A.; Hu, H.; Wu, Y.
2007-01-01
ITER nuclear analyses rely on the calculations with the three-dimensional (3D) Monte Carlo code e.g. the widely-used MCNP. However, continuous changes in the design of the components require the 3D neutronics model for nuclear analyses should be updated. Nevertheless, the modeling of a complex geometry with MCNP by hand is a very time-consuming task. It is an efficient way to develop CAD-based interface code for automatic conversion from CAD models to MCNP input files. Based on the latest CAD model and the available interface codes, the two approaches of updating 3D nuetronics model have been discussed by ITER IT (International Team): The first is to start with the existing MCNP model 'Brand' and update it through a combination of direct modification of the MCNP input file and generation of models for some components directly from the CAD data; The second is to start from the full CAD model, make the necessary simplifications, and generate the MCNP model by one of the interface codes. MCAM as an advanced CAD-based MCNP interface code developed by FDS Team in China has been successfully applied to update the ITER 3D neutronics model by adopting the above two approaches. The Brand model has been updated to generate portions of the geometry based on the newest CAD model by MCAM. MCAM has also successfully performed conversion to MCNP neutronics model from a full ITER CAD model which is simplified and issued by ITER IT to benchmark the above interface codes. Based on the two updated 3D neutronics models, the related nuclear analyses are performed. This paper presents the status of ITER 3D modeling by using MCAM and its nuclear analyses, as well as a brief introduction of advanced version of MCAM. (authors)
Comparison of CdZnTe neutron detector models using MCNP6 and Geant4
Wilson, Emma; Anderson, Mike; Prendergasty, David; Cheneler, David
2018-01-01
The production of accurate detector models is of high importance in the development and use of detectors. Initially, MCNP and Geant were developed to specialise in neutral particle models and accelerator models, respectively; there is now a greater overlap of the capabilities of both, and it is therefore useful to produce comparative models to evaluate detector characteristics. In a collaboration between Lancaster University, UK, and Innovative Physics Ltd., UK, models have been developed in both MCNP6 and Geant4 of Cadmium Zinc Telluride (CdZnTe) detectors developed by Innovative Physics Ltd. Herein, a comparison is made of the relative strengths of MCNP6 and Geant4 for modelling neutron flux and secondary γ-ray emission. Given the increasing overlap of the modelling capabilities of MCNP6 and Geant4, it is worthwhile to comment on differences in results for simulations which have similarities in terms of geometries and source configurations.
Comparison of EGS4 and MCNP Monte Carlo codes when calculating radiotherapy depth doses.
Love, P A; Lewis, D G; Al-Affan, I A; Smith, C W
1998-05-01
The Monte Carlo codes EGS4 and MCNP have been compared when calculating radiotherapy depth doses in water. The aims of the work were to study (i) the differences between calculated depth doses in water for a range of monoenergetic photon energies and (ii) the relative efficiency of the two codes for different electron transport energy cut-offs. The depth doses from the two codes agree with each other within the statistical uncertainties of the calculations (1-2%). The relative depth doses also agree with data tabulated in the British Journal of Radiology Supplement 25. A discrepancy in the dose build-up region may by attributed to the different electron transport algorithims used by EGS4 and MCNP. This discrepancy is considerably reduced when the improved electron transport routines are used in the latest (4B) version of MCNP. Timing calculations show that EGS4 is at least 50% faster than MCNP for the geometries used in the simulations.
Developing an interface between MCNP and McStas for simulation of neutron moderators
DEFF Research Database (Denmark)
Klinkby, Esben Bryndt; Lauritzen, Bent; Nonbøl, Erik
2012-01-01
Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using MCNP/X whereas simulations of neutron transport and instrument performance are carried out by neutron ray tracing codes such as McStas. The coupling between the two simulations suites...... typically consists of providing analytical fits from MCNP/X neutron spectra to McStas. This method is generally successful, but as will be discussed in the this paper, there are limitations and a more direct coupling between MCNP/X andMcStas could allow for more accurate simulations of e.g. complex...... moderator geometries, interference between beamlines as well as shielding requirements along the neutron guides. In this paper different possible interfaces between McStas and MCNP/X are discussed and first preliminary performance results are shown....
A Patch to MCNP5 for Multiplication Inference: Description and User Guide
Energy Technology Data Exchange (ETDEWEB)
Solomon, Jr., Clell J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
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.
MCNP-REN a Monte Carlo tool for neutron detector design
Abhold, M E
2002-01-01
The development of neutron detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model fails to accurately predict detector response in common applications. For this reason, the general Monte Carlo code developed at Los Alamos National Laboratory, Monte Carlo N-Particle (MCNP), was modified to simulate the pulse streams that would be generated by a neutron detector and normally analyzed by a shift register. This modified code, MCNP-Random Exponentially Distributed Neutron Source (MCNP-REN), along with the Time Analysis Program, predicts neutron detector response without using the point reactor model, making it unnecessary for the user to decide whether or not the assumptions of the point model are met for their application. MCNP-REN is capable of simulating standard neutron coincidence counting as well as neutron multiplicity counting. Measurements of mixed oxide fresh fuel w...
Determination of the detection efficiency of a HPGe detector by means of the MCNP 4A simulation code
International Nuclear Information System (INIS)
Leal, B.
2004-01-01
In the majority of the laboratories, the calibration in efficiency of the detector is carried out by means of the standard sources measurement of gamma photons that have a determined activity, or for matrices that contain a variety of radionuclides that can embrace the energy range of interest. Given the experimental importance that has the determination from the curves of efficiency to the effects of establishing the quantitative results, is appealed to the simulation of the response function of the detector used in the Regional Center of Nuclear Studies inside the energy range of 80 keV to 1400 keV varying the density of the matrix, by means of the application of the Monte Carlo code MCNP-4A. The adjustment obtained shows an acceptance grade in the range of 100 to 600 keV, with a smaller percentage discrepancy to 5%. (Author)
EJ2-MCNPlib. Contents of the JEF-2.2 based neutron cross-section library for MCNP4A
International Nuclear Information System (INIS)
Hogenbirk, A.; Oppe, J.
1995-05-01
In this report a description is given of the EJ2-MCNPlib library. The EJ2-MCNPlib library is to be used for reactivity/critically calculations and general neutron/photon transport calculations with the Monte Carlo code MCNP4A. The library is based on the European JEF-2.2 nuclear data evaluation and contains data for all (i.e. 313) nuclides available on this evaluation.The cross-section data were generated using the NJOY cross-section processing code system, version 91.118. For easy reference cross-section plots are given in this report for the total, elastic and absorption cross sections for all nuclides on the EJ2-MCNPlib library. Furthermore, for verification purposes a graphical intercomparison is given of the results of standard benchmark calculations performed with JEF-2.2 cross-section data and with ENDF/B-V cross-section data (whenever available). 6 refs
SWAT3.1 - the integrated burnup code system driving continuous energy Monte Carlo codes MVP and MCNP
International Nuclear Information System (INIS)
Suyama, Kenya; Mochizuki, Hiroki; Takada, Tomoyuki; Ryufuku, Susumu; Okuno, Hiroshi; Murazaki, Minoru; Ohkubo, Kiyoshi
2009-05-01
Integrated burnup calculation code system SWAT is a system that combines neutronics calculation code SRAC,which is widely used in Japan, and point burnup calculation code ORIGEN2. It has been used to evaluate the composition of the uranium, plutonium, minor actinides and the fission products in the spent nuclear fuel. Based on this idea, the integrated burnup calculation code system SWAT3.1 was developed by combining the continuous energy Monte Carlo code MVP and MCNP, and ORIGEN2. This enables us to treat the arbitrary fuel geometry and to generate the effective cross section data to be used in the burnup calculation with few approximations. This report describes the outline, input data instruction and several examples of the calculation. (author)
International Nuclear Information System (INIS)
Vargas E, S.; Esquivel E, J.; Ramirez S, J. R.
2013-10-01
The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)
CREPT-MCNP code for efficiency calibration of HPGe detectors with the representative point method.
Saegusa, Jun
2008-01-01
The representative point method for the efficiency calibration of volume samples has been previously proposed. For smoothly implementing the method, a calculation code named CREPT-MCNP has been developed. The code estimates the position of a representative point which is intrinsic to each shape of volume sample. The self-absorption correction factors are also given to make correction on the efficiencies measured at the representative point with a standard point source. Features of the CREPT-MCNP code are presented.
MCNP evaluation of top node control rod depletion below the core in KKL
International Nuclear Information System (INIS)
Beran, Tâm; Seltborg, Per; Lindahl, Sten-Örjan; Bieli, Roger; Ledergerber, Guido
2014-01-01
In previous studies, there has been identified a significant discrepancy in the BWR control rod top node depletion between the two core simulator nodal codes POLCA7 and PRESTO-2, which indicates that there is a large general uncertainty in nodal codes in calculating the top node depletion of fully withdrawn control rods. In this study, the stochastic Monte Carlo code MCNP has been used to calculate the top node control rod depletion for benchmarking the nodal codes. By using the TIP signal obtained from an extended TIP campaign below the core performed in the KKL reactor, the MCNP model has been verified by comparing the axial profile between the TIP data and the gamma flux calculated by MCNP. The MCNP results have also been compared with calculations from POLCA7, which was found to yield slightly higher depletion rates than MCNP. It was also found that the 10 B depletion in the top node is very sensitive to the exact axial location of the control rod top when it is fully withdrawn. By using the MCNP results, the neutron flux model below the core in the nodal codes can be improved by implementing an exponential function for the neutron flux. (author)
MCNP6 Simulation of Light and Medium Nuclei Fragmentation at Intermediate Energies
Energy Technology Data Exchange (ETDEWEB)
Mashnik, Stepan Georgievich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kerby, Leslie Marie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2015-05-22
MCNP6, the latest and most advanced LANL Monte Carlo transport code, representing a merger of MCNP5 and MCNPX, is actually much more than the sum of those two computer codes; MCNP6 is available to the public via RSICC at Oak Ridge, TN, USA. In the present work, MCNP6 was validated and verified (V&V) against different experimental data on intermediate-energy fragmentation reactions, and results by several other codes, using mainly the latest modifications of the Cascade-Exciton Model (CEM) and of the Los Alamos version of the Quark-Gluon String Model (LAQGSM) event generators CEM03.03 and LAQGSM03.03. It was found that MCNP6 using CEM03.03 and LAQGSM03.03 describes well fragmentation reactions induced on light and medium target nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below, and can serve as a reliable simulation tool for different applications, like cosmic-ray-induced single event upsets (SEU’s), radiation protection, and cancer therapy with proton and ion beams, to name just a few. Future improvements of the predicting capabilities of MCNP6 for such reactions are possible, and are discussed in this work.
Energy Technology Data Exchange (ETDEWEB)
Leal A, B.; Mireles G, F.; Quirino T, L.; Pinedo, J.L. [Universidad Autonoma de Zacatecas, Zacatecas (Mexico)]. e-mail: bleal79@yahoo.com.mx
2005-07-01
In the area of the Radiological Safety it is required of a calibrated detection system in energy and efficiency for the determination of the concentration in activity in samples that vary in chemical composition and by this in density. The area of Nuclear Engineering requires to find the grade of isotopic enrichment of the uranium of the Sub-critic Nuclear Chicago 9000 Mark. Given the experimental importance that has the determination from the curves of efficiency to the effects of establishing the quantitative results, is appealed to the simulation of the response function of the detector used in the Regional Center of Nuclear Studies inside the range of energy of 80 keV to 1400 keV varying the density of the matrix and the chemical composition by means of the application of the Monte Carlo code MCNP-4A. The obtained results in the simulation of the response function of the detector show a grade of acceptance in the range from 500 to 1400 keV energy, with a smaller percentage discrepancy to 10%, in the range of low energy that its go from 59 to 400 keV, the percentage discrepancy varies from 17% until 30%, which is manifested in the opposing isotopic relationship for 5 fuel rods of the Sub critic nuclear assemble. (Author)
International Nuclear Information System (INIS)
Vrban, B.; Lueley, J.; Farkas, G.; Hascik, J.; Hinca, R.; Petriska, M.; Slugen, V.
2012-01-01
The purpose of the analysis was the determination of critical H_3BO_3 concentrations for the first fuel loading into the reactor core of MO34 units using 2"n"d generation fuel during the first start-up of new unit using calculation code MCNP 1.60. H_3BO_3 concentrations were computed for the given temperature of the primary circuit and position of the 6"t"h safety control rod group. Because of the very first start-up of these units, detailed analyses of active-core parameters are required by National Regulatory Authority and needed for safe operation of nuclear facility. (authors)
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
International Nuclear Information System (INIS)
Leal A, B.; Mireles G, F.; Quirino T, L.; Pinedo, J.L.
2005-01-01
In the area of the Radiological Safety it is required of a calibrated detection system in energy and efficiency for the determination of the concentration in activity in samples that vary in chemical composition and by this in density. The area of Nuclear Engineering requires to find the grade of isotopic enrichment of the uranium of the Sub-critic Nuclear Chicago 9000 Mark. Given the experimental importance that has the determination from the curves of efficiency to the effects of establishing the quantitative results, is appealed to the simulation of the response function of the detector used in the Regional Center of Nuclear Studies inside the range of energy of 80 keV to 1400 keV varying the density of the matrix and the chemical composition by means of the application of the Monte Carlo code MCNP-4A. The obtained results in the simulation of the response function of the detector show a grade of acceptance in the range from 500 to 1400 keV energy, with a smaller percentage discrepancy to 10%, in the range of low energy that its go from 59 to 400 keV, the percentage discrepancy varies from 17% until 30%, which is manifested in the opposing isotopic relationship for 5 fuel rods of the Sub critic nuclear assemble. (Author)
Energy Technology Data Exchange (ETDEWEB)
Chapman, Bryan Scott [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Gough, Sean T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2016-12-05
This report documents a validation of the MCNP6 Version 1.0 computer code on the high performance computing platform Moonlight, for operations at Los Alamos National Laboratory (LANL) that involve plutonium metals, oxides, and solutions. The validation is conducted using the ENDF/B-VII.1 continuous energy group cross section library at room temperature. The results are for use by nuclear criticality safety personnel in performing analysis and evaluation of various facility activities involving plutonium materials.
Modelling of a proton spot scanning system using MCNP6
International Nuclear Information System (INIS)
Ardenfors, O; Gudowska, I; Dasu, A; Kopeć, M
2017-01-01
The aim of this work was to model the characteristics of a clinical proton spot scanning beam using Monte Carlo simulations with the code MCNP6. The proton beam was defined using parameters obtained from beam commissioning at the Skandion Clinic, Uppsala, Sweden. Simulations were evaluated against measurements for proton energies between 60 and 226 MeV with regard to range in water, lateral spot sizes in air and absorbed dose depth profiles in water. The model was also used to evaluate the experimental impact of lateral signal losses in an ionization chamber through simulations using different detector radii. Simulated and measured distal ranges agreed within 0.1 mm for R 90 and R 80 , and within 0.2 mm for R 50 . The average absolute difference of all spot sizes was 0.1 mm. The average agreement of absorbed dose integrals and Bragg-peak heights was 0.9%. Lateral signal losses increased with incident proton energy with a maximum signal loss of 7% for 226 MeV protons. The good agreement between simulations and measurements supports the assumptions and parameters employed in the presented Monte Carlo model. The characteristics of the proton spot scanning beam were accurately reproduced and the model will prove useful in future studies on secondary neutrons. (paper)
Monte Carlo modeling of ion chamber performance using MCNP.
Wallace, J D
2012-12-01
Ion Chambers have a generally flat energy response with some deviations at very low (2 MeV) energies. Some improvements in the low energy response can be achieved through use of high atomic number gases, such as argon and xenon, and higher chamber pressures. This work looks at the energy response of high pressure xenon-filled ion chambers using the MCNP Monte Carlo package to develop geometric models of a commercially available high pressure ion chamber (HPIC). The use of the F6 tally as an estimator of the energy deposited in a region of interest per unit mass, and the underlying assumptions associated with its use are described. The effect of gas composition, chamber gas pressure, chamber wall thickness, and chamber holder wall thicknesses on energy response are investigated and reported. The predicted energy response curve for the HPIC was found to be similar to that reported by other investigators. These investigations indicate that improvements to flatten the overall energy response of the HPIC down to 70 keV could be achieved through use of 3 mm-thick stainless steel walls for the ion chamber.
Study of bremsstrahlung photons in bulk target using MCNP code
Directory of Open Access Journals (Sweden)
S. Sangaroon
2017-11-01
Full Text Available The aim of this research was to study the feasibility of bremsstrahlung photon production in target bombarded by 1 GeV electrons. The calculations were performed by the Monte Carlo code MCNP. Six target materials with densities between 2 and 20 g/cm3 were studied. The bremsstrahlung photon flux is high for the target density above 8 g/cm3. Copper is the best target for 1 GeV electron beam due to high bremsstrahlung photon production, low scattering and low transmission electron flux. The copper target was altered to have different thicknesses between 0.01 and 2.5 cm. The results showed that the bremsstrahlung photon flux significantly increased when the target thickness increased from 0.01 to 1.5 cm. The angular distribution of the bremsstrahlung photons with angles between 0 and 120 degrees was determined for copper target. The maximum angle of the photon scattering was about 20 degree.
Monte Carlo modelling of large scale NORM sources using MCNP.
Wallace, J D
2013-12-01
The representative Monte Carlo modelling of large scale planar sources (for comparison to external environmental radiation fields) is undertaken using substantial diameter and thin profile planar cylindrical sources. The relative impact of source extent, soil thickness and sky-shine are investigated to guide decisions relating to representative geometries. In addition, the impact of source to detector distance on the nature of the detector response, for a range of source sizes, has been investigated. These investigations, using an MCNP based model, indicate a soil cylinder of greater than 20 m diameter and of no less than 50 cm depth/height, combined with a 20 m deep sky section above the soil cylinder, are needed to representatively model the semi-infinite plane of uniformly distributed NORM sources. Initial investigation of the effect of detector placement indicate that smaller source sizes may be used to achieve a representative response at shorter source to detector distances. Crown Copyright © 2013. Published by Elsevier Ltd. All rights reserved.
Nuclear power development and nuclear data activities in Malaysia
Energy Technology Data Exchange (ETDEWEB)
Gui Ah Auu [Malaysian Institute for Nuclear Technology Research, Ministry of Science, Technology and the Environment, Selangor (Malaysia)
1999-03-01
In this paper, research activities on nuclear power requirement carried out jointly by MINT and other organizations are described. Also discussed are activities on neutronics such as TRIGA reactor fuel management, storage pool criticality, and reactor fuel transfer cask calculations. In addition, recent work on radiation transport activities in MINT such as skyshine and photon phantom dose calculations using the MCNP and MRIPP computer codes are presented. Finally, nuclear data measurement works by researchers in Malaysian universities are described. (author)
Nuclear power development and nuclear data activities in Malaysia
International Nuclear Information System (INIS)
Gui Ah Auu
1999-01-01
In this paper, research activities on nuclear power requirement carried out jointly by MINT and other organizations are described. Also discussed are activities on neutronics such as TRIGA reactor fuel management, storage pool criticality, and reactor fuel transfer cask calculations. In addition, recent work on radiation transport activities in MINT such as skyshine and photon phantom dose calculations using the MCNP and MRIPP computer codes are presented. Finally, nuclear data measurement works by researchers in Malaysian universities are described. (author)
Comparison over the nuclear analysis of the HCLL blanket for the European DEMO
Energy Technology Data Exchange (ETDEWEB)
Jaboulay, Jean-Charles, E-mail: jean-charles.jaboulay@cea.fr [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Aiello, Giacomo; Aubert, Julien [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Villari, Rosaria [ENEA, UTFUS-TECN, Via E. Fermi 4, 00044 Frascati, Rome (Italy); Fischer, Ulrich [Karlsruhe Institute of Technology, 76344 Eggenstein-Leopoldshafen, Karlsruhe (Germany)
2016-11-01
Highlights: • A complete nuclear analysis of the DEMO HCLL has been carried out at CEA with the TRIPOLI-4{sup ®} Monte Carlo code. • The DEMO tokamak model was generated by the CAD import tool McCad. • The HCLL blankets were implemented using a previous MCNP model developed at ENEA. • A good agreement is observed between the results obtained at CEA with TRIPOLI-4 and JEFF-3.1.1 and whose obtained at ENEA with MCNP and FENDL-2.1. - Abstract: This paper presents the comparison over the nuclear analysis of the European DEMO with HCLL blanket carried out with the TRIPOLI-4{sup ®} Monte Carlo code and the JEFF-3.1.1 nuclear data library and with the MCNP5 Monte Carlo code and the FENDL-2.1 nuclear data library. The MCNP5 analysis was conducted firstly by ENEA with a detailed 3D model describing all the HCLL blanket internal structures. This MCNP5 model was converted into TRIPOLI-4{sup ®} representation for performing the nuclear analysis at CEA with the objective to demonstrate consistency between both analyses. A very good agreement was obtained for all of the relevant nuclear responses (neutron wall loading, tritium breeding ratio, nuclear heating, neutron flux distribution, etc.), validating CEA’s nuclear analysis approach, based on TRIPOLI-4{sup ®} Monte Carlo code and JEFF-3.1.1 nuclear data library, for the European DEMO.
Comparison over the nuclear analysis of the HCLL blanket for the European DEMO
International Nuclear Information System (INIS)
Jaboulay, Jean-Charles; Aiello, Giacomo; Aubert, Julien; Villari, Rosaria; Fischer, Ulrich
2016-01-01
Highlights: • A complete nuclear analysis of the DEMO HCLL has been carried out at CEA with the TRIPOLI-4"® Monte Carlo code. • The DEMO tokamak model was generated by the CAD import tool McCad. • The HCLL blankets were implemented using a previous MCNP model developed at ENEA. • A good agreement is observed between the results obtained at CEA with TRIPOLI-4 and JEFF-3.1.1 and whose obtained at ENEA with MCNP and FENDL-2.1. - Abstract: This paper presents the comparison over the nuclear analysis of the European DEMO with HCLL blanket carried out with the TRIPOLI-4"® Monte Carlo code and the JEFF-3.1.1 nuclear data library and with the MCNP5 Monte Carlo code and the FENDL-2.1 nuclear data library. The MCNP5 analysis was conducted firstly by ENEA with a detailed 3D model describing all the HCLL blanket internal structures. This MCNP5 model was converted into TRIPOLI-4"® representation for performing the nuclear analysis at CEA with the objective to demonstrate consistency between both analyses. A very good agreement was obtained for all of the relevant nuclear responses (neutron wall loading, tritium breeding ratio, nuclear heating, neutron flux distribution, etc.), validating CEA’s nuclear analysis approach, based on TRIPOLI-4"® Monte Carlo code and JEFF-3.1.1 nuclear data library, for the European DEMO.
MCNP to study the BF3 detection efficiency
International Nuclear Information System (INIS)
Castro, Vinicius A.; Cavalieri, Tassio A.; Siqueira, Paulo T.D.; Fedorenko, Giuliana G.; Coelho, Paulo R.P.; Madi Filho, Tufic
2011-01-01
One of the main parameters to monitor on the employment of the Boron Neutron Capture Therapy (BNCT) is the thermal neutron flux. It can be performed by different techniques such as the activation analysis and the detection by a Boron Trifluoride detector (BF 3 ). BF 3 detector is a real time neutron flux detector which retrieves results in real time. It is however necessary to study the efficiency of the BF 3 detectors when they are exposed to fields of different neutron energy spectra. BF 3 is known to have high efficiency for thermal neutrons (with energy up to 0.5 eV) due the presence of 10 B atoms in the detector. However, one must also understand how this detector interacts with other neutron energy ranges (epithermal and fast). This work shows the experiment and a set of associated simulations carried out in order to evaluate the BF 3 detector efficiency dependence on neutron energy spectra. A set of experiments was conducted in which a BF 3 detector was submitted to different mixed fields (field containing gamma rays and neutrons). These fields were generated by the interposition of paraffin layers with distinct thicknesses between the Am-Be source and the BF 3 detector. The BF 3 detector responses were recorded according to the number of paraffin planes used. MCNP simulations were also performed to study the detector responses on such experimental conditions. It has been possible to achieve the intended goal of evaluating the BF 3 detector response to different mixed irradiation fields. (author)
Analysis of neutron dose rates on RGTT200K core using MCNP5
International Nuclear Information System (INIS)
Suwoto; Zuhair
2016-01-01
The conceptual design of RGTT200K (High Temperature Gas-cooled Reactor of 200 MWth Cogeneration) is the non-annular cylindrical reactor core with TRISO kernel coated fuel particles in the form of balls called pebble and cooled by helium gas. The RGTT200K reactor core design adopts high temperature gas cooled reactor (HTGR) technology with inherent passive safety. The RGTT200K spherical fuel called pebble fuel containing thousand of TRISO-coated fuel particles of uranium oxide (UO 2 ) 10 % enriched. TRISO coating comprises four layers, namely: porous carbon buffer layer, inner pyrolytic carbon layer (IPyC, Inner Pyrolytic Carbon), silicon carbide layer (SiC) and a layer of pyrolytic carbon outer portion (OPyC, Outer Pyrolytic Carbon). Modeling and analysis of preliminary calculation of neutron dose rate on normal operating temperature (T kernel =1200K) and accident temperature (T kernel =1800K) of the RGTT200K core were performed using Monte Carlo MCNP5v1.2 code. The continuous energy nuclear data cross-sections was taken from ENDF/B-VII, JENDL-4 and JEFF-3.1 nuclear data files . Double heterogeneity model in TRISO-coated fuel particles kernel and the pebble of RGTT200K core. By utilizing EGS99304 code, the 640 amount of energy group structures (SAND-II neutron group structures) is used in the neutron fluxes and spectrum calculation in RGTT200K reactor. The RGTT200K reactor core is divided into 25 zones (5 zones in radial and 10 zones in axial directions), while the modeling of radiation and biological shielding reactor RGTT200K are used to determine of preliminary neutron dose rate emitted by the neutron source with tally cards are available in the MCNP5v1.2 code. The calculation result analyses of the neutron dose rate distributions are determined using a conversion factor of flux-to-dose taken from International Commission on Radiological Protection, ICRP. The preliminary calculations result show that the neutrons dose rate using ICRP-74 conversion factor for
Comparison of TITAN hybrid deterministic transport code and MCNP5 for simulation of SPECT
International Nuclear Information System (INIS)
Royston, K.; Haghighat, A.; Yi, C.
2010-01-01
Traditionally, Single Photon Emission Computed Tomography (SPECT) simulations use Monte Carlo methods. The hybrid deterministic transport code TITAN has recently been applied to the simulation of a SPECT myocardial perfusion study. The TITAN SPECT simulation uses the discrete ordinates formulation in the phantom region and a simplified ray-tracing formulation outside of the phantom. A SPECT model has been created in the Monte Carlo Neutral particle (MCNP)5 Monte Carlo code for comparison. In MCNP5 the collimator is directly modeled, but TITAN instead simulates the effect of collimator blur using a circular ordinate splitting technique. Projection images created using the TITAN code are compared to results using MCNP5 for three collimator acceptance angles. Normalized projection images for 2.97 deg, 1.42 deg and 0.98 deg collimator acceptance angles had maximum relative differences of 21.3%, 11.9% and 8.3%, respectively. Visually the images are in good agreement. Profiles through the projection images were plotted to find that the TITAN results followed the shape of the MCNP5 results with some differences in magnitude. A timing comparison on 16 processors found that the TITAN code completed the calculation 382 to 2787 times faster than MCNP5. Both codes exhibit good parallel performance. (author)
MCNP-REN: a Monte Carlo tool for neutron detector design
International Nuclear Information System (INIS)
Abhold, M.E.; Baker, M.C.
2002-01-01
The development of neutron detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model fails to accurately predict detector response in common applications. For this reason, the general Monte Carlo code developed at Los Alamos National Laboratory, Monte Carlo N-Particle (MCNP), was modified to simulate the pulse streams that would be generated by a neutron detector and normally analyzed by a shift register. This modified code, MCNP-Random Exponentially Distributed Neutron Source (MCNP-REN), along with the Time Analysis Program, predicts neutron detector response without using the point reactor model, making it unnecessary for the user to decide whether or not the assumptions of the point model are met for their application. MCNP-REN is capable of simulating standard neutron coincidence counting as well as neutron multiplicity counting. Measurements of mixed oxide fresh fuel were taken with the Underwater Coincidence Counter, and measurements of highly enriched uranium reactor fuel were taken with the active neutron interrogation Research Reactor Fuel Counter and compared to calculation. Simulations completed for other detector design applications are described. The method used in MCNP-REN is demonstrated to be fundamentally sound and shown to eliminate the need to use the point model for detector performance predictions
Natto, S A; Lewis, D G; Ryde, S J
1998-01-01
The Monte Carlo computer code MCNP (version 4A) has been used to develop a personal computer-based model of the Swansea in vivo neutron activation analysis (IVNAA) system. The model included specification of the neutron source (252Cf), collimators, reflectors and shielding. The MCNP model was 'benchmarked' against fast neutron and thermal neutron fluence data obtained experimentally from the IVNAA system. The Swansea system allows two irradiation geometries using 'short' and 'long' collimators, which provide alternative dose rates for IVNAA. The data presented here relate to the short collimator, although results of similar accuracy were obtained using the long collimator. The fast neutron fluence was measured in air at a series of depths inside the collimator. The measurements agreed with the MCNP simulation within the statistical uncertainty (5-10%) of the calculations. The thermal neutron fluence was measured and calculated inside the cuboidal water phantom. The depth of maximum thermal fluence was 3.2 cm (measured) and 3.0 cm (calculated). The width of the 50% thermal fluence level across the phantom at its mid-depth was found to be the same by both MCNP and experiment. This benchmarking exercise has given us a high degree of confidence in MCNP as a tool for the design of IVNAA systems.
International Nuclear Information System (INIS)
Galicia A, J.; Francois L, J. L.; Bastida O, G. E.; Esquivel E, J.
2016-09-01
The development of the AZTLAN platform for the analysis and design of nuclear reactors is led by Instituto Nacional de Investigaciones Nucleares (ININ) and divided into four working groups, which have well-defined activities to achieve significant progress in this project individually and jointly. Within these working groups is the users group, whose main task is to use the codes that make up the AZTLAN platform to provide feedback to the developers, and in this way to make the final versions of the codes are efficient and at the same time reliable and easy to understand. In this paper we present the results provided by the AZNHEX v.1.0 code when simulating the core of a fast reactor cooled with sodium at steady state. The validation of these results is a fundamental part of the platform development and responsibility of the users group, so in this research the results obtained with AZNHEX are compared and analyzed with those provided by the Monte Carlo code MCNP-5, software worldwide used and recognized. A description of the methodology used with MCNP-5 is also presented for the calculation of the interest variables and the difference that is obtained with respect to the calculated with AZNHEX. (Author)
International Nuclear Information System (INIS)
White, Morgan C.
2000-01-01
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 and 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
Energy Technology Data Exchange (ETDEWEB)
White, Morgan C. [Univ. of Florida, Gainesville, FL (United States)
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
MCNP calculation for calibration curve of X-ray fluorescence analysis
International Nuclear Information System (INIS)
Tan Chunming; Wu Zhifang; Guo Xiaojing; Xing Guilai; Wang Zhentao
2011-01-01
Due to the compositional variation of the sample, linear relationship between the element concentration and fluorescent intensity will not be well maintained in most X-ray fluorescence analysis. To overcome this, we use MCNP program to simulate fluorescent intensity of Fe (0∼100% concentration range) within binary mixture of Cr and O which represent typical strong absorption and weak absorption conditions respectively. The theoretic calculation shows that the relationship can be described as a curve determined by parameter p and value of p can be obtained with given absorption coefficient of substrate elements and element under detection. MCNP simulation results are consistent with theoretic calculation. Our research reveals that MCNP program can calculate the Calibration Curve of X-ray fluorescence very well. (authors)
Simulation of Photon energy Spectra Using MISC, SOURCES, MCNP and GADRAS
International Nuclear Information System (INIS)
Tucker, Lucas P.; Shores, Erik F.; Myers, Steven C.; Felsher, Paul D.; Garner, Scott E.; Solomon, Clell J. Jr.
2012-01-01
The detector response functions included in the Gamma Detector Response and Analysis Software (GADRAS) are a valuable resource for simulating radioactive source emission spectra. Application of these response functions to the results of three-dimensional transport calculations is a useful modeling capability. Using a 26.2 kg shell of depleted uranium (DU) as a simple test problem, this work illustrates a method for manipulating current tally results from MCNP into the GAM file format necessary for a practical link to GADRAS detector response functions. MISC (MCNP Intrinsic Source Constructor) and SOURCES 4C were used to develop photon and neutron source terms for subsequent MCNP transport, and the resultant spectrum is shown to be in good agreement with that from GADRAS. A 1 kg DU sphere was also modeled with the method described here and showed similarly encouraging results.
MCNP: a general Monte Carlo code for neutron and photon transport. Version 3A. Revision 2
International Nuclear Information System (INIS)
Briesmeister, J.F.
1986-09-01
This manual is a practical guide for the use of our general-purpose Monte Carlo code MCNP. The first chapter is a primer for the novice user. The second chapter describes the mathematics, data, physics, and Monte Carlo simulation found in MCNP. This discussion is not meant to be exhaustive - details of the particular techniques and of the Monte Carlo method itself will have to be found elsewhere. The third chapter shows the user how to prepare input for the code. The fourth chapter contains several examples, and the fifth chapter explains the output. The appendices show how to use MCNP on particular computer systems at the Los Alamos National Laboratory and also give details about some of the code internals that those who wish to modify the code may find useful. 57 refs
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. Copyright © 2012 Elsevier Ltd. All rights reserved.
International Nuclear Information System (INIS)
Mark Dennis Usang; Mohd Hairie Rabir; Mohd Amin Sharifuldin Salleh; Mohamad Puad Abu
2012-01-01
MPI parallelism are implemented on a SUN Workstation for running MCNPX and on the High Performance Computing Facility (HPC) for running MCNP5. 23 input less obtained from MCNP Criticality Validation Suite are utilized for the purpose of evaluating the amount of speed up achievable by using the parallel capabilities of MPI. More importantly, we will study the economics of using more processors and the type of problem where the performance gain are obvious. This is important to enable better practices of resource sharing especially for the HPC facilities processing time. Future endeavours in this direction might even reveal clues for best MCNP5/ MCNPX coding practices for optimum performance of MPI parallelisms. (author)
Simulation of Photon energy Spectra Using MISC, SOURCES, MCNP and GADRAS
Energy Technology Data Exchange (ETDEWEB)
Tucker, Lucas P. [Los Alamos National Laboratory; Shores, Erik F. [Los Alamos National Laboratory; Myers, Steven C. [Los Alamos National Laboratory; Felsher, Paul D. [Los Alamos National Laboratory; Garner, Scott E. [Los Alamos National Laboratory; Solomon, Clell J. Jr. [Los Alamos National Laboratory
2012-08-14
The detector response functions included in the Gamma Detector Response and Analysis Software (GADRAS) are a valuable resource for simulating radioactive source emission spectra. Application of these response functions to the results of three-dimensional transport calculations is a useful modeling capability. Using a 26.2 kg shell of depleted uranium (DU) as a simple test problem, this work illustrates a method for manipulating current tally results from MCNP into the GAM file format necessary for a practical link to GADRAS detector response functions. MISC (MCNP Intrinsic Source Constructor) and SOURCES 4C were used to develop photon and neutron source terms for subsequent MCNP transport, and the resultant spectrum is shown to be in good agreement with that from GADRAS. A 1 kg DU sphere was also modeled with the method described here and showed similarly encouraging results.
Gonzales, Matthew Alejandro
The calculation of the thermal neutron Doppler temperature reactivity feedback co-efficient, a key parameter in the design and safe operation of advanced reactors, using first order perturbation theory in continuous energy Monte Carlo codes is challenging as the continuous energy adjoint flux is not readily available. Traditional approaches of obtaining the adjoint flux attempt to invert the random walk process as well as require data corresponding to all temperatures and their respective temperature derivatives within the system in order to accurately calculate the Doppler temperature feedback. A new method has been developed using adjoint-weighted tallies and On-The-Fly (OTF) generated continuous energy cross sections within the Monte Carlo N-Particle (MCNP6) transport code. The adjoint-weighted tallies are generated during the continuous energy k-eigenvalue Monte Carlo calculation. The weighting is based upon the iterated fission probability interpretation of the adjoint flux, which is the steady state population in a critical nuclear reactor caused by a neutron introduced at that point in phase space. The adjoint-weighted tallies are produced in a forward calculation and do not require an inversion of the random walk. The OTF cross section database uses a high order functional expansion between points on a user-defined energy-temperature mesh in which the coefficients with respect to a polynomial fitting in temperature are stored. The coefficients of the fits are generated before run- time and called upon during the simulation to produce cross sections at any given energy and temperature. The polynomial form of the OTF cross sections allows the possibility of obtaining temperature derivatives of the cross sections on-the-fly. The use of Monte Carlo sampling of adjoint-weighted tallies and the capability of computing derivatives of continuous energy cross sections with respect to temperature are used to calculate the Doppler temperature coefficient in a research
MCNP Modeling Results for Location of Buried TRU Waste Drums
International Nuclear Information System (INIS)
Steinman, D K; Schweitzer, J S
2006-01-01
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
Installation of Monte Carlo neutron and photon transport code system MCNP4
International Nuclear Information System (INIS)
Takano, Makoto; Sasaki, Mikio; Kaneko, Toshiyuki; Yamazaki, Takao.
1993-03-01
The continuous energy Monte Carlo code MCNP-4 including its graphic functions has been installed on the Sun-4 sparc-2 work station with minor corrections. In order to validate the installed MCNP-4 code, 25 sample problems have been executed on the work station and these results have been compared with the original ones. And, the most of the graphic functions have been demonstrated by using 3 sample problems. Further, additional 14 nuclides have been included to the continuous cross section library edited from JENDL-3. (author)
International Nuclear Information System (INIS)
Hussein, M.S.; Bonin, H.W.; Lewis, B.J.
2013-01-01
The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as Deuterium Critical Assembly, (DCA). The variations of the criticality factors and the coupling coefficients were investigated by changing of the water levels in the inner and outer cores. The numerical results of the model developed with MCNP5 code were validated and verified against published results and the mathematical model based on coupled reactor theory. (author)
Energy Technology Data Exchange (ETDEWEB)
Hussein, M.S.; Bonin, H.W.; Lewis, B.J., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca, E-mail: lewis-b@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)
2013-07-01
The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as Deuterium Critical Assembly, (DCA). The variations of the criticality factors and the coupling coefficients were investigated by changing of the water levels in the inner and outer cores. The numerical results of the model developed with MCNP5 code were validated and verified against published results and the mathematical model based on coupled reactor theory. (author)
Gamma irradiator dose mapping simulation using the MCNP code and benchmarking with dosimetry
International Nuclear Information System (INIS)
Sohrabpour, M.; Hassanzadeh, M.; Shahriari, M.; Sharifzadeh, M.
2002-01-01
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
International Nuclear Information System (INIS)
Kotegawa, Hiroshi; Sasamoto, Nobuo; Tanaka, Shun-ichi
1987-02-01
Both ''measured radioactive inventory due to neutron activation in the shield concrete of JPDR'' and ''measured intermediate and low energy neutron spectra penetrating through a graphite sphere'' are analyzed using a continuous energy model Monte Carlo code MCNP so as to estimate calculational accuracy of the code for neutron transport in thermal and epithermal energy regions. Analyses reveal that MCNP calculates thermal neutron spectra fairly accurately, while it apparently over-estimates epithermal neutron spectra (of approximate 1/E distribution) as compared with the measurements. (author)
Acceleration of the MCNP branch of the OCTOPUS depletion code system
Energy Technology Data Exchange (ETDEWEB)
Pijlgroms, B.J.; Hogenbirk, A.; Oppe, J. [Section Nuclear and Reactor Physics, ECN Nuclear Research, Petten (Netherlands)
1998-09-01
OCTOPUS depletion calculations using the 3D Monte Carlo spectrum code MCNP (Monte Carlo Code for Neutron and Photon Transport) require much computing time. In a former implementation, the time required by OCTOPUS to perform multi-zone calculations, increased roughly proportional to the number of burnable zones. By using a different method the situation has improved considerably. In the new implementation described here, the dependence of the computing time on the number of zones has been moved from the MCNP code to a faster postprocessing code. By this, the overall computing time will reduce substantially. 11 refs.
Acceleration of the MCNP branch of the OCTOPUS depletion code system
International Nuclear Information System (INIS)
Pijlgroms, B.J.; Hogenbirk, A.; Oppe, J.
1998-09-01
OCTOPUS depletion calculations using the 3D Monte Carlo spectrum code MCNP (Monte Carlo Code for Neutron and Photon Transport) require much computing time. In a former implementation, the time required by OCTOPUS to perform multi-zone calculations, increased roughly proportional to the number of burnable zones. By using a different method the situation has improved considerably. In the new implementation described here, the dependence of the computing time on the number of zones has been moved from the MCNP code to a faster postprocessing code. By this, the overall computing time will reduce substantially. 11 refs
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. Copyright © 2014 Elsevier Ltd. All rights reserved.
Development of gamma-ray absorption and scattering simulation platform based on MCNP
International Nuclear Information System (INIS)
Lai Wanchang; Chen Henggui; Zhang Zhen; Chen Xiaoqiang
2010-01-01
It describes a γ-ray absorption and scattering simulation platform centering on MCNP, and developed corresponding accessories on the basis of the MCNP. Simulation of this simulation platform can be 93 kinds of single-quality materials and 2-3 kinds of multi-element mixture absorption experiment, simulating the absorption thickness of 0-100cm, and the thickness increment in 0.001cm. The media of Scattering Simulation is from the Li to the Am, the angle between the simulation measuring degree and incident ray direction is from-90 to 90, the angle in increments in 1 degree. (authors)
International Nuclear Information System (INIS)
Pashchenko, A.B.; Wienke, H.; Ganesan, S.
1996-01-01
This document summarizes a neutron activation cross-section database processed in two formats as generated by F.M. Mann within the project of the Fusion Evaluated Nuclear Data Library (FENDL): in continuous energy format as used by the Monte Carlo neutron/photon transport code MCNP; and in 175 group multigroup format with VIT-E weighting spectrum, as used by the transmutation code REAC*2/3. The data are available from the IAEA Nuclear Data Section online via INTERNET by FTP command, or on magnetic tape. (author). 2 refs, 1 tab
International Nuclear Information System (INIS)
Craig, D.S.
1989-03-01
The Monte Carlo code MCNP was used to check the accuracy of the WIMS calculation of the resolved resonance capture rate in CANDU-type lattices. Reactivities, relative conversion ratios, and fast fission factors are compared with experiments. Values of ρ 28 and reaction rates for U-238 are given as a function of position in the fuel bundle. A check was made on the correction made in WIMS to allow for endcaps on the fuel bundles. (26 refs)
Directory of Open Access Journals (Sweden)
Volmert Ben
2016-01-01
Full Text Available 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.
Pozzi, S A
2002-01-01
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 ...
International Nuclear Information System (INIS)
2014-01-01
This document proposes a presentation and discussion of the main notions, issues, principles, or characteristics related to nuclear energy: radioactivity (presence in the environment, explanation, measurement, periods and activities, low doses, applications), fuel cycle (front end, mining and ore concentration, refining and conversion, fuel fabrication, in the reactor, back end with reprocessing and recycling, transport), the future of the thorium-based fuel cycle (motivations, benefits and drawbacks), nuclear reactors (principles of fission reactors, reactor types, PWR reactors, BWR, heavy-water reactor, high temperature reactor of HTR, future reactors), nuclear wastes (classification, packaging and storage, legal aspects, vitrification, choice of a deep storage option, quantities and costs, foreign practices), radioactive releases of nuclear installations (main released radio-elements, radioactive releases by nuclear reactors and by La Hague plant, gaseous and liquid effluents, impact of releases, regulation), the OSPAR Convention, management and safety of nuclear activities (from control to quality insurance, to quality management and to sustainable development), national safety bodies (mission, means, organisation and activities of ASN, IRSN, HCTISN), international bodies, nuclear and medicine (applications of radioactivity, medical imagery, radiotherapy, doses in nuclear medicine, implementation, the accident in Epinal), nuclear and R and D (past R and D programmes and expenses, main actors in France and present funding, main R and D axis, international cooperation)
Antoni, Rodolphe; Bourgois, Laurent
2017-12-01
energy ranges. Accordingly, special care has to be taken in setting choice for calculating electron dose distribution with MCNP6, in particular with regards to dosimetry or nuclear medicine applications.
Program for the Generation of MCNP Inputs from State Files of CAREM
International Nuclear Information System (INIS)
Leszczynski, Francisco; Lopasso, Edmundo; Villarino, E
2000-01-01
The objective of this work is the development and tests of detailed input data for the Monte Carlo program MCNP, to be able of model the core of CAREM reactor, with the detail included on the updated models, for having available a calculation system that allow the production of confident results to be compared with results obtained with the system used today for designing the CAREM reactor core (CONDOR-CITVAP).The model includes the possibility of temperature and coolant density, and temperature and numeric densities of fuel.The detail consists of 21 different fuel elements (symmetry 3) and 14 axial zones.Results of comparisons of reactivity and power pick factors are presented, between MCNP and CONDOR-CITVAP.On average, these results show an acceptable agreement for all the compared parameters.It is described, also, the interface CONDOR-CITVAP-MCNP program, that has been developed for generating inputs of materials for MCNP, from outputs of CONDOR and CITVAP, for different reactor states
An improved algorithm to convert CAD model to MCNP geometry model based on STEP file
International Nuclear Information System (INIS)
Zhou, Qingguo; Yang, Jiaming; Wu, Jiong; Tian, Yanshan; Wang, Junqiong; Jiang, Hai; Li, Kuan-Ching
2015-01-01
Highlights: • Fully exploits common features of cells, making the processing efficient. • Accurately provide the cell position. • Flexible to add new parameters in the structure. • Application of novel structure in INP file processing, conveniently evaluate cell location. - Abstract: MCNP (Monte Carlo N-Particle Transport Code) is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron, or coupled neutron/photon/electron transport. Its input file, the INP file, has the characteristics of complicated form and is error-prone when describing geometric models. Due to this, a conversion algorithm that can solve the problem by converting general geometric model to MCNP model during MCNP aided modeling is highly needed. In this paper, we revised and incorporated a number of improvements over our previous work (Yang et al., 2013), which was proposed and targeted after STEP file and INP file were analyzed. Results of experiments show that the revised algorithm is more applicable and efficient than previous work, with the optimized extraction of geometry and topology information of the STEP file, as well as the production efficiency of output INP file. This proposed research is promising, and serves as valuable reference for the majority of researchers involved with MCNP-related researches
TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR
International Nuclear Information System (INIS)
Kurosawa, M.
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 54 Mn and 60 Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data. (authors)
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.
MCNP modelling of scintillation-detector gamma-ray spectra from natural radionuclides
Hendriks, Peter; Maucec, M; de Meijer, RJ
gamma-ray spectra of natural radionuclides are simulated for a BGO detector in a borehole geometry using the Monte Carlo code MCNP. All gamma-ray emissions of the decay of K-40 and the series of Th-232 and U-238 are used to describe the source. A procedure is proposed which excludes the
Studi Model Benchmark Mcnp6 Dalam Perhitungan Reaktivitas Batang Kendali Htr-10
Jupiter S.Pane, Zuhair, Suwoto, Putranto Ilham Yazid
2016-01-01
STUDI MODEL BENCHMARK MCNP6 DALAM PERHITUNGAN REAKTIVITAS BATANG KENDALI HTR-10. Dalam operasi reaktor nuklir, sistem batang kendali memainkan peranan yang sangat penting karena didesain untuk mengendalikan reaktivitas teras dan memadamkan reaktor. Nilai reaktivitas batang kendali harus diprediksi secara akurat melalui eksperimen dan perhitungan. Makalah ini mendiskusikan model Benchmark dalam perhitungan reaktivitas batang kendali reaktor HTR-10. Perhitungan dikerjakan dengan program transpo...
Considerations of MCNP Monte Carlo code to be used as a radiotherapy treatment planning tool.
Juste, B; Miro, R; Gallardo, S; Verdu, G; Santos, A
2005-01-01
The present work has simulated the photon and electron transport in a Theratron 780® (MDS Nordion)60Co radiotherapy unit, using the Monte Carlo transport code, MCNP (Monte Carlo N-Particle). This project explains mainly the different methodologies carried out to speedup calculations in order to apply this code efficiently in radiotherapy treatment planning.
Energy Technology Data Exchange (ETDEWEB)
Thanh, Tran Thien; Tao, Chau Van; Loan, Truong Thi Hong; Nhon, Mai Van; Chuong, Huynh Dinh; Au, Bui Hai [Vietnam National Univ., Ho Chi Minh City (Viet Nam). Dept. of Nuclear Physics
2012-12-15
The accuracy of the coincidence-summing corrections in gamma spectrometry depends on the total efficiency calibration that is hardly obtained over the whole energy as the required experimental conditions are not easily attained. Monte Carlo simulations using MCNP5 code was performed in order to estimate the affect of the shielding to total efficiency. The effect of HPGe response are also shown. (orig.)
Sogbadji, R B M; Abrefah, R G; Nyarko, B J B; Akaho, E H K; Odoi, H C; Attakorah-Birinkorang, S
2014-08-01
The americium-beryllium neutron irradiation facility at the National Nuclear Research Institute (NNRI), Ghana, was re-designed with four 20 Ci sources using Monte Carlo N-Particle (MCNP) code to investigate the maximum amount of flux that is produced by the combined sources. The results were compared with a single source Am-Be irradiation facility. The main objective was to enable us to harness the maximum amount of flux for the optimization of neutron activation analysis and to enable smaller sample sized samples to be irradiated. Using MCNP for the design construction and neutronic performance calculation, it was realized that the single-source Am-Be design produced a thermal neutron flux of (1.8±0.0007)×10(6) n/cm(2)s and the four-source Am-Be design produced a thermal neutron flux of (5.4±0.0007)×10(6) n/cm(2)s which is a factor of 3.5 fold increase compared to the single-source Am-Be design. The criticality effective, k(eff), of the single-source and the four-source Am-Be designs were found to be 0.00115±0.0008 and 0.00143±0.0008, respectively. Copyright © 2014 Elsevier Ltd. All rights reserved.
MCNP5 CRITICALITY VALIDATION AND BIAS FOR INTERMEDIATE ENRICHED URANIUM SYSTEMS
International Nuclear Information System (INIS)
Finfrock, S.H.
2009-01-01
The purpose of this analysis is to validate the Monte Carlo N-Particle 5 (MCNP5) code Version 1.40 (LA-UR-03-1987, 2005) and its cross-section database for k-code calculations of intermediate enriched uranium systems on INTEL(reg s ign) processor based PC's running any version of the WINDOWS operating system. Configurations with intermediate enriched uranium were modeled with the moderator range of 39 (le) H/Fissile (le) 1438. See Table 2-1 for brief descriptions of selected cases and Table 3-1 for the range of applicability for this validation. A total of 167 input cases were evaluated including bare and reflected systems in a single body or arrays. The 167 cases were taken directly from the previous (Version 4C [Lan 2005]) validation database. Section 2.0 list data used to calculate k-effective (k eff ) for the 167 experimental criticality benchmark cases using the MCNP5 code v1.40 and its cross section database. Appendix B lists the MCNP cross-section database entries validated for use in evaluating the intermediate enriched uranium systems for criticality safety. The dimensions and atom densities for the intermediate enriched uranium experiments were taken from NEA/NSC/DOC(95)03, September 2005, which will be referred to as the benchmark handbook throughout the report. For these input values, the experimental benchmark k eff is approximately 1.0. The MCNP validation computer runs ran to an accuracy of approximately ± 0.001. For the cases where the reported benchmark k eff was not equal to 1.0000 the MCNP calculational results were normalized. The difference between the MCNP validation computer runs and the experimentally measured k eff is the MCNP5 v1.40 bias. The USLSTATS code (ORNL 1998) was utilized to perform the statistical analysis and generate an acceptable maximum k eff limit for calculations of the intermediate enriched uranium type systems.
Energy Technology Data Exchange (ETDEWEB)
Galicia A, J.; Francois L, J. L.; Bastida O, G. E. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Esquivel E, J., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)
2016-09-15
The development of the AZTLAN platform for the analysis and design of nuclear reactors is led by Instituto Nacional de Investigaciones Nucleares (ININ) and divided into four working groups, which have well-defined activities to achieve significant progress in this project individually and jointly. Within these working groups is the users group, whose main task is to use the codes that make up the AZTLAN platform to provide feedback to the developers, and in this way to make the final versions of the codes are efficient and at the same time reliable and easy to understand. In this paper we present the results provided by the AZNHEX v.1.0 code when simulating the core of a fast reactor cooled with sodium at steady state. The validation of these results is a fundamental part of the platform development and responsibility of the users group, so in this research the results obtained with AZNHEX are compared and analyzed with those provided by the Monte Carlo code MCNP-5, software worldwide used and recognized. A description of the methodology used with MCNP-5 is also presented for the calculation of the interest variables and the difference that is obtained with respect to the calculated with AZNHEX. (Author)
Performance of the improved version of Monte Carlo code A 3MCNP for large-scale shielding problems
International Nuclear Information System (INIS)
Omura, M.; Miyake, Y.; Hasegawa, T.; Ueki, K.; Sato, O.; Haghighat, A.; Sjoden, G. E.
2005-01-01
A 3MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, which automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic 'importance' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A 3MCNP uses the three-dimensional (3-D) Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A 3MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A 3MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A 3MCNP (referred to as A 3MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A 3MCNPV for a concrete cask neutron and gamma-ray shielding problem, and a PWR dosimetry problem. (authors)
RBMK fuel channel blockage analysis by MCNP5, DRAGON and RELAP5-3D codes
International Nuclear Information System (INIS)
Parisi, C.; D'Auria, F.
2007-01-01
The aim of this work was to perform precise criticality analyses by Monte-Carlo code MCNP5 for a Fuel Channel (FC) flow blockage accident, considering as calculation domain a single FC and a 3x3 lattice of RBMK cells. Boundary conditions for MCNP5 input were derived by a previous transient calculation by state-of-the-art codes HELIOS/RELAP5-3D. In a preliminary phase, suitable MCNP5 models of a single cell and of a small lattice of RBMK cells were set-up; criticality analyses were performed at reference conditions for 2.0% and 2.4% enriched fuel. These analyses were compared with results obtained by University of Pisa (UNIPI) using deterministic transport code DRAGON and with results obtained by NIKIET Institute using MCNP4C. Then, the changes of the main physical parameters (e.g. fuel and water/steam temperature, water density, graphite temperature) at different time intervals of the FC blockage transient were evaluated by a RELAP5-3D calculation. This information was used to set up further MCNP5 inputs. Criticality analyses were performed for different systems (single channel and lattice) at those transient' states, obtaining global criticality versus transient time. Finally the weight of each parameter's change (fuel overheating and channel voiding) on global criticality was assessed. The results showed that reactivity of a blocked FC is always negative; nevertheless, when considering the effect of neighboring channels, the global reactivity trend reverts, becoming slightly positive or not changing at all, depending in inverse relation to the fuel enrichment. (author)
A graphical user interface for diagnostic radiology dosimetry using Monte Carlo (MCNP) simulation
International Nuclear Information System (INIS)
Collins, P.J.; Gorbatkov, D.; Schultz, F.W.
2000-01-01
Monte Carlo methods (for example, MCNP, EGGS4) are the 'gold standard' for both external and internal dosimetry in humans. These powerful simulation tools are, however, general-purpose codes and consequently do not provide a simple user interface for specific dosimetry tasks. We have developed a graphical user interface, for external radiation dosimetry (diagnostic radiology) using MCNP and an anthropomorphic mathematical phantom (Adam/Eva), which enables convenient modification and processing of the MCNP input and output files. The input form displays a colour coded, 3D representation of the phantom with a superimposed 'beam' for the required x-ray projection. The phantom can be rotated through 360 degrees and a transverse section at the level of the mid-point of the beam is also displayed. Text fields enable entry of input data (beam dimensions, source position, kVp, total filtration, focus-to-skin distance). A pull-down menu enables the user to select from 22 standard radiographic views. A standard projection can be modified, or new projection data entered if required. The input program modifies the MCNP input file and initiates processing. An output form displays the organ doses, normalised to unit skin entrance dose (with backscatter) (SED). The user can also enter the SED (calculated or measured) for a particular machine, to obtain the effective dose. To validate the program, the results for a PA Chest study (80 kVp, 2.5 mm Al total filtration) were compared with NRPB data (Jones and Wall, 1985). In conclusion, a convenient and reliable graphical user interface has been developed for MCNP, which enables dosimetry calculation for a full range of diagnostic radiological studies. (author)
International Nuclear Information System (INIS)
Boulaich, Y.; El Bardouni, T.; Erradi, L.; Chakir, E.; Boukhal, H.; Nacir, B.; El Younoussi, C.; El Bakkari, B.; Merroun, O.; Zoubair, M.
2011-01-01
Highlights: → In the present work, we have analyzed the CREOLE experiment on the reactivity temperature coefficient (RTC) by using the three-dimensional continuous energy code (MCNP5) and the last updated nuclear data evaluations. → Calculation-experiment discrepancies of the RTC were analyzed and the results have shown that the JENDL3.3 and JEFF3.1 evaluations give the most consistent values. → In order to specify the source of the relatively large discrepancy in the case of ENDF-BVII nuclear data evaluation, the k eff discrepancy between ENDF-BVII and JENDL3.3 was decomposed by using sensitivity and uncertainty analysis technique. - Abstract: In the present work, we analyze the CREOLE experiment on the reactivity temperature coefficient (RTC) by using the three-dimensional continuous energy code (MCNP5) and the last updated nuclear data evaluations. This experiment performed in the EOLE critical facility located at CEA/Cadarache, was mainly dedicated to the RTC studies for both UO 2 and UO 2 -PuO 2 PWR type lattices covering the whole temperature range from 20 deg. C to 300 deg. C. We have developed an accurate 3D model of the EOLE reactor by using the MCNP5 Monte Carlo code which guarantees a high level of fidelity in the description of different configurations at various temperatures taking into account their consequence on neutron cross section data and all thermal expansion effects. In this case, the remaining error between calculation and experiment will be awarded mainly to uncertainties on nuclear data. Our own cross section library was constructed by using NJOY99.259 code with point-wise nuclear data based on ENDF-BVII, JEFF3.1 and JENDL3.3 evaluation files. The MCNP model was validated through the axial and radial fission rate measurements at room and hot temperatures. Calculation-experiment discrepancies of the RTC were analyzed and the results have shown that the JENDL3.3 and JEFF3.1 evaluations give the most consistent values; the discrepancy is
International Nuclear Information System (INIS)
Vickers, Lisa
2003-01-01
The government of Mexico has expressed interest in utilizing the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed-oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18 to 30% depending on the fuel loading cycle. MOX fuel is expected to increase the neutron fluence, flux, fuel centerline temperature, reactor core pressure, and yield higher energy neutrons.There is concern that a core with a fraction of MOX fuel (i.e., increased 239 Pu wt%) would increase the radiation damage displacements per atom per second (dpa-s -1 ) in steel within the core shroud and vessel wall as compared to only conventional, enriched UOX fuel in the core. The evaluation of radiation damage within the core shroud and vessel wall is a concern because of the potentially adverse affect to personnel and public safety, environment, and operating life of the reactor.The primary uniqueness of this paper is the computation of radiation damage (dpa-s -1 ) using NJOY99-processed cross sections for steel within the core shroud and vessel wall. Specifically, the unique radiation damage results are several orders of magnitude greater than results of previous works. In addition, the conclusion of this paper was that the addition of the maximum fraction of one-third MOX fuel to the LV1 BWR core did significantly increase the radiation damage in steel within the core shroud and vessel wall such that without mitigation of radiation damage by periodic thermal annealing or reduction in operating parameters such as neutron fluence, core temperature, and pressure, it posed a potentially adverse affect to the personnel and public safety, environment, and operating life of the reactor
Modeling of LVRF critical experiments in ZED-2 using WIMS9A/PANTHER and MCNP5
International Nuclear Information System (INIS)
Sissaoui, M.T.; Carlson, P.A.; Lebenhaft, J.R.
2009-01-01
The accuracy of WIMS9A/PANTHER and MCNP5 in modeling D 2 O-moderated, and H 2 O-, D 2 O- or air-cooled, doubly heterogeneous lattices of fuel clusters was demonstrated using Low Void Reactivity Fuel (LVRF) substitution experiments in the ZED-2 critical facility. MCNP5 with ENDF/B-VI (Release 5) underpredicted k eff but gave excellent coolant void reactivity (CVR) bias values. WIMS9A/PANTHER with JEF-2.2 overpredicted k eff and underpredicted the CVR bias relative to MCNP5 by 100-200 pcm. Both codes reproduced the measured axial and radial flux shapes accurately
Zhang, Lei; Jia, Mingchun; Gong, Junjun; Xia, Wenming
2017-08-01
The linear attenuation coefficient, mass attenuation coefficient and mean free path of various Lead-Boron Polyethylene (PbBPE) samples which can be used as the photon shielding materials in marine reactor have been simulated using the Monte Carlo N-Particle (MCNP)-5 code. The MCNP simulation results are in good agreement with the XCOM values and the reported experimental data for source Cesium-137 and Cobalt-60. Thus, this method based on MCNP can be used to simulate the photon attenuation characteristics of various types of PbBPE materials.
Energy Technology Data Exchange (ETDEWEB)
Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Aguilar H, F., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)
2015-09-15
The main purpose of this paper is to obtain a model of the reactor core TRIGA Mark III that accurately represents the real operating conditions to 1 M Wth, using the Monte Carlo code MCNP5. To provide a more detailed analysis, different models of the reactor core were realized by simulating the control rods extracted and inserted in conditions in cold (293 K) also including an analysis for shutdown margin, so that satisfied the Operation Technical Specifications. The position they must have the control rods to reach a power equal to 1 M Wth, were obtained from practice entitled Operation in Manual Mode performed at Instituto Nacional de Investigaciones Nucleares (ININ). Later, the behavior of the K{sub eff} was analyzed considering different temperatures in the fuel elements, achieving calculate subsequently the values that best represent the actual reactor operation. Finally, the calculations in the developed model for to obtain the distribution of average flow of thermal, epithermal and fast neutrons in the six new experimental facilities are presented. (Author)
International Nuclear Information System (INIS)
Pecchia, Marco; Vasiliev, Alexander; Ferroukhi, Hakim; Pautz, Andreas
2017-01-01
Highlights: • Evaluation of neutron source importance for a given tally. • Assessment of ex-core detector response plus its uncertainty. • Direct use of neutron track evaluated by a Monte Carlo neutron transport code. - Abstract: The ex-core neutron detectors are commonly used to control reactor power in light water reactors. Therefore, it is relevant to understand the importance of a neutron source to the ex-core detectors response. In mathematical terms, this information is conveniently represented by the so called weighting functions. A new methodology based on the MCNP code for evaluating the weighting functions starting from the neutron history database is presented in this work. A simultaneous evaluation of the weighting functions in a user-given Cartesian coverage mesh is the main advantage of the method. The capability to generate weighting functions simultaneously in both spatial and energy ranges is the innovative part of this work. Then, an interpolation tool complements the methodology, allowing the generation of weighting functions up to the pin-by-pin fuel segment, where a direct evaluation is not possible due to low statistical precision. A comparison to reference results provides a verification of the methodology. Finally, an application to investigate the role of ex-core detectors spatial location and core burnup for a Swiss nuclear power plant is provided.
Yang, Zi-Yi; Tsai, Pi-En; Lee, Shao-Chun; Liu, Yen-Chiang; Chen, Chin-Cheng; Sato, Tatsuhiko; Sheu, Rong-Jiun
2017-09-01
The dose distributions from proton pencil beam scanning were calculated by FLUKA, GEANT4, MCNP, and PHITS, in order to investigate their applicability in proton radiotherapy. The first studied case was the integrated depth dose curves (IDDCs), respectively from a 100 and a 226-MeV proton pencil beam impinging a water phantom. The calculated IDDCs agree with each other as long as each code employs 75 eV for the ionization potential of water. The second case considered a similar condition of the first case but with proton energies in a Gaussian distribution. The comparison to the measurement indicates the inter-code differences might not only due to different stopping power but also the nuclear physics models. How the physics parameter setting affect the computation time was also discussed. In the third case, the applicability of each code for pencil beam scanning was confirmed by delivering a uniform volumetric dose distribution based on the treatment plan, and the results showed general agreement between each codes, the treatment plan, and the measurement, except that some deviations were found in the penumbra region. This study has demonstrated that the selected codes are all capable of performing dose calculations for therapeutic scanning proton beams with proper physics settings.
Neutronics modeling of TRIGA reactor at the University of Utah using agent, KENO6 and MCNP5 codes
International Nuclear Information System (INIS)
Yang, X.; Xiao, S.; Choe, D.; Jevremovic, T.
2010-01-01
The TRIGA reactor at the University of Utah is modelled in 2D using the AGENT state-of-the-art methodology based on the Method of Characteristics (MOC) and R-function theory supporting detailed reactor analysis of reactor geometries of any type. The TRIGA reactor is also modelled using KENO6 and MCNP5 for comparison. The spatial flux and reaction rates distribution are visualized by AGENT graphics support. All methodologies are in use in to study the effect of different fuel configurations in developing practical educational exercises for students studying reactor physics. At the University of Utah we train graduate and undergraduate students in obtaining the Nuclear Regulatory Commission license in operating the TRIGA reactor. The computational models as developed are in support of these extensive training classes and in helping students visualize the reactor core characteristics in regard to neutron transport under various operational conditions. Additionally, the TRIGA reactor is under the consideration for power uprate; this fleet of computational tools once benchmarked against real measurements will provide us with validated 3D simulation models for simulating operating conditions of TRIGA. (author)
Comparison between Nuclear Data Libraries of Different Density of Data for H in Light Water
International Nuclear Information System (INIS)
Torres, Lourdes; Gillette, Victor
2003-01-01
We introduce the results of comparison between nuclear data libraries at different density of data.Nuclear data libraries were produced for hydrogen (H) in light water at different density of data.These libraries were produced using the NJOY nuclear data processing system.With this code we produce pointwise cross sections and related quantities, in the ENDF format, and in the ACE format for MCNP.Experimental neutron spectrum was compared with MCNP4C simulations, based on the produced libraries and calculation time
Energy Technology Data Exchange (ETDEWEB)
Pozuelo, F.; Querol, A.; Gallardo, S.; Rodenas, J.; Verdu, G.
2012-07-01
In this case, used codes PENELOPE MCNP5, based on the Monte Carlo method for x-ray spectrum taking into account the characteristics of the x-ray tube. In order to achieve a greater fit of simulated by the theoretical spectrum. It carried out a sensitivity analysis of the parameters available in both codes. The obtaining of the simulated spectrum could lead to an improvement in quality control of the x-ray tube to incorporate it as a method complementary to techniques.
International Nuclear Information System (INIS)
Perkasa, Y. S.; Waris, A.; Kurniadi, R.; Su'ud, Z.
2014-01-01
Comparative studies of actinide and sub-actinide fission cross section calculation from MCNP6 and TALYS have been conducted. In this work, fission cross section resulted from MCNP6 prediction will be compared with result from TALYS calculation. MCNP6 with its event generator CEM03.03 and LAQGSM03.03 have been validated and verified for several intermediate and heavy nuclides fission reaction data and also has a good agreement with experimental data for fission reaction that induced by photons, pions, and nucleons at energy from several ten of MeV to about 1 TeV. The calculation that induced within TALYS will be focused mainly to several hundred MeV for actinide and sub-actinide nuclides and will be compared with MCNP6 code and several experimental data from other evaluator
International Nuclear Information System (INIS)
Park, W.S.; Lee, K.M.; Lee, C.S.; Lee, J.T.; Oh, S.K.
1992-01-01
In this work, the validity and quantitative uncertainty of WIMS (KAERI) - VENTURE code system for the design and analysis of KMRR core was tried to be inferred using a well known benchmark code, MCNP. WIMS (KAERI) showed an excellent agreement with MCNP code. For three different control rod positions at a simulated core which has a quarter symmetry, total peaking factors and three sub-factors (radial, axial, and local) obtained from VENTURE were compared with those of MCNP. The comparison proved the validity of VENTURE and showed better agreement in the order of radial, axial, and local factors. The uncertainty of WIMS (KAERI) - VENTURE system was inferred using the 2σ band of total peaking obtained by MCNP. The uncertainty of WIMS (KAERI) - VENTURE system were found to be 18.5 % for the operating condition. (author)
Castanier, Eric; Paterne, Loic; Louis, Céline
2017-09-01
In the nuclear engineering, you have to manage time and precision. Especially in shielding design, you have to be more accurate and efficient to reduce cost (shielding thickness optimization), and for this, you use 3D codes. In this paper, we want to see if we can easily applicate the CADIS methods for design shielding of small pipes which go through large concrete walls. We assess the impact of the WW generated by the 3D-deterministic code ATTILA versus WW directly generated by MCNP (iterative and manual process). The comparison is based on the quality of the convergence (estimated relative error (σ), Variance of Variance (VOV) and Figure of Merit (FOM)), on time (computer time + modelling) and on the implement for the engineer.
Energy Technology Data Exchange (ETDEWEB)
Mosleh-Shirazi, M. A.; Hadad, K.; Faghihi, R.; Baradaran-Ghahfarokhi, M.; Naghshnezhad, Z.; Meigooni, A. S. [Center for Research in Medical Physics and Biomedical Engineering and Physics Unit, Radiotherapy Department, Shiraz University of Medical Sciences, Shiraz 71936-13311 (Iran, Islamic Republic of); Radiation Research Center and Medical Radiation Department, School of Engineering, Shiraz University, Shiraz 71936-13311 (Iran, Islamic Republic of); Comprehensive Cancer Center of Nevada, Las Vegas, Nevada 89169 (United States)
2012-08-15
This study primarily aimed to obtain the dosimetric characteristics of the Model 6733 {sup 125}I seed (EchoSeed) with improved precision and accuracy using a more up-to-date Monte-Carlo code and data (MCNP5) compared to previously published results, including an uncertainty analysis. Its secondary aim was to compare the results obtained using the MCNP5, MCNP4c2, and PTRAN codes for simulation of this low-energy photon-emitting source. The EchoSeed geometry and chemical compositions together with a published {sup 125}I spectrum were used to perform dosimetric characterization of this source as per the updated AAPM TG-43 protocol. These simulations were performed in liquid water material in order to obtain the clinically applicable dosimetric parameters for this source model. Dose rate constants in liquid water, derived from MCNP4c2 and MCNP5 simulations, were found to be 0.993 cGyh{sup -1} U{sup -1} ({+-}1.73%) and 0.965 cGyh{sup -1} U{sup -1} ({+-}1.68%), respectively. Overall, the MCNP5 derived radial dose and 2D anisotropy functions results were generally closer to the measured data (within {+-}4%) than MCNP4c and the published data for PTRAN code (Version 7.43), while the opposite was seen for dose rate constant. The generally improved MCNP5 Monte Carlo simulation may be attributed to a more recent and accurate cross-section library. However, some of the data points in the results obtained from the above-mentioned Monte Carlo codes showed no statistically significant differences. Derived dosimetric characteristics in liquid water are provided for clinical applications of this source model.
International Nuclear Information System (INIS)
Krotov, A.D.; Son'ko, A.V.
2009-01-01
Calculation of neutron-physical properties and radiation protection of space power reactor was made by means of the MCNP code allowing simulation of neutron, γ- and electron transport by the Monte Carlo method in the systems with combined geometry. Universality of the MCNP code has been demonstrated both for the calculation of reactor-converter so for the optimization of radiation protection that allows to reserve a new level of complex simulation of SNPS [ru
International Nuclear Information System (INIS)
Ivanov, A.; Sanchez, V.; Imke, U.
2011-01-01
In order to increase the accuracy and the degree of spatial resolution of core design studies, coupled 3D neutronic (deterministic and Monte Carlo) and 3D thermal hydraulics (CFD and subchannel) codes are being developed worldwide. At KIT both deterministic and Monte Carlo codes were coupled with subchannel codes and applied to predict the safety-related design parameters such as pin power, maximal cladding and fuel temperature, DNB. These coupling approaches were revised and improved based on the experience gained. One particular example is replacing COBRA-TF with SUBCHANFLOW, in-house development subchannel code, in the COBRA-TF/MCNP coupling, accompanied with new way of radial mapping between the neutronic and thermal hydraulic domains. The new coupled system MCNP5/SUBCHANFLOW makes it possible to investigate variety of fuel assembly types (BWR, PWR or SCFR). Key issues in such a coupled system are the way in which thermal-hydraulic/neutronic feedbacks, accuracy of the Monte Carlo solutions and observation of convergence during the iterative solution are handled. Another key issue that might be considered is the optimal application of parallel computing. Using multi-processor computer architectures, it is possible to reduce the Monte- Carlo running time and obtain converged results within reasonable time limit. In particular it is shown that by exploiting the capabilities of multi-processor calculation, it is possible to investigate large fuel assemblies in a pin-by-pin manner with a resolution at pin and subchannel level. One of the most important issues addressed in the current work is the temperature effects on nuclear data. For the particular studies pseudo material approach was used, which produces interpolated results for Doppler broadened cross sections from NJOY pre-generated nuclear data. (author)
Validation of a new midway forward-adjoint coupling option in MCNP
Energy Technology Data Exchange (ETDEWEB)
Serov, I.V.; John, T.M.; Hoogenboom, J.E. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.
1996-09-01
The new midway Monte Carlo is based on the coupling of scores from a forward and an adjoint Monte Carlo calculation on a surface in between the source and the detector. The method is implemented in MCNP. The utilization of the method is fairly straight-forward and does not require any substantial expertise. The midway Monte Carlo method was tested against the gamma-ray skyshine MCNP benchmark problem. This problem involves deep penetration and streaming along complicated paths. The midway method supplied results, which agree with the results of the reference calculation within the limits of the estimated statistical uncertainties. The efficiency of the easy-to-implement midway calculation is higher than the efficiency of the reference calculation which is already optimized by use of an importance function. The midway method proves to be efficient in problems with complicated streaming paths towards small detectors. (author)
Validation of a new midway forward-adjoint coupling option in MCNP
International Nuclear Information System (INIS)
Serov, I.V.; John, T.M.; Hoogenboom, J.E.
1996-01-01
The new midway Monte Carlo is based on the coupling of scores from a forward and an adjoint Monte Carlo calculation on a surface in between the source and the detector. The method is implemented in MCNP. The utilization of the method is fairly straight-forward and does not require any substantial expertise. The midway Monte Carlo method was tested against the gamma-ray skyshine MCNP benchmark problem. This problem involves deep penetration and streaming along complicated paths. The midway method supplied results, which agree with the results of the reference calculation within the limits of the estimated statistical uncertainties. The efficiency of the easy-to-implement midway calculation is higher than the efficiency of the reference calculation which is already optimized by use of an importance function. The midway method proves to be efficient in problems with complicated streaming paths towards small detectors. (author)
MCNP simulation to optimise in-pile and shielding parts of the Portuguese SANS instrument.
Gonçalves, I F; Salgado, J; Falcão, A; Margaça, F M A; Carvalho, F G
2005-01-01
A Small Angle Neutron Scattering instrument is being installed at one end of the tangential beam tube of the Portuguese Research Reactor. The instrument is fed using a neutron scatterer positioned in the middle of the beam tube. The scatterer consists of circulating H2O contained in a hollow disc of Al. The in-pile shielding components and the shielding installed around the neutron selector have been the object of an MCNP simulation study. The quantities calculated were the neutron and gamma-ray fluxes in different positions, the energy deposited in the material by the neutron and gamma-ray fields, the material activation resulting from the neutron field and radiation doses at the exit wall of the shutter and around the shielding. The MCNP results are presented and compared with results of an analytical approach and with experimental data collected after installation.
Absorbed fractions in a voxel-based phantom calculated with the MCNP-4B code.
Yoriyaz, H; dos Santos, A; Stabin, M G; Cabezas, R
2000-07-01
A new approach for calculating internal dose estimates was developed through the use of a more realistic computational model of the human body. The present technique shows the capability to build a patient-specific phantom with tomography data (a voxel-based phantom) for the simulation of radiation transport and energy deposition using Monte Carlo methods such as in the MCNP-4B code. MCNP-4B absorbed fractions for photons in the mathematical phantom of Snyder et al. agreed well with reference values. Results obtained through radiation transport simulation in the voxel-based phantom, in general, agreed well with reference values. Considerable discrepancies, however, were found in some cases due to two major causes: differences in the organ masses between the phantoms and the occurrence of organ overlap in the voxel-based phantom, which is not considered in the mathematical phantom.
Using the MCNP Taylor series perturbation feature (efficiently) for shielding problems
Favorite, Jeffrey
2017-09-01
The Taylor series or differential operator perturbation method, implemented in MCNP and invoked using the PERT card, can be used for efficient parameter studies in shielding problems. This paper shows how only two PERT cards are needed to generate an entire parameter study, including statistical uncertainty estimates (an additional three PERT cards can be used to give exact statistical uncertainties). One realistic example problem involves a detailed helium-3 neutron detector model and its efficiency as a function of the density of its high-density polyethylene moderator. The MCNP differential operator perturbation capability is extremely accurate for this problem. A second problem involves the density of the polyethylene reflector of the BeRP ball and is an example of first-order sensitivity analysis using the PERT capability. A third problem is an analytic verification of the PERT capability.
Application of dose evaluation of the MCNP code for interim spent fuel cask storage facility
International Nuclear Information System (INIS)
Kosako, Toshiso; Iimoto, Takeshi; Ishikawa, Satoshi; Tsuboi, Takafumi; Teramura, Masahiro; Okamura, Tomomi; Narumiya, Yoshiyuki
2007-01-01
The interim storage facility for spent fuel metallic cask is designed as a concrete building structure with air inlet and outlet for circulating the natural cooling. The feature of the interim storage facility is big capacity of spent fuel at several thousands MTU and restricted site usage. It is important to evaluate realistic dose rate in shielding design of the interim storage facility, therefore the three-dimensional continuous-energy Monte Carlo radiation transport code MCNP that exactly treating the complicated geometry was applied. The validation of dose evaluation for interim storage facility by MCNP code were performed by three kinds of neutron shielding benchmark experiments; cask shadow shielding experiment, duct streaming experiment and concrete deep penetration experiment. Dose rate distributions at each benchmark were measured and compared with the calculated results. The comparison showed a good consistency between calculation and experiment results. (author)
International Nuclear Information System (INIS)
Pillon, M.; Martone, M.; Verschuur, K.A.; Jarvis, O.N.; Kaellne, J.
1989-01-01
Neutron transport calculations have been performed using fluence ray tracing (FURNACE code) and Monte Carlo particle trajectory sampling methods (MCNP code) in order to determine the neutron fluence and energy distributions at different locations in the JET tokamak. These calculations were used to calibrate the activation measurements used in the determination of the absolute fusion neutron yields from the JET plasma. We present here the neutron activation response coefficients calculated for three different materials. Comparison of the MCNP and FURNACE results helps identify the sources of error in these neutron transport calculations. The accuracy of these calculations was tested by comparing the total 2.5 MeV neutron yields derived from the activation measurements with those obtained with calibrated fission chambers; agreement at the ±15% level was demonstrate. (orig.)
MCNP6 Simulation of Light and Medium Nuclei Fragmentation at Intermediate Energies
Energy Technology Data Exchange (ETDEWEB)
Mashnik, Stepan Georgievich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kerby, Leslie Marie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Univ. of Idaho, Moscow, ID (United States)
2015-08-24
Fragmentation reactions induced on light and medium nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below are studied with the Los Alamos transport code MCNP6 and with its CEM03.03 and LAQGSM03.03 event generators. CEM and LAQGSM assume that intermediate-energy fragmentation reactions on light nuclei occur generally in two stages. The first stage is the intranuclear cascade (INC), followed by the second, Fermi breakup disintegration of light excited residual nuclei produced after the INC. CEM and LAQGSM account also for coalescence of light fragments (complex particles) up to sup>4He from energetic nucleons emitted during INC. We investigate the validity and performance of MCNP6, CEM, and LAQGSM in simulating fragmentation reactions at intermediate energies and discuss possible ways of further improving these codes.
International Nuclear Information System (INIS)
Mashnik, Stepan G.
2011-01-01
MCNP6, the latest and most advanced LANL transport code representing a recent merger of MCNP5 and MCNPX, has been Validated and Verified (V and V) against a variety of intermediate and high-energy experimental data and against results by different versions of MCNPX and other codes. In the present work, we V and V MCNP6 using mainly the latest modifications of the Cascade-Exciton Model (CEM) and of the Los Alamos version of the Quark-Gluon String Model (LAQGSM) event generators CEM03.02 and LAQGSM03.03. We found that MCNP6 describes reasonably well various reactions induced by particles and nuclei at incident energies from 18 MeV to about 1 TeV per nucleon measured on thin and thick targets and agrees very well with similar results obtained with MCNPX and calculations by CEM03.02, LAQGSM03.01 (03.03), INCL4 + ABLA, and Bertini INC + Dresner evaporation, EPAX, ABRABLA, HIPSE, and AMD, used as stand alone codes. Most of several computational bugs and more serious physics problems observed in MCNP6/X during our V and V have been fixed; we continue our work to solve all the known problems before MCNP6 is distributed to the public. (author)
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.
The study on neutron and photon distribution of AP1000 reactor by MCNP code
International Nuclear Information System (INIS)
Chen Defeng; Shen Mingqi
2014-01-01
The core and reactor structural of AP1000 was modeled by the MCNP calculation program which is based on the Monte Carlo method in this paper, the neutron and photon distribution of AP1000 reactor core was calculated by the conditions of reactor critical. The results show that the AP1000 reactor neutron and photon distribution is in accordance with the critical design of PWR. (authors)
The use of the MCNP code for the quantitative analysis of elements in geological formations
Energy Technology Data Exchange (ETDEWEB)
Cywicka-Jakiel, T.; Woynicka, U. [The Henryk Niewodniczanski Institute of Nuclear Physics, Krakow (Poland); Zorski, T. [University of Mining and Metallurgy, Faculty of Geology, Geophysics and Environmental Protection, Krakow (Poland)
2003-07-01
The Monte Carlo modelling calculations using the MCNP code have been performed, which support the spectrometric neutron-gamma (SNGL) borehole logging. The SNGL enables the lithology identification through the quantitative analysis of the elements in geological formations and thus can be very useful for the oil and gas industry as well as for prospecting of the potential host rocks for radioactive waste disposal. In the SNGL experiment, gamma-rays induced by the neutron interactions with the nuclei of the rock elements are detected using the gamma-ray probe of complex mechanical and electronic construction. The probe has to be calibrated for a wide range of the elemental concentrations, to assure the proper quantitative analysis. The Polish Calibration Station in Zielona Gora is equipped with a limited number of calibration standards. An extension of the experimental calibration and the evaluation of the effect of the so-called side effects (for example the borehole and formation salinity variation) on the accuracy of the SNGL method can be done by the use of the MCNP code. The preliminary MCNP results showing the effect of the borehole and formation fluids salinity variations on the accuracy of silicon (Si), calcium (Ca) and iron (Fe) content determination are presented in the paper. The main effort has been focused on a modelling of the complex SNGL probe situated in a fluid filled borehole, surrounded by a geological formation. Track length estimate of the photon flux from the (n,gamma) interactions as a function of gamma-rays energy was used. Calculations were run on the PC computer with AMD Athlon 1.33 GHz processor. Neutron and photon cross-sections libraries were taken from the MCNP4c package and based mainly on the ENDF/B-6, ENDF/B-5 and MCPLIB02 data. The results of simulated experiment are in conformity with results of the real experiment performed with the use of the main lithology models (sandstones, limestones and dolomite). (authors)
The use of the MCNP code for the quantitative analysis of elements in geological formations
International Nuclear Information System (INIS)
Cywicka-Jakiel, T.; Woynicka, U.; Zorski, T.
2003-01-01
The Monte Carlo modelling calculations using the MCNP code have been performed, which support the spectrometric neutron-gamma (SNGL) borehole logging. The SNGL enables the lithology identification through the quantitative analysis of the elements in geological formations and thus can be very useful for the oil and gas industry as well as for prospecting of the potential host rocks for radioactive waste disposal. In the SNGL experiment, gamma-rays induced by the neutron interactions with the nuclei of the rock elements are detected using the gamma-ray probe of complex mechanical and electronic construction. The probe has to be calibrated for a wide range of the elemental concentrations, to assure the proper quantitative analysis. The Polish Calibration Station in Zielona Gora is equipped with a limited number of calibration standards. An extension of the experimental calibration and the evaluation of the effect of the so-called side effects (for example the borehole and formation salinity variation) on the accuracy of the SNGL method can be done by the use of the MCNP code. The preliminary MCNP results showing the effect of the borehole and formation fluids salinity variations on the accuracy of silicon (Si), calcium (Ca) and iron (Fe) content determination are presented in the paper. The main effort has been focused on a modelling of the complex SNGL probe situated in a fluid filled borehole, surrounded by a geological formation. Track length estimate of the photon flux from the (n,gamma) interactions as a function of gamma-rays energy was used. Calculations were run on the PC computer with AMD Athlon 1.33 GHz processor. Neutron and photon cross-sections libraries were taken from the MCNP4c package and based mainly on the ENDF/B-6, ENDF/B-5 and MCPLIB02 data. The results of simulated experiment are in conformity with results of the real experiment performed with the use of the main lithology models (sandstones, limestones and dolomite). (authors)
Conversion of Input Data between KENO and MCNP File Formats for Computer Criticality Assessments
International Nuclear Information System (INIS)
Schwarz, Randolph A.; Carter, Leland L.; Schwarz Alysia L.
2006-01-01
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
MCNP modelling of scintillation-detector gamma-ray spectra from natural radionuclides.
Hendriks, P H G M; Maucec, M; de Meijer, R J
2002-09-01
gamma-ray spectra of natural radionuclides are simulated for a BGO detector in a borehole geometry using the Monte Carlo code MCNP. All gamma-ray emissions of the decay of 40K and the series of 232Th and 238U are used to describe the source. A procedure is proposed which excludes the time-consuming electron tracking in less relevant areas of the geometry. The simulated gamma-ray spectra are benchmarked against laboratory data.
Energy Technology Data Exchange (ETDEWEB)
Kim, Seung Jun [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Buechler, Cynthia Eileen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2017-07-17
The current study aims to predict the steady state power of a generic solution vessel and to develop a corresponding heat transfer coefficient correlation for a Moly99 production facility by conducting a fully coupled multi-physics simulation. A prediction of steady state power for the current application is inherently interconnected between thermal hydraulic characteristics (i.e. Multiphase computational fluid dynamics solved by ANSYS-Fluent 17.2) and the corresponding neutronic behavior (i.e. particle transport solved by MCNP6.2) in the solution vessel. Thus, the development of a coupling methodology is vital to understand the system behavior at a variety of system design and postulated operating scenarios. In this study, we report on the k-effective (keff) calculation for the baseline solution vessel configuration with a selected solution concentration using MCNP K-code modeling. The associated correlation of thermal properties (e.g. density, viscosity, thermal conductivity, specific heat) at the selected solution concentration are developed based on existing experimental measurements in the open literature. The numerical coupling methodology between multiphase CFD and MCNP is successfully demonstrated, and the detailed coupling procedure is documented. In addition, improved coupling methods capturing realistic physics in the solution vessel thermal-neutronic dynamics are proposed and tested further (i.e. dynamic height adjustment, mull-cell approach). As a key outcome of the current study, a multi-physics coupling methodology between MCFD and MCNP is demonstrated and tested for four different operating conditions. Those different operating conditions are determined based on the neutron source strength at a fixed geometry condition. The steady state powers for the generic solution vessel at various operating conditions are reported, and a generalized correlation of the heat transfer coefficient for the current application is discussed. The assessment of multi
MCNP modeling of NORM dosimetry in the oil and gas industry
International Nuclear Information System (INIS)
Siqiu Wang
2016-01-01
Naturally-occurring radioactive materials wastes in the oil and gas industry create a radioactive environment for the workers in the field. MCNP simulation conducted in this work provides a useful tool in terms of radiation safety design of the oil field, as well as validation and an important addition to in situ measurements. Furthermore, phantoms are employed to observe the dose distribution throughout the human body, demonstrating radiation effects on each individual organ. (author)
International Nuclear Information System (INIS)
Pashchenko, A.B.; Wienke, H.
1997-01-01
This document summarizes the libraries of neutron activation cross-section data processed into the following three formats: continuous energy format as used by the Monte Carlo neutron/photon transport code MCNP4A; VITAMIN-J 175 multigroup format weighted with the VITAMIN-E weighting spectrum as used by the transmutation codes REAC*2/3 and FOUR ACES; VITAMIN-J 175 multigroup ENDF-6 format, with a flat weighting spectrum. The data are available from the IAEA Nuclear Data Section online via INTERNET by FTP command, or on magnetic tape. (author)
International Nuclear Information System (INIS)
Vega C, H. R.
2014-08-01
A Subcritical Nuclear Assembly is a device where the nuclear-fission chain reaction is initiated and maintained using an external neutron source. It is a valuable educational and research tool where in a safe way many reactor parameters can be measured. Here, we have used the Wigner-Seitz method in the six-factor formula to calculate the effective multiplication factor of a subcritical nuclear reactor Nuclear Chicago model 9000. This reactor has approximately 2500 kg of natural uranium heterogeneously distributed in slugs. The reactor uses a 239 PuBe neutron source that is located in the center of an hexagonal array. Using Monte Carlo methods, with the MCNP5 code, a three-dimensional model of the subcritical reactor was designed to estimate the effective multiplication factor, the neutron spectra, the total and thermal neutron fluences along the radial and axial axis. With the neutron spectra in two locations outside the reactor the ambient dose equivalent were estimated. (Author)
International Nuclear Information System (INIS)
Anon.
2000-01-01
The first text deals with a new circular concerning the collect of the medicine radioactive wastes, containing radium. This campaign wants to incite people to let go their radioactive wastes (needles, tubes) in order to suppress any danger. The second text presents a decree of the 31 december 1999, relative to the limitations of noise and external risks resulting from the nuclear facilities exploitation: noise, atmospheric pollution, water pollution, wastes management and fire prevention. (A.L.B.)
Comparison of thermal scattering processing options for S(α,β) cards in MCNP
International Nuclear Information System (INIS)
Čerba, Štefan; Damian, Jose Ignacio Marquez; Lüley, Jakub; Vrban, Branislav; Farkas, Gabriel; Nečas, Vladimír; Haščík, Jan
2013-01-01
Highlights: ► Determination of MCNP calculation bias for WWER-440. ► Specific scattering law S(α,β). ► Benchmark cases investigated. ► Three methods to process material cards for hydrogen bound in light water. - Abstract: The MCNP distributions include sets of pre-calculated thermal scattering libraries but these libraries are available for several temperature steps only. In order to achieve reliable results it is suitable to process the cross section libraries for the desired temperature. In general, there are three methods to process these thermal scattering libraries for the desired temperatures. This paper deals with the comparison of these three methods on the basis of several benchmarks and on the basis of a thermal transient experiment of a WWER-440 reactor. The choice is up to the MCNP user but unfortunately very few studies concerning the comparison have been published so far. Therefore conclusions and results presented in this paper may help the user to choose the most appropriate method for his calculation
A simulation of a pebble bed reactor core by the MCNP-4C computer code
Directory of Open Access Journals (Sweden)
Bakhshayesh Moshkbar Khalil
2009-01-01
Full Text Available Lack of energy is a major crisis of our century; the irregular increase of fossil fuel costs has forced us to search for novel, cheaper, and safer sources of energy. Pebble bed reactors - an advanced new generation of reactors with specific advantages in safety and cost - might turn out to be the desired candidate for the role. The calculation of the critical height of a pebble bed reactor at room temperature, while using the MCNP-4C computer code, is the main goal of this paper. In order to reduce the MCNP computing time compared to the previously proposed schemes, we have devised a new simulation scheme. Different arrangements of kernels in fuel pebble simulations were investigated and the best arrangement to decrease the MCNP execution time (while keeping the accuracy of the results, chosen. The neutron flux distribution and control rods worth, as well as their shadowing effects, have also been considered in this paper. All calculations done for the HTR-10 reactor core are in good agreement with experimental results.
Implementation of 3D models in the Monte Carlo code MCNP
International Nuclear Information System (INIS)
Lopes, Vivaldo; Millian, Felix M.; Guevara, Maria Victoria M.; Garcia, Fermin; Sena, Isaac; Menezes, Hugo
2009-01-01
On the area of numerical dosimetry Applied to medical physics, the scientific community focuses on the elaboration of new hybrids models based on 3D models. But different steps of the process of simulation with 3D models needed improvement and optimization in order to expedite the calculations and accuracy using this methodology. This project was developed with the aim of optimize the process of introduction of 3D models within the simulation code of radiation transport by Monte Carlo (MCNP). The fast implementation of these models on the simulation code allows the estimation of the dose deposited on the patient organs on a more personalized way, increasing the accuracy with this on the estimates and reducing the risks to health, caused by ionizing radiations. The introduction o these models within the MCNP was made through a input file, that was constructed through a sequence of images, bi-dimensional in the 3D model, generated using the program '3DSMAX', imported by the program 'TOMO M C' and thus, introduced as INPUT FILE of the MCNP code. (author)
Validation of MCNP and WIMS-AECL/DRAGON/RFSP for ACR-1000 applications
International Nuclear Information System (INIS)
Bromley, Blair P.; Adams, Fred P.; Zeller, Michael B.; Watts, David G.; Shukhman, Boris V.; Pencer, Jeremy
2008-01-01
This paper gives a summary of the validation of the reactor physics codes WIMS-AECL, DRAGON, RFSP and MCNP5, which are being used in the design, operation, and safety analysis of the ACR-1000 R . The standards and guidelines being followed for code validation of the suite are established in CSA Standard N286.7-99 and ANS Standard ANS-19.3-2005. These codes are being validated for the calculation of key output parameters associated with various reactor physics phenomena of importance during normal operations and postulated accident conditions in an ACR-1000 reactor. Experimental data from a variety of sources are being used for validation. The bulk of the validation data is from critical experiments in the ZED-2 research reactor with ACR-type lattices. To supplement and complement ZED-2 data, qualified and applicable data are being taken from other power and research reactors, such as existing CANDU R units, FUGEN, NRU and SPERT research reactors, and the DCA critical facility. MCNP simulations of the ACR-1000 are also being used for validating WIMS-AECL/ DRAGON/RFSP, which involves extending the validation results for MCNP through the assistance of TSUNAMI analyses. Code validation against commissioning data in the first-build ACR-1000 will be confirmatory. The code validation is establishing the biases and uncertainties in the calculations of the WIMS-AECL/DRAGON/RFSP suite for the evaluation of various key parameters of importance in the reactor physics analysis of the ACR-1000. (authors)
MCNP6 simulation of reactions of interest to FRIB, medical, and space applications
International Nuclear Information System (INIS)
Mashnik, Stepan G.
2015-01-01
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. (author)
Simulation of the GCR spectrum in the Mars curiosity rover's RAD detector using MCNP6
Ratliff, Hunter N.; Smith, Michael B. R.; Heilbronn, Lawrence
2017-08-01
The paper presents results from MCNP6 simulations of galactic cosmic ray (GCR) propagation down through the Martian atmosphere to the surface and comparison with RAD measurements made there. This effort is part of a collaborative modeling workshop for space radiation hosted by Southwest Research Institute (SwRI). All modeling teams were tasked with simulating the galactic cosmic ray (GCR) spectrum through the Martian atmosphere and the Radiation Assessment Detector (RAD) on-board the Curiosity rover. The detector had two separate particle acceptance angles, 4π and 30 ° off zenith. All ions with Z = 1 through Z = 28 were tracked in both scenarios while some additional secondary particles were only tracked in the 4π cases. The MCNP6 4π absorbed dose rate was 307.3 ± 1.3 μGy/day while RAD measured 233 μGy/day. Using the ICRP-60 dose equivalent conversion factors built into MCNP6, the simulated 4π dose equivalent rate was found to be 473.1 ± 2.4 μSv/day while RAD reported 710 μSv/day.
An analysis of MCNP cross-sections and tally methods for low-energy photon emitters.
Demarco, John J; Wallace, Robert E; Boedeker, Kirsten
2002-04-21
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.
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.
Simulation of the GCR spectrum in the Mars curiosity rover's RAD detector using MCNP6.
Ratliff, Hunter N; Smith, Michael B R; Heilbronn, Lawrence
2017-08-01
The paper presents results from MCNP6 simulations of galactic cosmic ray (GCR) propagation down through the Martian atmosphere to the surface and comparison with RAD measurements made there. This effort is part of a collaborative modeling workshop for space radiation hosted by Southwest Research Institute (SwRI). All modeling teams were tasked with simulating the galactic cosmic ray (GCR) spectrum through the Martian atmosphere and the Radiation Assessment Detector (RAD) on-board the Curiosity rover. The detector had two separate particle acceptance angles, 4π and 30 ° off zenith. All ions with Z = 1 through Z = 28 were tracked in both scenarios while some additional secondary particles were only tracked in the 4π cases. The MCNP6 4π absorbed dose rate was 307.3 ± 1.3 µGy/day while RAD measured 233 µGy/day. Using the ICRP-60 dose equivalent conversion factors built into MCNP6, the simulated 4π dose equivalent rate was found to be 473.1 ± 2.4 µSv/day while RAD reported 710 µSv/day. Copyright © 2017 The Committee on Space Research (COSPAR). Published by Elsevier Ltd. All rights reserved.
Utilization of the MCNP-3A code for criticality safety analysis
International Nuclear Information System (INIS)
Maragni, M.G.; Moreira, J.M.L.
1996-01-01
In the last decade, Brazil started to operate facilities for processing and storing uranium in different forms. The necessity of criticality safety analysis appeared in the design phase of the uranium pilot process plants and also in the licensing of transportation and storage of fissile materials. The 2-MW research reactor and the Angra I power plant also required criticality safety assessments because their spent-fuel storage was approaching full-capacity utilization. The criticality safety analysis in Brazil has been based on KENO IV code calculations, which present some difficulties for correct geometry representation. The MCNP-3A code is not reported to be used frequently for criticality safety analysis in Brazil, but its good geometry representation makes it a possible tool for treating problems of complex geometry. A set of benchmark tests was performed to verify its applicability for criticality safety analysis in Brazil. This paper presents several benchmark tests aimed at selecting a set of options available in the MCNP-3A code that would be adequate for criticality safety analysis. The MCNP-3A code is also compared with the KENO-IV code regarding its performance for criticality safety analysis
Au-coated X-ray Anti-scattering Grid Performance Test by MCNP
Energy Technology Data Exchange (ETDEWEB)
Bae, JunWoo; Yoo, Dong Han; Kim, Hee Reyoung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)
2014-10-15
It is required to protect individual against the dangers of ionizing radiation from medical exposure. And increasing of resolution for x-ray radiography tools can give radiation protectoral benefits. Because the image device has higher resolution in same energy source, it requires low energy level source and it can reduce individual dose. The anti-scattering grid is sub-device that is attached in front of detector (direction of source). It is square lattice shape generally. It is composed of penetration parts and shielding parts. Penetration part is generally air (the void) and in some studies it uses wood or aluminum. Shielding part is composed of various materials such as lead or copper. In this study, it is focused on the gold as one of X-ray grid materials, where gold is generally known as excellent shielding material and the performance test on the gold coated anti-scattering grid is carried out by MCNP simulation. X-ray grid was simulated by using MCNP code and its performance was investigated. It was understood that glass based and Au-coated grid could lessen the scattered photons more where the reduction was about two third. In further study, geometry optimization or material selection will be conducted by MCNP simulation for giving benefits to design proper grid for various instruments.
The Use of MCNP in Flash Radiographic Applications at AWE
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.
V&V of MCNP 6.1.1 Beta Against Intermediate and High-Energy Experimental Data
Energy Technology Data Exchange (ETDEWEB)
Mashnik, Stepan G [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2014-09-08
This report presents a set of validation and verification (V&V) MCNP 6.1.1 beta results calculated in parallel, with MPI, obtained using its event generators at intermediate and high-energies compared against various experimental data. It also contains several examples of results using the models at energies below 150 MeV, down to 10 MeV, where data libraries are normally used. This report can be considered as the forth part of a set of MCNP6 Testing Primers, after its first, LA-UR-11-05129, and second, LA-UR-11-05627, and third, LA-UR-26944, publications, but is devoted to V&V with the latest, 1.1 beta version of MCNP6. The MCNP6 test-problems discussed here are presented in the /VALIDATION_CEM/and/VALIDATION_LAQGSM/subdirectories in the MCNP6/Testing/directory. README files that contain short descriptions of every input file, the experiment, the quantity of interest that the experiment measures and its description in the MCNP6 output files, and the publication reference of that experiment are presented for every test problem. Templates for plotting the corresponding results with xmgrace as well as pdf files with figures representing the final results of our V&V efforts are presented. Several technical “bugs” in MCNP 6.1.1 beta were discovered during our current V&V of MCNP6 while running it in parallel with MPI using its event generators. These “bugs” are to be fixed in the following version of MCNP6. Our results show that MCNP 6.1.1 beta using its CEM03.03, LAQGSM03.03, Bertini, and INCL+ABLA, event generators describes, as a rule, reasonably well different intermediate- and high-energy measured data. This primer isn’t meant to be read from cover to cover. Readers may skip some sections and go directly to any test problem in which they are interested.
MCNP4C2, Coupled Neutron, Electron Gamma 3-D Time-Dependent Monte Carlo Transport Calculations
International Nuclear Information System (INIS)
2002-01-01
1 - Description of program or function: MCNP is a general-purpose, continuous-energy, generalized geometry, time-dependent, coupled neutron-photon-electron Monte Carlo transport code system. MCNP4C2 is an interim release of MCNP4C with distribution restricted to the Criticality Safety community and attendees of the LANL MCNP workshops. The major new features of MCNP4C2 include: - Photonuclear physics; - Interactive plotting; - Plot superimposed weight window mesh; - Implement remaining macro-body surfaces; - Upgrade macro-bodies to surface sources and other capabilities; - Revised summary tables; - Weight window improvements. See the MCNP home page more information http://www-xdiv.lanl.gov/XCI/PROJECTS/MCNP with a link to the MCNP Forum. See the Electronic Notebook at http://www-rsicc.ornl.gov/rsic.html for information on user experiences with MCNP. 2 - Methods:MCNP treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces. Pointwise continuous-energy cross section data are used, although multigroup data may also be used. Fixed-source adjoint calculations may be made with the multigroup data option. For neutrons, all reactions in a particular cross-section evaluation are accounted for. Both free gas and S(alpha, beta) thermal treatments are used. Criticality sources as well as fixed and surface sources are available. For photons, the code takes account of incoherent and coherent scattering with and without electron binding effects, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. A very general source and tally structure is available. The tallies have extensive statistical analysis of convergence. Rapid convergence is enabled by a wide variety of variance reduction methods. Energy ranges are 0-60 MeV for neutrons (data generally only available up to
Energy Technology Data Exchange (ETDEWEB)
Poškus, Andrius, E-mail: andrius.poskus@ff.vu.lt
2016-02-01
This work evaluates the accuracy of the single-event (SE) and condensed-history (CH) models of electron transport in Monte Carlo simulations of electron backscattering from thick layers of Be, C, Al, Cu, Ag, Au and U at incident electron energies from 200 eV to 15 MeV. The CH method is used in simulations performed with MCNP6.1, and the SE method is used in simulations performed with an open-source single-event code MCNelectron written by the author of this paper. Both MCNP6.1 and MCNelectron use mainly ENDF/B-VI.8 library data, but MCNelectron allows replacing cross sections of certain types of interactions by alternative datasets from other sources. The SE method is evaluated both using only ENDF/B-VI.8 cross sections (the “SE-ENDF/B method”, which is equivalent to using MCNP6.1 in SE mode) and with an alternative set of elastic scattering cross sections obtained from relativistic (Dirac) partial-wave (DPW) calculations (the “SE-DPW method”). It is shown that at energies from 200 eV to 300 keV the estimates of the backscattering coefficients obtained using the SE-DPW method are typically within 10% of the experimental data, which is approximately the same accuracy that is achieved using MCNP6.1 in CH mode. At energies below 1 keV and above 300 keV, the SE-DPW method is much more accurate than the SE-ENDF/B method due to lack of angular distribution data in the ENDF/B library in those energy ranges. At energies from 500 keV to 15 MeV, the CH approximation is roughly twice more accurate than the SE-DPW method, with the average relative errors equal 7% and 14%, respectively. The energy probability density functions (PDFs) of backscattered electrons for Al and Cu, calculated using the SE method with DPW cross sections when energy of incident electrons is 20 keV, have an average absolute error as low as 4% of the average PDF. This error is approximately twice less than the error of the corresponding PDF calculated using the CH approximation. It is concluded
Nuclear criticality research at the University of New Mexico
International Nuclear Information System (INIS)
Busch, R.D.
1997-01-01
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 2 F 2 solution critical assembly at Los Alamos National Laboratory. The results obtained differed slightly from what was predicted by the TWODANT code
2014-03-27
Vehicle Code System (VCS), the Monte Carlo Adjoint SHielding (MASH), and the Monte Carlo n- Particle ( MCNP ) code. Of the three, the oldest and still most...widely utilized radiation transport code is MCNP . First created at Los Alamos National Laboratory (LANL) in 1957, the code simulated neutral...particle types, and previous versions of MCNP were repeatedly validated using both simple and complex 10 geometries [12, 13]. Much greater discussion and
Energy Technology Data Exchange (ETDEWEB)
Xolocostli M, J. V.; Ramirez S, J. R., E-mail: vicente.xolocostli@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)
2013-10-15
A main concern exists inside the nuclear power plants in operation around the world that is the with respect to the storage capacity of the spent fuel, due to the useful life of the plant and the storage capacity in the spent fuel pool. In diverse countries is believed that one of the best alternatives for the spent fuel is the reprocessing of the same one since exists a great quantity of fissile material that can be profitable as the Pu-239, but even so the costs for the reprocessing continue being high, what limits taking this process to great scale. Is for that reason the importance of the containers for storage of spent fuel in dry which has had a great apogee in the last years, since they represent an alternative to store the spent fuel before making a decision on the reprocessing of the same one or the final disposal. In this work an evaluation of the criticality of a concrete container for storage of spent fuel in dry commercially available is made, and which is useful for fuel assemblies type PWR like BWR, in our case only the type BWR is considered. For the analysis of the evaluation was used the code MCNP5, considering the characteristics of the concrete container according to the available data, although the type of fuel assembly is BWR one of the models of the ABB company was considered with which the comparative of the results is made. The made calculations were carried out considering the inundation of the gap that exist and the external cavity, being this the most extreme condition to arrive to the criticality or in the case of happening an accident to have the filtration of the water toward the space of the gap. (author)
Energy Technology Data Exchange (ETDEWEB)
Flores O, F.E.; Mireles G, F.; Davila R, J.I.; Pinedo V, J.L.; Risorios M, C.; Lopez del Rio, H. [UAZ, Unidad Academica de Estudios Nucleares, 98068 Zacatecas (Mexico)
2008-07-01
The MCNP code is used to simulate the radiation transport taking as tools the transport physics of each particle, either photon, neutron or electron, and the generation of random numbers. Developed in the Los Alamos National Laboratory, this code has been used thoroughly with great success, because the results of the simulations are broadly validated with representative experiments. In the one present work the room of radiotherapy of the Institute Zacatecano of the Tumor it is simulated, located in the city of Zacatecas where one is Theratron 780C machine manufactured by MSD Nordion, with the purpose of estimating the contribution to the dose that would be received in different points of the structure, included three directly under the source. Three results of analytical calculations for points located at different distances from the source are presented, and they are compared against those obtained by the simulation. Its are also presented results for the simulation of 10 points more distributed around the source. (Author)
International Nuclear Information System (INIS)
Irhas; Andang Widi Harto; Yohannes Sardjono
2014-01-01
Boron Neutron Capture Therapy (BNCT) using physics principle when B 10 (Boron-10) irradiated by low energy neutron (thermal neutron). Boron and thermal neutron reaction produced B 11m (Boron-11m) (t 1/2 =10 -2 s). B 11m decay emitted alpha, Li 7 (Lithium-7) particle and gamma ray. Irradiated time needed to ensure cancer dose enough. Liver cancer was primary malignant who located in liver (Hepatocellular carcinoma). Malignant in liver were different to metastatic from Breast, Colon Cancer, and the other. This condition was Metastatic Liver Cancer. Monte Carlo method used by Monte Carlo N-Particle (MCNP) Software. Probabilistic approach used for probability of interaction occurred and record refers to characteristic of particle and material. In this case, thermal neutron produced by model of Collimated Thermal Column Kartini Research Nuclear Reactor, Yogyakarta. Modelling organ and source used liver organ that contain of cancer tissue and research reactor. Variation of boron concentration was 20, 25, 30, 35, 40, 45, and 47 µg/g cancers. Output of MCNP calculation were neutron scattering dose, gamma ray dose and neutron flux from reactor. Neutron flux used to calculate alpha, proton and gamma ray dose from interaction of tissue material and thermal neutron. Variation of boron concentration result dose rate to every variation were 0,059; 0,072; 0,084; 0,098; 0.108; 0,12; 0,125 Gy/sec. Irradiation time who need to every concentration were 841,5 see (14 min 1 sec); 696,07 sec(11 min 36 sec); 593.11 sec (9 min 53 sec); 461,35 sec (8 min 30 sec); 461,238 sec (7 min 41 sec); 414,23 sec (6 min 54 sec); 398,38 sec (6 min 38 sec). Irradiating time could shortly when boron concentration more high. (author)
Muratov, V. G.; Lopatkin, A. V.
An important aspect in the verification of the engineering techniques used in the safety analysis of MOX-fuelled reactors, is the preparation of test calculations to determine nuclide composition variations under irradiation and analysis of burnup problem errors resulting from various factors, such as, for instance, the effect of nuclear data uncertainties on nuclide concentration calculations. So far, no universally recognized tests have been devised. A calculation technique has been developed for solving the problem using the up-to-date calculation tools and the latest versions of nuclear libraries. Initially, in 1997, a code was drawn up in an effort under ISTC Project No. 116 to calculate the burnup in one VVER-1000 fuel rod, using the MCNP Code. Later on, the authors developed a computation technique which allows calculating fuel burnup in models of a fuel rod, or a fuel assembly, or the whole reactor. It became possible to apply it to fuel burnup in all types of nuclear reactors and subcritical blankets.
Directory of Open Access Journals (Sweden)
Ralind Re Marla
2015-03-01
Full Text Available Telah dilakukan desain teras Pembangkit Listrik Tenaga Nuklir (PLTN untuk jenis Pebble Bed Modular Reactor (PBMR dengan daya 70 MWe untuk keperluan proses smelter pada keadaan beginning of life (BOL. Analisis ini bertujuan untuk mengetahui persen pengkayaan, distribusi suhu dan nilai keselamatan dengan koefisien reaktivitas teras yang negatif pada reaktor jenis PBMR apabila daya reaktor 70 MWe. Analisis menggunakan program Monte Carlo N-Particle-5 (MCNP5 dan dari hasil analisis ini diharapkan dapat memenuhi syarat dalam mendukung program percepatan pembangunan kelistrikan batubara 10.000 MWe khususnya untuk proses smelter, yang tersebar merata di wilayah Indonesia. Hasil penelitian menunjukkan bahwa, faktor perlipatan efektif (k-eff Reaktor jenis PBMR daya 70 MWe mengalami kondisi kritis pada pengkayaan 5,626 % dengan nilai faktor perlipatan efektif 1,00031±0,00087 dan nilai koefisien reaktivitas suhu pada -10,0006 pcm/K. Dari hasil analisis daat disimpulkan bahwa reaktor jenis PBMR daya 70 MWe adalah aman. ABSTRACT The core design of Nuclear Power Plant for Pebble Bed Modular Reactor (PBMR type with 70 MWe capacity power in Beginning of Life (BOL has been performed. The aim of this analysis, to know percent enrichment, temperature distribution and safety value by negative temperature coefficient at type PBMR if reactor power become lower equal to 70 MWe. This analysis was expected become one part of overview project development the power plant with 10.000 MWe of total capacity, spread evenly in territory of Indonesia especially to support of smelter industries. The results showed that, effective multiplication factor (keff with power 70 MWe critical condition at enrichment 5,626 %is 1,00031±0,00087, based on enrichment result, a value of the temperature coefficient reactivity is - 10,0006 pcm/K. Based on the results of these studies, it can beconcluded that the PBMR 70 MWe design is theoritically safe.
PWR Facility Dose Modeling Using MCNP5 and the CADIS/ADVANTG Variance-Reduction Methodology
Energy Technology Data Exchange (ETDEWEB)
Blakeman, Edward D [ORNL; Peplow, Douglas E. [ORNL; Wagner, John C [ORNL; Murphy, Brian D [ORNL; Mueller, Don [ORNL
2007-09-01
The feasibility of modeling a pressurized-water-reactor (PWR) facility and calculating dose rates at all locations within the containment and adjoining structures using MCNP5 with mesh tallies is presented. Calculations of dose rates resulting from neutron and photon sources from the reactor (operating and shut down for various periods) and the spent fuel pool, as well as for the photon source from the primary coolant loop, were all of interest. Identification of the PWR facility, development of the MCNP-based model and automation of the run process, calculation of the various sources, and development of methods for visually examining mesh tally files and extracting dose rates were all a significant part of the project. Advanced variance reduction, which was required because of the size of the model and the large amount of shielding, was performed via the CADIS/ADVANTG approach. This methodology uses an automatically generated three-dimensional discrete ordinates model to calculate adjoint fluxes from which MCNP weight windows and source bias parameters are generated. Investigative calculations were performed using a simple block model and a simplified full-scale model of the PWR containment, in which the adjoint source was placed in various regions. In general, it was shown that placement of the adjoint source on the periphery of the model provided adequate results for regions reasonably close to the source (e.g., within the containment structure for the reactor source). A modification to the CADIS/ADVANTG methodology was also studied in which a global adjoint source is weighted by the reciprocal of the dose response calculated by an earlier forward discrete ordinates calculation. This method showed improved results over those using the standard CADIS/ADVANTG approach, and its further investigation is recommended for future efforts.
Electron absorbed dose comparison between MCNP5 and Penelope Monte Carlo code for microdosimetry
International Nuclear Information System (INIS)
Cintra, Felipe B. de; Yoriyaz, Helio
2009-01-01
The objective of the present work was to compare electron absorbed dose results between two widespread used codes in international scientific community: MCNP5 and Penelope-2003. Individual water spheres with masses between 10 -9 g up to 10 -3 g immersed in an infinite water medium (density of 1g/cm 3 ) and monoenergetic electron sources with energy from 0.002 MeV to 0.1 MeV have been considered. The absorbed dose in the spheres was evaluated by both codes and the relative differences have been quantified. The results shown that Penelope gives, in general, higher results that, in some cases saturate or reach a maximum point and then rapidly drops. Particularly, for the 40 keV electron source we have done additional tests in three different scenarios: more points in the region of lower masses to a better definition of the curve behavior; MCNP used 200 substeps and Penelope was set to a full detail history methodology, and almost same parameters of case B but with the density of exterior medium increased to 10 g/cm 3 . The three cases show the influence of the backscattering that contribute with an important fraction of absorbed dose, finally we can infer a range of reliability to use the codes in this kind of simulations: both codes can calculate close results for up to 10 -4 g.Even though MCNP5 uses the condensed history method, if simulation parameters are chosen carefully it can reproduce results very close to those obtained using detailed history mode. In some cases, the use of higher number of electron substeps causes significant differences in the result. (author)
PWR Facility Dose Modeling Using MCNP5 and the CADIS/ADVANTG Variance-Reduction Methodology
International Nuclear Information System (INIS)
Blakeman, Edward D.; Peplow, Douglas E.; Wagner, John C.; Murphy, Brian D.; Mueller, Don
2007-01-01
The feasibility of modeling a pressurized-water-reactor (PWR) facility and calculating dose rates at all locations within the containment and adjoining structures using MCNP5 with mesh tallies is presented. Calculations of dose rates resulting from neutron and photon sources from the reactor (operating and shut down for various periods) and the spent fuel pool, as well as for the photon source from the primary coolant loop, were all of interest. Identification of the PWR facility, development of the MCNP-based model and automation of the run process, calculation of the various sources, and development of methods for visually examining mesh tally files and extracting dose rates were all a significant part of the project. Advanced variance reduction, which was required because of the size of the model and the large amount of shielding, was performed via the CADIS/ADVANTG approach. This methodology uses an automatically generated three-dimensional discrete ordinates model to calculate adjoint fluxes from which MCNP weight windows and source bias parameters are generated. Investigative calculations were performed using a simple block model and a simplified full-scale model of the PWR containment, in which the adjoint source was placed in various regions. In general, it was shown that placement of the adjoint source on the periphery of the model provided adequate results for regions reasonably close to the source (e.g., within the containment structure for the reactor source). A modification to the CADIS/ADVANTG methodology was also studied in which a global adjoint source is weighted by the reciprocal of the dose response calculated by an earlier forward discrete ordinates calculation. This method showed improved results over those using the standard CADIS/ADVANTG approach, and its further investigation is recommended for future efforts
Calibration curves of a PGNAA system for cement raw material analysis using the MCNP code
International Nuclear Information System (INIS)
Oliveira, Carlos; Salgado, Jose
1998-01-01
In large samples, the γ-ray count rate of a prompt gamma neutron activation analysis system is a multi-variable function of the elemental dry composition, density, water content and thickness of the material. The experimental calibration curves require tremendous laboratory work, using a great number of standards with well-known compositions. Although a Monte Carlo simulation study does not avoid the experimental calibration work, it reduces the number of experimental calibration standards. This paper is part of a feasibility study for a PGNAA system for on-line continuous characterisation of cement raw material conveyed on a belt (Oliveira, C., Salgado, J. and Carvalho, F. G. (1997) Optimisation of PGNAA instrument design for cement raw materials using the MCNP code. J. Radioanal. Nucl. Chem. 216(2), 191-198; Oliveira, C., Salgado, J., Goncalves, I. F., Carvalho, F. G. and Leitao, F. (1997a) A Monte Carlo study of the influence of geometry arrangements and structural materials on a PGNAA system performance for cement raw materials analysis. Appl. Radiat. Isot. (accepted); Oliveira, C., Salgado, J. and Leitao, F. (1997b) Density and water content corrections in the gamma count rate of a PGNAA system for cement raw material analysis using the MCNP code. Appl. Radiat. Isot. (accepted).]. It reports on the influence of the density, mass water content and thickness on the calibration curves of the PGNAA system. The MCNP-4A code, running in a Pentium-PC and in a DEC workstation, was used to simulate the PGNAA configuration system
MCNP: a general Monte Carlo code for neutron and photon transport
International Nuclear Information System (INIS)
1979-11-01
The general-purpose Monte Carlo code MCNP ca be used for neutron, photon, or coupled neutron-photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation are accounted for. Thermal neutrons are described by both the free-gas and S(α,β) models. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. MCNP includes an elaborate, interactive plotting capability that allows the user to view his input geometry to help check for setup errors. Standard features which are available to improve computational efficiency include geometry splitting and Russian roulette, weight cutoff with Russian roulette, correlated sampling, analog capture or capture by weight reduction, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point or ring detectors, deterministically transporting pseudo-particles to designated regions, track-length estimators, source biasing, and several parameter cutoffs. Extensive summary information is provided to help the user better understand the physics and Monte Carlo simulation of his problem. The standard, user-defined output of MCNP includes two-way current as a function of direction across any set of surfaces or surface segments in the problem. Flux across any set of surfaces or surface segments is available. 58 figures, 28 tables
International Nuclear Information System (INIS)
Yoon, W.Y.; Meachum, T.R.; Blackwood, L.G.; Harker, Y.D.
2000-01-01
The Idaho National Engineering and Environmental Laboratory Stored Waste Examination Pilot Plant (SWEPP) passive active neutron (PAN) radioassay system is used to certify transuranic (TRU) waste drums in terms of quantifying plutonium and other TRU element activities. Depending on the waste form involved, significant systematic and random errors need quantification in addition to the counting statistics. To determine the total uncertainty of the radioassay results, a statistical sampling and verification approach has been developed. In this approach, the total performance of the PAN nondestructive assay system is simulated using the computer models of the assay system, and the resultant output is compared with the known input to assess the total uncertainty. The supporting steps in performing the uncertainty analysis for the passive assay measurements in particular are as follows: (1) Create simulated waste drums and associated conditions; (2) Simulate measurements to determine the basic counting data that would be produced by the PAN assay system under the conditions specified; and (3) Apply the PAN assay system analysis algorithm to the set of counting data produced by simulating measurements to determine the measured plutonium mass. The validity of this simulation approach was verified by comparing simulated output against results from actual measurements using known plutonium sources and surrogate waste drums. The computer simulation of the PAN system performance uses the Monte Carlo N-Particle (MCNP) Code System to produce a neutron transport calculation for a simulated waste drum. Specifically, the passive system uses the neutron coincidence counting technique, utilizing the spontaneous fission of 240 Pu. MCNP application to the SWEPP PAN assay system uncertainty analysis has been very useful for a variety of waste types contained in 208-ell drums measured by a passive radioassay system. The application of MCNP to the active radioassay system is also feasible
γ radiation level simulation and analysis with MCNP in EPR containment during severe accident
International Nuclear Information System (INIS)
Zeng Jun; Liu Shuhuan; Wang Yang; Zhai Liang
2013-01-01
The γ dosimetry model based on the EPR core structure, material composition and the designed shielding system was established. The γ-ray dose rate distributions in EPR containment under different conditions including normal operation state, loss-of-coolant accident and core melt severe accident were simulated with MCNP5, and the calculation results under normal operation state and severe accident were compared and analyzed respectively with that of the designed limit. The study results may provide some relative data reference for EPR core accident prediction and reactor accident emergency decision making. (authors)
A comparison study for mass attenuation coefficients of some amino acids using MCNP code
Energy Technology Data Exchange (ETDEWEB)
Vahabi, Seyed Milad; Bahreynipour, Mostean; Shamsaie-Zafarghandi, Mojtaba [Amirkabir Univ. of Technology, Tehran (Iran, Islamic Republic of). Dept. of Energy Engineering and Physics
2017-07-15
In this study, a novel model of MCNP4C code reported recently was used to determine the photon mass attenuation coefficients of some amino acids at energies, 123, 360, 511, 662, 1170, 1280 and 1330 keV. The simulation results were compared with the XCOM data. It was indicated that the results were highly close to the calculated XCOM values. Obtained results were used to calculate the molar extinction coefficient. All the results showed the convenience and usefulness of the model in calculation of mass attenuation coefficients of amino acids.
Shielding analysis of high level waste water storage facilities using MCNP code
Energy Technology Data Exchange (ETDEWEB)
Yabuta, Naohiro [Mitsubishi Research Inst., Inc., Tokyo (Japan)
2001-01-01
The neutron and gamma-ray transport analysis for the facility as a reprocessing facility with large buildings having thick shielding was made. Radiation shielding analysis consists of a deep transmission calculation for the concrete wall and a skyshine calculation for the space out of the buildings. An efficient analysis with a short running time and high accuracy needs a variance reduction technique suitable for all the calculation regions and structures. In this report, the shielding analysis using MCNP and a discrete ordinate transport code is explained and the idea and procedure of decision of variance reduction parameter is completed. (J.P.N.)
Calculation of the effective dose from natural radioactivity sources in soil using MCNP code
International Nuclear Information System (INIS)
Krstic, D.; Nikezic, D.
2008-01-01
Full text: Effective dose delivered by photon emitted from natural radioactivity in soil was calculated in this report. Calculations have been done for the most common natural radionuclides in soil as 238 U, 232 Th series and 40 K. A ORNL age-dependent phantom and the Monte Carlo transport code MCNP-4B were employed to calculate the energy deposited in all organs of phantom.The effective dose was calculated according to ICRP74 recommendations. Conversion coefficients of effective dose per air kerma were determined. Results obtained here were compared with other authors
Optimization study of ultracold neutron sources at TRIGA reactors using MCNP
International Nuclear Information System (INIS)
Pokotilovskij, Yu.N.; Rogov, A.D.
1997-01-01
Monte Carlo simulation for the optimization of ultracold and very cold neutron sources for TRIGA reactors is performed. The calculations of thermal and cold neutron fluxes from the TRIGA reactor for different positions and configurations of a very cold solid methane moderator were performed with using the MCNP program. The production of neutrons in the ultracold and very cold energy range was calculated for the most promising final moderators (converters): very cold solid deuterium and heavy methane. The radiation energy deposition was calculated for the optimized solid methane-heavy methane cold neutron moderator
Simulation of the field dose of the irradiator PX γ 30 using MCNP
International Nuclear Information System (INIS)
Torres, N.; Prieto, E. F.; Chavez, A.; Rosales, J.
2011-01-01
Given the acceptance and actual application of radiation technology for research purposes, industrial and trade is increasing, and that safety and quality of the product being treated by radiation technology is a function of absorbed dose, it becomes necessary to have a good characterization of the radiation field at processing volume, avoiding in this way that the product receives a different dose that affects its properties, or failure to reach the desired effect, which in many cases would be embarrassing. The simulation using the MCNP program, which uses probabilistic Monte Carlo code, can correctly characterize the dose field in the irradiation chamber of research irradiator PX γ 30 used in the CEADEN. (Author)
Dose mapping using MCNP code and experiment for SVST-Co-60/B irradiator in Vietnam.
Tran, Van Hung; Tran, Khac An
2010-06-01
By using MCNP code and ethanol-chlorobenzene (ECB) dosimeters the simulations and measurements of absorbed dose distribution in a tote-box of the Cobalt-60 irradiator, SVST-Co60/B at VINAGAMMA have been done. Based on the results Dose Uniformity Ratios (DUR), positions and values of minimum and maximum dose extremes in a tote-box, and efficiency of the irradiator for the different dummy densities have been gained. There is a good agreement between simulation and experimental results in comparison and they have valuable meanings for operation of the irradiator. Copyright 2010 Elsevier Ltd. All rights reserved.
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. Copyright 2009 Elsevier Ltd. All rights reserved.
Oliveira, C
2001-01-01
A systematic study of isodose distributions and dose uniformity in sample carriers of the Portuguese Gamma Irradiation Facility was carried out using the MCNP code. The absorbed dose rate, gamma flux per energy interval and average gamma energy were calculated. For comparison purposes, boxes filled with air and 'dummy' boxes loaded with layers of folded and crumpled newspapers to achieve a given value of density were used. The magnitude of various contributions to the total photon spectra, including source-dependent factors, irradiator structures, sample material and other origins were also calculated.
Calculation of self–shielding factor for neutron activation experiments using GEANT4 and MCNP
Energy Technology Data Exchange (ETDEWEB)
Romero–Barrientos, Jaime, E-mail: jaromero@ing.uchile.cl [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, Las Condes, Santiago (Chile); Universidad de Chile, DFI, Facultad de Ciencias Físicas Y Matemáticas, Avenida Blanco Encalada 2008, Santiago (Chile); Molina, F. [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, Las Condes, Santiago (Chile); Aguilera, Pablo [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, Las Condes, Santiago (Chile); Universidad de Chile, Depto. de Física, Facultad de Ciencias, Las Palmeras 3425, Ñuñoa, Santiago (Chile); Arellano, H. F. [Universidad de Chile, DFI, Facultad de Ciencias Físicas Y Matemáticas, Avenida Blanco Encalada 2008, Santiago (Chile)
2016-07-07
The neutron self–shielding factor G as a function of the neutron energy was obtained for 14 pure metallic samples in 1000 isolethargic energy bins from 1·10{sup −5}eV to 2·10{sup 7}eV using Monte Carlo simulations in GEANT4 and MCNP6. The comparison of these two Monte Carlo codes shows small differences in the final self–shielding factor mostly due to the different cross section databases that each program uses.
Dose calculation for 40K ingestion in samples of beans using spectrometry and MCNP
International Nuclear Information System (INIS)
Garcez, R.W.D.; Lopes, J.M.; Silva, A.X.; Domingues, A.M.; Lima, M.A.F.
2014-01-01
A method based on gamma spectroscopy and on the use of voxel phantoms to calculate dose due to ingestion of 40 K contained in bean samples are presented in this work. To quantify the activity of radionuclide, HPGe detector was used and the data entered in the input file of MCNP code. The highest value of equivalent dose was 7.83 μSv.y -1 in the stomach for white beans, whose activity 452.4 Bq.Kg -1 was the highest of the five analyzed. The tool proved to be appropriate when you want to calculate the dose in organs due to ingestion of food. (author)
International Nuclear Information System (INIS)
Bykov, V.
2014-08-01
The Swiss National Cooperative for the Disposal of Radioactive Waste (NAGRA) regularly performs analysis of cost estimates associated with the NPP decommissioning. For this purpose, NAGRA has over the past ten years developed a NPP activation analysis methodology based on MCNP models of Swiss NPPs. The validation of these models is accomplished using measurements from oil activation campaigns, in which foil samples are activated at key locations inside the NPP for the duration of one cycle. The measurement campaigns have already been carried out at the Gösgen PWR (KKG) and the Mühleberg BWR (KKM). The first validation has already been successfully conducted for the KKG MCNP model. This thesis describes the efforts to validate the KKM MCNP model. This process included modifications, such as modeling of steam separators individually and improving the definition of jet pumps. Furthermore, the core definition was completely redefined, going from a 6-cell cylindrical model to a 940-cell model, shaped like the actual KKM core, which more accurately represented the void distribution. In order to benchmark the new model, the locations of samples during the two KKM foil activation campaigns were implemented into the model using the GSAM code. The interface between the MCNP model and GSAM was improved by creating a new energy group structure, optimized specifically for the activation of the three foil materials. Their activation was stimulated the state of the art hybrid VR code ADVANTG. The calculated results were then compared against the measured values for each foil material separately. The numerous improvements introduced in the 2014 model led to good agreement in many areas. The agreement is within the factor of two on the inner side of the bioshield, at the core height and above, and factor of three above the bioshield. Furthermore, distinct suggestion for improving the agreement in other areas was presented. This includes modeling of pipes extending from the RPV
Definition of neutron lifespan and neutron lifetime in MCNP4B
International Nuclear Information System (INIS)
Busch, R.D.; Spriggs, G.D.; Hendricks, J.S.
1997-01-01
MCNP4B was released in early 1997. In this new version, several major changes were made to the underlying theory used to estimate the non-adjoint-weighted removal, fission, capture, and escape prompt-neutron lifetimes. These four lifetimes are now being calculated in accordance to the neutron-balance theory described by Spriggs et al. in which the non-adjoint-weighted lifetime for a particular type of reaction (i.e., fission, capture, escape, removal, etc.) is defined as the total neutron population in the system divided by that reaction rate
MCNP4c JEFF-3.1 Based Libraries. Eccolib-Jeff-3.1 libraries
International Nuclear Information System (INIS)
Sublet, J.Ch.
2006-01-01
Continuous-energy and multi-temperatures MCNP Ace types libraries, derived from the Joint European Fusion-Fission JEFF-3.1 evaluations, have been generated using the NJOY-99.111 processing code system. They include the continuous-energy neutron JEFF-3.1/General Purpose, JEFF-3.1/Activation-Dosimetry and thermal S(α,β) JEFF-3.1/Thermal libraries and data tables. The processing steps and features are explained together with the Quality Assurance processes and records linked to the generation of such multipurpose libraries. (author)
New Tools to Prepare ACE Cross-section Files for MCNP Analytic Test Problems
International Nuclear Information System (INIS)
Brown, Forrest B.
2016-01-01
Monte Carlo calculations using one-group cross sections, multigroup cross sections, or simple continuous energy cross sections are often used to: (1) verify production codes against known analytical solutions, (2) verify new methods and algorithms that do not involve detailed collision physics, (3) compare Monte Carlo calculation methods with deterministic methods, and (4) teach fundamentals to students. In this work we describe 2 new tools for preparing the ACE cross-section files to be used by MCNP ® for these analytic test problems, simple a ce.pl and simple a ce m g.pl.
International Nuclear Information System (INIS)
Boia, Leonardo S.; Silva, Ademir X.
2009-01-01
It is possible nowadays to make changes in any digital image format due to the advancement of editing systems for images, with a little definition loss. Intending to increase the degrees of freedom on computer simulation fields, a process of integration of irregular geometries in the structure of medical DICOM images of the Anthropomorphic Rando Phantom making it so a cell is developed in this work and, therefore, the inclusion or change of the TLD's location in phantom for dosimetric studies, become a more dynamic simulation in MCNP. At first, creation and processing of the desired geometry are proceeded. It was coupled to the geometry in the study area of the DICOM image and the image's conversion into a MCNP input file was performed by software Scan2MCNP. Using the proposed computational process, a case of a clot and its ramifications was studied in Alderson Rando Phantom's left side brain area. (author)
International Nuclear Information System (INIS)
Hussein, M.S; Lewis, B.J.; Bonin, H.W.
2013-01-01
The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5 and the continuous-energy Monte Carlo reactor physics burnup calculation Serpent code. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as the Deuterium Critical Assembly, DCA. The multiplication factors k eff calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k eff calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k eff calculated with the Serpent code, with MCNP5, and from the coupled reactor theory. This analysis demonstrates that the Serpent code is valid for the multipoint coupled reactor calculations. (author)
Energy Technology Data Exchange (ETDEWEB)
Hussein, M.S, E-mail: mohamed.hussein@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Technology, Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada); Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)
2013-07-01
The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5 and the continuous-energy Monte Carlo reactor physics burnup calculation Serpent code. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as the Deuterium Critical Assembly, DCA. The multiplication factors k{sub eff} calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k{sub eff} calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k{sub eff} calculated with the Serpent code, with MCNP5, and from the coupled reactor theory. This analysis demonstrates that the Serpent code is valid for the multipoint coupled reactor calculations. (author)
Neutrons Flux Distributions of the Pu-Be Source and its Simulation by the MCNP-4B Code
Faghihi, F.; Mehdizadeh, S.; Hadad, K.
Neutron Fluence rate of a low intense Pu-Be source is measured by Neutron Activation Analysis (NAA) of 197Au foils. Also, the neutron fluence rate distribution versus energy is calculated using the MCNP-4B code based on ENDF/B-V library. Theoretical simulation as well as our experimental performance are a new experience for Iranians to make reliability with the code for further researches. In our theoretical investigation, an isotropic Pu-Be source with cylindrical volume distribution is simulated and relative neutron fluence rate versus energy is calculated using MCNP-4B code. Variation of the fast and also thermal neutrons fluence rate, which are measured by NAA method and MCNP code, are compared.
International Nuclear Information System (INIS)
Abdel Munem, E.; Shukri, A.; Tajuddin, A.A.
2006-01-01
A study of the thermal column of the Malaysian TRIGA Mark II reactor, forming part of a feasibility study for BNCT was proposed in 2001. In the current study, pure metals were used to measure the neutron flux at selected points in the thermal column and the neutron flux determined using SAND-II. Monte Carlo simulation of the thermal column was also carried out. The reactor core was homogenized and calculations of the neutron flux through the graphite stringers performed using MCNP5. The results show good agreement between the measured flux and the MCNP calculated flux. An obvious extension from this is that the MCNP neutron flux output can be utilized as an input spectrum for SAND-II for the flux iteration. (author)
MCNP Simulations of End Flux Peaking in ACR-1000, 2.4 wt % {sup 235}U Fuel Bundles
Energy Technology Data Exchange (ETDEWEB)
Hill, Ian; Donnelly, Jim [Atomic Energy of Canada Limited (AECL), 2251 Speakman Drive, Mississauga, ON, L5K 1B2 (Canada)
2008-07-01
This paper examines the end flux peaking in ACR-1000 fuel bundles. Reactor physics simulations are performed with MCNP to assess the steady state end-flux peaking in an infinite lattice of ACR fuel, as well as to quantify the peaking that occurs during refuelling. 3-dimensional MCNP models are created based on the detailed geometry of the fuel bundle. Detailed position-dependent fuel compositions are obtained from MONTEBURNS which couples MCNP and ORIGIN2.2. Axial and radial power profiles are obtained for both fresh and mid-burnup fuel bundles in an infinite lattice. Subsequently an assessment of the impact of a refuelling transient on the power profiles is performed. The refuelling transient is found to increase the end flux peaking in the region adjacent to light water. (authors)
Modeling of LVRF Critical Experiments in ZED-2 Using WIMS9A/PANTHER and MCNP5
International Nuclear Information System (INIS)
Sissaoui, M.T.; Lebenhaft, J.R; Carlson, P.A.
2008-01-01
The accuracy of WIMS9A/PANTHER and MCNP5 in modeling D 2 O-moderated, and H 2 O-, D 2 O- or air-cooled, doubly heterogeneous lattices of fuel clusters was demonstrated using Low Void Reactivity Fuel (LVRF) substitution experiments in the ZED-2 critical facility. MCNP5 with ENDF/B-VI (Release 5) under-predicted k eff but gave excellent coolant void reactivity (CVR) bias values. WIMS9A/PANTHER with JEF-2.2 over-predicted k eff and under-predicted the CVR bias relative to MCNP5 by 100 pcm to 200 pcm. Both codes reproduced the measured axial and radial flux shapes accurately. (authors)
Conceptual Nuclear Design Of Two Models Of Research Reactor Proposed For Vietnam
International Nuclear Information System (INIS)
Nguyen Nhi Dien; Huynh Ton Nghiem; Le Vinh Vinh; Vo Doan Hai Dang
2007-01-01
The joint study on the development of a new research reactor model for Vietnam was done. The KAERI (Korea Atomic Energy Research Institute) experts and DNRI (Dalat Nuclear Research Institute) researchers developed an advanced HANARO reactor (AHR), a 20-MW open-tank-in-pool type reactor, upward cooled and moderated by light water, reflected by heavy water and rod type fuel assemblies used. Based on the AHR model, a MTR reactor with plate fuel assemblies was developed. Computer codes named MCNP and MVP/BURN were used. Major analyses have been done for the relevant nuclear design parameters such as the neutron flux and power distributions, reactivity coefficients, control rod worth, etc. in both with clean, unperturbed core and equilibrium core condition. In case of AHR model, calculation results using MVP/BURN and MCNP codes were compared with the results using HELIOS and MCNP codes by KAERI experts and they are in a good agreement. (author)
Cai, Zhongli; Kwon, Yongkyu Luke; Reilly, Raymond M
2017-02-01
64 Cu emits positrons as well as β - particles and Auger and internal conversion electrons useful for radiotherapy. Our objective was to model the cellular dosimetry of 64 Cu under different geometries commonly used to study the cytotoxic effects of 64 Cu. Monte Carlo N-Particle (MCNP) was used to simulate the transport of all particles emitted by 64 Cu from the cell surface (CS), cytoplasm (Cy), or nucleus (N) of a single cell; monolayer in a well (radius = 0.32-1.74 cm); or a sphere (radius = 50-6,000 μm) of cells to calculate S values. The radius of the cell and N ranged from 5 to 12 μm and 2 to 11 μm, respectively. S values were obtained by MIRDcell for comparison. MCF7/HER2-18 cells were exposed in vitro to 64 Cu-labeled trastuzumab. The subcellular distribution of 64 Cu was measured by cell fractionation. The surviving fraction was determined in a clonogenic assay. The relative differences of MCNP versus MIRDcell self-dose S values (S self ) for 64 Cu ranged from -0.2% to 3.6% for N to N (S N←N ), 2.3% to 8.6% for Cy to N (S N←Cy ), and -12.0% to 7.3% for CS to N (S N←CS ). The relative differences of MCNP versus MIRDcell cross-dose S values were 25.8%-30.6% for a monolayer and 30%-34% for a sphere, respectively. The ratios of S N←N versus S N←Cy and S N←Cy versus S N←CS decreased with increasing ratio of the N of the cell versus radius of the cell and the size of the monolayer or sphere. The surviving fraction of MCF7 /: HER2-18 cells treated with 64 Cu-labeled trastuzumab (0.016-0.368 MBq/μg, 67 nM) for 18 h versus the absorbed dose followed a linear survival curve with α = 0.51 ± 0.05 Gy -1 and R 2 = 0.8838. This is significantly different from the linear quadratic survival curve of MCF7 /: HER2-18 cells exposed to γ-rays. MCNP- and MIRDcell-calculated S values agreed well. 64 Cu in the N increases the dose to the N in isolated single cells but has less effect in a cell monolayer or small cluster of cells simulating a micrometastasis
Zaker, Neda; Sina, Sedigheh; Koontz, Craig; Meigooni1, Ali S.
2016-01-01
Monte Carlo simulations are widely used for calculation of the dosimetric parameters of brachytherapy sources. MCNP4C2, MCNP5, MCNPX, EGS4, EGSnrc, PTRAN, and GEANT4 are among the most commonly used codes in this field. Each of these codes utilizes a cross‐sectional library for the purpose of simulating different elements and materials with complex chemical compositions. The accuracies of the final outcomes of these simulations are very sensitive to the accuracies of the cross‐sectional libraries. Several investigators have shown that inaccuracies of some of the cross section files have led to errors in 125I and 103Pd parameters. The purpose of this study is to compare the dosimetric parameters of sample brachytherapy sources, calculated with three different versions of the MCNP code — MCNP4C, MCNP5, and MCNPX. In these simulations for each source type, the source and phantom geometries, as well as the number of the photons, were kept identical, thus eliminating the possible uncertainties. The results of these investigations indicate that for low‐energy sources such as 125I and 103Pd there are discrepancies in gL(r) values. Discrepancies up to 21.7% and 28% are observed between MCNP4C and other codes at a distance of 6 cm for 103Pd and 10 cm for 125I from the source, respectively. However, for higher energy sources, the discrepancies in gL(r) values are less than 1.1% for 192Ir and less than 1.2% for 137Cs between the three codes. PACS number(s): 87.56.bg PMID:27074460
Zaker, Neda; Zehtabian, Mehdi; Sina, Sedigheh; Koontz, Craig; Meigooni, Ali S
2016-03-08
Monte Carlo simulations are widely used for calculation of the dosimetric parameters of brachytherapy sources. MCNP4C2, MCNP5, MCNPX, EGS4, EGSnrc, PTRAN, and GEANT4 are among the most commonly used codes in this field. Each of these codes utilizes a cross-sectional library for the purpose of simulating different elements and materials with complex chemical compositions. The accuracies of the final outcomes of these simulations are very sensitive to the accuracies of the cross-sectional libraries. Several investigators have shown that inaccuracies of some of the cross section files have led to errors in 125I and 103Pd parameters. The purpose of this study is to compare the dosimetric parameters of sample brachytherapy sources, calculated with three different versions of the MCNP code - MCNP4C, MCNP5, and MCNPX. In these simulations for each source type, the source and phantom geometries, as well as the number of the photons, were kept identical, thus eliminating the possible uncertainties. The results of these investigations indicate that for low-energy sources such as 125I and 103Pd there are discrepancies in gL(r) values. Discrepancies up to 21.7% and 28% are observed between MCNP4C and other codes at a distance of 6 cm for 103Pd and 10 cm for 125I from the source, respectively. However, for higher energy sources, the discrepancies in gL(r) values are less than 1.1% for 192Ir and less than 1.2% for 137Cs between the three codes.
Chibani, Omar; Li, X Allen
2002-05-01
Three Monte Carlo photon/electron transport codes (GEPTS, EGSnrc, and MCNP) are bench-marked against dose measurements in homogeneous (both low- and high-Z) media as well as at interfaces. A brief overview on physical models used by each code for photon and electron (positron) transport is given. Absolute calorimetric dose measurements for 0.5 and 1 MeV electron beams incident on homogeneous and multilayer media are compared with the predictions of the three codes. Comparison with dose measurements in two-layer media exposed to a 60Co gamma source is also performed. In addition, comparisons between the codes (including the EGS4 code) are done for (a) 0.05 to 10 MeV electron beams and positron point sources in lead, (b) high-energy photons (10 and 20 MeV) irradiating a multilayer phantom (water/steel/air), and (c) simulation of a 90Sr/90Y brachytherapy source. A good agreement is observed between the calorimetric electron dose measurements and predictions of GEPTS and EGSnrc in both homogeneous and multilayer media. MCNP outputs are found to be dependent on the energy-indexing method (Default/ITS style). This dependence is significant in homogeneous media as well as at interfaces. MCNP(ITS) fits more closely the experimental data than MCNP(DEF), except for the case of Be. At low energy (0.05 and 0.1 MeV), MCNP(ITS) dose distributions in lead show higher maximums in comparison with GEPTS and EGSnrc. EGS4 produces too penetrating electron-dose distributions in high-Z media, especially at low energy (MCNP results depend significantly on the electron energy-indexing method.
International Nuclear Information System (INIS)
Ibrahim, Ahmad M.; Polunovskiy, Eduard; Loughlin, Michael J.; Grove, Robert E.; Sawan, Mohamed E.
2016-01-01
Highlights: • Assess the detailed distribution of the nuclear heating among the components of the ITER toroidal field coils. • Utilize the FW-CADIS method to dramatically accelerate the calculation of detailed nuclear analysis. • Compare the efficiency and reliability of the FW-CADIS method and the MCNP weight window generator. - Abstract: Because the superconductivity of the ITER toroidal field coils (TFC) must be protected against local overheating, detailed spatial distribution of the TFC nuclear heating is needed to assess the acceptability of the designs of the blanket, vacuum vessel (VV), and VV thermal shield. Accurate Monte Carlo calculations of the distributions of the TFC nuclear heating are challenged by the small volumes of the tally segmentations and by the thick layers of shielding provided by the blanket and VV. To speed up the MCNP calculation of the nuclear heating distribution in different segments of the coil casing, ground insulation, and winding packs of the ITER TFC, the ITER Organization (IO) used the MCNP weight window generator (WWG). The maximum relative uncertainty of the tallies in this calculation was 82.7%. In this work, this MCNP calculation was repeated using variance reduction parameters generated by the Oak Ridge National Laboratory AutomateD VAriaNce reducTion Generator (ADVANTG) code and both MCNP calculations were compared in terms of computational efficiency and reliability. Even though the ADVANTG MCNP calculation used less than one-sixth of the computational resources of the IO calculation, the relative uncertainties of all the tallies in the ADVANTG MCNP calculation were less than 6.1%. The nuclear heating results of the two calculations were significantly different by factors between 1.5 and 2.3 in some of the segments of the furthest winding pack turn from the plasma neutron source. Even though the nuclear heating in this turn may not affect the ITER design because it is much smaller than the nuclear heating in the
Energy Technology Data Exchange (ETDEWEB)
Ibrahim, Ahmad M., E-mail: ibrahimam@ornl.gov [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Polunovskiy, Eduard; Loughlin, Michael J. [ITER Organization, Route de Vinon Sur Verdon, 13067 St. Paul Lez Durance (France); Grove, Robert E. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Sawan, Mohamed E. [University of Wisconsin-Madison, 1500 Engineering Dr., Madison, WI 53706 (United States)
2016-11-01
Highlights: • Assess the detailed distribution of the nuclear heating among the components of the ITER toroidal field coils. • Utilize the FW-CADIS method to dramatically accelerate the calculation of detailed nuclear analysis. • Compare the efficiency and reliability of the FW-CADIS method and the MCNP weight window generator. - Abstract: Because the superconductivity of the ITER toroidal field coils (TFC) must be protected against local overheating, detailed spatial distribution of the TFC nuclear heating is needed to assess the acceptability of the designs of the blanket, vacuum vessel (VV), and VV thermal shield. Accurate Monte Carlo calculations of the distributions of the TFC nuclear heating are challenged by the small volumes of the tally segmentations and by the thick layers of shielding provided by the blanket and VV. To speed up the MCNP calculation of the nuclear heating distribution in different segments of the coil casing, ground insulation, and winding packs of the ITER TFC, the ITER Organization (IO) used the MCNP weight window generator (WWG). The maximum relative uncertainty of the tallies in this calculation was 82.7%. In this work, this MCNP calculation was repeated using variance reduction parameters generated by the Oak Ridge National Laboratory AutomateD VAriaNce reducTion Generator (ADVANTG) code and both MCNP calculations were compared in terms of computational efficiency and reliability. Even though the ADVANTG MCNP calculation used less than one-sixth of the computational resources of the IO calculation, the relative uncertainties of all the tallies in the ADVANTG MCNP calculation were less than 6.1%. The nuclear heating results of the two calculations were significantly different by factors between 1.5 and 2.3 in some of the segments of the furthest winding pack turn from the plasma neutron source. Even though the nuclear heating in this turn may not affect the ITER design because it is much smaller than the nuclear heating in the
S values at voxels level for 188Re and 90Y calculated with the MCNP-4C code
International Nuclear Information System (INIS)
Coca Perez, Marco Antonio; Torres Aroche, Leonel Alberto; Cornejo, Nestor; Martin Hernandez, Guido
2003-01-01
The main objective of this work was estimate the voxels S values for 188 Re at cubical geometry using the MCNP-4C code for the simulation of radiation transport and energy deposition. Mean absorbed dose to target voxels per radioactive decay in a source voxels were estimated and reported for 188 Re and Y 90 . A comparison of voxels S values computed with the MCNP code the data reported in MIRD pamphlet 17 for 90 Y was performed in order to evaluate our results
International Nuclear Information System (INIS)
Yamazaki, Takao; Fujisaki, Masahide; Okuda, Motoi; Takano, Makoto; Masukawa, Fumihiro; Naito, Yoshitaka
1993-01-01
The general purpose Monte Carlo code MCNP4 has been implemented on the Fujitsu AP1000 distributed memory highly parallel computer. Parallelization techniques developed and studied are reported. A shielding analysis function of the MCNP4 code is parallelized in this study. A technique to map a history to each processor dynamically and to map control process to a certain processor was applied. The efficiency of parallelized code is up to 80% for a typical practical problem with 512 processors. These results demonstrate the advantages of a highly parallel computer to the conventional computers in the field of shielding analysis by Monte Carlo method. (orig.)
International Nuclear Information System (INIS)
Cramer, S.N.
1985-09-01
An overview of the RSIC-distributed version of the MCNP code (a soupled Monte Carlo neutron-photon code) is presented. All general features of the code, from machine hardware requirements to theoretical details, are discussed. The current nuclide cross-section and other libraries available in the standard code package are specified, and a realistic example of the flexible geometry input is given. Standard and nonstandard source, estimator, and variance-reduction procedures are outlined. Examples of correct usage and possible misuse of certain code features are presented graphically and in standard output listings. Finally, itemized summaries of sample problems, various MCNP code documentation, and future work are given
MCNP simulations of a glass display used in a mobile phone as an accident dosimeter
International Nuclear Information System (INIS)
Discher, Michael; Hiller, Mauritius; Woda, Clemens
2015-01-01
It has been demonstrated that glass display of mobile phones can be used as a device for accident dosimetry. Published studies concentrated on the experimental investigation of parts of the glass display. In the work presented here, the experimental results are compared with results of radiation transport calculations using the Monte Carlo code MCNP5. An experimental setup of an irradiation of an extracted glass display is simulated. The simulation is then extended to a simulation of a modern day mobile phone consisting of all major parts. Simulations are performed for various irradiation conditions and different geometric and material properties. The results of the simulation show a good agreement with the experiments for an extracted glass sample as well as for an actual modern mobile phone. The glass display is exposed to radiation in various angular and energy distributions. Simulated results were compared to experimentally determined results. The effects of the irradiation condition on the photon energy dependence were investigated and variations in the material constants of the display glass composition were discussed. This work affirms the usability of a mobile phone as a versatile and flexible accident radiation detector. - Highlights: • Simulations of a modern day mobile phone using MCNP are carried out. • Results of the simulation show a good agreement with the experiments. • Photon energy dependence and angular response for display glass are verified
Extensions of the MCNP5 and TRIPOLI4 Monte Carlo Codes for Transient Reactor Analysis
Hoogenboom, J. Eduard; Sjenitzer, Bart L.
2014-06-01
To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.
Simulation of gamma-ray irradiation of lettuce leaves in a 137Cs irradiator using MCNP
International Nuclear Information System (INIS)
Kim, Jongsoon; Moreira, Rosana G.; Braby, Leslie A.
2010-01-01
Ionizing radiation effectively reduces the number of common microbial pathogens in fresh produce. However, the efficacy of the process for pathogens internalized into produce tissue is unknown. The objective of this study was to understand gamma irradiation of lettuce leaf structure exposed in a 137 Cs irradiator using MCNP. The simulated 137 Cs irradiator is a self-shielded device, and its geometry and sources are described in the MCNP input file. When the irradiation chamber is filled with water, lower doses are found at the center of the irradiation volume and the dose uniformity ratio (maximum dose/minimum dose) is 1.76. For randomly oriented rectangular lettuce leaf segments in the irradiation chamber, the dose uniformity ratio is 1.25. It shows that dose uniformity in the Cs irradiator is strongly dependent of the density of the sample. To understand dose distribution inside the leaf, we divided a lettuce leaf into a low density (flat) region (0.72 g/cm 3 ) and high density (rib) region (0.86 g/cm 3 ). Calculated doses to the rib are 61% higher than doses to the flat region of the leaf. This indicates that internalized microorganisms can be inactivated more easily than organisms on the surface. This study shows that irradiation can effectively reduce viable microorganism internalized in lettuce. (author)
Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system
International Nuclear Information System (INIS)
Iga, Kiminori; Takada, Hiroshi; Nagao, Tadashi.
1998-01-01
In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B 4 C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)
A group of neutronics calculations in the MNSR using the MCNP-4C code
International Nuclear Information System (INIS)
Khattab, K.; Sulieman, I.
2009-11-01
The MCNP-4C code was used to model the 3-D core configuration for the Syrian Miniature Neutron Source Reactor (MNSR). The continuous energy neutron cross sections were evaluated from ENDF/B-VI library to calculate the thermal and fast neutron fluxes in the MNSR inner and outer irradiation sites. The thermal fluxes in the MNSR inner irradiation sites were measured for the first time using the multiple foil activation method. Good agreements were noticed between the calculated and measured results. This model is used as well to calculate neutron flux spectrum in the reactor inner and outer irradiation sites and the reactor thermal power. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed also to assess the possibility of fuel conversion from 89.87 % HEU fuel (UAl 4 -Al) to 19.75 % LEU fuel (UO 2 ). This model is used in this paper to calculate the following reactor core physics parameters: clean cold core excess reactivity, calibration of the control rod worth and calculation its shut down margin, calibration of the top beryllium shim plate reflector, axial neutron flux distributions in the inner and outer irradiation sites and the kinetics parameters ( ι p l and β e ff). (authors)
Schweda, K
2002-01-01
The analysis of (e,e'n) experiments at the Darmstadt superconducting electron linear accelerator S-DALINAC required the calculation of neutron response functions for the NE213 liquid scintillation detectors used. In an open geometry, these response functions can be obtained using the Monte Carlo codes NRESP7 and NEFF7. However, for more complex geometries, an extended version of the Monte Carlo code MCNP exists. This extended version of the MCNP code was improved upon by adding individual light-output functions for charged particles. In addition, more than one volume can be defined as a scintillator, thus allowing the simultaneous calculation of the response for multiple detector setups. With the implementation of sup 1 sup 2 C(n,n'3 alpha) reactions, all relevant reactions for neutron energies E sub n <20 MeV are now taken into consideration. The results of these calculations were compared to experimental data using monoenergetic neutrons in an open geometry and a sup 2 sup 5 sup 2 Cf neutron source in th...
Shahmohammadi Beni, Mehrdad; Ng, C Y P; Krstic, D; Nikezic, D; Yu, K N
2017-01-01
Radiotherapy is a common cancer treatment module, where a certain amount of dose will be delivered to the targeted organ. This is achieved usually by photons generated by linear accelerator units. However, radiation scattering within the patient's body and the surrounding environment will lead to dose dispersion to healthy tissues which are not targets of the primary radiation. Determination of the dispersed dose would be important for assessing the risk and biological consequences in different organs or tissues. In the present work, the concept of conversion coefficient (F) of the dispersed dose was developed, in which F = (Dd/Dt), where Dd was the dispersed dose in a non-targeted tissue and Dt is the absorbed dose in the targeted tissue. To quantify Dd and Dt, a comprehensive model was developed using the Monte Carlo N-Particle (MCNP) package to simulate the linear accelerator head, the human phantom, the treatment couch and the radiotherapy treatment room. The present work also demonstrated the feasibility and power of parallel computing through the use of the Message Passing Interface (MPI) version of MCNP5.
MCNP modelling of the wall effects observed in tissue-equivalent proportional counters.
Hoff, J L; Townsend, L W
2002-01-01
Tissue-equivalent proportional counters (TEPCs) utilise tissue-equivalent materials to depict homogeneous microscopic volumes of human tissue. Although both the walls and gas simulate the same medium, they respond to radiation differently. Density differences between the two materials cause distortions, or wall effects, in measurements, with the most dominant effect caused by delta rays. This study uses a Monte Carlo transport code, MCNP, to simulate the transport of secondary electrons within a TEPC. The Rudd model, a singly differential cross section with no dependence on electron direction, is used to describe the energy spectrum obtained by the impact of two iron beams on water. Based on the models used in this study, a wall-less TEPC had a higher lineal energy (keV.micron-1) as a function of impact parameter than a solid-wall TEPC for the iron beams under consideration. An important conclusion of this study is that MCNP has the ability to model the wall effects observed in TEPCs.
International Nuclear Information System (INIS)
Quade, U.
1994-01-01
Neutron- und Gamma dose rate calculations were performed for the storage containers filled with plutonium nitrate of the MOX fabrication facility of Siemens. For the particle transport calculations the Monte Carlo Code MCNP 4.2 was used. The calculated results were compared with experimental dose rate measurements. It can be stated that the choice of the code system was appropriate since all aspects of the many facettes of the problem were well reproduced in the calculations. The position dependency as well as the influence of the shieldings, the reflections and the mutual influences of the sources were well described by the calculations for the gamma and for the neutron dose rates. However, good agreement with the experimental results on the gamma dose rates could only be reached when the lead shielding of the detector was integrated into the geometry modelling of the calculations. For some few cases of thick shieldings and soft gamma ray sources the statistics of the calculational results were not sufficient. In such cases more elaborate variance reduction methods must be applied in future calculations. Thus the MCNP code in connection with NGSRC has been proven as an effective tool for the solution of this type of problems. (orig./HP) [de
A Monte-Carlo Benchmark of TRIPOLI-4® and MCNP on ITER neutronics
Blanchet, David; Pénéliau, Yannick; Eschbach, Romain; Fontaine, Bruno; Cantone, Bruno; Ferlet, Marc; Gauthier, Eric; Guillon, Christophe; Letellier, Laurent; Proust, Maxime; Mota, Fernando; Palermo, Iole; Rios, Luis; Guern, Frédéric Le; Kocan, Martin; Reichle, Roger
2017-09-01
Radiation protection and shielding studies are often based on the extensive use of 3D Monte-Carlo neutron and photon transport simulations. ITER organization hence recommends the use of MCNP-5 code (version 1.60), in association with the FENDL-2.1 neutron cross section data library, specifically dedicated to fusion applications. The MCNP reference model of the ITER tokamak, the `C-lite', is being continuously developed and improved. This article proposes to develop an alternative model, equivalent to the 'C-lite', but for the Monte-Carlo code TRIPOLI-4®. A benchmark study is defined to test this new model. Since one of the most critical areas for ITER neutronics analysis concerns the assessment of radiation levels and Shutdown Dose Rates (SDDR) behind the Equatorial Port Plugs (EPP), the benchmark is conducted to compare the neutron flux through the EPP. This problem is quite challenging with regard to the complex geometry and considering the important neutron flux attenuation ranging from 1014 down to 108 n•cm-2•s-1. Such code-to-code comparison provides independent validation of the Monte-Carlo simulations, improving the confidence in neutronic results.
Cagnazzo, M; Borio di Tigliole, A; Böck, H; Villa, M
2018-05-01
Aim of this work was the detection of fission products activity distribution along the axial dimension of irradiated fuel elements (FEs) at the TRIGA Mark II research reactor of the Technische Universität (TU) Wien. The activity distribution was measured by means of a customized fuel gamma scanning device, which includes a vertical lifting system to move the fuel rod along its vertical axis. For each investigated FE, a gamma spectrum measurement was performed along the vertical axis, with steps of 1 cm, in order to determine the axial distribution of the fission products. After the fuel elements underwent a relatively short cooling down period, different fission products were detected. The activity concentration was determined by calibrating the gamma detector with a standard calibration source of known activity and by MCNP6 simulations for the evaluation of self-absorption and geometric effects. Given the specific TRIGA fuel composition, a correction procedure is developed and used in this work for the measurement of the fission product Zr 95 . This measurement campaign is part of a more extended project aiming at the modelling of the TU Wien TRIGA reactor by means of different calculation codes (MCNP6, Serpent): the experimental results presented in this paper will be subsequently used for the benchmark of the models developed with the calculation codes. Copyright © 2018 Elsevier Ltd. All rights reserved.
An experimental test on large animals of MCNP application for whole body counting
International Nuclear Information System (INIS)
Borisov, N.; Yatsenko, V.; Kochetkov, O.; Gusev, I.; Vlasov, P.; Kalistratova, V.; Nisimov, P.; Levochkin, F.; Borovkov, M.; Stolyarov, V.; Tsedish, S.; Tyurin, I.; Franck, D.; Carlan, L. de
2005-01-01
Measurements of actinide body burden using whole body counting spectrometry is hampered due to intensive absorption of γ-rays inside the patient's body, which depends on the anatomy of a patient. To establish the correspondence between pulse-height-spectra intensity and radionuclide activity, Monte Carlo calculations are widely used. For such calculations, the radiation transport geometry is usually described in terms of small rectangular boxes (voxels) retrieved from computed tomography or magnetic resonance images. The software for Monte Carlo-assisted calibration of whole body counting, which performs automatic creation of individual MCNP voxel phantoms, was checked in a quasi-in vivo experiment on large animals. During the experiment, pigs of 35-40 kg body mass were used as phantoms for measurement of actinides body burden. 241 Am was administered (via injection of a radioactive solution or via implantation of plastic capsules containing the radioactive material) into the lungs of pigs. The pigs were measured using the pure germanium low-energy γ-spectrometers. The images of animals were obtained using the computed tomography machine. On the base of these tomograms, MCNP4c2 calculations were done to obtain the pulse-height-spectra of the whole body counters. The experimental results were reproduced in calculations with error of less than 30% for 241 Am administered via injection and less than 10% for 241 Am administered inside the capsules. (authors)
Comparison of ATTILA{sup TM} and MCNP{sup TM} for fusion applications
Energy Technology Data Exchange (ETDEWEB)
Loughlin, M. [UKAEA Fusion, Culham Science Centre, Abingdon, Oxfordshire, OX (United Kingdom); Wareing, T.; Barnett, A.; Failla, G.; McGhee, J. [Transpire Inc., Gig Harbor WA (United States)
2005-07-01
This paper describes comparison of the results of neutron transport calculations using two very different codes. ATTILA{sup TM} is a discrete ordinates radiation transport code which models complex 3-D geometries using arbitrary tetrahedra. MCNP{sup TM} is a Monte-Carlo radiation transport code which models the geometry using a combinatorial representation. This code is more widely known within the fusion community where it has been extensively used. In contrast, this is the first reporting of the use of ATTILA for fusion applications. The purpose of the work described herein was to compare calculations by each code of the neutron spectra at points around a greatly simplified representation of a typical fusion experiment. Spectra, in twenty-seven energy groups, were calculated at five locations which are typical of fusion neutronics problems; these are i) within the torus wall, ii) opposite a port, iii) near the torus hall floor, iv) at a straight penetration through the torus hall roof, and v) at the exit of a labyrinth through the wall. A solution was obtained from ATTILA in one 24 hour run on a single processor. An MCNP run of a similar duration was required on 18 parallel processors. Excellent agreement was obtained at all locations with only some minor disparities at thermal neutron energies. (authors)
International Nuclear Information System (INIS)
Khattab, K.; Boush, M.; Alkassiri, H.
2013-01-01
Highlights: • The MCNP4C was used to calculate the gamma ray dose rate spatial distribution in for the SGIF. • Measurement of the gamma ray dose rate spatial distribution using the Chlorobenzene dosimeter was conducted as well. • Good agreements were noticed between the calculated and measured results. • The maximum relative differences were less than 7%, 4% and 4% in the x, y and z directions respectively. - Abstract: A three dimensional model for the Syrian gamma irradiation facility (SGIF) is developed in this paper to calculate the gamma ray dose rate spatial distribution in the irradiation room at the 60 Co source board using the MCNP-4C code. Measurement of the gamma ray dose rate spatial distribution using the Chlorobenzene dosimeter is conducted as well to compare the calculated and measured results. Good agreements are noticed between the calculated and measured results with maximum relative differences less than 7%, 4% and 4% in the x, y and z directions respectively. This agreement indicates that the established model is an accurate representation of the SGIF and can be used in the future to make the calculation design for a new irradiation facility
Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system
Energy Technology Data Exchange (ETDEWEB)
Iga, Kiminori [Kyushu Univ., Fukuoka (Japan); Takada, Hiroshi; Nagao, Tadashi
1998-01-01
In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B{sub 4}C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)
International Nuclear Information System (INIS)
Deng Li; Xie Zhongsheng
1999-01-01
The coupled neutron and photon transport Monte Carlo code MCNP (version 3B) has been parallelized in parallel virtual machine (PVM) and message passing interface (MPI) by modifying a previous serial code. The new code has been verified by solving sample problems. The speedup increases linearly with the number of processors and the average efficiency is up to 99% for 12-processor. (author)
International Nuclear Information System (INIS)
Kim Jung-Do; Gil Choong-Sup
1996-01-01
JEF-1-based 50-group cross section library for fast reactor applications and point data library for continuous-energy Monte Carlo code MCNP have been generated using NJOY91.38 system. They have been examined by analyzing measured integral quantities such as criticality and central reaction rate ratios for 8 small fast critical assemblies. (author). 9 refs, 2 figs, 10 tabs
International Nuclear Information System (INIS)
Abella, Vicente; Miro, Rafael; Juste, Belen; Verdu, Gumersindo
2010-01-01
Multileaf collimators are used on linear accelerators to provide conformal shaping of radiotherapy treatment beams, being an important tool for radiation therapy dose delivery. In this work, a multileaf collimator has been designed and implemented in the MCNP model of an Elekta Precise Linear Accelerator and introduced in PLUNC, a set of software tools for radiotherapy treatment planning (RTP) which was coupled in previous works with MCNP5 (Monte Carlo N-Particle transport code), with the purpose of comparing its effect on deterministic and Monte Carlo dose calculations. A 3D Shepp-Logan phantom was utilized as the patient model for validation purposes. Once the multileaf collimator model is implemented in the PLUNC LINAC model, a series of Matlab interfaces extract phantom and beam information created with PLUNC during the treatment plan and write it in MCNP5 input deck format. After the Monte Carlo simulation is performed, results are input back again in PLUNC in order to continue with the plan evaluation. The comparison is made via mapping of dose distribution inside the phantom with different field sizes, utilizing the MCNP5 tool EMESH, superimposed mesh tally, which allows registering the results over the problem geometry. This work follows a valid methodology for multileaf LINAC MC calculations during radiation treatment plans. (author)
International Nuclear Information System (INIS)
Hossny, K.
2015-01-01
The purpose of this work is to validate MCNP5 libraries by simulating 4 detailed benchmark experiments and comparing MCNP5 results (each library) with the experimental results and also the previously validated codes for the same experiments MORET 4.A coupled with APOLLO2 (France), and MONK8 (UK). The reasons for difference between libraries are also investigated in this work. Investigating the reason for the differences between libraries will be done by specifying a different library for specific part (clad, fuel, light water) and checking the result deviation than the previously calculated result (with all parts of the same library). The investigated benchmark experiments are of single fuel rods arrays that are water-moderated and water-reflected. Rods contained low-enriched (4.738 wt.% 92 235 U)uranium dioxide (UO 2 ) fuel were clad with aluminum alloy AGS. These experiments were subcritical approaches extrapolated to critical, with the multiplication factor reached being very close to 1.000 (within 0.1%); the subcritical approach parameter was the water level. The studied four cases differ from each other in pitch, number of fuel rods and of course critical height of water. The results show that although library ENDF/B-IV lacks light water treatment card, however its results can be reliable as light water treatment library does not have significant differences from library to another, so it will not be necessary to specify light water treatment card. The main reason for differences between ENDF/B-V and ENDF/B-VI is light water material, especially the Hydrogen element. Specifying the library of Uranium is necessary in case of using library ENDF/B-IV. On the other hand it is not necessary to specify library of cladding material whatever the used library. Validated libraries are ENDF/BIV, ENDF/B-V and ENDF/B-VI with codes in MCNP 42C, 50C and 60C respectively. The presentation slides have been added to the article
Coupled MCNP - SAS-SFR calculations for sodium fast reactor core at steady-state - 15460
International Nuclear Information System (INIS)
Ponomarev, A.; Travleev, A.; Pfrang, W.; Sanchez, V.
2015-01-01
The prediction of core parameters at steady state is the first step when studying core accident transient behaviour. At this step thermal hydraulics (TH) and core geometry parameters are calculated corresponding to initial operating conditions. In this study we present the coupling of the SAS-SFR code to the Monte-Carlo neutron transport code MCNP at steady state together with application to the European Sodium Fast Reactor (ESFR). The SAS-SFR code employs a multi-channel core representation where each channel represents subassemblies with similar power, thermal-hydraulics and pin mechanics conditions. For every axial node of every channel the individual geometry and material compositions parameters are calculated in accord with power and cooling conditions. This requires supplying the SAS-SFR-code with nodal power values which should be calculated by neutron physics code with given realistic core parameters. In the conventional approach the neutron physics model employs some core averaged TH and geometry data (fuel temperature, coolant density, core axial and radial expansion). In this study we organize a new approach coupling the MCNP neutron physics models and the SAS-SFR models, so that calculations of power can be improved by using distributed core parameters (TH and geometry) taken from SAS-SFR. The MCNP code is capable to describe cores with distributed TH parameters and even to model non-uniform axial expansion of fuel subassemblies. In this way, core TH and geometrical data calculated by SAS-SFR are taken into account accurately in the neutronics model. The coupling implementation is done by data exchange between two codes with help of processing routines managed by driver routine. Currently it is model-specific and realized for the ESFR 'Reference Oxide' core. The Beginning-Of-Life core state is considered with 10 channel representation for fuel subassemblies. For this model several sets of coupled calculations are performed, in which different
A new effective Monte Carlo Midway coupling method in MCNP applied to a well logging problem
Energy Technology Data Exchange (ETDEWEB)
Serov, I.V.; John, T.M.; Hoogenboom, J.E
1998-12-01
The background of the Midway forward-adjoint coupling method including the black absorber technique for efficient Monte Carlo determination of radiation detector responses is described. The method is implemented in the general purpose MCNP Monte Carlo code. The utilization of the method is fairly straightforward and does not require any substantial extra expertise. The method was applied to a standard neutron well logging porosity tool problem. The results exhibit reliability and high efficiency of the Midway method. For the studied problem the efficiency gain is considerably higher than for a normal forward calculation, which is already strongly optimized by weight-windows. No additional effort is required to adjust the Midway model if the position of the detector or the porosity of the formation is changed. Additionally, the Midway method can be used with other variance reduction techniques if extra gain in efficiency is desired.
Estimation of Amount of Scattered Neutrons at Devices PFZ and GIT-12 by MCNP Simulations
Directory of Open Access Journals (Sweden)
Ondrej Šíla
2013-01-01
Full Text Available Our work is dedicated to pinch effect occurring during current discharge in deuterium plasma, and our results are connected with two devices – plasma focus PFZ, situated in the Faculty of Electrical Engineering, CTU, Prague, and Z-pinch GIT-12, which is situated in the Institute of High Current Electronics, Tomsk. During fusion reactions that proceed in plasma during discharge, neutrons are produced. We use neutrons as instrument for plasma diagnostics. Despite of the advantage that neutrons do not interact with electric and magnetic fields inside device, they are inevitably scattered by materials that are placed between their source and probe, and information about plasma from which they come from is distorted. For estimation of rate of neutron scattering we use MCNP code.
Parallel processing of Monte Carlo code MCNP for particle transport problem
Energy Technology Data Exchange (ETDEWEB)
Higuchi, Kenji; Kawasaki, Takuji
1996-06-01
It is possible to vectorize or parallelize Monte Carlo codes (MC code) for photon and neutron transport problem, making use of independency of the calculation for each particle. Applicability of existing MC code to parallel processing is mentioned. As for parallel computer, we have used both vector-parallel processor and scalar-parallel processor in performance evaluation. We have made (i) vector-parallel processing of MCNP code on Monte Carlo machine Monte-4 with four vector processors, (ii) parallel processing on Paragon XP/S with 256 processors. In this report we describe the methodology and results for parallel processing on two types of parallel or distributed memory computers. In addition, we mention the evaluation of parallel programming environments for parallel computers used in the present work as a part of the work developing STA (Seamless Thinking Aid) Basic Software. (author)
NaI(Tl) detectors modeling in MCNP-X and Gate/Geant4 codes
Energy Technology Data Exchange (ETDEWEB)
Affonso, Renato Raoni Werneck; Silva, Ademir Xavier da, E-mail: raoniwa@yahoo.com.br, E-mail: ademir@nuclear.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Salgado, Cesar Marques, E-mail: otero@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)
2017-07-01
NaI (Tl) detectors are widely used in gamma-ray densitometry, but their modeling in Monte Carlo codes, such as MCNP-X and Gate/Geant4, needs a lot of work and does not yield comparable results with experimental arrangements, possibly due to non-simulated physical phenomena, such as light transport within the scintillator. Therefore, it is necessary a methodology that positively impacts the results of the simulations while maintaining the real dimensions of the detectors and other objects to allow validating a modeling that matches up with the experimental arrangement. Thus, the objective of this paper is to present the studies conducted with the MCNPX and Gate/Geant4 codes, in which the comparisons of their results were satisfactory, showing that both can be used for the same purposes. (author)
EBR-II Static Neutronic Calculations by PHISICS / MCNP6 codes
Energy Technology Data Exchange (ETDEWEB)
Paolo Balestra; Carlo Parisi; Andrea Alfonsi
2016-02-01
The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) on the Shutdown Heat Removal Tests (SHRT) performed in the '80s at the Experimental fast Breeder Reactor EBR-II, USA. The scope of the CRP is to improve and validate the simulation tools for the study and the design of the liquid metal cooled fast reactors. Moreover, training of the next generation of fast reactor analysts is being also considered the other scope of the CRP. In this framework, a static neutronic model was developed, using state-of-the art neutron transport codes like SCALE/PHISICS (deterministic solution) and MCNP6 (stochastic solution). Comparison between both solutions is briefly illustrated in this summary.
Solution of the 'MIDICORE' WWER-1000 core periphery power distribution benchmark by KARATE and MCNP
International Nuclear Information System (INIS)
Temesvari, E.; Hegyi, G.; Hordosy, G.; Maraczy, C.
2011-01-01
The 'MIDICORE' WWER-1000 core periphery power distribution benchmark was proposed by Mr. Mikolas on the twentieth Symposium of AER in Finland in 2010. This MIDICORE benchmark is a two-dimensional calculation benchmark based on the WWER-1000 reactor core cold state geometry with taking into account the geometry of explicit radial reflector. The main task of the benchmark is to test the pin by pin power distribution in selected fuel assemblies at the periphery of the WWER-1000 core. In this paper we present our results (k eff , integral fission power) calculated by MCNP and the KARATE code system in KFKI-AEKI and the comparison to the preliminary reference Monte Carlo calculation results made by NRI, Rez. (Authors)
Calculation of age-dependent effective doses for external exposure using the MCNP code
International Nuclear Information System (INIS)
Hung, Tran Van
2013-01-01
Age-dependent effective dose for external exposure to photons uniformly distributed in air were calculated. Firstly, organ doses were calculated with a series of age-specific MIRD-5 type phantoms using the Monte Carlo code MCNP. The calculations were performed for mono-energetic photon sources with source energies from 10 keV to 5 MeV and for phantoms of newborn, 1, 5, 10, and 15 years-old and adult. Then, the effective doses to the different age-phantoms from the mono-energetic photon sources were estimated based on the obtained organ doses. From the calculated results, it is shown that the effective doses depend on the body size; the effective doses in younger phantoms are higher than those in the older phantoms, especially below 100 keV. (orig.)
International Nuclear Information System (INIS)
Sakurai, Kiyoshi; Yamamoto, Toshihiro; Arakawa, Takuya; Naito, Yoshitaka
1998-01-01
Experiments on coupled cores performed at TCA were analysed using continuous energy Monte Carlo calculation code MCNP 4A. Errors of neutron multiplication factors are evaluated using Indirect Bias Estimation Method proposed by authors. Calculation for simulation of pulsed neutron method was performed for 17 X 17 + 5G + 17 x 17 core system and its of exponential experiment method was also performed for 16 x 9 + 3G + 16 x 9 and 16 x 9 + 5G + 16 x 9 core systems. Errors of neutron multiplication factors are estimated to be (-1.5) - (-0.6)% evaluated by Indirect Bias Estimation Method. Its errors evaluated by conventional pulsed neutron method and exponential experiment method are estimated to be 7%, but it is below 1% for estimation of subcriticality with the computed values by applying Indirect Bias Estimation Method. Feasibility of subcriticality management is higher by application of the method to full scale fuel strage facility. (author)
MCNP study for epithermal neutron irradiation of an isolated liver at the Finnish BNCT facility.
Kotiluoto, P; Auterinen, I
2004-11-01
A successful boron neutron capture treatment (BNCT) of a patient with multiple liver metastases has been first given in Italy, by placing the removed organ into the thermal neutron column of the Triga research reactor of the University of Pavia. In Finland, FiR 1 Triga reactor with an epithermal neutron beam well suited for BNCT has been extensively used to irradiate patients with brain tumors such as glioblastoma and recently also head and neck tumors. In this work we have studied by MCNP Monte Carlo simulations, whether it would be beneficial to treat an isolated liver with epithermal neutrons instead of thermal ones. The results show, that the epithermal field penetrates deeper into the liver and creates a build-up distribution of the boron dose. Our results strongly encourage further studying of irradiation arrangement of an isolated liver with epithermal neutron fields.
Pebble bed reactors simulation using MCNP: The Chinese HTR-10 reactor
Directory of Open Access Journals (Sweden)
SA Hosseini
2013-09-01
Full Text Available Given the role of Gas-Graphite reactors as the fourth generation reactors and their recently renewed importance, in 2002 the IAEA proposed a set of Benchmarking problems. In this work, we propose a model both efficient in time and resources and exact to simulate the HTR-10 reactor using MCNP-4C code. During the present work, all of the pressing factors in PBM reactor design such as the inter-pebble leakage, fuel particle distribution and fuel pebble packing fraction effects have been taken into account to obtain an exact and easy to run model. Finally, the comparison between the results of the present work and other calculations made at INEEL proves the exactness of the proposed model.
Analysis of MCNP simulated gamma spectra of CdTe detectors for boron neutron capture therapy.
Winkler, Alexander; Koivunoro, Hanna; Savolainen, Sauli
2017-06-01
The next step in the boron neutron capture therapy (BNCT) is the real time imaging of the boron concentration in healthy and tumor tissue. Monte Carlo simulations are employed to predict the detector response required to realize single-photon emission computed tomography in BNCT, but have failed to correctly resemble measured data for cadmium telluride detectors. In this study we have tested the gamma production cross-section data tables of commonly used libraries in the Monte Carlo code MCNP in comparison to measurements. The cross section data table TENDL-2008-ACE is reproducing measured data best, whilst the commonly used ENDL92 and other studied libraries do not include correct tables for the gamma production from the cadmium neutron capture reaction that is occurring inside the detector. Furthermore, we have discussed the size of the annihilation peaks of spectra obtained by cadmium telluride and germanium detectors. Copyright © 2017 Elsevier Ltd. All rights reserved.
Calculation of conversion coefficients for clinical photon spectra using the MCNP code.
Lima, M A F; Silva, A X; Crispim, V R
2004-01-01
In this work, the MCNP4B code has been employed to calculate conversion coefficients from air kerma to the ambient dose equivalent, H*(10)/Ka, for monoenergetic photon energies from 10 keV to 50 MeV, assuming the kerma approximation. Also estimated are the H*(10)/Ka for photon beams produced by linear accelerators, such as Clinac-4 and Clinac-2500, after transmission through primary barriers of radiotherapy treatment rooms. The results for the conversion coefficients for monoenergetic photon energies, with statistical uncertainty <2%, are compared with those in ICRP publication 74 and good agreements were obtained. The conversion coefficients calculated for real clinic spectra transmitted through walls of concrete of 1, 1.5 and 2 m thick, are in the range of 1.06-1.12 Sv Gy(-1).
Simulations of neutron transport at low energy: a comparison between GEANT and MCNP.
Colonna, N; Altieri, S
2002-06-01
The use of the simulation tool GEANT for neutron transport at energies below 20 MeV is discussed, in particular with regard to shielding and dose calculations. The reliability of the GEANT/MICAP package for neutron transport in a wide energy range has been verified by comparing the results of simulations performed with this package in a wide energy range with the prediction of MCNP-4B, a code commonly used for neutron transport at low energy. A reasonable agreement between the results of the two codes is found for the neutron flux through a slab of material (iron and ordinary concrete), as well as for the dose released in soft tissue by neutrons. These results justify the use of the GEANT/MICAP code for neutron transport in a wide range of applications, including health physics problems.
Calculation of age-dependent effective doses for external exposure using the MCNP code
Energy Technology Data Exchange (ETDEWEB)
Hung, Tran Van [Research and Development Center for Radiation Technology, ThuDuc, HoChiMinh City (VT)
2013-07-15
Age-dependent effective dose for external exposure to photons uniformly distributed in air were calculated. Firstly, organ doses were calculated with a series of age-specific MIRD-5 type phantoms using the Monte Carlo code MCNP. The calculations were performed for mono-energetic photon sources with source energies from 10 keV to 5 MeV and for phantoms of newborn, 1, 5, 10, and 15 years-old and adult. Then, the effective doses to the different age-phantoms from the mono-energetic photon sources were estimated based on the obtained organ doses. From the calculated results, it is shown that the effective doses depend on the body size; the effective doses in younger phantoms are higher than those in the older phantoms, especially below 100 keV. (orig.)
Image enhancement using MCNP5 code and MATLAB in neutron radiography.
Tharwat, Montaser; Mohamed, Nader; Mongy, T
2014-07-01
This work presents a method that can be used to enhance the neutron radiography (NR) image for objects with high scattering materials like hydrogen, carbon and other light materials. This method used Monte Carlo code, MCNP5, to simulate the NR process and get the flux distribution for each pixel of the image and determines the scattered neutron distribution that caused image blur, and then uses MATLAB to subtract this scattered neutron distribution from the initial image to improve its quality. This work was performed before the commissioning of digital NR system in Jan. 2013. The MATLAB enhancement method is quite a good technique in the case of static based film neutron radiography, while in neutron imaging (NI) technique, image enhancement and quantitative measurement were efficient by using ImageJ software. The enhanced image quality and quantitative measurements were presented in this work. Copyright © 2014 Elsevier Ltd. All rights reserved.
Application of MCNP code in shielding calculation of minitype fast reactor
International Nuclear Information System (INIS)
He Keyu; Han Weishi
2008-01-01
An accurate shielding calculation model has been set up for the minitype sodium-cooled fast reactor (MFR) based on MCNP code and particular calculation of its primary shielding parameters has been carried out. The results indicate that the photon and neutron flux density of MFR has rapidly fallen to a low-level. The material for the shielding layer outside of main container is primarily of carbon steel, which can be design as a shielding structure satisfying the safety code. The sodium activation in primary circuit is extremely limited and it is simple to shield from. Both the output of helium in reflector and burn up of boron-10 in control rod are very small. These materials can be used for several cycle lives. (authors)
Shielding property of bismuth glass based on MCNP 5 and WINXCOM simulated calculation
International Nuclear Information System (INIS)
Zhang Zhicheng; Zhang Jinzhao; Liu Ze; Lu Chunhai; Chen Min
2013-01-01
Background: Currently, lead glass is widely used as observation window, while lead is toxic heavy metal. Purpose: Non-toxic materials and their shielding effects are researched in order to find a new material to replace lead containing material. Methods: The mass attenuation coefficients of bismuth silicate glass were investigated with gamma-ray's energy at 0.662 MeV, 1.17 MeV and 1.33 MeV, respectively, by MCNP 5 (Monte Carlo) and WINXCOM program, and compared with those of the lead glass. Results: With attenuation factor K, shielding and mechanical properties taken into consideration bismuth glass containing 50% bismuth oxide might be selected as the right material. Dose rate distributions of water phantom were calculated with 2-cm and 10-cm thick glass, respectively, irradiated by 137 Cs and 60 Co in turn. Conclusion: Results show that the bismuth glass may replace lead glass for radiation shielding with appropriate energy. (authors)
Skyshine analysis using various nuclear data files
Energy Technology Data Exchange (ETDEWEB)
Zharkov, V.P.; Dikareva, O.F.; Kartashev, I.A.; Kiselev, A.N. [Research and Development Inst. of Power Engineering, Moscow (Russian Federation); Nomura, Y.; Tsubosaka, A. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)
2000-03-01
The calculations of the spacial distributions of dose rate for neutron and secondary photons, thermal neutron fluxes and space-energy distributions of neutron and photons near the air-ground interface were performed by MCNP and DORT codes. Different nuclear data files were used (ENDF/B-IV, ENDF/B-VI, FENDL-2, JENDL-3.2). Either the standard pointwise libraries (MCNP) or special libraries prepared by NJOY code from ENDF/B and others' files were used. Prepared multigroup coupled neutron and photon cross sections libraries for DORT code had CASK-40 group energy structures. The libraries contain pointwise or multigroup cross sections data for all elements included in the atmosphere and ground composition. The validation of the calculated results was performed with using the experimental data obtained for the series of measurements at RA reactor. (author)
Nuclear materials identification by photon interrogation
International Nuclear Information System (INIS)
Pozzi, S.A.; Monville, M.; Padovani, E.
2005-01-01
We describe a preliminary modification to the Monte Carlo codes MCNP-X and MCNP-PoliMi that is aimed at simulating the neutron and photon field generated by interrogating fissile (and non-fissile) material with a high energy photon source. Photo-atomic and photo-nuclear collisions are modeled, with particular emphasis on the generation of secondary particles that are emitted as a result of these interactions. The simulations can be used to design and analyze measurements that are performed in a wide variety of scenarios. An application of the methodology to the interrogation of packages on a luggage belt conveyor is presented. Preliminary results show that it is possible to detect 5 Kg of highly enriched uranium in a package by measuring the correlation function between 2 detectors. This correlation function is based on the detection of prompt radiation from photonuclear events
Skyshine analysis using various nuclear data files
International Nuclear Information System (INIS)
Zharkov, V.P.; Dikareva, O.F.; Kartashev, I.A.; Kiselev, A.N.; Nomura, Y.; Tsubosaka, A.
2000-01-01
The calculations of the spacial distributions of dose rate for neutron and secondary photons, thermal neutron fluxes and space-energy distributions of neutron and photons near the air-ground interface were performed by MCNP and DORT codes. Different nuclear data files were used (ENDF/B-IV, ENDF/B-VI, FENDL-2, JENDL-3.2). Either the standard pointwise libraries (MCNP) or special libraries prepared by NJOY code from ENDF/B and others' files were used. Prepared multigroup coupled neutron and photon cross sections libraries for DORT code had CASK-40 group energy structures. The libraries contain pointwise or multigroup cross sections data for all elements included in the atmosphere and ground composition. The validation of the calculated results was performed with using the experimental data obtained for the series of measurements at RA reactor. (author)
Computational methods for nuclear criticality safety analysis
International Nuclear Information System (INIS)
Maragni, M.G.
1992-01-01
Nuclear criticality safety analyses require the utilization of methods which have been tested and verified against benchmarks results. In this work, criticality calculations based on the KENO-IV and MCNP codes are studied aiming the qualification of these methods at the IPEN-CNEN/SP and COPESP. The utilization of variance reduction techniques is important to reduce the computer execution time, and several of them are analysed. As practical example of the above methods, a criticality safety analysis for the storage tubes for irradiated fuel elements from the IEA-R1 research has been carried out. This analysis showed that the MCNP code is more adequate for problems with complex geometries, and the KENO-IV code shows conservative results when it is not used the generalized geometry option. (author)
MCNP Code in Assessment of Variations of Effective Dose with Torso Adipose Tissue Thickness
International Nuclear Information System (INIS)
Massoud, E.
2005-01-01
The effective dose is the unite used in the field of radiation protection. It is a well defined doubly weighted uantity involving both physical and biological variables. Several factors may induce variation in the effective dose in different individuals of similar exposure data. One of these factors is the variation of adipose tissue thickness in different exposed individuals. This study essentially concenrs the assessment of the possible variation in the effective dose due to variation in the thickness of adipose tissue. The study was done using MCNP4b code to perform mathematical model of the human body depending on that given to the reference man developed by International Commission of Radiological Protection (ICRP), and calculate the effective dose with different thicknessess of adipose tissues. The study includes a comprehensive appraisal of the Monte Cario simulation, the Medical Internal Radiation Dose (MIRD) model for the human body, and the various mathematical considerations involved in the radiation dose calculations for the various pertinent parts of the human body. The radiation energies considered were 80 KeV, 300 KeV and I MeV, applying two exposure positions; anteroposterior (AP), postero-anterior (PA) with different adipose tissue thickness. This study is a theoretical approach based on detailed mathematical calculations of great precision that deals with all considerations involved in the mechanisms of radiation energy absorption in biological system depending on the variation in the densities of the particular in biological system depending on the variation in the densities of the particular tissues. The results obtained indicate that maximum decrease in effective dose occures with the lowest energy at 5cm adipose tissues thickeness for both AP and PA exposure positions. The results obtained were compared to similar work previsouly done using MCNP4 b showing very good agreement
Extensions of the MCNP5 and TRIPOLI4 Monte Carlo codes for transient reactor analysis
International Nuclear Information System (INIS)
Hoogenboom, J.E.
2013-01-01
To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branch-less collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires the coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3*3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3*3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail. (authors)
Assessment of doses caused by electrons in thin layers of tissue-equivalent materials, using MCNP.
Heide, Bernd
2013-10-01
Absorbed doses caused by electron irradiation were calculated with Monte Carlo N-Particle transport code (MCNP) for thin layers of tissue-equivalent materials. The layers were so thin that the calculation of energy deposition was on the border of the scope of MCNP. Therefore, in this article application of three different methods of calculation of energy deposition is discussed. This was done by means of two scenarios: in the first one, electrons were emitted from the centre of a sphere of water and also recorded in that sphere; and in the second, an irradiation with the PTB Secondary Standard BSS2 was modelled, where electrons were emitted from an (90)Sr/(90)Y area source and recorded inside a cuboid phantom made of tissue-equivalent material. The speed and accuracy of the different methods were of interest. While a significant difference in accuracy was visible for one method in the first scenario, the difference in accuracy of the three methods was insignificant for the second one. Considerable differences in speed were found for both scenarios. In order to demonstrate the need for calculating the dose in thin small zones, a third scenario was constructed and simulated as well. The third scenario was nearly equal to the second one, but a pike of lead was assumed to be inside the phantom in addition. A dose enhancement (caused by the pike of lead) of ∼113 % was recorded for a thin hollow cylinder at a depth of 0.007 cm, which the basal-skin layer is referred to in particular. Dose enhancements between 68 and 88 % were found for a slab with a radius of 0.09 cm for all depths. All dose enhancements were hardly noticeable for a slab with a cross-sectional area of 1 cm(2), which is usually applied to operational radiation protection.
MCNP: a general Monte Carlo code for neutron and photon transport
International Nuclear Information System (INIS)
1978-07-01
The general-purpose Monte Carlo code MCNP can be used for neutron, photon, or coupled neutron--photon transport. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation (such as ENDF/B-IV) are accounted for. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. Standard optional variance reduction schemes include geometry splitting and Russian roulette, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point detectors, track-length estimators, and source biasing. The standard output of MCNP includes two-way current as a function of energy, time, and angle with the normal, across any subset of bounding surfaces in the problem. Fluxes across any set of bounding surfaces are available as a function of time and energy. Similarly, the flux at designated points and the average flux in a cell (track length per unit volume) are standard tallies. Reactions such as fissions or absorptions may be obtained in a subset of geometric cells. The heating tallies give the energy deposition per starting particle. In addition, particles may be flagged when they cross specified surfaces or enter designated cells, and the contributions of these flagged particles to certain of the tallies are listed separately. All quantities printed out have their relative errors listed also. 11 figures, 27 tables
Energy Technology Data Exchange (ETDEWEB)
Nagao, Y. E-mail: nagao@jmtr.oarai.jaeri.go.jp; Nakamichi, K.; Tsuchiya, M.; Ishitsuka, E.; Kawamura, H
2000-11-01
To evaluate exactly the total amount of tritium production in tritium breeding materials during in-pile test with JMTR, the 'tritium monitor' has been produced and evaluation of total tritium generation was done by using 'tritium monitor' in preliminary in-pile mock-up, and verification of procedure concerning tritium production evaluation was conducted by using Monte Carlo code MCNP and nuclear cross section library of FSXLIBJ3R2. Li-Al alloy (Li 3.4 wt.%, 95.5% enrichment of {sup 6}Li) was selected as tritium monitor material for the evaluation on the total amount of tritium production in high {sup 6}Li enriched materials. From the results of preliminary experiment, calculated amounts of total tritium production at each 'tritium monitor', which was installed in the preliminary in-pile mock-up, were about 50-290% higher than the measured values. Concerning tritium measurement, increase of measurement error in tritium leak form measuring system to measure small amount of tritium (0.2-0.7 mCi in tritium monitor) was found in the results of present experiment. The tendency for overestimation of calculated thermal neutron flux in the range of 1-6x10{sup 13} n cm{sup -2} per s was found in JMTR and the reason may be due to the beryllium cross section data base in JENDL3.2.