Preliminary results of 3D dose calculations with MCNP-4B code from a SPECT image.

PubMed

Rodríguez Gual, M; Lima, F F; Sospedra Alfonso, R; González González, J; Calderón Marín, C

2004-01-01

Interface software was developed to generate the input file to run Monte Carlo MCNP-4B code from medical image in Interfile format version 3.3. The software was tested using a spherical phantom of tomography slides with known cumulated activity distribution in Interfile format generated with IMAGAMMA medical image processing system. The 3D dose calculation obtained with Monte Carlo MCNP-4B code was compared with the voxel S factor method. The results show a relative error between both methods less than 1 %.

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

PubMed

Henry, R; Tiselj, I; Snoj, L

2015-03-01

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

MCNP4A: Features and philosophy

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

Hendricks, J.S.

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, newmore » 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.« less

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

NASA Astrophysics Data System (ADS)

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

2017-08-01

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

Modification and benchmarking of MCNP for low-energy tungsten spectra.

PubMed

Mercier, J R; Kopp, D T; McDavid, W D; Dove, S B; Lancaster, J L; Tucker, D M

2000-12-01

The MCNP Monte Carlo radiation transport code was modified for diagnostic medical physics applications. In particular, the modified code was thoroughly benchmarked for the production of polychromatic tungsten x-ray spectra in the 30-150 kV range. Validating the modified code for coupled electron-photon transport with benchmark spectra was supplemented with independent electron-only and photon-only transport benchmarks. Major revisions to the code included the proper treatment of characteristic K x-ray production and scoring, new impact ionization cross sections, and new bremsstrahlung cross sections. Minor revisions included updated photon cross sections, electron-electron bremsstrahlung production, and K x-ray yield. The modified MCNP code is benchmarked to electron backscatter factors, x-ray spectra production, and primary and scatter photon transport.

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

DTIC Science & Technology

2014-01-01

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

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

PubMed

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

2012-10-01

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

MCNP6 Simulation of Light and Medium Nuclei Fragmentation at Intermediate Energies

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

Mashnik, Stepan Georgievich; Kerby, Leslie Marie

2015-05-22

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

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

PubMed

Li, Fusheng; Han, Xiaogang

2012-07-01

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

Performance of MCNP4A on seven computing platforms

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

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

1994-12-31

The performance of seven computer platforms has been evaluated with the MCNP4A Monte Carlo radiation transport code. For the first time we report timing results using MCNP4A and its new test set and libraries. Comparisons are made on platforms not available to us in previous MCNP timing studies. By using MCNP4A and its 325-problem test set, a widely-used and readily-available physics production code is used; the timing comparison is not limited to a single ``typical`` problem, demonstrating the problem dependence of timing results; the results are reproducible at the more than 100 installations around the world using MCNP; comparison ofmore » performance of other computer platforms to the ones tested in this study is possible because we present raw data rather than normalized results; and a measure of the increase in performance of computer hardware and software over the past two years is possible. The computer platforms reported are the Cray-YMP 8/64, IBM RS/6000-560, Sun Sparc10, Sun Sparc2, HP/9000-735, 4 processor 100 MHz Silicon Graphics ONYX, and Gateway 2000 model 4DX2-66V PC. In 1991 a timing study of MCNP4, the predecessor to MCNP4A, was conducted using ENDF/B-V cross-section libraries, which are export protected. The new study is based upon the new MCNP 25-problem test set which utilizes internationally available data. MCNP4A, its test problems and the test data library are available from the Radiation Shielding and Information Center in Oak Ridge, Tennessee, or from the NEA Data Bank in Saclay, France. Anyone with the same workstation and compiler can get the same test problem sets, the same library files, and the same MCNP4A code from RSIC or NEA and replicate our results. And, because we report raw data, comparison of the performance of other compute platforms and compilers can be made.« less

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

NASA Astrophysics Data System (ADS)

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

2018-01-01

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

A comparison of the COG and MCNP codes in computational neutron capture therapy modeling, Part I: boron neutron capture therapy models.

PubMed

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

2005-08-01

The goal of this study was to evaluate the COG Monte Carlo radiation transport code, developed and tested by Lawrence Livermore National Laboratory, for neutron capture therapy related modeling. A boron neutron capture therapy model was analyzed comparing COG calculational results to results from the widely used MCNP4B (Monte Carlo N-Particle) transport code. The approach for computing neutron fluence rate and each dose component relevant in boron neutron capture therapy is described, and calculated values are shown in detail. The differences between the COG and MCNP predictions are qualified and quantified. The differences are generally small and suggest that the COG code can be applied for BNCT research related problems.

MCNP-model for the OAEP Thai Research Reactor

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

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

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

Considerations of MCNP Monte Carlo code to be used as a radiotherapy treatment planning tool.

PubMed

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

2005-01-01

The present work has simulated the photon and electron transport in a Theratron 780® (MDS Nordion)⁶⁰Co radiotherapy unit, using the Monte Carlo transport code, MCNP (Monte Carlo N-Particle). This project explains mainly the different methodologies carried out to speedup calculations in order to apply this code efficiently in radiotherapy treatment planning.

Gamma irradiator dose mapping simulation using the MCNP code and benchmarking with dosimetry.

PubMed

Sohrabpour, M; Hassanzadeh, M; Shahriari, M; Sharifzadeh, M

2002-10-01

The Monte Carlo transport code, MCNP, has been applied in simulating dose rate distribution in the IR-136 gamma irradiator system. Isodose curves, cumulative dose values, and system design data such as throughputs, over-dose-ratios, and efficiencies have been simulated as functions of product density. Simulated isodose curves, and cumulative dose values were compared with dosimetry values obtained using polymethyle-methacrylate, Fricke, ethanol-chlorobenzene, and potassium dichromate dosimeters. The produced system design data were also found to agree quite favorably with those of the system manufacturer's data. MCNP has thus been found to be an effective transport code for handling of various dose mapping excercises for gamma irradiators.

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

PubMed

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

2006-02-01

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

How to Build MCNP 6.2

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

Bull, Jeffrey S.

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

Criticality Calculations with MCNP6 - Practical Lectures

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

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

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 modelmore » 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.« less

MCNP6 Fission Multiplicity with FMULT Card

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

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

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

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

NASA Astrophysics Data System (ADS)

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

2018-05-01

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

Elaborate SMART MCNP Modelling Using ANSYS and Its Applications

NASA Astrophysics Data System (ADS)

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

2017-09-01

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

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

NASA Astrophysics Data System (ADS)

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

2011-10-01

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

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

PubMed

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

2005-08-01

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

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

PubMed

Chibani, Omar; Li, X Allen

2002-05-01

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

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

PubMed

Eakins, Jonathan

2009-02-01

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

Features of MCNP6 Relevant to Medical Radiation Physics

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

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

2012-08-29

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

Source terms, shielding calculations and soil activation for a medical cyclotron.

PubMed

Konheiser, J; Naumann, B; Ferrari, A; Brachem, C; Müller, S E

2016-12-01

Calculations of the shielding and estimates of soil activation for a medical cyclotron are presented in this work. Based on the neutron source term from the 18 O(p,n) 18 F reaction produced by a 28 MeV proton beam, neutron and gamma dose rates outside the building were estimated with the Monte Carlo code MCNP6 (Goorley et al 2012 Nucl. Technol. 180 298-315). The neutron source term was calculated with the MCNP6 code and FLUKA (Ferrari et al 2005 INFN/TC_05/11, SLAC-R-773) code as well as with supplied data by the manufacturer. MCNP and FLUKA calculations yielded comparable results, while the neutron yield obtained using the manufacturer-supplied information is about a factor of 5 smaller. The difference is attributed to the missing channels in the manufacturer-supplied neutron source terms which considers only the 18 O(p,n) 18 F reaction, whereas the MCNP and FLUKA calculations include additional neutron reaction channels. Soil activation was performed using the FLUKA code. The estimated dose rate based on MCNP6 calculations in the public area is about 0.035 µSv h -1 and thus significantly below the reference value of 0.5 µSv h -1 (2011 Strahlenschutzverordnung, 9 Auflage vom 01.11.2011, Bundesanzeiger Verlag). After 5 years of continuous beam operation and a subsequent decay time of 30 d, the activity concentration of the soil is about 0.34 Bq g -1 .

Evaluation of RAPID for a UNF cask benchmark problem

NASA Astrophysics Data System (ADS)

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

2017-09-01

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

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

PubMed

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

2008-01-01

In this work, the general purpose Monte Carlo N-particle radiation transport computer code (MCNP-4C) was used for the simulation of X-ray spectra in diagnostic radiology. The electron's path in the target was followed until its energy was reduced to 10 keV. A user-friendly interface named 'diagnostic X-ray spectra by Monte Carlo simulation (DXRaySMCS)' was developed to facilitate the application of MCNP-4C code for diagnostic radiology spectrum prediction. The program provides a user-friendly interface for: (i) modifying the MCNP input file, (ii) launching the MCNP program to simulate electron and photon transport and (iii) processing the MCNP output file to yield a summary of the results (relative photon number per energy bin). In this article, the development and characteristics of DXRaySMCS are outlined. As part of the validation process, output spectra for 46 diagnostic radiology system settings produced by DXRaySMCS were compared with the corresponding IPEM78. Generally, there is a good agreement between the two sets of spectra. No statistically significant differences have been observed between IPEM78 reported spectra and the simulated spectra generated in this study.

The MCNP6 Analytic Criticality Benchmark Suite

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

Brown, Forrest B.

2016-06-16

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

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

PubMed

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

2004-02-07

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

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

NASA Astrophysics Data System (ADS)

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

2014-06-01

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

Prompt Radiation Protection Factors

DTIC Science & Technology

2018-02-01

dimensional Monte-Carlo radiation transport code MCNP (Monte Carlo N-Particle) and the evaluation of the protection factors (ratio of dose in the open to...radiation was performed using the three dimensional Monte- Carlo radiation transport code MCNP (Monte Carlo N-Particle) and the evaluation of the protection...by detonation of a nuclear device have placed renewed emphasis on evaluation of the consequences in case of such an event. The Defense Threat

Geometry creation for MCNP by Sabrina and XSM

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

Van Riper, K.A.

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

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

PubMed

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

1998-01-01

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

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

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

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

1999-07-25

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

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

Pinilla, Maria Isabel

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

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

DOE PAGES

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

2017-03-23

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

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

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

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

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

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

PubMed

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

2011-03-01

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

Radiation shielding quality assurance

NASA Astrophysics Data System (ADS)

Um, Dallsun

For the radiation shielding quality assurance, the validity and reliability of the neutron transport code MCNP, which is now one of the most widely used radiation shielding analysis codes, were checked with lot of benchmark experiments. And also as a practical example, follows were performed in this thesis. One integral neutron transport experiment to measure the effect of neutron streaming in iron and void was performed with Dog-Legged Void Assembly in Knolls Atomic Power Laboratory in 1991. Neutron flux was measured six different places with the methane detectors and a BF-3 detector. The main purpose of the measurements was to provide benchmark against which various neutron transport calculation tools could be compared. Those data were used in verification of Monte Carlo Neutron & Photon Transport Code, MCNP, with the modeling for that. Experimental results and calculation results were compared in both ways, as the total integrated value of neutron fluxes along neutron energy range from 10 KeV to 2 MeV and as the neutron spectrum along with neutron energy range. Both results are well matched with the statistical error +/-20%. MCNP results were also compared with those of TORT, a three dimensional discrete ordinates code which was developed by Oak Ridge National Laboratory. MCNP results are superior to the TORT results at all detector places except one. This means that MCNP is proved as a very powerful tool for the analysis of neutron transport through iron & air and further it could be used as a powerful tool for the radiation shielding analysis. For one application of the analysis of variance (ANOVA) to neutron and gamma transport problems, uncertainties for the calculated values of critical K were evaluated as in the ANOVA on statistical data.

Using NJOY to Create MCNP ACE Files and Visualize Nuclear Data

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

Kahler, Albert Comstock

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

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

PubMed

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

2003-03-01

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

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

PubMed

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)

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

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

Burns, T.D. Jr.

1996-05-01

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

Treating voxel geometries in radiation protection dosimetry with a patched version of the Monte Carlo codes MCNP and MCNPX.

PubMed

Burn, K W; Daffara, C; Gualdrini, G; Pierantoni, M; Ferrari, P

2007-01-01

The question of Monte Carlo simulation of radiation transport in voxel geometries is addressed. Patched versions of the MCNP and MCNPX codes are developed aimed at transporting radiation both in the standard geometry mode and in the voxel geometry treatment. The patched code reads an unformatted FORTRAN file derived from DICOM format data and uses special subroutines to handle voxel-to-voxel radiation transport. The various phases of the development of the methodology are discussed together with the new input options. Examples are given of employment of the code in internal and external dosimetry and comparisons with results from other groups are reported.

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

PubMed

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

2005-01-01

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

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

PubMed

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

2001-03-01

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

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

NASA Astrophysics Data System (ADS)

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

2017-09-01

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

Lecture Notes on Criticality Safety Validation Using MCNP & Whisper

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

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

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

Total reaction cross sections in CEM and MCNP6 at intermediate energies

DOE PAGES

Kerby, Leslie M.; Mashnik, Stepan G.

2015-05-14

Accurate total reaction cross section models are important to achieving reliable predictions from spallation and transport codes. The latest version of the Cascade Exciton Model (CEM) as incorporated in the code CEM03.03, and the Monte Carlo N-Particle transport code (MCNP6), both developed at Los Alamos National Laboratory (LANL), each use such cross sections. Having accurate total reaction cross section models in the intermediate energy region (50 MeV to 5 GeV) is very important for different applications, including analysis of space environments, use in medical physics, and accelerator design, to name just a few. The current inverse cross sections used inmore » the preequilibrium and evaporation stages of CEM are based on the Dostrovsky et al. model, published in 1959. Better cross section models are now available. Implementing better cross section models in CEM and MCNP6 should yield improved predictions for particle spectra and total production cross sections, among other results.« less

Total reaction cross sections in CEM and MCNP6 at intermediate energies

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

Kerby, Leslie M.; Mashnik, Stepan G.

Accurate total reaction cross section models are important to achieving reliable predictions from spallation and transport codes. The latest version of the Cascade Exciton Model (CEM) as incorporated in the code CEM03.03, and the Monte Carlo N-Particle transport code (MCNP6), both developed at Los Alamos National Laboratory (LANL), each use such cross sections. Having accurate total reaction cross section models in the intermediate energy region (50 MeV to 5 GeV) is very important for different applications, including analysis of space environments, use in medical physics, and accelerator design, to name just a few. The current inverse cross sections used inmore » the preequilibrium and evaporation stages of CEM are based on the Dostrovsky et al. model, published in 1959. Better cross section models are now available. Implementing better cross section models in CEM and MCNP6 should yield improved predictions for particle spectra and total production cross sections, among other results.« less

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

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

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

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

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

PubMed

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

2006-11-01

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

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

NASA Astrophysics Data System (ADS)

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

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

Performance Study of Monte Carlo Codes on Xeon Phi Coprocessors — Testing MCNP 6.1 and Profiling ARCHER Geometry Module on the FS7ONNi Problem

NASA Astrophysics Data System (ADS)

Liu, Tianyu; Wolfe, Noah; Lin, Hui; Zieb, Kris; Ji, Wei; Caracappa, Peter; Carothers, Christopher; Xu, X. George

2017-09-01

This paper contains two parts revolving around Monte Carlo transport simulation on Intel Many Integrated Core coprocessors (MIC, also known as Xeon Phi). (1) MCNP 6.1 was recompiled into multithreading (OpenMP) and multiprocessing (MPI) forms respectively without modification to the source code. The new codes were tested on a 60-core 5110P MIC. The test case was FS7ONNi, a radiation shielding problem used in MCNP's verification and validation suite. It was observed that both codes became slower on the MIC than on a 6-core X5650 CPU, by a factor of 4 for the MPI code and, abnormally, 20 for the OpenMP code, and both exhibited limited capability of strong scaling. (2) We have recently added a Constructive Solid Geometry (CSG) module to our ARCHER code to provide better support for geometry modelling in radiation shielding simulation. The functions of this module are frequently called in the particle random walk process. To identify the performance bottleneck we developed a CSG proxy application and profiled the code using the geometry data from FS7ONNi. The profiling data showed that the code was primarily memory latency bound on the MIC. This study suggests that despite low initial porting e_ort, Monte Carlo codes do not naturally lend themselves to the MIC platform — just like to the GPUs, and that the memory latency problem needs to be addressed in order to achieve decent performance gain.

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

PubMed

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

2006-01-01

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

AN ASSESSMENT OF MCNP WEIGHT WINDOWS

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

J. S. HENDRICKS; C. N. CULBERTSON

2000-01-01

The weight window variance reduction method in the general-purpose Monte Carlo N-Particle radiation transport code MCNPTM has recently been rewritten. In particular, it is now possible to generate weight window importance functions on a superimposed mesh, eliminating the need to subdivide geometries for variance reduction purposes. Our assessment addresses the following questions: (1) Does the new MCNP4C treatment utilize weight windows as well as the former MCNP4B treatment? (2) Does the new MCNP4C weight window generator generate importance functions as well as MCNP4B? (3) How do superimposed mesh weight windows compare to cell-based weight windows? (4) What are the shortcomingsmore » of the new MCNP4C weight window generator? Our assessment was carried out with five neutron and photon shielding problems chosen for their demanding variance reduction requirements. The problems were an oil well logging problem, the Oak Ridge fusion shielding benchmark problem, a photon skyshine problem, an air-over-ground problem, and a sample problem for variance reduction.« less

Brachytherapy dosimetry of 125I and 103Pd sources using an updated cross section library for the MCNP Monte Carlo transport code.

PubMed

Bohm, Tim D; DeLuca, Paul M; DeWerd, Larry A

2003-04-01

Permanent implantation of low energy (20-40 keV) photon emitting radioactive seeds to treat prostate cancer is an important treatment option for patients. In order to produce accurate implant brachytherapy treatment plans, the dosimetry of a single source must be well characterized. Monte Carlo based transport calculations can be used for source characterization, but must have up to date cross section libraries to produce accurate dosimetry results. This work benchmarks the MCNP code and its photon cross section library for low energy photon brachytherapy applications. In particular, we calculate the emitted photon spectrum, air kerma, depth dose in water, and radial dose function for both 125I and 103Pd based seeds and compare to other published results. Our results show that MCNP's cross section library differs from recent data primarily in the photoelectric cross section for low energies and low atomic number materials. In water, differences as large as 10% in the photoelectric cross section and 6% in the total cross section occur at 125I and 103Pd photon energies. This leads to differences in the dose rate constant of 3% and 5%, and differences as large as 18% and 20% in the radial dose function for the 125I and 103Pd based seeds, respectively. Using a partially updated photon library, calculations of the dose rate constant and radial dose function agree with other published results. Further, the use of the updated photon library allows us to verify air kerma and depth dose in water calculations performed using MCNP's perturbation feature to simulate updated cross sections. We conclude that in order to most effectively use MCNP for low energy photon brachytherapy applications, we must update its cross section library. Following this update, the MCNP code system will be a very effective tool for low energy photon brachytherapy dosimetry applications.

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

NASA Astrophysics Data System (ADS)

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

2017-09-01

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

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

NASA Astrophysics Data System (ADS)

Karriem, Veronica V.

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

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

PubMed

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

2018-05-04

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

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

PubMed

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.

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

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

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

2002-09-11

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

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

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

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

2005-05-24

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

LIBMAKER

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

2015-08-01

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

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

PubMed

Krstic, D; Nikezic, D

2010-01-01

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

Accelerating Pseudo-Random Number Generator for MCNP on GPU

NASA Astrophysics Data System (ADS)

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

2010-09-01

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

An Electron/Photon/Relaxation Data Library for MCNP6

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

Hughes, III, H. Grady

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

Benchmarking study of the MCNP code against cold critical experiments

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

Sitaraman, S.

1991-01-01

The purpose of this study was to benchmark the widely used Monte Carlo code MCNP against a set of cold critical experiments with a view to using the code as a means of independently verifying the performance of faster but less accurate Monte Carlo and deterministic codes. The experiments simulated consisted of both fast and thermal criticals as well as fuel in a variety of chemical forms. A standard set of benchmark cold critical experiments was modeled. These included the two fast experiments, GODIVA and JEZEBEL, the TRX metallic uranium thermal experiments, the Babcock and Wilcox oxide and mixed oxidemore » experiments, and the Oak Ridge National Laboratory (ORNL) and Pacific Northwest Laboratory (PNL) nitrate solution experiments. The principal case studied was a small critical experiment that was performed with boiling water reactor bundles.« less

Performance upgrades to the MCNP6 burnup capability for large scale depletion calculations

DOE PAGES

Fensin, M. L.; Galloway, J. D.; James, M. R.

2015-04-11

The first MCNP based inline Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. With the merger of MCNPX and MCNP5, MCNP6 combined the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. The new MCNP6 depletion capability was first showcased at the International Congress for Advancements in Nuclear Power Plants (ICAPP) meeting in 2012. At that conference the new capabilities addressed included the combined distributive and shared memory parallel architecture for the burnup capability, improved memory management, physics enhancements, and newmore » predictability as compared to the H.B Robinson Benchmark. At Los Alamos National Laboratory, a special purpose cluster named “tebow,” was constructed such to maximize available RAM per CPU, as well as leveraging swap space with solid state hard drives, to allow larger scale depletion calculations (allowing for significantly more burnable regions than previously examined). As the MCNP6 burnup capability was scaled to larger numbers of burnable regions, a noticeable slowdown was realized.This paper details two specific computational performance strategies for improving calculation speedup: (1) retrieving cross sections during transport; and (2) tallying mechanisms specific to burnup in MCNP. To combat this slowdown new performance upgrades were developed and integrated into MCNP6 1.2.« less

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

NASA Astrophysics Data System (ADS)

Jaradat, Safwan Qasim Mohammad

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

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

PubMed

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

2004-01-01

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

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

PubMed

Tran, Van Hung; Tran, Khac An

2010-06-01

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

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

NASA Astrophysics Data System (ADS)

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

2003-06-01

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

Enhancements to the MCNP6 background source

DOE PAGES

McMath, Garrett E.; McKinney, Gregg W.

2015-10-19

The particle transport code MCNP has been used to produce a background radiation data file on a worldwide grid that can easily be sampled as a source in the code. Location-dependent cosmic showers were modeled by Monte Carlo methods to produce the resulting neutron and photon background flux at 2054 locations around Earth. An improved galactic-cosmic-ray feature was used to model the source term as well as data from multiple sources to model the transport environment through atmosphere, soil, and seawater. A new elevation scaling feature was also added to the code to increase the accuracy of the cosmic neutronmore » background for user locations with off-grid elevations. Furthermore, benchmarking has shown the neutron integral flux values to be within experimental error.« less

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

PubMed

Duggan, Dennis M

2004-12-01

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

LANL Summer 2016 Report

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

Mendoza, Paul Michael

The Monte Carlo N-Particle (MCNP) transport code developed at Los Alamos National Laboratory (LANL) utilizes nuclear cross-section data in a compact ENDF (ACE) format. The accuracy of MCNP calculations depends on the accuracy of nuclear ACE data tables, which depends on the accuracy of the original ENDF files. There are some noticeable differences in ENDF files from one generation to the next, even among the more common fissile materials. As the next generation of ENDF files is being prepared, several software tools were developed to simulate a large number of benchmarks in MCNP (over 1000), collect data from these simulations,more » and visually represent the results.« less

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

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard; Sjenitzer, Bart L.

2014-06-01

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

Sensitivity-Uncertainty Based Nuclear Criticality Safety Validation

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

Brown, Forrest B.

2016-09-20

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

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

PubMed

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

2003-10-01

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

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

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

West, J.T. III

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

Characterization of Filters Loaded With Reactor Strontium Carbonate - 13203

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

Josephson, Walter S.; Steen, Franciska H.

A collection of three highly radioactive filters containing reactor strontium carbonate were being prepared for disposal. All three filters were approximately characterized at the time of manufacture by gravimetric methods. The first filter had been partially emptied, and the quantity of residual activity was uncertain. Dose rate to activity modeling using the Monte-Carlo N Particle (MCNP) code was selected to confirm the gravimetric characterization of the full filters, and to fully characterize the partially emptied filter. Although dose rate to activity modeling using MCNP is a common technique, it is not often used for Bremsstrahlung-dominant materials such as reactor strontium.more » As a result, different MCNP modeling options were compared to determine the optimum approach. This comparison indicated that the accuracy of the results were heavily dependent on the MCNP modeling details and the location of the dose rate measurement point. The optimum model utilized a photon spectrum generated by the Oak Ridge Isotope Generation and Depletion (ORIGEN) code and dose rates measured at 30 cm. Results from the optimum model agreed with the gravimetric estimates within 15%. It was demonstrated that dose rate to activity modeling can be successful for Bremsstrahlung-dominant radioactive materials. However, the degree of success is heavily dependent on the choice of modeling techniques. (authors)« less

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

NASA Astrophysics Data System (ADS)

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

2015-06-01

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

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

NASA Astrophysics Data System (ADS)

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

2018-06-01

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

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

PubMed

Colonna, N; Altieri, S

2002-06-01

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

Comparison of scientific computing platforms for MCNP4A Monte Carlo calculations

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

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

1994-04-01

The performance of seven computer platforms is evaluated with the widely used and internationally available MCNP4A Monte Carlo radiation transport code. All results are reproducible and are presented in such a way as to enable comparison with computer platforms not in the study. The authors observed that the HP/9000-735 workstation runs MCNP 50% faster than the Cray YMP 8/64. Compared with the Cray YMP 8/64, the IBM RS/6000-560 is 68% as fast, the Sun Sparc10 is 66% as fast, the Silicon Graphics ONYX is 90% as fast, the Gateway 2000 model 4DX2-66V personal computer is 27% as fast, and themore » Sun Sparc2 is 24% as fast. In addition to comparing the timing performance of the seven platforms, the authors observe that changes in compilers and software over the past 2 yr have resulted in only modest performance improvements, hardware improvements have enhanced performance by less than a factor of [approximately]3, timing studies are very problem dependent, MCNP4Q runs about as fast as MCNP4.« less

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

NASA Astrophysics Data System (ADS)

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

2014-06-01

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

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

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

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

Voxel2MCNP: a framework for modeling, simulation and evaluation of radiation transport scenarios for Monte Carlo codes.

PubMed

Pölz, Stefan; Laubersheimer, Sven; Eberhardt, Jakob S; Harrendorf, Marco A; Keck, Thomas; Benzler, Andreas; Breustedt, Bastian

2013-08-21

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.

MCNP/X TRANSPORT IN THE TABULAR REGIME

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

HUGHES, H. GRADY

2007-01-08

The authors review the transport capabilities of the MCNP and MCNPX Monte Carlo codes in the energy regimes in which tabular transport data are available. Giving special attention to neutron tables, they emphasize the measures taken to improve the treatment of a variety of difficult aspects of the transport problem, including unresolved resonances, thermal issues, and the availability of suitable cross sections sets. They also briefly touch on the current situation in regard to photon, electron, and proton transport tables.

MCNP modelling of scintillation-detector gamma-ray spectra from natural radionuclides.

PubMed

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.

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

PubMed Central

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

2013-01-01

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

Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

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

White, Morgan C.

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 selectmore » group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V&V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second, the ability to calculate radiation dose due to the neutron environment around a MEA is shown. An uncertainty of a factor of three in the MEA calculations is shown to be due to uncertainties in the geometry modeling. It is believed that the methodology is sound and that good agreement between simulation and experiment has been demonstrated.« less

EBR-II Static Neutronic Calculations by PHISICS / MCNP6 codes

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

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).more » Comparison between both solutions is briefly illustrated in this summary.« less

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

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

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

2016-07-07

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

Monte Carlo Techniques for Nuclear Systems - Theory Lectures

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

Brown, Forrest B.

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

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

DOE PAGES

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

2016-10-01

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

Using Machine Learning to Predict MCNP Bias

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

Grechanuk, Pavel Aleksandrovi

For many real-world applications in radiation transport where simulations are compared to experimental measurements, like in nuclear criticality safety, the bias (simulated - experimental k eff) in the calculation is an extremely important quantity used for code validation. The objective of this project is to accurately predict the bias of MCNP6 [1] criticality calculations using machine learning (ML) algorithms, with the intention of creating a tool that can complement the current nuclear criticality safety methods. In the latest release of MCNP6, the Whisper tool is available for criticality safety analysts and includes a large catalogue of experimental benchmarks, sensitivity profiles,more » and nuclear data covariance matrices. This data, coming from 1100+ benchmark cases, is used in this study of ML algorithms for criticality safety bias predictions.« less

Benchmark study for charge deposition by high energy electrons in thick slabs

NASA Technical Reports Server (NTRS)

Jun, I.

2002-01-01

The charge deposition profiles created when highenergy (1, 10, and 100 MeV) electrons impinge ona thick slab of elemental aluminum, copper, andtungsten are presented in this paper. The chargedeposition profiles were computed using existing representative Monte Carlo codes: TIGER3.0 (1D module of ITS3.0) and MCNP version 4B. The results showed that TIGER3.0 and MCNP4B agree very well (within 20% of each other) in the majority of the problem geometry. The TIGER results were considered to be accurate based on previous studies. Thus, it was demonstrated that MCNP, with its powerful geometry capability and flexible source and tally options, could be used in calculations of electron charging in high energy electron-rich space radiation environments.

Monte Carlo calculation of dose rate conversion factors for external exposure to photon emitters in soil

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

Clovas, A.; Zanthos, S.; Antonopoulos-Domis, M.

2000-03-01

The dose rate conversion factors {dot D}{sub CF} (absorbed dose rate in air per unit activity per unit of soil mass, nGy h{sup {minus}1} per Bq kg{sup {minus}1}) are calculated 1 m above ground for photon emitters of natural radionuclides uniformly distributed in the soil. Three Monte Carlo codes are used: (1) The MCNP code of Los Alamos; (2) The GEANT code of CERN; and (3) a Monte Carlo code developed in the Nuclear Technology Laboratory of the Aristotle University of Thessaloniki. The accuracy of the Monte Carlo results is tested by the comparison of the unscattered flux obtained bymore » the three Monte Carlo codes with an independent straightforward calculation. All codes and particularly the MCNP calculate accurately the absorbed dose rate in air due to the unscattered radiation. For the total radiation (unscattered plus scattered) the {dot D}{sub CF} values calculated from the three codes are in very good agreement between them. The comparison between these results and the results deduced previously by other authors indicates a good agreement (less than 15% of difference) for photon energies above 1,500 keV. Antithetically, the agreement is not as good (difference of 20--30%) for the low energy photons.« less

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

NASA Astrophysics Data System (ADS)

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

2017-12-01

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

Neutron flux and power in RTP core-15

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

Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Zin, Muhammad Rawi Md; Usang, Mark Dennis

PUSPATI TRIGA Reactor 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 of TRIGA core. 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 withmore » literally no physical approximation. The consistency and accuracy of the developed RTP MCNP model was established by comparing calculations to the available experimental results and TRIGLAV code calculation.« less

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

NASA Astrophysics Data System (ADS)

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

2018-06-01

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

Simulation the spatial resolution of an X-ray imager based on zinc oxide nanowires in anodic aluminium oxide membrane by using MCNP and OPTICS Codes

NASA Astrophysics Data System (ADS)

Samarin, S. N.; Saramad, S.

2018-05-01

The spatial resolution of a detector is a very important parameter for x-ray imaging. A bulk scintillation detector because of spreading of light inside the scintillator does't have a good spatial resolution. The nanowire scintillators because of their wave guiding behavior can prevent the spreading of light and can improve the spatial resolution of traditional scintillation detectors. The zinc oxide (ZnO) scintillator nanowire, with its simple construction by electrochemical deposition in regular hexagonal structure of Aluminum oxide membrane has many advantages. The three dimensional absorption of X-ray energy in ZnO scintillator is simulated by a Monte Carlo transport code (MCNP). The transport, attenuation and scattering of the generated photons are simulated by a general-purpose scintillator light response simulation code (OPTICS). The results are compared with a previous publication which used a simulation code of the passage of particles through matter (Geant4). The results verify that this scintillator nanowire structure has a spatial resolution less than one micrometer.

MCNP and GADRAS Comparisons

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

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

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

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

PubMed

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

2004-01-01

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

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

NASA Astrophysics Data System (ADS)

Castanier, Eric; Paterne, Loic; Louis, Céline

2017-09-01

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

A method to optimize the shield compact and lightweight combining the structure with components together by genetic algorithm and MCNP code.

PubMed

Cai, Yao; Hu, Huasi; Pan, Ziheng; Hu, Guang; Zhang, Tao

2018-05-17

To optimize the shield for neutrons and gamma rays compact and lightweight, a method combining the structure and components together was established employing genetic algorithms and MCNP code. As a typical case, the fission energy spectrum of 235 U which mixed neutrons and gamma rays was adopted in this study. Six types of materials were presented and optimized by the method. Spherical geometry was adopted in the optimization after checking the geometry effect. Simulations have made to verify the reliability of the optimization method and the efficiency of the optimized materials. To compare the materials visually and conveniently, the volume and weight needed to build a shield are employed. The results showed that, the composite multilayer material has the best performance. Copyright © 2018 Elsevier Ltd. All rights reserved.

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

PubMed

Khattab, K; Sulieman, I

2009-04-01

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

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

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

Kim, Seung Jun; Buechler, Cynthia Eileen

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

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

NASA Astrophysics Data System (ADS)

Burns, Kimberly Ann

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

Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS

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

Connaway, H. M.; Lee, C. H.

The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy’s resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To supportmore » this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SN is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.« less

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

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

Brown, Forrest B.

Monte Carlo calculations using one-group cross sections, multigroup cross sections, or simple continuous energy cross sections are often used to: (1) verify production codes against known analytical solutions, (2) verify new methods and algorithms that do not involve detailed collision physics, (3) compare Monte Carlo calculation methods with deterministic methods, and (4) teach fundamentals to students. In this work we describe 2 new tools for preparing the ACE cross-section files to be used by MCNP ® for these analytic test problems, simple_ace.pl and simple_ace_mg.pl.

Validation of MCNP6 Version 1.0 with the ENDF/B-VII.1 Cross Section Library for Plutonium Metals, Oxides, and Solutions on the High Performance Computing Platform Moonlight

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

Chapman, Bryan Scott; Gough, Sean T.

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.

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

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

Crawford, C.; Palmer, B.

1992-11-01

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

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

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

Crawford, C.; Palmer, B.

1992-11-01

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

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

NASA Astrophysics Data System (ADS)

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

2017-09-01

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

Benchmark study for total enery electrons in thick slabs

NASA Technical Reports Server (NTRS)

Jun, I.

2002-01-01

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

Image enhancement using MCNP5 code and MATLAB in neutron radiography.

PubMed

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.

Monte Carlo dose calculations of beta-emitting sources for intravascular brachytherapy: a comparison between EGS4, EGSnrc, and MCNP.

PubMed

Wang, R; Li, X A

2001-02-01

The dose parameters for the beta-particle emitting 90Sr/90Y source for intravascular brachytherapy (IVBT) have been calculated by different investigators. At a distant distance from the source, noticeable differences are seen in these parameters calculated using different Monte Carlo codes. The purpose of this work is to quantify as well as to understand these differences. We have compared a series of calculations using an EGS4, an EGSnrc, and the MCNP Monte Carlo codes. Data calculated and compared include the depth dose curve for a broad parallel beam of electrons, and radial dose distributions for point electron sources (monoenergetic or polyenergetic) and for a real 90Sr/90Y source. For the 90Sr/90Y source, the doses at the reference position (2 mm radial distance) calculated by the three code agree within 2%. However, the differences between the dose calculated by the three codes can be over 20% in the radial distance range interested in IVBT. The difference increases with radial distance from source, and reaches 30% at the tail of dose curve. These differences may be partially attributed to the different multiple scattering theories and Monte Carlo models for electron transport adopted in these three codes. Doses calculated by the EGSnrc code are more accurate than those by the EGS4. The two calculations agree within 5% for radial distance <6 mm.

Evaluation and Testing of the ADVANTG Code on SNM Detection

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

Shaver, Mark W.; Casella, Andrew M.; Wittman, Richard S.

2013-09-24

Pacific Northwest National Laboratory (PNNL) has been tasked with evaluating the effectiveness of ORNL’s new hybrid transport code, ADVANTG, on scenarios of interest to our NA-22 sponsor, specifically of detection of diversion of special nuclear material (SNM). PNNL staff have determined that acquisition and installation of ADVANTG was relatively straightforward for a code in its phase of development, but probably not yet sufficient for mass distribution to the general user. PNNL staff also determined that with little effort, ADVANTG generated weight windows that typically worked for the problems and generated results consistent with MCNP. With slightly greater effort of choosingmore » a finer mesh around detectors or sample reaction tally regions, the figure of merit (FOM) could be further improved in most cases. This does take some limited knowledge of deterministic transport methods. The FOM could also be increased by limiting the energy range for a tally to the energy region of greatest interest. It was then found that an MCNP run with the full energy range for the tally showed improved statistics in the region used for the ADVANTG run. The specific case of interest chosen by the sponsor is the CIPN project from Las Alamos National Laboratory (LANL), which is an active interrogation, non-destructive assay (NDA) technique to quantify the fissile content in a spent fuel assembly and is also sensitive to cases of material diversion. Unfortunately, weight windows for the CIPN problem cannot currently be properly generated with ADVANTG due to inadequate accommodations for source definition. ADVANTG requires that a fixed neutron source be defined within the problem and cannot account for neutron multiplication. As such, it is rendered useless in active interrogation scenarios. It is also interesting to note that this is a difficult problem to solve and that the automated weight windows generator in MCNP actually slowed down the problem. Therefore, PNNL had determined that there is not an effective tool available for speeding up MCNP for problems such as the CIPN scenario. With regard to the Benchmark scenarios, ADVANTG performed very well for most of the difficult, long-running, standard radiation detection scenarios. Specifically, run time speedups were observed for spatially large scenarios, or those having significant shielding or scattering geometries. ADVANTG performed on par with existing codes for moderate sized scenarios, or those with little to moderate shielding, or multiple paths to the detectors. ADVANTG ran slower than MCNP for very simply, spatially small cases with little to no shielding that run very quickly anyway. Lastly, ADVANTG could not solve problems that did not consist of fixed source to detector geometries. For example, it could not solve scenarios with multiple detectors or secondary particles, such as active interrogation, neutron induced gamma, or fission neutrons.« less

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

NASA Astrophysics Data System (ADS)

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

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

Efficiency of whole-body counter for various body size calculated by MCNP5 software.

PubMed

Krstic, D; Nikezic, D

2012-11-01

The efficiency of a whole-body counter for (137)Cs and (40)K was calculated using the MCNP5 code. The ORNL phantoms of a human body of different body sizes were applied in a sitting position in front of a detector. The aim was to investigate the dependence of efficiency on the body size (age) and the detector position with respect to the body and to estimate the accuracy of real measurements. The calculation work presented here is related to the NaI detector, which is available in the Serbian Whole-body Counter facility in Vinca Institute.

Validation of absolute axial neutron flux distribution calculations with MCNP with 197Au(n,γ)198Au reaction rate distribution measurements at the JSI TRIGA Mark II reactor.

PubMed

Radulović, Vladimir; Štancar, Žiga; Snoj, Luka; Trkov, Andrej

2014-02-01

The calculation of axial neutron flux distributions with the MCNP code at the JSI TRIGA Mark II reactor has been validated with experimental measurements of the (197)Au(n,γ)(198)Au reaction rate. The calculated absolute reaction rate values, scaled according to the reactor power and corrected for the flux redistribution effect, are in good agreement with the experimental results. The effect of different cross-section libraries on the calculations has been investigated and shown to be minor. Copyright © 2013 Elsevier Ltd. All rights reserved.

MCNP calculations for container inspection with tagged neutrons

NASA Astrophysics Data System (ADS)

Boghen, G.; Donzella, A.; Filippini, V.; Fontana, A.; Lunardon, M.; Moretto, S.; Pesente, S.; Zenoni, A.

2005-12-01

We are developing an innovative tagged neutrons inspection system (TNIS) for cargo containers: the system will allow us to assay the chemical composition of suspect objects, previously identified by a standard X-ray radiography. The operation of the system is extensively being simulated by using the MCNP Monte Carlo code to study different inspection geometries, cargo loads and hidden threat materials. Preliminary simulations evaluating the signal and the signal over background ratio expected as a function of the system parameters are presented. The results for a selection of cases are briefly discussed and demonstrate that the system can operate successfully in different filling conditions.

Skyshine line-beam response functions for 20- to 100-MeV photons

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

Brockhoff, R.C.; Shultis, J.K.; Faw, R.E.

1996-06-01

The line-beam response function, needed for skyshine analyses based on the integral line-beam method, was evaluated with the MCNP Monte Carlo code for photon energies from 20 to 100 MeV and for source-to-detector distances out to 1,000 m. These results are compared with point-kernel results, and the effects of bremsstrahlung and positron transport in the air are found to be important in this energy range. The three-parameter empirical formula used in the integral line-beam skyshine method was fit to the MCNP results, and values of these parameters are reported for various source energies and angles.

Addressing Fission Product Validation in MCNP Burnup Credit Criticality Calculations

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

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

2015-01-01

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

Physical models, cross sections, and numerical approximations used in MCNP and GEANT4 Monte Carlo codes for photon and electron absorbed fraction calculation.

PubMed

Yoriyaz, Hélio; Moralles, Maurício; Siqueira, Paulo de Tarso Dalledone; Guimarães, Carla da Costa; Cintra, Felipe Belonsi; dos Santos, Adimir

2009-11-01

Radiopharmaceutical applications in nuclear medicine require a detailed dosimetry estimate of the radiation energy delivered to the human tissues. Over the past years, several publications addressed the problem of internal dose estimate in volumes of several sizes considering photon and electron sources. Most of them used Monte Carlo radiation transport codes. Despite the widespread use of these codes due to the variety of resources and potentials they offered to carry out dose calculations, several aspects like physical models, cross sections, and numerical approximations used in the simulations still remain an object of study. Accurate dose estimate depends on the correct selection of a set of simulation options that should be carefully chosen. This article presents an analysis of several simulation options provided by two of the most used codes worldwide: MCNP and GEANT4. For this purpose, comparisons of absorbed fraction estimates obtained with different physical models, cross sections, and numerical approximations are presented for spheres of several sizes and composed as five different biological tissues. Considerable discrepancies have been found in some cases not only between the different codes but also between different cross sections and algorithms in the same code. Maximum differences found between the two codes are 5.0% and 10%, respectively, for photons and electrons. Even for simple problems as spheres and uniform radiation sources, the set of parameters chosen by any Monte Carlo code significantly affects the final results of a simulation, demonstrating the importance of the correct choice of parameters in the simulation.

MCNP5 CALCULATIONS REPLICATING ARH-600 NITRATE DATA

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

FINFROCK SH

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

Intrinsic Radiation Source Generation with the ISC Package: Data Comparisons and Benchmarking

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

Solomon, Clell J. Jr.

The characterization of radioactive emissions from unstable isotopes (intrinsic radiation) is necessary for shielding and radiological-dose calculations from radioactive materials. While most radiation transport codes, e.g., MCNP [X-5 Monte Carlo Team, 2003], provide the capability to input user prescribed source definitions, such as radioactive emissions, they do not provide the capability to calculate the correct radioactive-source definition given the material compositions. Special modifications to MCNP have been developed in the past to allow the user to specify an intrinsic source, but these modification have not been implemented into the primary source base [Estes et al., 1988]. To facilitate the descriptionmore » of the intrinsic radiation source from a material with a specific composition, the Intrinsic Source Constructor library (LIBISC) and MCNP Intrinsic Source Constructor (MISC) utility have been written. The combination of LIBISC and MISC will be herein referred to as the ISC package. LIBISC is a statically linkable C++ library that provides the necessary functionality to construct the intrinsic-radiation source generated by a material. Furthermore, LIBISC provides the ability use different particle-emission databases, radioactive-decay databases, and natural-abundance databases allowing the user flexibility in the specification of the source, if one database is preferred over others. LIBISC also provides functionality for aging materials and producing a thick-target bremsstrahlung photon source approximation from the electron emissions. The MISC utility links to LIBISC and facilitates the description of intrinsic-radiation sources into a format directly usable with the MCNP transport code. Through a series of input keywords and arguments the MISC user can specify the material, age the material if desired, and produce a source description of the radioactive emissions from the material in an MCNP readable format. Further details of using the MISC utility can be obtained from the user guide [Solomon, 2012]. The remainder of this report presents a discussion of the databases available to LIBISC and MISC, a discussion of the models employed by LIBISC, a comparison of the thick-target bremsstrahlung model employed, a benchmark comparison to plutonium and depleted-uranium spheres, and a comparison of the available particle-emission databases.« less

Comparison of Three Methods of Calculation, Experimental and Monte Carlo Simulation in Investigation of Organ Doses (Thyroid, Sternum, Cervical Vertebra) in Radioiodine Therapy

PubMed Central

Shahbazi-Gahrouei, Daryoush; Ayat, Saba

2012-01-01

Radioiodine therapy is an effective method for treating thyroid cancer carcinoma, but it has some affects on normal tissues, hence dosimetry of vital organs is important to weigh the risks and benefits of this method. The aim of this study is to measure the absorbed doses of important organs by Monte Carlo N Particle (MCNP) simulation and comparing the results of different methods of dosimetry by performing a t-paired test. To calculate the absorbed dose of thyroid, sternum, and cervical vertebra using the MCNP code, *F8 tally was used. Organs were simulated by using a neck phantom and Medical Internal Radiation Dosimetry (MIRD) method. Finally, the results of MCNP, MIRD, and Thermoluminescent dosimeter (TLD) measurements were compared by SPSS software. The absorbed dose obtained by Monte Carlo simulations for 100, 150, and 175 mCi administered 131I was found to be 388.0, 427.9, and 444.8 cGy for thyroid, 208.7, 230.1, and 239.3 cGy for sternum and 272.1, 299.9, and 312.1 cGy for cervical vertebra. The results of paired t-test were 0.24 for comparing TLD dosimetry and MIRD calculation, 0.80 for MCNP simulation and MIRD, and 0.19 for TLD and MCNP. The results showed no significant differences among three methods of Monte Carlo simulations, MIRD calculation and direct experimental dosimetry using TLD. PMID:23717806

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

PubMed

Heuel-Fabianek, Burkhard; Hille, Ralf

2005-01-01

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

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

PubMed

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

2012-01-01

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

Fission products detection in irradiated TRIGA fuel by means of gamma spectroscopy and MCNP calculation.

PubMed

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.

a Proposed Benchmark Problem for Scatter Calculations in Radiographic Modelling

NASA Astrophysics Data System (ADS)

Jaenisch, G.-R.; Bellon, C.; Schumm, A.; Tabary, J.; Duvauchelle, Ph.

2009-03-01

Code Validation is a permanent concern in computer modelling, and has been addressed repeatedly in eddy current and ultrasonic modeling. A good benchmark problem is sufficiently simple to be taken into account by various codes without strong requirements on geometry representation capabilities, focuses on few or even a single aspect of the problem at hand to facilitate interpretation and to avoid that compound errors compensate themselves, yields a quantitative result and is experimentally accessible. In this paper we attempt to address code validation for one aspect of radiographic modeling, the scattered radiation prediction. Many NDT applications can not neglect scattered radiation, and the scatter calculation thus is important to faithfully simulate the inspection situation. Our benchmark problem covers the wall thickness range of 10 to 50 mm for single wall inspections, with energies ranging from 100 to 500 keV in the first stage, and up to 1 MeV with wall thicknesses up to 70 mm in the extended stage. A simple plate geometry is sufficient for this purpose, and the scatter data is compared on a photon level, without a film model, which allows for comparisons with reference codes like MCNP. We compare results of three Monte Carlo codes (McRay, Sindbad and Moderato) as well as an analytical first order scattering code (VXI), and confront them to results obtained with MCNP. The comparison with an analytical scatter model provides insights into the application domain where this kind of approach can successfully replace Monte-Carlo calculations.

SMITHERS: An object-oriented modular mapping methodology for MCNP-based neutronic–thermal hydraulic multiphysics

DOE PAGES

Richard, Joshua; Galloway, Jack; Fensin, Michael; ...

2015-04-04

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. In addition, it performs the basis mapping from themore » 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.« less

MCNP modelling of the wall effects observed in tissue-equivalent proportional counters.

PubMed

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.

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

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

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

2011-07-01

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

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

PubMed Central

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

2010-01-01

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

Radioactive ion beams produced by neutron-induced fission at ISOLDE

NASA Astrophysics Data System (ADS)

Catherall, R.; Lettry, J.; Gilardoni, S.; Köster, U.; Isolde Collaboration

2003-05-01

The production rates of neutron-rich fission products for the next-generation radioactive beam facility EURISOL [EU-RTD Project EURISOL (HPRI-CT-1999-50001)] are mainly limited by the maximum amount of power deposited by protons in the target. An alternative approach is to use neutron beams to induce fission in actinide targets. This has the advantage of reducing: the energy deposited by the proton beam in the target; contamination from neutron-deficient isobars that would be produced by spallation; and mechanical stress on the target. At ISOLDE CERN [E. Kugler, Hyperfine Interact. 129 (2000) 23], tests have been made on standard ISOLDE actinide targets using fast-neutron bunches produced by bombarding thick, high- Z metal converters with 1 and 1.4 GeV proton pulses. This paper reviews the first applications of converters used at ISOLDE. It highlights the different geometries and the techniques used to compare fission yields produced by the proton beam directly on the target with neutron-induced fission. Results from the six targets already tested, namely UC 2/graphite and ThO 2 targets with tungsten and tantalum converters, are presented. To gain further knowledge for the design of a dedicated target as required by the TARGISOL project [EU-RTD Project TARGISOL (HPRI-CT-2001-50033)], the results are compared to simulations, using the MARS [N.V. Mokhov, S.I. Striganov, A. Van Ginneken, S.G. Mashnik, A.J. Sierk, J. Ranft, MARS code developments, in: 4th Workshop on Simulating Accelerator Radiation Environments, SARE-4, Knoxville, USA, 14-15.9.1998, FERMILAB-PUB-98-379, nucl-th/9812038; N.V. Mokhov, The Mars Code System User's Guide, Fermilab-FN-628, 1995; N.V. Mokhov, MARS Code Developments, Benchmarking and Applications, Fermilab-Conf-00-066, 2000; O.E. Krivosheev, N.V. Mokhov, A New MARS and its Applications, Fermilab-Conf-98/43, 1998] code interfaced with MCNP [J.S. Hendrics, MCNP4C LANL Memo X-5; JSH-2000-3; J.F. Briemesteir (Ed.), MCNP - A General Montecarlo N-Particle Transport Code, Version 4C, LA-13709-M] libraries, of the neutron flux from the converters interacting with the actinide targets.

Radioactive ion beams produced by neutron-induced fission at ISOLDE

NASA Astrophysics Data System (ADS)

Isolde Collaboration; Catherall, R.; Lettry, J.; Gilardoni, S.; Köster, U.

2003-05-01

The production rates of neutron-rich fission products for the next-generation radioactive beam facility EURISOL [EU-RTD Project EURISOL (HPRI-CT-1999-50001)] are mainly limited by the maximum amount of power deposited by protons in the target. An alternative approach is to use neutron beams to induce fission in actinide targets. This has the advantage of reducing: the energy deposited by the proton beam in the target; contamination from neutron-deficient isobars that would be produced by spallation; and mechanical stress on the target. At ISOLDE CERN [E. Kugler, Hyperfine Interact. 129 (2000) 23], tests have been made on standard ISOLDE actinide targets using fast-neutron bunches produced by bombarding thick, high-/Z metal converters with 1 and 1.4 GeV proton pulses. This paper reviews the first applications of converters used at ISOLDE. It highlights the different geometries and the techniques used to compare fission yields produced by the proton beam directly on the target with neutron-induced fission. Results from the six targets already tested, namely UC2/graphite and ThO2 targets with tungsten and tantalum converters, are presented. To gain further knowledge for the design of a dedicated target as required by the TARGISOL project [EU-RTD Project TARGISOL (HPRI-CT-2001-50033)], the results are compared to simulations, using the MARS [N.V. Mokhov, S.I. Striganov, A. Van Ginneken, S.G. Mashnik, A.J. Sierk, J. Ranft, MARS code developments, in: 4th Workshop on Simulating Accelerator Radiation Environments, SARE-4, Knoxville, USA, 14-15.9.1998, FERMILAB-PUB-98-379, nucl-th/9812038; N.V. Mokhov, The Mars Code System User's Guide, Fermilab-FN-628, 1995; N.V. Mokhov, MARS Code Developments, Benchmarking and Applications, Fermilab-Conf-00-066, 2000; O.E. Krivosheev, N.V. Mokhov, A New MARS and its Applications, Fermilab-Conf-98/43, 1998] code interfaced with MCNP [J.S. Hendrics, MCNP4C LANL Memo X-5; JSH-2000-3; J.F. Briemesteir (Ed.), MCNP - A General Montecarlo N-Particle Transport Code, Version 4C, LA-13709-M] libraries, of the neutron flux from the converters interacting with the actinide targets.

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

NASA Astrophysics Data System (ADS)

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

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

"ON ALGEBRAIC DECODING OF Q-ARY REED-MULLER AND PRODUCT REED-SOLOMON CODES"

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

SANTHI, NANDAKISHORE

We consider a list decoding algorithm recently proposed by Pellikaan-Wu for q-ary Reed-Muller codes RM{sub q}({ell}, m, n) of length n {le} q{sup m} when {ell} {le} q. A simple and easily accessible correctness proof is given which shows that this algorithm achieves a relative error-correction radius of {tau} {le} (1-{radical}{ell}q{sup m-1}/n). This is an improvement over the proof using one-point Algebraic-Geometric decoding method given in. The described algorithm can be adapted to decode product Reed-Solomon codes. We then propose a new low complexity recursive aJgebraic decoding algorithm for product Reed-Solomon codes and Reed-Muller codes. This algorithm achieves a relativemore » error correction radius of {tau} {le} {Pi}{sub i=1}{sup m} (1 - {radical}k{sub i}/q). This algorithm is then proved to outperform the Pellikaan-Wu algorithm in both complexity and error correction radius over a wide range of code rates.« less

MCNP simulation of a Theratron 780 radiotherapy unit.

PubMed

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.

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

PubMed

Hordósy, G

2005-01-01

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

The viability of ADVANTG deterministic method for synthetic radiography generation

NASA Astrophysics Data System (ADS)

Bingham, Andrew; Lee, Hyoung K.

2018-07-01

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

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

NASA Astrophysics Data System (ADS)

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

2015-05-01

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

Benchmark of neutron production cross sections with Monte Carlo codes

NASA Astrophysics Data System (ADS)

Tsai, Pi-En; Lai, Bo-Lun; Heilbronn, Lawrence H.; Sheu, Rong-Jiun

2018-02-01

Aiming to provide critical information in the fields of heavy ion therapy, radiation shielding in space, and facility design for heavy-ion research accelerators, the physics models in three Monte Carlo simulation codes - PHITS, FLUKA, and MCNP6, were systematically benchmarked with comparisons to fifteen sets of experimental data for neutron production cross sections, which include various combinations of 12C, 20Ne, 40Ar, 84Kr and 132Xe projectiles and natLi, natC, natAl, natCu, and natPb target nuclides at incident energies between 135 MeV/nucleon and 600 MeV/nucleon. For neutron energies above 60% of the specific projectile energy per nucleon, the LAQGMS03.03 in MCNP6, the JQMD/JQMD-2.0 in PHITS, and the RQMD-2.4 in FLUKA all show a better agreement with data in heavy-projectile systems than with light-projectile systems, suggesting that the collective properties of projectile nuclei and nucleon interactions in the nucleus should be considered for light projectiles. For intermediate-energy neutrons whose energies are below the 60% projectile energy per nucleon and above 20 MeV, FLUKA is likely to overestimate the secondary neutron production, while MCNP6 tends towards underestimation. PHITS with JQMD shows a mild tendency for underestimation, but the JQMD-2.0 model with a modified physics description for central collisions generally improves the agreement between data and calculations. For low-energy neutrons (below 20 MeV), which are dominated by the evaporation mechanism, PHITS (which uses GEM linked with JQMD and JQMD-2.0) and FLUKA both tend to overestimate the production cross section, whereas MCNP6 tends to underestimate more systems than to overestimate. For total neutron production cross sections, the trends of the benchmark results over the entire energy range are similar to the trends seen in the dominate energy region. Also, the comparison of GEM coupled with either JQMD or JQMD-2.0 in the PHITS code indicates that the model used to describe the first stage of a nucleus-nucleus collision also affects the low-energy neutron production. Thus, in this case, a proper combination of two physics models is desired to reproduce the measured results. In addition, code users should be aware that certain models consistently produce secondary neutrons within a constant fraction of another model in certain energy regions, which might be correlated to different physics treatments in different models.

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

Haghighat, A.; Sjoden, G.E.; Wagner, J.C.

In the past 10 yr, the Penn State Transport Theory Group (PSTTG) has concentrated its efforts on developing accurate and efficient particle transport codes to address increasing needs for efficient and accurate simulation of nuclear systems. The PSTTG's efforts have primarily focused on shielding applications that are generally treated using multigroup, multidimensional, discrete ordinates (S{sub n}) deterministic and/or statistical Monte Carlo methods. The difficulty with the existing public codes is that they require significant (impractical) computation time for simulation of complex three-dimensional (3-D) problems. For the S{sub n} codes, the large memory requirements are handled through the use of scratchmore » files (i.e., read-from and write-to-disk) that significantly increases the necessary execution time. Further, the lack of flexible features and/or utilities for preparing input and processing output makes these codes difficult to use. The Monte Carlo method becomes impractical because variance reduction (VR) methods have to be used, and normally determination of the necessary parameters for the VR methods is very difficult and time consuming for a complex 3-D problem. For the deterministic method, the authors have developed the 3-D parallel PENTRAN (Parallel Environment Neutral-particle TRANsport) code system that, in addition to a parallel 3-D S{sub n} solver, includes pre- and postprocessing utilities. PENTRAN provides for full phase-space decomposition, memory partitioning, and parallel input/output to provide the capability of solving large problems in a relatively short time. Besides having a modular parallel structure, PENTRAN has several unique new formulations and features that are necessary for achieving high parallel performance. For the Monte Carlo method, the major difficulty currently facing most users is the selection of an effective VR method and its associated parameters. For complex problems, generally, this process is very time consuming and may be complicated due to the possibility of biasing the results. In an attempt to eliminate this problem, the authors have developed the A{sup 3}MCNP (automated adjoint accelerated MCNP) code that automatically prepares parameters for source and transport biasing within a weight-window VR approach based on the S{sub n} adjoint function. A{sup 3}MCNP prepares the necessary input files for performing multigroup, 3-D adjoint S{sub n} calculations using TORT.« less

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

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

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

2002-05-15

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

Evaluation of the DRAGON code for VHTR design analysis.

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

Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division

2006-01-12

This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by themore » IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR.« less

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

NASA Astrophysics Data System (ADS)

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

2017-09-01

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

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

PubMed

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

2005-01-01

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

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

PubMed

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

2017-01-01

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

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

PubMed Central

Krstic, D.; Nikezic, D.

2017-01-01

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

Gamma-ray dose from an overhead plume

DOE PAGES

McNaughton, Michael W.; Gillis, Jessica McDonnel; Ruedig, Elizabeth; ...

2017-05-01

Standard plume models can underestimate the gamma-ray dose when most of the radioactive material is above the heads of the receptors. Typically, a model is used to calculate the air concentration at the height of the receptor, and the dose is calculated by multiplying the air concentration by a concentration-to-dose conversion factor. Models indicate that if the plume is emitted from a stack during stable atmospheric conditions, the lower edges of the plume may not reach the ground, in which case both the ground-level concentration and the dose are usually reported as zero. However, in such cases, the dose frommore » overhead gamma-emitting radionuclides may be substantial. Such underestimates could impact decision making in emergency situations. The Monte Carlo N-Particle code, MCNP, was used to calculate the overhead shine dose and to compare with standard plume models. At long distances and during unstable atmospheric conditions, the MCNP results agree with the standard models. As a result, at short distances, where many models calculate zero, the true dose (as modeled by MCNP) can be estimated with simple equations.« less

Path Toward a Unified Geometry for Radiation Transport

NASA Astrophysics Data System (ADS)

Lee, Kerry

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

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

PubMed Central

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

2012-01-01

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

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

PubMed Central

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

2011-01-01

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

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

PubMed

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

2012-10-01

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

Neutron streaming studies along JET shielding penetrations

NASA Astrophysics Data System (ADS)

Stamatelatos, Ion E.; Vasilopoulou, Theodora; Batistoni, Paola; Obryk, Barbara; Popovichev, Sergey; Naish, Jonathan

2017-09-01

Neutronic benchmark experiments are carried out at JET aiming to assess the neutronic codes and data used in ITER analysis. Among other activities, experiments are performed in order to validate neutron streaming simulations along long penetrations in the JET shielding configuration. In this work, neutron streaming calculations along the JET personnel entrance maze are presented. Simulations were performed using the MCNP code for Deuterium-Deuterium and Deuterium- Tritium plasma sources. The results of the simulations were compared against experimental data obtained using thermoluminescence detectors and activation foils.

Parameter dependence of the MCNP electron transport in determining dose distributions.

PubMed

Reynaert, N; Palmans, H; Thierens, H; Jeraj, R

2002-10-01

In this paper, a detailed study of the electron transport in MCNP is performed, separating the effects of the energy binning technique on the energy loss rate, the scattering angles, and the sub-step length as a function of energy. As this problem is already well known, in this paper we focus on the explanation as to why the default mode of MCNP can lead to large deviations. The resolution dependence was investigated as well. An error in the MCNP code in the energy binning technique in the default mode (DBCN 18 card = 0) was revealed, more specific in the updating of cross sections when a sub-step is performed corresponding to a high-energy loss. This updating error is not present in the ITS mode (DBCN 18 card = 1) and leads to a systematically lower dose deposition rate in the default mode. The effect is present for all energies studied (0.5-10 MeV) and depends on the geometrical resolution of the scoring regions and the energy grid resolution. The effect of the energy binning technique is of the same order of that of the updating error for energies below 2 MeV, and becomes less important for higher energies. For a 1 MeV point source surrounded by homogeneous water, the deviation of the default MCNP results at short distances attains 9% and remains approximately the same for all energies. This effect could be corrected by removing the completion of an energy step each time an electron changes from an energy bin during a sub-step. Another solution consists of performing all calculations in the ITS mode. Another problem is the resolution dependence, even in the ITS mode. The higher the resolution is chosen (the smaller the scoring regions) the faster the energy is deposited along the electron track. It is proven that this is caused by starting a new energy step when crossing a surface. The resolution effect should be investigated for every specific case when calculating dose distributions around beta sources. The resolution should not be higher than 0.85*(1-EFAC)*CSDA, where EFAC is the energy loss per energy step and CSDA a continuous slowing down approximation range. This effect could as well be removed by determining the cross sections for energy loss and multiple scattering at the average energy of an energy step and by sampling the cross sections for each sub-step. Overall, we conclude that MCNP cannot be used without a caution due to possible errors in the electron transport. When care is taken, it is possible to obtain correct results that are in agreement with other Monte Carlo codes.

Monte Carlo calculation for the development of a BNCT neutron source (1eV-10KeV) using MCNP code.

PubMed

El Moussaoui, F; El Bardouni, T; Azahra, M; Kamili, A; Boukhal, H

2008-09-01

Different materials have been studied in order to produce the epithermal neutron beam between 1eV and 10KeV, which are extensively used to irradiate patients with brain tumors such as GBM. For this purpose, we have studied three different neutrons moderators (H(2)O, D(2)O and BeO) and their combinations, four reflectors (Al(2)O(3), C, Bi, and Pb) and two filters (Cd and Bi). Results of calculation showed that the best obtained assembly configuration corresponds to the combination of the three moderators H(2)O, BeO and D(2)O jointly to Al(2)O(3) reflector and two filter Cd+Bi optimize the spectrum of the epithermal neutron at 72%, and minimize the thermal neutron to 4% and thus it can be used to treat the deep tumor brain. The calculations have been performed by means of the Monte Carlo N (particle code MCNP 5C). Our results strongly encourage further studying of irradiation of the head with epithermal neutron fields.

Assessment and Mitigation of Radiation, EMP, Debris & Shrapnel Impacts at Megajoule-Class Laser Facilities

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

Eder, D C; Anderson, R W; Bailey, D S

2009-10-05

The generation of neutron/gamma radiation, electromagnetic pulses (EMP), debris and shrapnel at mega-Joule class laser facilities (NIF and LMJ) impacts experiments conducted at these facilities. The complex 3D numerical codes used to assess these impacts range from an established code that required minor modifications (MCNP - calculates neutron and gamma radiation levels in complex geometries), through a code that required significant modifications to treat new phenomena (EMSolve - calculates EMP from electrons escaping from laser targets), to a new code, ALE-AMR, that is being developed through a joint collaboration between LLNL, CEA, and UC (UCSD, UCLA, and LBL) for debrismore » and shrapnel modelling.« less

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

NASA Astrophysics Data System (ADS)

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

2014-10-01

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

Treating electron transport in MCNP{sup trademark}

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

Hughes, H.G.

1996-12-31

The transport of electrons and other charged particles is fundamentally different from that of neutrons and photons. A neutron, in aluminum slowing down from 0.5 MeV to 0.0625 MeV will have about 30 collisions; a photon will have fewer than ten. An electron with the same energy loss will undergo 10{sup 5} individual interactions. This great increase in computational complexity makes a single- collision Monte Carlo approach to electron transport unfeasible for many situations of practical interest. Considerable theoretical work has been done to develop a variety of analytic and semi-analytic multiple-scattering theories for the transport of charged particles. Themore » theories used in the algorithms in MCNP are the Goudsmit-Saunderson theory for angular deflections, the Landau an theory of energy-loss fluctuations, and the Blunck-Leisegang enhancements of the Landau theory. In order to follow an electron through a significant energy loss, it is necessary to break the electron`s path into many steps. These steps are chosen to be long enough to encompass many collisions (so that multiple-scattering theories are valid) but short enough that the mean energy loss in any one step is small (for the approximations in the multiple-scattering theories). The energy loss and angular deflection of the electron during each step can then be sampled from probability distributions based on the appropriate multiple- scattering theories. This subsumption of the effects of many individual collisions into single steps that are sampled probabilistically constitutes the ``condensed history`` Monte Carlo method. This method is exemplified in the ETRAN series of electron/photon transport codes. The ETRAN codes are also the basis for the Integrated TIGER Series, a system of general-purpose, application-oriented electron/photon transport codes. The electron physics in MCNP is similar to that of the Integrated TIGER Series.« less

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

PubMed

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

2010-03-01

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

FY2012 summary of tasks completed on PROTEUS-thermal work.

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

Lee, C.H.; Smith, M.A.

2012-06-06

PROTEUS is a suite of the neutronics codes, both old and new, that can be used within the SHARP codes being developed under the NEAMS program. Discussion here is focused on updates and verification and validation activities of the SHARP neutronics code, DeCART, for application to thermal reactor analysis. As part of the development of SHARP tools, the different versions of the DeCART code created for PWR, BWR, and VHTR analysis were integrated. Verification and validation tests for the integrated version were started, and the generation of cross section libraries based on the subgroup method was revisited for the targetedmore » reactor types. The DeCART code has been reorganized in preparation for an efficient integration of the different versions for PWR, BWR, and VHTR analysis. In DeCART, the old-fashioned common blocks and header files have been replaced by advanced memory structures. However, the changing of variable names was minimized in order to limit problems with the code integration. Since the remaining stability problems of DeCART were mostly caused by the CMFD methodology and modules, significant work was performed to determine whether they could be replaced by more stable methods and routines. The cross section library is a key element to obtain accurate solutions. Thus, the procedure for generating cross section libraries was revisited to provide libraries tailored for the targeted reactor types. To improve accuracy in the cross section library, an attempt was made to replace the CENTRM code by the MCNP Monte Carlo code as a tool obtaining reference resonance integrals. The use of the Monte Carlo code allows us to minimize problems or approximations that CENTRM introduces since the accuracy of the subgroup data is limited by that of the reference solutions. The use of MCNP requires an additional set of libraries without resonance cross sections so that reference calculations can be performed for a unit cell in which only one isotope of interest includes resonance cross sections, among the isotopes in the composition. The OECD MHTGR-350 benchmark core was simulated using DeCART as initial focus of the verification/validation efforts. Among the benchmark problems, Exercise 1 of Phase 1 is a steady-state benchmark case for the neutronics calculation for which block-wise cross sections were provided in 26 energy groups. This type of problem was designed for a homogenized geometry solver like DIF3D rather than the high-fidelity code DeCART. Instead of the homogenized block cross sections given in the benchmark, the VHTR-specific 238-group ENDF/B-VII.0 library of DeCART was directly used for preliminary calculations. Initial results showed that the multiplication factors of a fuel pin and a fuel block with or without a control rod hole were off by 6, -362, and -183 pcm Dk from comparable MCNP solutions, respectively. The 2-D and 3-D one-third core calculations were also conducted for the all-rods-out (ARO) and all-rods-in (ARI) configurations, producing reasonable results. Figure 1 illustrates the intermediate (1.5 eV - 17 keV) and thermal (below 1.5 eV) group flux distributions. As seen from VHTR cores with annular fuels, the intermediate group fluxes are relatively high in the fuel region, but the thermal group fluxes are higher in the inner and outer graphite reflector regions than in the fuel region. To support the current project, a new three-year I-NERI collaboration involving ANL and KAERI was started in November 2011, focused on performing in-depth verification and validation of high-fidelity multi-physics simulation codes for LWR and VHTR. The work scope includes generating improved cross section libraries for the targeted reactor types, developing benchmark models for verification and validation of the neutronics code with or without thermo-fluid feedback, and performing detailed comparisons of predicted reactor parameters against both Monte Carlo solutions and experimental measurements. The following list summarizes the work conducted so far for PROTEUS-Thermal Tasks: Unification of different versions of DeCART was initiated, and at the same time code modernization was conducted to make code unification efficient; (2) Regeneration of cross section libraries was attempted for the targeted reactor types, and the procedure for generating cross section libraries was updated by replacing CENTRM with MCNP for reference resonance integrals; (3) The MHTGR-350 benchmark core was simulated using DeCART with VHTR-specific 238-group ENDF/B-VII.0 library, and MCNP calculations were performed for comparison; and (4) Benchmark problems for PWR and BWR analysis were prepared for the DeCART verification/validation effort. In the coming months, the work listed above will be completed. Cross section libraries will be generated with optimized group structures for specific reactor types.« less

SUMCOR: Cascade summing correction for volumetric sources applying MCNP6.

PubMed

Dias, M S; Semmler, R; Moreira, D S; de Menezes, M O; Barros, L F; Ribeiro, R V; Koskinas, M F

2018-04-01

The main features of code SUMCOR developed for cascade summing correction for volumetric sources are described. MCNP6 is used to track histories starting from individual points inside the volumetric source, for each set of cascade transitions from the radionuclide. Total and FEP efficiencies are calculated for all gamma-rays and X-rays involved in the cascade. Cascade summing correction is based on the matrix formalism developed by Semkow et al. (1990). Results are presented applying the experimental data sent to the participants of two intercomparisons organized by the ICRM-GSWG and coordinated by Dr. Marie-Cristine Lépy from the Laboratoire National Henri Becquerel (LNE-LNHB), CEA, in 2008 and 2010, respectively and compared to the other participants in the intercomparisons. Copyright © 2017 Elsevier Ltd. All rights reserved.

An analysis of MCNP cross-sections and tally methods for low-energy photon emitters.

PubMed

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.

Los Alamos radiation transport code system on desktop computing platforms

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

Briesmeister, J.F.; Brinkley, F.W.; Clark, B.A.

The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. These codes were originally developed many years ago and have undergone continual improvement. With a large initial effort and continued vigilance, the codes are easily portable from one type of hardware to another. The performance of scientific work-stations (SWS) has evolved to the point that such platforms can be used routinely to perform sophisticated radiation transport calculations. As the personal computer (PC) performance approaches that of the SWS, the hardware options for desk-top radiation transport calculations expands considerably. Themore » current status of the radiation transport codes within the LARTCS is described: MCNP, SABRINA, LAHET, ONEDANT, TWODANT, TWOHEX, and ONELD. Specifically, the authors discuss hardware systems on which the codes run and present code performance comparisons for various machines.« less

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

NASA Astrophysics Data System (ADS)

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

2018-01-01

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

Path Toward a Unifid Geometry for Radiation Transport

NASA Technical Reports Server (NTRS)

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

2014-01-01

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

Verification of MCNP6.2 for Nuclear Criticality Safety Applications

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

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

2017-05-10

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

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

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

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

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

An approach to design a 90Sr radioisotope thermoelectric generator using analytical and Monte Carlo methods with ANSYS, COMSOL, and MCNP.

PubMed

Khajepour, Abolhasan; Rahmani, Faezeh

2017-01-01

In this study, a 90 Sr radioisotope thermoelectric generator (RTG) with power of milliWatt was designed to operate in the determined temperature (300-312K). For this purpose, the combination of analytical and Monte Carlo methods with ANSYS and COMSOL software as well as the MCNP code was used. This designed RTG contains 90 Sr as a radioisotope heat source (RHS) and 127 coupled thermoelectric modules (TEMs) based on bismuth telluride. Kapton (2.45mm in thickness) and Cryotherm sheets (0.78mm in thickness) were selected as the thermal insulators of the RHS, as well as a stainless steel container was used as a generator chamber. The initial design of the RHS geometry was performed according to the amount of radioactive material (strontium titanate) as well as the heat transfer calculations and mechanical strength considerations. According to the Monte Carlo simulation performed by the MCNP code, approximately 0.35 kCi of 90 Sr is sufficient to generate heat power in the RHS. To determine the optimal design of the RTG, the distribution of temperature as well as the dissipated heat and input power to the module were calculated in different parts of the generator using the ANSYS software. Output voltage according to temperature distribution on TEM was calculated using COMSOL. Optimization of the dimension of the RHS and heat insulator was performed to adapt the average temperature of the hot plate of TEM to the determined hot temperature value. This designed RTG generates 8mW in power with an efficiency of 1%. This proposed approach of combination method can be used for the precise design of various types of RTGs. Copyright © 2016 Elsevier Ltd. All rights reserved.

Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation

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

Pecchia, M.; D'Auria, F.; Mazzantini, O.

2012-07-01

Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3D{sup C}/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI formore » performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)« less

Feasibility study for wax deposition imaging in oil pipelines by PGNAA technique.

PubMed

Cheng, Can; Jia, Wenbao; Hei, Daqian; Wei, Zhiyong; Wang, Hongtao

2017-10-01

Wax deposition in pipelines is a crucial problem in the oil industry. A method based on the prompt gamma-ray neutron activation analysis technique was applied to reconstruct the image of wax deposition in oil pipelines. The 2.223MeV hydrogen capture gamma rays were used to reconstruct the wax deposition image. To validate the method, both MCNP simulation and experiments were performed for wax deposited with a maximum thickness of 20cm. The performance of the method was simulated using the MCNP code. The experiment was conducted with a 252 Cf neutron source and a LaBr 3 : Ce detector. A good correspondence between the simulations and the experiments was observed. The results obtained indicate that the present approach is efficient for wax deposition imaging in oil pipelines. Copyright © 2017 Elsevier Ltd. All rights reserved.

Proton Dose Assessment to the Human Eye Using Monte Carlo N-Particle Transport Code (MCNPX)

DTIC Science & Technology

2006-08-01

current treatments are applied using an infrared diode laser 10 (projecting a spot size of 2-3 mm), used for about 1 minute per exposure. The laser heats...1983. Shultis J, Faw R. An MCNP Primer. Available at: http:// ww2 .mne.ksu.edu/-jks/MCNPprmr.pdf. Accessed 3 January 2006. Stys P, Lopachin R

Estimation of coolant void reactivity for CANDU-NG lattice using DRAGON and validation using MCNP5 and TRIPOLI-4.3

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

Karthikeyan, R.; Tellier, R. L.; Hebert, A.

2006-07-01

The Coolant Void Reactivity (CVR) is an important safety parameter that needs to be estimated at the design stage of a nuclear reactor. It helps to have an a priori knowledge of the behavior of the system during a transient initiated by the loss of coolant. In the present paper, we have attempted to estimate the CVR for a CANDU New Generation (CANDU-NG) lattice, as proposed at an early stage of the Advanced CANDU Reactor (ACR) development. We have attempted to estimate the CVR with development version of the code DRAGON, using the method of characteristics. DRAGON has several advancedmore » self-shielding models incorporated in it, each of them compatible with the method of characteristics. This study will bring to focus the performance of these self-shielding models, especially when there is voiding of such a tight lattice. We have also performed assembly calculations in 2 x 2 pattern for the CANDU-NG fuel, with special emphasis on checkerboard voiding. The results obtained have been validated against Monte Carlo codes MCNP5 and TRIPOLI-4.3. (authors)« less

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

PubMed

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

2000-05-01

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

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

PubMed

Shaaban, Ismail; Albarhoum, Mohamad

2016-08-01

The MCNP4C2 code was used to simulate the MTR-22 MW research reactor and perform the neutronic analysis for a new fuel namely: a MOX (U3O8&PuO2) fuel dispersed in an Al matrix for One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) in its core. Its new characteristics were compared to its original characteristics based on the U3O8-Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U3O8-Al fuel at nominal power were used to validate the calculated values and were found acceptable. The achieved results seem to confirm that the use of MOX fuel in the MTR-22 MW will not degrade the safe operational conditions of the reactor. In addition, the use of MOX fuel in the MTR-22 MW core leads to reduce the uranium fuel enrichment with (235)U and the amount of loaded (235)U in the core by about 34.84% and 15.21% for the ONT and TNTs cases, respectively. Copyright © 2016 Elsevier Ltd. All rights reserved.

Advanced Modulation and Coding Technology Conference

NASA Technical Reports Server (NTRS)

1992-01-01

The objectives, approach, and status of all current LeRC-sponsored industry contracts and university grants are presented. The following topics are covered: (1) the LeRC Space Communications Program, and Advanced Modulation and Coding Projects; (2) the status of four contracts for development of proof-of-concept modems; (3) modulation and coding work done under three university grants, two small business innovation research contracts, and two demonstration model hardware development contracts; and (4) technology needs and opportunities for future missions.

Neutron dose rate analysis on HTGR-10 reactor using Monte Carlo code

NASA Astrophysics Data System (ADS)

Suwoto; Adrial, H.; Hamzah, A.; Zuhair; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The HTGR-10 reactor is cylinder-shaped core fuelled with kernel TRISO coated fuel particles in the spherical pebble with helium cooling system. The outlet helium gas coolant temperature outputted from the reactor core is designed to 700 °C. One advantage HTGR type reactor is capable of co-generation, as an addition to generating electricity, the reactor was designed to produce heat at high temperature can be used for other processes. The spherical fuel pebble contains 8335 TRISO UO2 kernel coated particles with enrichment of 10% and 17% are dispersed in a graphite matrix. The main purpose of this study was to analysis the distribution of neutron dose rates generated from HTGR-10 reactors. The calculation and analysis result of neutron dose rate in the HTGR-10 reactor core was performed using Monte Carlo MCNP5v1.6 code. The problems of double heterogeneity in kernel fuel coated particles TRISO and spherical fuel pebble in the HTGR-10 core are modelled well with MCNP5v1.6 code. The neutron flux to dose conversion factors taken from the International Commission on Radiological Protection (ICRP-74) was used to determine the dose rate that passes through the active core, reflectors, core barrel, reactor pressure vessel (RPV) and a biological shield. The calculated results of neutron dose rate with MCNP5v1.6 code using a conversion factor of ICRP-74 (2009) for radiation workers in the radial direction on the outside of the RPV (radial position = 220 cm from the center of the patio HTGR-10) provides the respective value of 9.22E-4 μSv/h and 9.58E-4 μSv/h for enrichment 10% and 17%, respectively. The calculated values of neutron dose rates are compliant with BAPETEN Chairman’s Regulation Number 4 Year 2013 on Radiation Protection and Safety in Nuclear Energy Utilization which sets the limit value for the average effective dose for radiation workers 20 mSv/year or 10μSv/h. Thus the protection and safety for radiation workers to be safe from the radiation source has been fulfilled. From the result analysis, it can be concluded that the model of calculation result of neutron dose rate for HTGR-10 core has met the required radiation safety standards.

Methodes iteratives paralleles: Applications en neutronique et en mecanique des fluides

NASA Astrophysics Data System (ADS)

Qaddouri, Abdessamad

Dans cette these, le calcul parallele est applique successivement a la neutronique et a la mecanique des fluides. Dans chacune de ces deux applications, des methodes iteratives sont utilisees pour resoudre le systeme d'equations algebriques resultant de la discretisation des equations du probleme physique. Dans le probleme de neutronique, le calcul des matrices des probabilites de collision (PC) ainsi qu'un schema iteratif multigroupe utilisant une methode inverse de puissance sont parallelises. Dans le probleme de mecanique des fluides, un code d'elements finis utilisant un algorithme iteratif du type GMRES preconditionne est parallelise. Cette these est presentee sous forme de six articles suivis d'une conclusion. Les cinq premiers articles traitent des applications en neutronique, articles qui representent l'evolution de notre travail dans ce domaine. Cette evolution passe par un calcul parallele des matrices des PC et un algorithme multigroupe parallele teste sur un probleme unidimensionnel (article 1), puis par deux algorithmes paralleles l'un mutiregion l'autre multigroupe, testes sur des problemes bidimensionnels (articles 2--3). Ces deux premieres etapes sont suivies par l'application de deux techniques d'acceleration, le rebalancement neutronique et la minimisation du residu aux deux algorithmes paralleles (article 4). Finalement, on a mis en oeuvre l'algorithme multigroupe et le calcul parallele des matrices des PC sur un code de production DRAGON ou les tests sont plus realistes et peuvent etre tridimensionnels (article 5). Le sixieme article (article 6), consacre a l'application a la mecanique des fluides, traite la parallelisation d'un code d'elements finis FES ou le partitionneur de graphe METIS et la librairie PSPARSLIB sont utilises.

Multiple Detector Optimization for Hidden Radiation Source Detection

DTIC Science & Technology

2015-03-26

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

Calculations of skyshine from an intense portable electron linac

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

Estes, G.P.; Hughes, H.G.; Fry, D.A.

1994-12-31

The MCNP Monte carlo code has been used at Los Alamos to calculate skyshine and terrain albedo efects from an intense portable electron linear accelerator that is to be used by the Russian Federation to radiograph nuclear weapons that may have been damaged by accidents. Relative dose rate profiles have been calculated. The design of the accelerator, along with a diagram, is presented.

Implementation and testing of the on-the-fly thermal scattering Monte Carlo sampling method for graphite and light water in MCNP6

DOE PAGES

Pavlou, Andrew T.; Ji, Wei; Brown, Forrest B.

2016-01-23

Here, a proper treatment of thermal neutron scattering requires accounting for chemical binding through a scattering law S(α,β,T). Monte Carlo codes sample the secondary neutron energy and angle after a thermal scattering event from probability tables generated from S(α,β,T) tables at discrete temperatures, requiring a large amount of data for multiscale and multiphysics problems with detailed temperature gradients. We have previously developed a method to handle this temperature dependence on-the-fly during the Monte Carlo random walk using polynomial expansions in 1/T to directly sample the secondary energy and angle. In this paper, the on-the-fly method is implemented into MCNP6 andmore » tested in both graphite-moderated and light water-moderated systems. The on-the-fly method is compared with the thermal ACE libraries that come standard with MCNP6, yielding good agreement with integral reactor quantities like k-eigenvalue and differential quantities like single-scatter secondary energy and angle distributions. The simulation runtimes are comparable between the two methods (on the order of 5–15% difference for the problems tested) and the on-the-fly fit coefficients only require 5–15 MB of total data storage.« less

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

PubMed

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

2014-08-01

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

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

PubMed

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

2018-07-01

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

Simulation of irradiation exposure of electronic devices due to heavy ion therapy with Monte Carlo Code MCNP6

NASA Astrophysics Data System (ADS)

Lapins, Janis; Guilliard, Nicole; Bernnat, Wolfgang; Buck, Arnulf

2017-09-01

During heavy ion irradiation therapy the patient has to be located exactly at the right position to make sure that the Bragg peak occurs in the tumour. The patient has to be moved in the range of millimetres to scan the ill tissue. For that reason a special table was developed which allows exact positioning. The electronic control can be located outside the surgery. But that has some disadvantage for the construction. To keep the system compact it would be much more comfortable to put the electronic control inside the surgery. As a lot of high energetic secondary particles are produced during the therapy causing a high dose in the room it is important to find positions with low dose rates. Therefore, investigations are needed where the electronic devices should be located to obtain a minimum of radiation, help to prevent the failure of sensitive devices. The dose rate was calculated for carbon ions with different initial energy and protons over the entire therapy room with Monte Carlo particle tracking using MCNP6. The types of secondary particles were identified and the dose rate for a thin silicon layer and an electronic mixture material was determined. In addition, the shielding effect of several selected material layers was calculated using MCNP6.

Characterizing scintillator detector response for correlated fission experiments with MCNP and associated packages

DOE PAGES

Andrews, M. T.; Rising, M. E.; Meierbachtol, K.; ...

2018-06-15

Wmore » hen multiple neutrons are emitted in a fission event they are correlated in both energy and their relative angle, which may impact the design of safeguards equipment and other instrumentation for non-proliferation applications. The most recent release of MCNP 6 . 2 contains the capability to simulate correlated fission neutrons using the event generators CGMF and FREYA . These radiation transport simulations will be post-processed by the detector response code, DRiFT , and compared directly to correlated fission measurements. DRiFT has been previously compared to single detector measurements, its capabilities have been recently expanded with correlated fission simulations in mind. Finally, this paper details updates to DRiFT specific to correlated fission measurements, including tracking source particle energy of all detector events (and non-events), expanded output formats, and digitizer waveform generation.« less

Spectral unfolding of fast neutron energy distributions

NASA Astrophysics Data System (ADS)

Mosby, Michelle; Jackman, Kevin; Engle, Jonathan

2015-10-01

The characterization of the energy distribution of a neutron flux is difficult in experiments with constrained geometry where techniques such as time of flight cannot be used to resolve the distribution. The measurement of neutron fluxes in reactors, which often present similar challenges, has been accomplished using radioactivation foils as an indirect probe. Spectral unfolding codes use statistical methods to adjust MCNP predictions of neutron energy distributions using quantified radioactive residuals produced in these foils. We have applied a modification of this established neutron flux characterization technique to experimentally characterize the neutron flux in the critical assemblies at the Nevada National Security Site (NNSS) and the spallation neutron flux at the Isotope Production Facility (IPF) at Los Alamos National Laboratory (LANL). Results of the unfolding procedure are presented and compared with a priori MCNP predictions, and the implications for measurements using the neutron fluxes at these facilities are discussed.

Methodology of full-core Monte Carlo calculations with leakage parameter evaluations for benchmark critical experiment analysis

NASA Astrophysics Data System (ADS)

Sboev, A. G.; Ilyashenko, A. S.; Vetrova, O. A.

1997-02-01

The method of bucking evaluation, realized in the MOnte Carlo code MCS, is described. This method was applied for calculational analysis of well known light water experiments TRX-1 and TRX-2. The analysis of this comparison shows, that there is no coincidence between Monte Carlo calculations, obtained by different ways: the MCS calculations with given experimental bucklings; the MCS calculations with given bucklings evaluated on base of full core MCS direct simulations; the full core MCNP and MCS direct simulations; the MCNP and MCS calculations, where the results of cell calculations are corrected by the coefficients taking into the account the leakage from the core. Also the buckling values evaluated by full core MCS calculations have differed from experimental ones, especially in the case of TRX-1, when this difference has corresponded to 0.5 percent increase of Keff value.

Characterizing scintillator detector response for correlated fission experiments with MCNP and associated packages

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

Andrews, M. T.; Rising, M. E.; Meierbachtol, K.

Wmore » hen multiple neutrons are emitted in a fission event they are correlated in both energy and their relative angle, which may impact the design of safeguards equipment and other instrumentation for non-proliferation applications. The most recent release of MCNP 6 . 2 contains the capability to simulate correlated fission neutrons using the event generators CGMF and FREYA . These radiation transport simulations will be post-processed by the detector response code, DRiFT , and compared directly to correlated fission measurements. DRiFT has been previously compared to single detector measurements, its capabilities have been recently expanded with correlated fission simulations in mind. Finally, this paper details updates to DRiFT specific to correlated fission measurements, including tracking source particle energy of all detector events (and non-events), expanded output formats, and digitizer waveform generation.« less

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

Bily, T.

Thermoluminescent dosimeters represent very useful tool for gamma fields parameters measurements at nuclear research reactors, especially at zero power ones. {sup 7}LiF:Mg,Ti and {sup 7}LiF:Mg,Cu,P type TL dosimeters enable determination of only gamma component in mixed neutron - gamma field. At VR-1 reactor operated within the Faculty of Nuclear Sciences and Physical Engineering at the Czech Technical University in Prague the integral characteristics of gamma rays field were investigated, especially its spatial distribution and time behaviour, i.e. the non-saturated delayed gamma ray emission influence. Measured spatial distributions were compared with monte carlo code MCNP5 calculations. Although MCNP cannot generate delayedmore » gamma rays from fission, the relative gamma dose rate distribution is within {+-} 15% with measured values. The experiments were carried out with core configuration C1 consisting of LEU fuel IRT-4M (19.7 %). (author)« less

Gas Core Reactor Numerical Simulation Using a Coupled MHD-MCNP Model

NASA Technical Reports Server (NTRS)

Kazeminezhad, F.; Anghaie, S.

2008-01-01

Analysis is provided in this report of using two head-on magnetohydrodynamic (MHD) shocks to achieve supercritical nuclear fission in an axially elongated cylinder filled with UF4 gas as an energy source for deep space missions. The motivation for each aspect of the design is explained and supported by theory and numerical simulations. A subsequent report will provide detail on relevant experimental work to validate the concept. Here the focus is on the theory of and simulations for the proposed gas core reactor conceptual design from the onset of shock generations to the supercritical state achieved when the shocks collide. The MHD model is coupled to a standard nuclear code (MCNP) to observe the neutron flux and fission power attributed to the supercritical state brought about by the shock collisions. Throughout the modeling, realistic parameters are used for the initial ambient gaseous state and currents to ensure a resulting supercritical state upon shock collisions.

Calibration with MCNP of NaI detector for the determination of natural radioactivity levels in the field.

PubMed

Cinelli, Giorgia; Tositti, Laura; Mostacci, Domiziano; Baré, Jonathan

2016-05-01

In view of assessing natural radioactivity with on-site quantitative gamma spectrometry, efficiency calibration of NaI(Tl) detectors is investigated. A calibration based on Monte Carlo simulation of detector response is proposed, to render reliable quantitative analysis practicable in field campaigns. The method is developed with reference to contact geometry, in which measurements are taken placing the NaI(Tl) probe directly against the solid source to be analyzed. The Monte Carlo code used for the simulations was MCNP. Experimental verification of the calibration goodness is obtained by comparison with appropriate standards, as reported. On-site measurements yield a quick quantitative assessment of natural radioactivity levels present ((40)K, (238)U and (232)Th). On-site gamma spectrometry can prove particularly useful insofar as it provides information on materials from which samples cannot be taken. Copyright © 2016 The Authors. Published by Elsevier Ltd.. All rights reserved.

MCNP modelling of vaginal and uterine applicators used in intracavitary brachytherapy and comparison with radiochromic film measurements

NASA Astrophysics Data System (ADS)

Ceccolini, E.; Gerardy, I.; Ródenas, J.; van Dycke, M.; Gallardo, S.; Mostacci, D.

Brachytherapy is an advanced cancer treatment that is minimally invasive, minimising radiation exposure to the surrounding healthy tissues. Microselectron© Nucletron devices with 192Ir source can be used for gynaecological brachytherapy, in patients with vaginal or uterine cancer. Measurements of isodose curves have been performed in a PMMA phantom and compared with Monte Carlo calculations and TPS (Plato software of Nucletron BPS 14.2) evaluation. The isodose measurements have been performed with radiochromic films (Gafchromic EBT©). The dose matrix has been obtained after digitalisation and use of a dose calibration curve obtained with a 6 MV photon beam provided by a medical linear accelerator. A comparison between the calculated and the measured matrix has been performed. The calculated dose matrix is obtained with a simulation using the MCNP5 Monte Carlo code (F4MESH tally).

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

NASA Astrophysics Data System (ADS)

Ardiyati, Tanti; Rozali, Bang; Kasmudin

2018-02-01

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

Los Alamos and Lawrence Livermore National Laboratories Code-to-Code Comparison of Inter Lab Test Problem 1 for Asteroid Impact Hazard Mitigation

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

Weaver, Robert P.; Miller, Paul; Howley, Kirsten

The NNSA Laboratories have entered into an interagency collaboration with the National Aeronautics and Space Administration (NASA) to explore strategies for prevention of Earth impacts by asteroids. Assessment of such strategies relies upon use of sophisticated multi-physics simulation codes. This document describes the task of verifying and cross-validating, between Lawrence Livermore National Laboratory (LLNL) and Los Alamos National Laboratory (LANL), modeling capabilities and methods to be employed as part of the NNSA-NASA collaboration. The approach has been to develop a set of test problems and then to compare and contrast results obtained by use of a suite of codes, includingmore » MCNP, RAGE, Mercury, Ares, and Spheral. This document provides a short description of the codes, an overview of the idealized test problems, and discussion of the results for deflection by kinetic impactors and stand-off nuclear explosions.« less

Validation of an advanced analytical procedure applied to the measurement of environmental radioactivity.

PubMed

Thanh, Tran Thien; Vuong, Le Quang; Ho, Phan Long; Chuong, Huynh Dinh; Nguyen, Vo Hoang; Tao, Chau Van

2018-04-01

In this work, an advanced analytical procedure was applied to calculate radioactivity in spiked water samples in a close geometry gamma spectroscopy. It included MCNP-CP code in order to calculate the coincidence summing correction factor (CSF). The CSF results were validated by a deterministic method using ETNA code for both p-type HPGe detectors. It showed that a good agreement for both codes. Finally, the validity of the developed procedure was confirmed by a proficiency test to calculate the activities of various radionuclides. The results of the radioactivity measurement with both detectors using the advanced analytical procedure were received the ''Accepted'' statuses following the proficiency test. Copyright © 2018 Elsevier Ltd. All rights reserved.

Copper benchmark experiment for the testing of JEFF-3.2 nuclear data for fusion applications

NASA Astrophysics Data System (ADS)

Angelone, M.; Flammini, D.; Loreti, S.; Moro, F.; Pillon, M.; Villar, R.; Klix, A.; Fischer, U.; Kodeli, I.; Perel, R. L.; Pohorecky, W.

2017-09-01

A neutronics benchmark experiment on a pure Copper block (dimensions 60 × 70 × 70 cm3) aimed at testing and validating the recent nuclear data libraries for fusion applications was performed in the frame of the European Fusion Program at the 14 MeV ENEA Frascati Neutron Generator (FNG). Reaction rates, neutron flux spectra and doses were measured using different experimental techniques (e.g. activation foils techniques, NE213 scintillator and thermoluminescent detectors). This paper first summarizes the analyses of the experiment carried-out using the MCNP5 Monte Carlo code and the European JEFF-3.2 library. Large discrepancies between calculation (C) and experiment (E) were found for the reaction rates both in the high and low neutron energy range. The analysis was complemented by sensitivity/uncertainty analyses (S/U) using the deterministic and Monte Carlo SUSD3D and MCSEN codes, respectively. The S/U analyses enabled to identify the cross sections and energy ranges which are mostly affecting the calculated responses. The largest discrepancy among the C/E values was observed for the thermal (capture) reactions indicating severe deficiencies in the 63,65Cu capture and elastic cross sections at lower rather than at high energy. Deterministic and MC codes produced similar results. The 14 MeV copper experiment and its analysis thus calls for a revision of the JEFF-3.2 copper cross section and covariance data evaluation. A new analysis of the experiment was performed with the MCNP5 code using the revised JEFF-3.3-T2 library released by NEA and a new, not yet distributed, revised JEFF-3.2 Cu evaluation produced by KIT. A noticeable improvement of the C/E results was obtained with both new libraries.

A comparison of skyshine computational methods.

PubMed

Hertel, Nolan E; Sweezy, Jeremy E; Shultis, J Kenneth; Warkentin, J Karl; Rose, Zachary J

2005-01-01

A variety of methods employing radiation transport and point-kernel codes have been used to model two skyshine problems. The first problem is a 1 MeV point source of photons on the surface of the earth inside a 2 m tall and 1 m radius silo having black walls. The skyshine radiation downfield from the point source was estimated with and without a 30-cm-thick concrete lid on the silo. The second benchmark problem is to estimate the skyshine radiation downfield from 12 cylindrical canisters emplaced in a low-level radioactive waste trench. The canisters are filled with ion-exchange resin with a representative radionuclide loading, largely 60Co, 134Cs and 137Cs. The solution methods include use of the MCNP code to solve the problem by directly employing variance reduction techniques, the single-scatter point kernel code GGG-GP, the QADMOD-GP point kernel code, the COHORT Monte Carlo code, the NAC International version of the SKYSHINE-III code, the KSU hybrid method and the associated KSU skyshine codes.

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

NASA Astrophysics Data System (ADS)

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

2004-11-01

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

MCNP Output Data Analysis with ROOT (MODAR)

NASA Astrophysics Data System (ADS)

Carasco, C.

2010-06-01

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

Theory and Performance of AIMS for Active Interrogation

NASA Astrophysics Data System (ADS)

Walters, William J.; Royston, Katherine E. K.; Haghighat, Alireza

2014-06-01

A hybrid Monte Carlo and deterministic methodology has been developed for application to active interrogation systems. The methodology consists of four steps: i) determination of neutron flux distribution due to neutron source transport and subcritical multiplication; ii) generation of gamma source distribution from (n, γ) interactions; iii) determination of gamma current at a detector window; iv) detection of gammas by the detector. This paper discusses the theory and results of the first three steps for the case of a cargo container with a sphere of HEU in third-density water. In the first step, a response-function formulation has been developed to calculate the subcritical multiplication and neutron flux distribution. Response coefficients are pre-calculated using the MCNP5 Monte Carlo code. The second step uses the calculated neutron flux distribution and Bugle-96 (n, γ) cross sections to find the resulting gamma source distribution. Finally, in the third step the gamma source distribution is coupled with a pre-calculated adjoint function to determine the gamma flux at a detector window. A code, AIMS (Active Interrogation for Monitoring Special-Nuclear-materials), has been written to output the gamma current for an source-detector assembly scanning across the cargo using the pre-calculated values and takes significantly less time than a reference MCNP5 calculation.

Application of the MCNP5 code to the Modeling of vaginal and intra-uterine applicators used in intracavitary brachytherapy: a first approach

NASA Astrophysics Data System (ADS)

Gerardy, I.; Rodenas, J.; Van Dycke, M.; Gallardo, S.; Tondeur, F.

2008-02-01

Brachytherapy is a radiotherapy treatment where encapsulated radioactive sources are introduced within a patient. Depending on the technique used, such sources can produce high, medium or low local dose rates. The Monte Carlo method is a powerful tool to simulate sources and devices in order to help physicists in treatment planning. In multiple types of gynaecological cancer, intracavitary brachytherapy (HDR Ir-192 source) is used combined with other therapy treatment to give an additional local dose to the tumour. Different types of applicators are used in order to increase the dose imparted to the tumour and to limit the effect on healthy surrounding tissues. The aim of this work is to model both applicator and HDR source in order to evaluate the dose at a reference point as well as the effect of the materials constituting the applicators on the near field dose. The MCNP5 code based on the Monte Carlo method has been used for the simulation. Dose calculations have been performed with *F8 energy deposition tally, taking into account photons and electrons. Results from simulation have been compared with experimental in-phantom dose measurements. Differences between calculations and measurements are lower than 5%.The importance of the source position has been underlined.

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

NASA Astrophysics Data System (ADS)

Dixon, David A.; Hughes, H. Grady

2017-09-01

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

Computer codes developed and under development at Lewis

NASA Technical Reports Server (NTRS)

Chamis, Christos C.

1992-01-01

The objective of this summary is to provide a brief description of: (1) codes developed or under development at LeRC; and (2) the development status of IPACS with some typical early results. The computer codes that have been developed and/or are under development at LeRC are listed in the accompanying charts. This list includes: (1) the code acronym; (2) select physics descriptors; (3) current enhancements; and (4) present (9/91) code status with respect to its availability and documentation. The computer codes list is grouped by related functions such as: (1) composite mechanics; (2) composite structures; (3) integrated and 3-D analysis; (4) structural tailoring; and (5) probabilistic structural analysis. These codes provide a broad computational simulation infrastructure (technology base-readiness) for assessing the structural integrity/durability/reliability of propulsion systems. These codes serve two other very important functions: they provide an effective means of technology transfer; and they constitute a depository of corporate memory.

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

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

Taneja, S; Bartol, L; Culberson, W

2015-06-15

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

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

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

Mashnik, Stepan Georgievich

2015-08-04

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

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

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

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

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

MUFFSgenMC: An Open Source MUon Flexible Framework for Spectral GENeration for Monte Carlo Applications

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

Chatzidakis, Stylianos; Greulich, Christopher

A cosmic ray Muon Flexible Framework for Spectral GENeration for Monte Carlo Applications (MUFFSgenMC) has been developed to support state-of-the-art cosmic ray muon tomographic applications. The flexible framework allows for easy and fast creation of source terms for popular Monte Carlo applications like GEANT4 and MCNP. This code framework simplifies the process of simulations used for cosmic ray muon tomography.

COMPTEL neutron response at 17 MeV

NASA Technical Reports Server (NTRS)

Oneill, Terrence J.; Ait-Ouamer, Farid; Morris, Joann; Tumer, O. Tumay; White, R. Stephen; Zych, Allen D.

1992-01-01

The Compton imaging telescope (COMPTEL) instrument of the Gamma Ray Observatory was exposed to 17 MeV d,t neutrons prior to launch. These data were analyzed and compared with Monte Carlo calculations using the MCNP(LANL) code. Energy and angular resolutions are compared and absolute efficiencies are calculated at 0 and 30 degrees incident angle. The COMPTEL neutron responses at 17 MeV and higher energies are needed to understand solar flare neutron data.

Distributed multitasking ITS with PVM

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

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

1995-12-31

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

Coupled Monte Carlo neutronics and thermal hydraulics for power reactors

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

Bernnat, W.; Buck, M.; Mattes, M.

The availability of high performance computing resources enables more and more the use of detailed Monte Carlo models even for full core power reactors. The detailed structure of the core can be described by lattices, modeled by so-called repeated structures e.g. in Monte Carlo codes such as MCNP5 or MCNPX. For cores with mainly uniform material compositions, fuel and moderator temperatures, there is no problem in constructing core models. However, when the material composition and the temperatures vary strongly a huge number of different material cells must be described which complicate the input and in many cases exceed code ormore » memory limits. The second problem arises with the preparation of corresponding temperature dependent cross sections and thermal scattering laws. Only if these problems can be solved, a realistic coupling of Monte Carlo neutronics with an appropriate thermal-hydraulics model is possible. In this paper a method for the treatment of detailed material and temperature distributions in MCNP5 is described based on user-specified internal functions which assign distinct elements of the core cells to material specifications (e.g. water density) and temperatures from a thermal-hydraulics code. The core grid itself can be described with a uniform material specification. The temperature dependency of cross sections and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. Applications will be shown for the stationary part of the Purdue PWR benchmark using ATHLET for thermal- hydraulics and for a generic Modular High Temperature reactor using THERMIX for thermal- hydraulics. (authors)« less

Benchmarking the MCNP Monte Carlo code with a photon skyshine experiment

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

Olsher, R.H.; Hsu, Hsiao Hua; Harvey, W.F.

1993-07-01

The MCNP Monte Carlo transport code is used by the Los Alamos National Laboratory Health and Safety Division for a broad spectrum of radiation shielding calculations. One such application involves the determination of skyshine dose for a variety of photon sources. To verify the accuracy of the code, it was benchmarked with the Kansas State Univ. (KSU) photon skyshine experiment of 1977. The KSU experiment for the unshielded source geometry was simulated in great detail to include the contribution of groundshine, in-silo photon scatter, and the effect of spectral degradation in the source capsule. The standard deviation of the KSUmore » experimental data was stated to be 7%, while the statistical uncertainty of the simulation was kept at or under 1%. The results of the simulation agreed closely with the experimental data, generally to within 6%. At distances of under 100 m from the silo, the modeling of the in-silo scatter was crucial to achieving close agreement with the experiment. Specifically, scatter off the top layer of the source cask accounted for [approximately]12% of the dose at 50 m. At distance >300m, using the [sup 60]Co line spectrum led to a dose overresponse as great as 19% at 700 m. It was necessary to use the actual source spectrum, which includes a Compton tail from photon collisions in the source capsule, to achieve close agreement with experimental data. These results highlight the importance of using Monte Carlo transport techniques to account for the nonideal features of even simple experiments''.« less

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

PubMed Central

Thanh, Minh‐Tri Ho; Munro, John J.

2015-01-01

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

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

PubMed

Ajaj, F A; Ghassal, N M

2003-09-01

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

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

PubMed

Winkler, Alexander; Koivunoro, Hanna; Savolainen, Sauli

2017-06-01

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

Development of low level 226Ra analysis for live fish using gamma-ray spectrometry

NASA Astrophysics Data System (ADS)

Chandani, Z.; Prestwich, W. V.; Byun, S. H.

2017-06-01

A low level 226Ra analysis method for live fish was developed using a 4π NaI(Tl) gamma-ray spectrometer. In order to find out the best algorithm for accomplishing the lowest detection limit, the gamma-ray spectrum from a 226Ra point was collected and nine different methods were attempted for spectral analysis. The lowest detection limit of 0.99 Bq for an hour counting occurred when the spectrum was integrated in the energy region of 50-2520 keV. To extend 226Ra analysis to live fish, a Monte Carlo simulation model with a cylindrical fish in a water container was built using the MCNP code. From simulation results, the spatial distribution of the efficiency and the efficiency correction factor for the live fish model were determined. The MCNP model will be able to be conveniently modified when a different fish or container geometry is employed as fish grow up in real experiments.

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

PubMed

Khabaz, Rahim

2012-12-01

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

Multi-group Fokker-Planck proton transport in MCNP{trademark}

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

Adams, K.J.

1997-11-01

MCNP has been enhanced to perform proton transport using a multigroup Fokker Planck (MGFP) algorithm with primary emphasis on proton radiography simulations. The new method solves the Fokker Planck approximation to the Boltzmann transport equation for the small angle multiple scattering portion of proton transport. Energy loss is accounted for by applying a group averaged stopping power over each transport step. Large angle scatter and non-inelastic events are treated as extinction. Comparisons with the more rigorous LAHET code show agreement to a few per cent for the total transmitted currents. The angular distributions through copper and low Z compounds showmore » good agreement between LAHET and MGFP with the MGFP method being slightly less forward peaked and without the large angle tails apparent in the LAHET simulation. Suitability of this method for proton radiography simulations is shown for a simple problem of a hole in a copper slab. LAHET and MGFP calculations of position, angle and energy through more complex objects are presented.« less

MCNP-based computational model for the Leksell gamma knife.

PubMed

Trnka, Jiri; Novotny, Josef; Kluson, Jaroslav

2007-01-01

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

Development of the 3DHZETRN code for space radiation protection

NASA Astrophysics Data System (ADS)

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

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

Assay of the Martian Regolith with Neutrons

NASA Technical Reports Server (NTRS)

Drake, Darrell M.

1997-01-01

The purpose of the research is to combine experiments and Monte Carlo transport of neutrons through volume of soil in an attempt to model neutron leakage from planetary surfaces. Emphasis is given to the change of neutron spectra as a function of water content and location. During the first stage of effort, two experiments were conducted in which leakage of neutrons from a Pu-Be source through about 30 g/cm(exp 2) of soil were measured with several counters. A Monte Carlo code, MCNP, has been used to model many of the 100 individual runs of the experiment. Hydrogen is the element that has the most dramatic effect on the neutron spectrum and its effect on the neutron spectrum is almost the same whether it is in the form of water or polyethylene. In order to simulate various water configurations, sheets of polyethylene have been used between layers of soil as well as water in several concentrations up to 18%. Comparison of experimental results to theoretical predictions made with the MCNP code were disappointing for low concentrations of water. We have made extensive calculations to see if room return could be the cause of the discrepancies. Water concentrations of the 'dry' soil were measured by two different laboratories and differed only by 0.5%. We have made calculations to optimize the next experiment and are investigating other methods of determining the water content of 'dry' soil.

NUCLEAR HEATING IN LIF DOSEMETERS IN A FUSION NEUTRON FIELD, TRIAL OF DIRECT COMPARISON OF EXPERIMENTAL AND SIMULATED RESULTS.

PubMed

Pohorecki, Wladyslaw; Obryk, Barbara

2017-09-29

The results of nuclear heating measured by means of thermoluminescent dosemeters (TLD-LiF) in a Cu block irradiated by 14 MeV neutrons are presented. The integral Cu experiment relevant for verification of copper nuclear data at neutron energies characteristic for fusion facilities was performed in the ENEA FNG Laboratory at Frascati. Five types of TLDs were used: highly photon sensitive LiF:Mg,Cu,P (MCP-N), 7LiF:Mg,Cu,P (MCP-7) and standard, lower sensitivity LiF:Mg,Ti (MTS-N), 7LiF:Mg,Ti (MTS-7) and 6LiF:Mg,Ti (MTS-6). Calibration of the detectors was performed with gamma rays in terms of air-kerma (10 mGy of 137Cs air-kerma). Nuclear heating in the Cu block was also calculated with the use of MCNP transport code Nuclear heating in Cu and air in TLD's positions was calculated as well. The nuclear heating contribution from all simulated by MCNP6 code particles including protons, deuterons, alphas tritons and heavier ions produced by the neutron interactions were calculated. A trial of the direct comparison between experimental results and results of simulation was performed. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

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

PubMed

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

2011-02-07

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

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

NASA Astrophysics Data System (ADS)

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

2018-04-01

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

Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Fast Salt Reactor (MFSR)

NASA Astrophysics Data System (ADS)

Laureau, A.; Rubiolo, P. R.; Heuer, D.; Merle-Lucotte, E.; Brovchenko, M.

2014-06-01

Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated.

Analysis of activation and shutdown contact dose rate for EAST neutral beam port

NASA Astrophysics Data System (ADS)

Chen, Yuqing; Wang, Ji; Zhong, Guoqiang; Li, Jun; Wang, Jinfang; Xie, Yahong; Wu, Bin; Hu, Chundong

2017-12-01

For the safe operation and maintenance of neutral beam injector (NBI), specific activity and shutdown contact dose rate of the sample material SS316 are estimated around the experimental advanced superconducting tokamak (EAST) neutral beam port. Firstly, the neutron emission intensity is calculated by TRANSP code while the neutral beam is co-injected to EAST. Secondly, the neutron activation and shutdown contact dose rates for the neutral beam sample materials SS316 are derived by the Monte Carlo code MCNP and the inventory code FISPACT-2007. The simulations indicate that the primary radioactive nuclides of SS316 are 58Co and 54Mn. The peak contact dose rate is 8.52 × 10-6 Sv/h after EAST shutdown one second. That is under the International Thermonuclear Experimental Reactor (ITER) design values 1 × 10-5 Sv/h.

Validation of updated neutronic calculation models proposed for Atucha-II PHWR. Part I: Benchmark comparisons of WIMS-D5 and DRAGON cell and control rod parameters with MCNP5

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

Mollerach, R.; Leszczynski, F.; Fink, J.

2006-07-01

In 2005 the Argentine Government took the decision to complete the construction of the Atucha-II nuclear power plant, which has been progressing slowly during the last ten years. Atucha-II is a 745 MWe nuclear station moderated and cooled with heavy water, of German (Siemens) design located in Argentina. It has a pressure-vessel design with 451 vertical coolant channels, and the fuel assemblies (FA) are clusters of 37 natural UO{sub 2} rods with an active length of 530 cm. For the reactor physics area, a revision and update calculation methods and models (cell, supercell and reactor) was recently carried out coveringmore » cell, supercell (control rod) and core calculations. As a validation of the new models some benchmark comparisons were done with Monte Carlo calculations with MCNP5. This paper presents comparisons of cell and supercell benchmark problems based on a slightly idealized model of the Atucha-I core obtained with the WIMS-D5 and DRAGON codes with MCNP5 results. The Atucha-I core was selected because it is smaller, similar from a neutronic point of view, and more symmetric than Atucha-II Cell parameters compared include cell k-infinity, relative power levels of the different rings of fuel rods, and some two-group macroscopic cross sections. Supercell comparisons include supercell k-infinity changes due to the control rods (tubes) of steel and hafnium. (authors)« less

Rates for neutron-capture reactions on tungsten isotopes in iron meteorites. [Abstract only

NASA Technical Reports Server (NTRS)

Masarik, J.; Reedy, R. C.

1994-01-01

High-precision W isotopic analyses by Harper and Jacobsen indicate the W-182/W-183 ratio in the Toluca iron meteorite is shifted by -(3.0 +/- 0.9) x 10(exp -4) relative to a terrestrial standard. Possible causes of this shift are neutron-capture reactions on W during Toluca's approximately 600-Ma exposure to cosmic ray particles or radiogenic growth of W-182 from 9-Ma Hf-182 in the silicate portion of the Earth after removal of W to the Earth's core. Calculations for the rates of neutron-capture reactions on W isotopes were done to study the first possibility. The LAHET Code System (LCS) which consists of the Los Alamos High Energy Transport (LAHET) code and the Monte Carlo N-Particle(MCNP) transport code was used to numerically simulate the irradiation of the Toluca iron meteorite by galactic-cosmic-ray (GCR) particles and to calculate the rates of W(n, gamma) reactions. Toluca was modeled as a 3.9-m-radius sphere with the composition of a typical IA iron meteorite. The incident GCR protons and their interactions were modeled with LAHET, which also handled the interactions of neutrons with energies above 20 MeV. The rates for the capture of neutrons by W-182, W-183, and W-186 were calculated using the detailed library of (n, gamma) cross sections in MCNP. For this study of the possible effect of W(n, gamma) reactions on W isotope systematics, we consider the peak rates. The calculated maximum change in the normalized W-182/W-183 ratio due to neutron-capture reactions cannot account for more than 25% of the mass 182 deficit observed in Toluca W.

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

MacFarlane, R. E.

An accurate representation of the scattering of neutrons by the materials used to build cold sources at neutron scattering facilities is important for the initial design and optimization of a cold source, and for the analysis of experimental results obtained using the cold source. In practice, this requires a good representation of the physics of scattering from the material, a method to convert this into observable quantities (such as scattering cross sections), and a method to use the results in a neutron transport code (such as the MCNP Monte Carlo code). At Los Alamos, the authors have been developing thesemore » capabilities over the last ten years. The final set of cold-moderator evaluations, together with evaluations for conventional moderator materials, was released in 1994. These materials have been processed into MCNP data files using the NJOY Nuclear Data Processing System. Over the course of this work, they were able to develop a new module for NJOY called LEAPR based on the LEAP + ADDELT code from the UK as modified by D.J. Picton for cold-moderator calculations. Much of the physics for methane came from Picton`s work. The liquid hydrogen work was originally based on a code using the Young-Koppel approach that went through a number of hands in Europe (including Rolf Neef and Guy Robert). It was generalized and extended for LEAPR, and depends strongly on work by Keinert and Sax of the University of Stuttgart. Thus, their collection of cold-moderator scattering kernels is truly an international effort, and they are glad to be able to return the enhanced evaluations and processing techniques to the international community. In this paper, they give sections on the major cold moderator materials (namely, solid methane, liquid methane, and liquid hydrogen) using each section to introduce the relevant physics for that material and to show typical results.« less

WE-DE-201-05: Evaluation of a Windowless Extrapolation Chamber Design and Monte Carlo Based Corrections for the Calibration of Ophthalmic Applicators

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

Hansen, J; Culberson, W; DeWerd, L

Purpose: To test the validity of a windowless extrapolation chamber used to measure surface dose rate from planar ophthalmic applicators and to compare different Monte Carlo based codes for deriving correction factors. Methods: Dose rate measurements were performed using a windowless, planar extrapolation chamber with a {sup 90}Sr/{sup 90}Y Tracerlab RA-1 ophthalmic applicator previously calibrated at the National Institute of Standards and Technology (NIST). Capacitance measurements were performed to estimate the initial air gap width between the source face and collecting electrode. Current was measured as a function of air gap, and Bragg-Gray cavity theory was used to calculate themore » absorbed dose rate to water. To determine correction factors for backscatter, divergence, and attenuation from the Mylar entrance window found in the NIST extrapolation chamber, both EGSnrc Monte Carlo user code and Monte Carlo N-Particle Transport Code (MCNP) were utilized. Simulation results were compared with experimental current readings from the windowless extrapolation chamber as a function of air gap. Additionally, measured dose rate values were compared with the expected result from the NIST source calibration to test the validity of the windowless chamber design. Results: Better agreement was seen between EGSnrc simulated dose results and experimental current readings at very small air gaps (<100 µm) for the windowless extrapolation chamber, while MCNP results demonstrated divergence at these small gap widths. Three separate dose rate measurements were performed with the RA-1 applicator. The average observed difference from the expected result based on the NIST calibration was −1.88% with a statistical standard deviation of 0.39% (k=1). Conclusion: EGSnrc user code will be used during future work to derive correction factors for extrapolation chamber measurements. Additionally, experiment results suggest that an entrance window is not needed in order for an extrapolation chamber to provide accurate dose rate measurements for a planar ophthalmic applicator.« less

Using the Monte Carlo method for assessing the tissue and organ doses of patients in dental radiography

NASA Astrophysics Data System (ADS)

Makarevich, K. O.; Minenko, V. F.; Verenich, K. A.; Kuten, S. A.

2016-05-01

This work is dedicated to modeling dental radiographic examinations to assess the absorbed doses of patients and effective doses. For simulating X-ray spectra, the TASMIP empirical model is used. Doses are assessed on the basis of the Monte Carlo method by using MCNP code for voxel phantoms of ICRP. The results of the assessment of doses to individual organs and effective doses for different types of dental examinations and features of X-ray tube are presented.

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

NASA Astrophysics Data System (ADS)

Antoni, Rodolphe; Bourgois, Laurent

2017-12-01

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

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

NASA Astrophysics Data System (ADS)

Poškus, A.

2016-09-01

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

LeRC-HT: NASA Lewis Research Center General Multiblock Navier-Stokes Heat Transfer Code Developed

NASA Technical Reports Server (NTRS)

Heidmann, James D.; Gaugler, Raymond E.

1999-01-01

For the last several years, LeRC-HT, a three-dimensional computational fluid dynamics (CFD) computer code for analyzing gas turbine flow and convective heat transfer, has been evolving at the NASA Lewis Research Center. The code is unique in its ability to give a highly detailed representation of the flow field very close to solid surfaces. This is necessary for an accurate representation of fluid heat transfer and viscous shear stresses. The code has been used extensively for both internal cooling passage flows and hot gas path flows--including detailed film cooling calculations, complex tip-clearance gap flows, and heat transfer. In its current form, this code has a multiblock grid capability and has been validated for a number of turbine configurations. The code has been developed and used primarily as a research tool (at least 35 technical papers have been published relative to the code and its application), but it should be useful for detailed design analysis. We now plan to make this code available to selected users for further evaluation.

Developpement et validation d'un outil base sur l'acoustique geometrique pour le diagnostic du bruit de nacelle

NASA Astrophysics Data System (ADS)

Minard, Benoit

De nos jours, la problématique du bruit généré par les avions est devenue un point de développement important dans le domaine de l'aéronautique. C'est ainsi que de nombreuses études sont faites dans le domaine et une première approche consiste à modéliser de façon numérique ce bruit de manière à réduire de façon conséquente les coûts lors de la conception. C'est dans ce contexte qu'un motoriste a demandé à l'université de Sherbrooke, et plus particulièrement au groupe d'acoustique de l'Université de Sherbrooke (GAUS), de développer un outil de calcul de la propagation des ondes acoustiques dans les nacelles mais aussi pour l'étude des effets d'installation. Cet outil de prédiction leur permet de réaliser des études afin d'optimiser les traitements acoustiques (« liners »), la géométrie de ces nacelles pour des études portant sur l'intérieur de la nacelle et des études de positionnement des moteurs et de design pour les effets d'installation. L'objectif de ce projet de maîtrise était donc de poursuivre le travail réalisé par [gousset, 2011] sur l'utilisation d'une méthode de lancer de rayons pour l'étude des effets d'installation des moteurs d'avion. L'amélioration du code, sa rapidité, sa fiabilité et sa généralité étaient les objectifs principaux. Le code peut être utilisé avec des traitements acoustiques de surfaces («liners») et peut prendre en compte le phénomène de la diffraction par les arêtes et enfin peut être utilisé pour réaliser des études dans des environnements complexes tels que les nacelles d'avion. Le code développé fonctionne en 3D et procéde en 3 étapes : (1) Calcul des faisceaux initiaux (division d'une sphère, demi-sphère, maillage des surfaces de la géométrie) (2) Propagation des faisceaux dans l'environnement d'étude : calcul de toutes les caractéristiques des rayons convergents (amplitude, phase, nombre de réflexions, ...) (3) Reconstruction du champ de pression en un ou plusieurs points de l'espace à partir de rayons convergents (sommation des contributions de chaque rayon) : sommation cohérente. Le code (GA3DP) permet de prendre en compte les traitements de surface des parois, la directivité de la source, l'atténuation atmosphérique et la diffraction d'ordre 1. Le code a été validé en utilisant différentes méthodes telles que la méthode des sources-images, la méthode d'analyse modale ou encore la méthode des éléments finis de frontière. Un module Matlab a été créé spécialement pour l'étude des effets d'installation et intégré au code existant chez Pratt & Whitney Canada. Mots-clés : Acoustique géométrique - Ray-Tracing - Lancer de faisceaux - Diffraction - Sommation Cohérente - Niveau de Pression.

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

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

Mashnik, Stepan G

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

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

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

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

Design of the radiation shielding for the time of flight enhanced diagnostics neutron spectrometer at Experimental Advanced Superconducting Tokamak

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

Du, T. F.; Chen, Z. J.; Peng, X. Y.

A radiation shielding has been designed to reduce scattered neutrons and background gamma-rays for the new double-ring Time Of Flight Enhanced Diagnostics (TOFED). The shielding was designed based on simulation with the Monte Carlo code MCNP5. Dedicated model of the EAST tokamak has been developed together with the emission neutron source profile and spectrum; the latter were simulated with the Nubeam and GENESIS codes. Significant reduction of background radiation at the detector can be achieved and this satisfies the requirement of TOFED. The intensities of the scattered and direct neutrons in the line of sight of the TOFED neutron spectrometermore » at EAST are studied for future data interpretation.« less

Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors.

PubMed

Mohammadi, A; Hassanzadeh, M; Gharib, M

2016-02-01

In this study, shielding calculation and criticality safety analysis were carried out for general material testing reactor (MTR) research reactors interim storage and relevant transportation cask. During these processes, three major terms were considered: source term, shielding, and criticality calculations. The Monte Carlo transport code MCNP5 was used for shielding calculation and criticality safety analysis and ORIGEN2.1 code for source term calculation. According to the results obtained, a cylindrical cask with body, top, and bottom thicknesses of 18, 13, and 13 cm, respectively, was accepted as the dual-purpose cask. Furthermore, it is shown that the total dose rates are below the normal transport criteria that meet the standards specified. Copyright © 2015 Elsevier Ltd. All rights reserved.

High-Fidelity Coupled Monte-Carlo/Thermal-Hydraulics Calculations

NASA Astrophysics Data System (ADS)

Ivanov, Aleksandar; Sanchez, Victor; Ivanov, Kostadin

2014-06-01

Monte Carlo methods have been used as reference reactor physics calculation tools worldwide. The advance in computer technology allows the calculation of detailed flux distributions in both space and energy. In most of the cases however, those calculations are done under the assumption of homogeneous material density and temperature distributions. The aim of this work is to develop a consistent methodology for providing realistic three-dimensional thermal-hydraulic distributions by coupling the in-house developed sub-channel code SUBCHANFLOW with the standard Monte-Carlo transport code MCNP. In addition to the innovative technique of on-the fly material definition, a flux-based weight-window technique has been introduced to improve both the magnitude and the distribution of the relative errors. Finally, a coupled code system for the simulation of steady-state reactor physics problems has been developed. Besides the problem of effective feedback data interchange between the codes, the treatment of temperature dependence of the continuous energy nuclear data has been investigated.

Neutron flux measurements on a mock-up of a storage cask for high-level nuclear waste using 2.5 MeV neutrons.

PubMed

Suárez, H Saurí; Becker, F; Klix, A; Pang, B; Döring, T

2018-06-07

To store and dispose spent nuclear fuel, shielding casks are employed to reduce the emitted radiation. To evaluate the exposure of employees handling such casks, Monte Carlo radiation transport codes can be employed. Nevertheless, to assess the reliability of these codes and nuclear data, experimental checks are required. In this study, a neutron generator (NG) producing neutrons of 2.5 MeV was employed to simulate neutrons produced in spent nuclear fuel. Different configurations of shielding layers of steel and polyethylene were positioned between the target of the NG and a NE-213 detector. The results of the measurements of neutron and γ radiation and the corresponding simulations with the code MCNP6 are presented. Details of the experimental set-up as well as neutron and photon flux spectra are provided as reference points for such NG investigations with shielding structures.

Monte Carlo simulations in Nuclear Medicine

NASA Astrophysics Data System (ADS)

Loudos, George K.

2007-11-01

Molecular imaging technologies provide unique abilities to localise signs of disease before symptoms appear, assist in drug testing, optimize and personalize therapy, and assess the efficacy of treatment regimes for different types of cancer. Monte Carlo simulation packages are used as an important tool for the optimal design of detector systems. In addition they have demonstrated potential to improve image quality and acquisition protocols. Many general purpose (MCNP, Geant4, etc) or dedicated codes (SimSET etc) have been developed aiming to provide accurate and fast results. Special emphasis will be given to GATE toolkit. The GATE code currently under development by the OpenGATE collaboration is the most accurate and promising code for performing realistic simulations. The purpose of this article is to introduce the non expert reader to the current status of MC simulations in nuclear medicine and briefly provide examples of current simulated systems, and present future challenges that include simulation of clinical studies and dosimetry applications.

Benchmark test of neutron transport calculations: indium, nickel, gold, europium, and cobalt activation with and without energy moderated fission neutrons by iron simulating the Hiroshima atomic bomb casing.

PubMed

Iwatani, K; Hoshi, M; Shizuma, K; Hiraoka, M; Hayakawa, N; Oka, T; Hasai, H

1994-10-01

A benchmark test of the Monte Carlo neutron and photon transport code system (MCNP) was performed using a bare- and energy-moderated 252Cf fission neutron source which was obtained by transmission through 10-cm-thick iron. An iron plate was used to simulate the effect of the Hiroshima atomic bomb casing. This test includes the activation of indium and nickel for fast neutrons and gold, europium, and cobalt for thermal and epithermal neutrons, which were inserted in the moderators. The latter two activations are also to validate 152Eu and 60Co activity data obtained from the atomic bomb-exposed specimens collected at Hiroshima and Nagasaki, Japan. The neutron moderators used were Lucite and Nylon 6 and the total thickness of each moderator was 60 cm or 65 cm. Measured activity data (reaction yield) of the neutron-irradiated detectors in these moderators decreased to about 1/1,000th or 1/10,000th, which corresponds to about 1,500 m ground distance from the hypocenter in Hiroshima. For all of the indium, nickel, and gold activity data, the measured and calculated values agreed within 25%, and the corresponding values for europium and cobalt were within 40%. From this study, the MCNP code was found to be accurate enough for the bare- and energy-moderated 252Cf neutron activation calculations of these elements using moderators containing hydrogen, carbon, nitrogen, and oxygen.

Dosimetric study of a brachytherapy treatment of esophagus with Brazilian 192Ir sources using an anthropomorphic phantom

NASA Astrophysics Data System (ADS)

Neves, Lucio P.; Santos, William S.; Gorski, Ronan; Perini, Ana P.; Maia, Ana F.; Caldas, Linda V. E.; Orengo, Gilberto

2014-11-01

Several radioisotopes are produced at Instituto de Pesquisas Energéticas e Nucleares for the use in medical treatments, including the activation of 192Ir sources. These sources are suitable for brachytherapy treatments, due to their low or high activity, depending on the concentration of 192Ir, easiness to manufacture, small size, stable daughter products and the possibility of re-utilization. They may be used for the treatment of prostate, cervix, head and neck, skin, breast, gallbladder, uterus, vagina, lung, rectum, and eye cancer treatment. In this work, the use of some 192Ir sources was studied for the treatment of esophagus cancer, especially the dose determination of important structures, such as those on the mediastinum. This was carried out utilizing a FASH anthropomorphic phantom and the MCNP5 Monte Carlo code to transport the radiation through matter. It was possible to observe that the doses at lungs, breast, esophagus, thyroid and heart were the highest, which was expected due to their proximity to the source. Therefore, the data are useful to assess the representative dose specific to brachytherapy treatments on the esophagus for radiation protection purposes. The use of brachytherapy sources was studied for the treatment of esophagus cancer. FASH anthropomorphic phantom and MCNP5 Monte Carlo code were employed. The doses at lungs, breast, esophagus, thyroid and heart were the highest. The data is useful to assess the representative doses of treatments on the esophagus.

Shielding calculations for industrial 5/7.5MeV electron accelerators using the MCNP Monte Carlo Code

NASA Astrophysics Data System (ADS)

Peri, Eyal; Orion, Itzhak

2017-09-01

High energy X-rays from accelerators are used to irradiate food ingredients to prevent growth and development of unwanted biological organisms in food, and by that extend the shelf life of the products. The production of X-rays is done by accelerating 5 MeV electrons and bombarding them into a heavy target (high Z). Since 2004, the FDA has approved using 7.5 MeV energy, providing higher production rates with lower treatments costs. In this study we calculated all the essential data needed for a straightforward concrete shielding design of typical food accelerator rooms. The following evaluation is done using the MCNP Monte Carlo code system: (1) Angular dependence (0-180°) of photon dose rate for 5 MeV and 7.5 MeV electron beams bombarding iron, aluminum, gold, tantalum, and tungsten targets. (2) Angular dependence (0-180°) spectral distribution simulations of bremsstrahlung for gold, tantalum, and tungsten bombarded by 5 MeV and 7.5 MeV electron beams. (3) Concrete attenuation calculations in several photon emission angles for the 5 MeV and 7.5 MeV electron beams bombarding a tantalum target. Based on the simulation, we calculated the expected increase in dose rate for facilities intending to increase the energy from 5 MeV to 7.5 MeV, and the concrete width needed to be added in order to keep the existing dose rate unchanged.

INDIVIDUAL DOSIMETRY IN DISPOSAL REPOSITORY OF HEAT-GENERATING NUCLEAR WASTE.

PubMed

Pang, Bo; Saurí Suárez, Héctor; Becker, Frank

2016-09-01

Certain working scenarios in a disposal facility of heat-generating nuclear waste might lead to an enhanced level of radiation exposure for workers in such facilities. Hence, a realistic estimation of the personal dose during individual working scenarios is desired. In this study, the general-purpose Monte Carlo N-Particle code MCNP6 (Pelowitz, D. B. (ed). MCNP6 user manual LA-CP-13-00634, Rev. 0 (2013)) was applied to simulate a representative radiation field in a disposal facility. A tool to estimate the personal dose was then proposed by taking into account the influence of individual motion sequences during working scenarios. As basis for this approach, a movable whole-body phantom was developed to describe individual body gestures of the workers during motion sequences. In this study, the proposed method was applied to the German concept of geological disposal in rock salt. The feasibility of the proposed approach was demonstrated with an example of working scenario in an emplacement drift of a rock salt mine. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

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

PubMed

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

2018-05-01

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

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

DOE PAGES

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

2017-02-01

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

Commissioning and initial acceptance tests for a commercial convolution dose calculation algorithm for radiotherapy treatment planning in comparison with Monte Carlo simulation and measurement

PubMed Central

Moradi, Farhad; Mahdavi, Seyed Rabi; Mostaar, Ahmad; Motamedi, Mohsen

2012-01-01

In this study the commissioning of a dose calculation algorithm in a currently used treatment planning system was performed and the calculation accuracy of two available methods in the treatment planning system i.e., collapsed cone convolution (CCC) and equivalent tissue air ratio (ETAR) was verified in tissue heterogeneities. For this purpose an inhomogeneous phantom (IMRT thorax phantom) was used and dose curves obtained by the TPS (treatment planning system) were compared with experimental measurements and Monte Carlo (MCNP code) simulation. Dose measurements were performed by using EDR2 radiographic films within the phantom. Dose difference (DD) between experimental results and two calculation methods was obtained. Results indicate maximum difference of 12% in the lung and 3% in the bone tissue of the phantom between two methods and the CCC algorithm shows more accurate depth dose curves in tissue heterogeneities. Simulation results show the accurate dose estimation by MCNP4C in soft tissue region of the phantom and also better results than ETAR method in bone and lung tissues. PMID:22973081

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

PubMed

Ferrari, P; Gualdrini, G

2005-09-21

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

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

NASA Astrophysics Data System (ADS)

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

2017-09-01

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

Device for Detection of Explosives, Nuclear and Other Hazardous Materials in Luggage and Cargo Containers

NASA Astrophysics Data System (ADS)

Kuznetsov, Andrey; Evsenin, Alexey; Gorshkov, Igor; Osetrov, Oleg; Vakhtin, Dmitry

2009-12-01

Device for detection of explosives, radioactive and heavily shielded nuclear materials in luggage and cargo containers based on Nanosecond Neutron Analysis/Associated Particles Technique (NNA/APT) is under construction. Detection module consists of a small neutron generator with built-in position-sensitive detector of associated alpha-particles, and several scintillator-based gamma-ray detectors. Explosives and other hazardous chemicals are detected by analyzing secondary high-energy gamma-rays from reactions of fast neutrons with materials inside a container. The same gamma-ray detectors are used to detect unshielded radioactive and nuclear materials. An array of several neutron detectors is used to detect fast neutrons from induced fission of nuclear materials. Coincidence and timing analysis allows one to discriminate between fission neutrons and scattered probing neutrons. Mathematical modeling by MCNP5 and MCNP-PoliMi codes was used to estimate the sensitivity of the device and its optimal configuration. Comparison of the features of three gamma detector types—based on BGO, NaI and LaBr3 crystals is presented.

Reanalysis of tritium production in a sphere of /sup 6/LiD irradiated by 14-MeV neutrons

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

Fawcett, L.R. Jr.

1985-08-01

Tritium production and activation of radiochemical detector foils in a sphere of /sup 6/LiD irradiated by a central source of 14-MeV neutrons has been reanalyzed. The /sup 6/LiD sphere consisted of 10 solid hemispherical nested shells with ampules of /sup 6/LiH, /sup 7/LiH, and activation foils located 2.2, 5, 7.7, 12.6, 20, and 30 cm from the center. The Los Alamos Monte Carlo Neutron Photon Transport Code (MCNP) was used to calculate neutron transport through the /sup 6/LiD, tritium production in the ampules, and foil activation. The MCNP input model was three-dimensional and employed ENDF/B-V cross sections for transport, tritiummore » production, and (where available) foil activation. The reanalyzed experimentally observed-to-calculated values of tritium production were 1.053 +- 2.1% in /sup 6/LiH and 0.999 +- 2.1% in /sup 7/LiH. The recalculated foil activation observed-to-calculated ratios were not generally improved over those reported in the original analysis.« less

MCNP5 evaluation of photoneutron production from the Alexandria University 15 MV Elekta Precise medical LINAC.

PubMed

Abou-Taleb, W M; Hassan, M H; El Mallah, E A; Kotb, S M

2018-05-01

Photoneutron production, and the dose equivalent, in the head assembly of the 15 MV Elekta Precise medical linac; operating in the faculty of Medicine at Alexandria University were estimated with the MCNP5 code. Photoneutron spectra were calculated in air and inside a water phantom to different depths as a function of the radiation field sizes. The maximum neutron fluence is 3.346×10 -9 n/cm 2 -e for a 30×30 cm 2 field size to 2-4 cm-depth in the phantom. The dose equivalent due to fast neutron increases as the field size increases, being a maximum of 0.912 ± 0.05 mSv/Gy at depth between 2 and 4 cm in the water phantom for 40×40 cm 2 field size. Photoneutron fluence and dose equivalent are larger to 100 cm from the isocenter than to 35 cm from the treatment room wall. Copyright © 2018 Elsevier Ltd. All rights reserved.

Genome Sequence of Novosphingobium lindaniclasticum LE124T, Isolated from a Hexachlorocyclohexane Dumpsite

PubMed Central

Saxena, Anjali; Nayyar, Namita; Sangwan, Naseer; Kumari, Rashmi; Khurana, J. P.

2013-01-01

Novosphingobium lindaniclasticum LE124T is a hexachlorocyclohexane (HCH)-degrading bacterium isolated from a high-dosage-point HCH dumpsite (450 mg HCH/g soil) located in Lucknow, India (27°00′N and 81°09′E). Here, we present the annotated draft genome sequence of strain LE124T, which has an estimated size of 4.86 Mb and is comprised of 4,566 coding sequences. PMID:24029761

Manifold Learning in MR spectroscopy using nonlinear dimensionality reduction and unsupervised clustering.

PubMed

Yang, Guang; Raschke, Felix; Barrick, Thomas R; Howe, Franklyn A

2015-09-01

To investigate whether nonlinear dimensionality reduction improves unsupervised classification of (1) H MRS brain tumor data compared with a linear method. In vivo single-voxel (1) H magnetic resonance spectroscopy (55 patients) and (1) H magnetic resonance spectroscopy imaging (MRSI) (29 patients) data were acquired from histopathologically diagnosed gliomas. Data reduction using Laplacian eigenmaps (LE) or independent component analysis (ICA) was followed by k-means clustering or agglomerative hierarchical clustering (AHC) for unsupervised learning to assess tumor grade and for tissue type segmentation of MRSI data. An accuracy of 93% in classification of glioma grade II and grade IV, with 100% accuracy in distinguishing tumor and normal spectra, was obtained by LE with unsupervised clustering, but not with the combination of k-means and ICA. With (1) H MRSI data, LE provided a more linear distribution of data for cluster analysis and better cluster stability than ICA. LE combined with k-means or AHC provided 91% accuracy for classifying tumor grade and 100% accuracy for identifying normal tissue voxels. Color-coded visualization of normal brain, tumor core, and infiltration regions was achieved with LE combined with AHC. The LE method is promising for unsupervised clustering to separate brain and tumor tissue with automated color-coding for visualization of (1) H MRSI data after cluster analysis. © 2014 Wiley Periodicals, Inc.

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

NASA Astrophysics Data System (ADS)

Gonzales, Matthew Alejandro

The calculation of the thermal neutron Doppler temperature reactivity feedback co-efficient, a key parameter in the design and safe operation of advanced reactors, using first order perturbation theory in continuous energy Monte Carlo codes is challenging as the continuous energy adjoint flux is not readily available. Traditional approaches of obtaining the adjoint flux attempt to invert the random walk process as well as require data corresponding to all temperatures and their respective temperature derivatives within the system in order to accurately calculate the Doppler temperature feedback. A new method has been developed using adjoint-weighted tallies and On-The-Fly (OTF) generated continuous energy cross sections within the Monte Carlo N-Particle (MCNP6) transport code. The adjoint-weighted tallies are generated during the continuous energy k-eigenvalue Monte Carlo calculation. The weighting is based upon the iterated fission probability interpretation of the adjoint flux, which is the steady state population in a critical nuclear reactor caused by a neutron introduced at that point in phase space. The adjoint-weighted tallies are produced in a forward calculation and do not require an inversion of the random walk. The OTF cross section database uses a high order functional expansion between points on a user-defined energy-temperature mesh in which the coefficients with respect to a polynomial fitting in temperature are stored. The coefficients of the fits are generated before run- time and called upon during the simulation to produce cross sections at any given energy and temperature. The polynomial form of the OTF cross sections allows the possibility of obtaining temperature derivatives of the cross sections on-the-fly. The use of Monte Carlo sampling of adjoint-weighted tallies and the capability of computing derivatives of continuous energy cross sections with respect to temperature are used to calculate the Doppler temperature coefficient in a research version of MCNP6. Temperature feedback results from the cross sections themselves, changes in the probability density functions, as well as changes in the density of the materials. The focus of this work is specific to the Doppler temperature feedback which result from Doppler broadening of cross sections as well as changes in the probability density function within the scattering kernel. This method is compared against published results using Mosteller's numerical benchmark to show accurate evaluations of the Doppler temperature coefficient, fuel assembly calculations, and a benchmark solution based on the heavy gas model for free-gas elastic scattering. An infinite medium benchmark for neutron free gas elastic scattering for large scattering ratios and constant absorption cross section has been developed using the heavy gas model. An exact closed form solution for the neutron energy spectrum is obtained in terms of the confluent hypergeometric function and compared against spectra for the free gas scattering model in MCNP6. Results show a quick increase in convergence of the analytic energy spectrum to the MCNP6 code with increasing target size, showing absolute relative differences of less than 5% for neutrons scattering with carbon. The analytic solution has been generalized to accommodate piecewise constant in energy absorption cross section to produce temperature feedback. Results reinforce the constraints in which heavy gas theory may be applied resulting in a significant target size to accommodate increasing cross section structure. The energy dependent piecewise constant cross section heavy gas model was used to produce a benchmark calculation of the Doppler temperature coefficient to show accurate calculations when using the adjoint-weighted method. Results show the Doppler temperature coefficient using adjoint weighting and cross section derivatives accurately obtains the correct solution within statistics as well as reduce computer runtimes by a factor of 50.

Release Notes for Whisper-1.1

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

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

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

ASTROP2-LE: A Mistuned Aeroelastic Analysis System Based on a Two Dimensional Linearized Euler Solver

NASA Technical Reports Server (NTRS)

Reddy, T. S. R.; Srivastava, R.; Mehmed, Oral

2002-01-01

An aeroelastic analysis system for flutter and forced response analysis of turbomachines based on a two-dimensional linearized unsteady Euler solver has been developed. The ASTROP2 code, an aeroelastic stability analysis program for turbomachinery, was used as a basis for this development. The ASTROP2 code uses strip theory to couple a two dimensional aerodynamic model with a three dimensional structural model. The code was modified to include forced response capability. The formulation was also modified to include aeroelastic analysis with mistuning. A linearized unsteady Euler solver, LINFLX2D is added to model the unsteady aerodynamics in ASTROP2. By calculating the unsteady aerodynamic loads using LINFLX2D, it is possible to include the effects of transonic flow on flutter and forced response in the analysis. The stability is inferred from an eigenvalue analysis. The revised code, ASTROP2-LE for ASTROP2 code using Linearized Euler aerodynamics, is validated by comparing the predictions with those obtained using linear unsteady aerodynamic solutions.

The NASA-LeRC wind turbine sound prediction code

NASA Technical Reports Server (NTRS)

Viterna, L. A.

1981-01-01

Since regular operation of the DOE/NASA MOD-1 wind turbine began in October 1979 about 10 nearby households have complained of noise from the machine. Development of the NASA-LeRC with turbine sound prediction code began in May 1980 as part of an effort to understand and reduce the noise generated by MOD-1. Tone sound levels predicted with this code are in generally good agreement with measured data taken in the vicinity MOD-1 wind turbine (less than 2 rotor diameters). Comparison in the far field indicates that propagation effects due to terrain and atmospheric conditions may be amplifying the actual sound levels by about 6 dB. Parametric analysis using the code has shown that the predominant contributions to MOD-1 rotor noise are: (1) the velocity deficit in the wake of the support tower; (2) the high rotor speed; and (3) off column operation.

MCNP HPGe detector benchmark with previously validated Cyltran model.

PubMed

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

2009-05-01

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

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

PubMed

Rezaeian, M; Kamali, J

2017-01-01

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

A Pulsed Sphere Tutorial

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

Cullen, Dermott E.

2017-01-30

Here I attempt to explain what physically happens when we pulse an object with neutrons, specifically what we expect the time dependent behavior of the neutron population to look like. Emphasis is on the time dependent emission of both prompt and delayed neutrons. I also describe how the TART Monte Carlo transport code models this situation; see the appendix for a complete description of the model used by TART. I will also show that, as we expect, MCNP and MERCURY, produce similar results using the same delayed neutron model (again, see the appendix).

Delta-ray Production in MCNP 6.2.0

NASA Astrophysics Data System (ADS)

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

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

Benchmarking comparison and validation of MCNP photon interaction data

NASA Astrophysics Data System (ADS)

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

2017-09-01

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

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

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

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

Integral experiments on thorium assemblies with D-T neutron source

NASA Astrophysics Data System (ADS)

Liu, Rong; Yang, Yiwei; Feng, Song; Zheng, Lei; Lai, Caifeng; Lu, Xinxin; Wang, Mei; Jiang, Li

2017-09-01

To validate nuclear data and code in the neutronics design of a hybrid reactor with thorium, integral experiments in two kinds of benchmark thorium assemblies with a D-T fusion neutron source have been performed. The one kind of 1D assemblies consists of polyethylene and depleted uranium shells. The other kind of 2D assemblies consists of three thorium oxide cylinders. The capture reaction rates, fission reaction rates, and (n, 2n) reaction rates in 232Th in the assemblies are measured by ThO2 foils. The leakage neutron spectra from the ThO2 cylinders are measured by a liquid scintillation detector. The experimental uncertainties in all the results are analyzed. The measured results are compared to the calculated ones with MCNP code and ENDF/B-VII.0 library data.

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

DOE PAGES

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

2016-08-13

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

Investigation of some possible changes in Am-Be neutron source configuration in order to increase the thermal neutron flux using Monte Carlo code

NASA Astrophysics Data System (ADS)

Basiri, H.; Tavakoli-Anbaran, H.

2018-01-01

Am-Be neutrons source is based on (α, n) reaction and generates neutrons in the energy range of 0-11 MeV. Since the thermal neutrons are widely used in different fields, in this work, we investigate how to improve the source configuration in order to increase the thermal flux. These suggested changes include a spherical moderator instead of common cylindrical geometry, a reflector layer and an appropriate materials selection in order to achieve the maximum thermal flux. All calculations were done by using MCNP1 Monte Carlo code. Our final results indicated that a spherical paraffin moderator, a layer of beryllium as a reflector can efficiently increase the thermal neutron flux of Am-Be source.

Status Report on the MCNP 2020 Initiative

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

Brown, Forrest B.; Rising, Michael Evan

2017-10-02

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

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

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

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

2014-11-01

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

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

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

Franco, Manuel

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

Delta-ray Production in MCNP 6.2.0

DOE PAGES

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

2017-10-26

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

Association of Injury History and Incident Injury in Cadet Basic Military Training

PubMed Central

Kucera, Kristen L.; Marshall, Stephen W.; Wolf, Susanne H.; Padua, Darin A.; Cameron, Kenneth L.; Beutler, Anthony I.

2016-01-01

Purpose To determine the association between injury history at enrollment and incident lower extremity (LE) injury during cadet basic training among first-year military cadets. Methods Medically treated LE injuries during cadet basic training documented in the Defense Medical Surveillance System (DMSS) were ascertained in a prospective cohort study of three large U.S. military academies from 2005–2008. Both acute injuries (ICD-9 codes in the 800–900s, including fracture, dislocations, sprains/strains) and injury-related musculoskeletal injuries (ICD-9 codes in the 700s, including inflammation and pain, joint derangement, stress fracture, sprain/strain/rupture, and dislocation) were included. Risk ratios (RR) and 95% confidence intervals (CI) were computed using multivariate log-binomial models stratified by gender. Results During basic training there were 1,438 medically treated acute and 1,719 musculoskeletal-related LE injuries in the 9,811 cadets. The most frequent LE injuries were sprains/strains (73.6% of acute) and inflammation and pain (89.6% of musculoskeletal-related). The overall risk of incident LE injury was 23.2% [95%CI: 22.3%, 24.0%]. Cadets with a previous history of LE injury were at increased risk for incident LE injury. This association was identical in males (RR=1.74 [1.55, 1.94]) and females (RR=1.74 [1.52, 1.99]). In site-specific analyses, strong associations between injury history and incident injury were observed for hip, knee ligament, stress fracture, and ankle sprain. Injury risk was greater (p<0.01) for females (39.1%) compared to males (18.0%). The elevated injury risk in females (RR=2.19 [2.04, 2.36]) was independent of injury history (adjusted RR=2.09 [1.95, 2.24]). Conclusion Injury history upon entry to the military is associated with incidence of LE injuries sustained during cadet basic training. Prevention programs targeted at modifiable factors in cadets with a prior history of LE injury should be considered. PMID:26765627

Un accumulateur echangeur de chaleur hybride pour la gestion simultanee des energies solaire et electrique

NASA Astrophysics Data System (ADS)

Ait Hammou, Zouhair

Cette etude porte sur la conception d'un accumulateur echangeur de chaleur hybride (AECH) pour la gestion simultanee des energies solaire et electrique. Un modele mathematique reposant sur les equations de conservation de la quantite d'energie est expose. Il est developpe pour tester differents materiaux de stockage, entre autres, les materiaux a changement de phase (solide/liquide) et les materiaux de stockage sensible. Un code de calcul est mis en eeuvre sur ordinateur, puis valide a l'aide des resultats analytiques et numeriques de la litterature. En parallele, un prototype experimental a echelle reduite est concu au laboratoire afin de valider le code de calcul. Des simulations sont effectuees pour etudier les effets des parametres de conception et des materiaux de stockage sur le comportement thermique de l'AECH et sur la consommation d'energie electrique. Les resultats des simulations sur quatre mois d'hiver montrent que la paraffine n-octadecane et l'acide caprique sont deux candidats souhaitables pour le stockage d'energie destine au chauffage des habitats. L'utilisation de ces deux materiaux dans l'AECH permet de reduire la consommation d'energie electrique de 32% et d'aplanir le probleme de pointe electrique puisque 90% de l'energie electrique est consommee durant les heures creuses. En plus, en adoptant un tarif preferentiel, le calcul des couts lies a la consommation d'energie electrique montre que le consommateur adoptant ce systeme beneficie d'une reduction de 50% de la facture d'electricite.

Development of Parallel Computing Framework to Enhance Radiation Transport Code Capabilities for Rare Isotope Beam Facility Design

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

Kostin, Mikhail; Mokhov, Nikolai; Niita, Koji

A parallel computing framework has been developed to use with general-purpose radiation transport codes. The framework was implemented as a C++ module that uses MPI for message passing. It is intended to be used with older radiation transport codes implemented in Fortran77, Fortran 90 or C. The module is significantly independent of radiation transport codes it can be used with, and is connected to the codes by means of a number of interface functions. The framework was developed and tested in conjunction with the MARS15 code. It is possible to use it with other codes such as PHITS, FLUKA andmore » MCNP after certain adjustments. Besides the parallel computing functionality, the framework offers a checkpoint facility that allows restarting calculations with a saved checkpoint file. The checkpoint facility can be used in single process calculations as well as in the parallel regime. The framework corrects some of the known problems with the scheduling and load balancing found in the original implementations of the parallel computing functionality in MARS15 and PHITS. The framework can be used efficiently on homogeneous systems and networks of workstations, where the interference from the other users is possible.« less

Distributed multitasking ITS with PVM

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

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

1995-02-01

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

Modeling Sodium Iodide Detector Response Using Parametric Equations

DTIC Science & Technology

2013-03-22

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

Developpement d'une methode de Monte Carlo dependante du temps et application au reacteur de type CANDU-6

NASA Astrophysics Data System (ADS)

Mahjoub, Mehdi

La resolution de l'equation de Boltzmann demeure une etape importante dans la prediction du comportement d'un reacteur nucleaire. Malheureusement, la resolution de cette equation presente toujours un defi pour une geometrie complexe (reacteur) tout comme pour une geometrie simple (cellule). Ainsi, pour predire le comportement d'un reacteur nucleaire,un schema de calcul a deux etapes est necessaire. La premiere etape consiste a obtenir les parametres nucleaires d'une cellule du reacteur apres une etape d'homogeneisation et condensation. La deuxieme etape consiste en un calcul de diffusion pour tout le reacteur en utilisant les resultats de la premiere etape tout en simplifiant la geometrie du reacteur a un ensemble de cellules homogenes le tout entoure de reflecteur. Lors des transitoires (accident), ces deux etapes sont insuffisantes pour pouvoir predire le comportement du reacteur. Comme la resolution de l'equation de Boltzmann dans sa forme dependante du temps presente toujours un defi de taille pour tous types de geometries,un autre schema de calcul est necessaire. Afin de contourner cette difficulte, l'hypothese adiabatique est utilisee. Elle se concretise en un schema de calcul a quatre etapes. La premiere et deuxieme etapes demeurent les memes pour des conditions nominales du reacteur. La troisieme etape se resume a obtenir les nouvelles proprietes nucleaires de la cellule a la suite de la perturbation pour les utiliser, au niveau de la quatrieme etape, dans un nouveau calcul de reacteur et obtenir l'effet de la perturbation sur le reacteur. Ce projet vise a verifier cette hypothese. Ainsi, un nouveau schema de calcul a ete defini. La premiere etape de ce projet a ete de creer un nouveau logiciel capable de resoudre l'equation de Boltzmann dependante du temps par la methode stochastique Monte Carlo dans le but d'obtenir des sections efficaces qui evoluent dans le temps. Ce code a ete utilise pour simuler un accident LOCA dans un reacteur nucleaire de type CANDU-6. Les sections efficaces dependantes du temps ont ete par la suite utilisees dans un calcul de diffusion espace-temps pour un reacteur CANDU-6 subissant un accident de type LOCA affectant la moitie du coeur afin d'observer son comportement durant toutes les phases de la perturbation. Dans la phase de developpement, nous avons choisi de demarrer avec le code OpenMC, developpe au MIT,comme plateforme initiale de developpement. L'introduction et le traitement des neutrons retardes durant la simulation ont presente un grand defi a surmonter. Il est important de noter que le code developpe utilisant la methode Monte Carlo peut etre utilise a grande echelle pour la simulation de tous les types des reacteurs nucleaires si les supports informatiques sont disponibles.

Control of the Low-energy X-rays by Using MCNP5 and Numerical Analysis for a New Concept Intra-oral X-ray Imaging System

NASA Astrophysics Data System (ADS)

Huh, Jangyong; Ji, Yunseo; Lee, Rena

2018-05-01

An X-ray control algorithm to modulate the X-ray intensity distribution over the FOV (field of view) has been developed by using numerical analysis and MCNP5, a particle transport simulation code on the basis of the Monte Carlo method. X-rays, which are widely used in medical diagnostic imaging, should be controlled in order to maximize the performance of the X-ray imaging system. However, transporting X-rays, like a liquid or a gas is conveyed through a physical form such as pipes, is not possible. In the present study, an X-ray control algorithm and technique to uniformize the Xray intensity projected on the image sensor were developed using a flattening filter and a collimator in order to alleviate the anisotropy of the distribution of X-rays due to intrinsic features of the X-ray generator. The proposed method, which is combined with MCNP5 modeling and numerical analysis, aimed to optimize a flattening filter and a collimator for a uniform distribution of X-rays. Their size and shape were estimated from the method. The simulation and the experimental results both showed that the method yielded an intensity distribution over an X-ray field of 6×4 cm2 at SID (source to image-receptor distance) of 5 cm with a uniformity of more than 90% when the flattening filter and the collimator were mounted on the system. The proposed algorithm and technique are not only confined to flattening filter development but can also be applied for other X-ray related research and development efforts.

Quantitative basis for component factors of gas flow proportional counting efficiencies

NASA Astrophysics Data System (ADS)

Nichols, Michael C.

This dissertation investigates the counting efficiency calibration of a gas flow proportional counter with beta-particle emitters in order to (1) determine by measurements and simulation the values of the component factors of beta-particle counting efficiency for a proportional counter, (2) compare the simulation results and measured counting efficiencies, and (3) determine the uncertainty of the simulation and measurements. Monte Carlo simulation results by the MCNP5 code were compared with measured counting efficiencies as a function of sample thickness for 14C, 89Sr, 90Sr, and 90Y. The Monte Carlo model simulated strontium carbonate with areal thicknesses from 0.1 to 35 mg cm-2. The samples were precipitated as strontium carbonate with areal thicknesses from 3 to 33 mg cm-2 , mounted on membrane filters, and counted on a low background gas flow proportional counter. The estimated fractional standard deviation was 2--4% (except 6% for 14C) for efficiency measurements of the radionuclides. The Monte Carlo simulations have uncertainties estimated to be 5 to 6 percent for carbon-14 and 2.4 percent for strontium-89, strontium-90, and yttrium-90. The curves of simulated counting efficiency vs. sample areal thickness agreed within 3% of the curves of best fit drawn through the 25--49 measured points for each of the four radionuclides. Contributions from this research include development of uncertainty budgets for the analytical processes; evaluation of alternative methods for determining chemical yield critical to the measurement process; correcting a bias found in the MCNP normalization of beta spectra histogram; clarifying the interpretation of the commonly used ICRU beta-particle spectra for use by MCNP; and evaluation of instrument parameters as applied to the simulation model to obtain estimates of the counting efficiency from simulated pulse height tallies.

Calculated criticality for sup 235 U/graphite systems using the VIM Monte Carlo code

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

Collins, P.J.; Grasseschi, G.L.; Olsen, D.N.

1992-01-01

Calculations for highly enriched uranium and graphite systems gained renewed interest recently for the new production modular high-temperature gas-cooled reactor (MHTGR). Experiments to validate the physics calculations for these systems are being prepared for the Transient Reactor Test Facility (TREAT) reactor at Argonne National Laboratory (ANL-West) and in the Compact Nuclear Power Source facility at Los Alamos National Laboratory. The continuous-energy Monte Carlo code VIM, or equivalently the MCNP code, can utilize fully detailed models of the MHTGR and serve as benchmarks for the approximate multigroup methods necessary in full reactor calculations. Validation of these codes and their associated nuclearmore » data did not exist for highly enriched {sup 235}U/graphite systems. Experimental data, used in development of more approximate methods, dates back to the 1960s. The authors have selected two independent sets of experiments for calculation with the VIM code. The carbon-to-uranium (C/U) ratios encompass the range of 2,000, representative of the new production MHTGR, to the ratio of 10,000 in the fuel of TREAT. Calculations used the ENDF/B-V data.« less

MuSim, a Graphical User Interface for Multiple Simulation Programs

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

Roberts, Thomas; Cummings, Mary Anne; Johnson, Rolland

2016-06-01

MuSim is a new user-friendly program designed to interface to many different particle simulation codes, regardless of their data formats or geometry descriptions. It presents the user with a compelling graphical user interface that includes a flexible 3-D view of the simulated world plus powerful editing and drag-and-drop capabilities. All aspects of the design can be parametrized so that parameter scans and optimizations are easy. It is simple to create plots and display events in the 3-D viewer (with a slider to vary the transparency of solids), allowing for an effortless comparison of different simulation codes. Simulation codes: G4beamline, MAD-X,more » and MCNP; more coming. Many accelerator design tools and beam optics codes were written long ago, with primitive user interfaces by today's standards. MuSim is specifically designed to make it easy to interface to such codes, providing a common user experience for all, and permitting the construction and exploration of models with very little overhead. For today's technology-driven students, graphical interfaces meet their expectations far better than text-based tools, and education in accelerator physics is one of our primary goals.« less

Simulation of Charge Collection in Diamond Detectors Irradiated with Deuteron-Triton Neutron Sources

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

Milocco, Alberto; Trkov, Andrej; Pillon, Mario

2011-12-13

Diamond-based neutron spectrometers exhibit outstanding properties such as radiation hardness, low sensitivity to gamma rays, fast response and high-energy resolution. They represent a very promising application of diamonds for plasma diagnostics in fusion devices. The measured pulse height spectrum is obtained from the collection of helium and beryllium ions produced by the reactions on {sup 12}C. An original code is developed to simulate the production and the transport of charged particles inside the diamond detector. The ion transport methodology is based on the well-known TRIM code. The reactions of interest are triggered using the ENDF/B-VII.0 nuclear data for the neutronmore » interactions on carbon. The model is implemented in the TALLYX subroutine of the MCNP5 and MCNPX codes. Measurements with diamond detectors in a {approx}14 MeV neutron field have been performed at the FNG (Rome, Italy) and IRMM (Geel, Belgium) facilities. The comparison of the experimental data with the simulations validates the proposed model.« less

MCNP6 unstructured mesh application to estimate the photoneutron distribution and induced activity inside a linac bunker

NASA Astrophysics Data System (ADS)

Juste, B.; Morató, S.; Miró, R.; Verdú, G.; Díez, S.

2017-08-01

Unwanted neutrons in radiation therapy treatments are typically generated by photonuclear reactions. High-energy beams emitted by medical Linear Accelerators (LinAcs) interact with high atomic number materials situated in the accelerator head and release neutrons. Since neutrons have a high relative biological effectiveness, even low neutron doses may imply significant exposure of patients. It is also important to study radioactivity induced by these photoneutrons when interacting with the different materials and components of the treatment head facility and the shielding room walls, since persons not present during irradiation (e.g. medical staff) may be exposed to them even when the accelerator is not operating. These problems are studied in this work in order to contribute to challenge the radiation protection in these treatment locations. The work has been performed by simulation using the latest state of the art of Monte-Carlo computer code MCNP6. To that, a detailed model of particles transport inside the bunker and treatment head has been carried out using a meshed geometry model. The LinAc studied is an Elekta Precise accelerator with a treatment photon energy of 15 MeV used at the Hospital Clinic Universitari de Valencia, Spain.

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

DTIC Science & Technology

2010-03-01

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

Comparison of CdZnTe neutron detector models using MCNP6 and Geant4

NASA Astrophysics Data System (ADS)

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

2018-01-01

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

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

PubMed

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

2004-01-01

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

Impact of ASTM Standard E722 update on radiation damage metrics

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

DePriest, Kendall Russell

2014-06-01

The impact of recent changes to the ASTM Standard E722 is investigated. The methodological changes in the production of the displacement kerma factors for silicon has significant impact for some energy regions of the 1-MeV(Si) equivalent fluence response function. When evaluating the integral over all neutrons energies in various spectra important to the SNL electronics testing community, the change in the response results in an increase in the total 1-MeV(Si) equivalent fluence of 2 7%. Response functions have been produced and are available for users of both the NuGET and MCNP codes.

Calibration of a portable HPGe detector using MCNP code for the determination of 137Cs in soils.

PubMed

Gutiérrez-Villanueva, J L; Martín-Martín, A; Peña, V; Iniguez, M P; de Celis, B; de la Fuente, R

2008-10-01

In situ gamma spectrometry provides a fast method to determine (137)Cs inventories in soils. To improve the accuracy of the estimates, one can use not only the information on the photopeak count rates but also on the peak to forward-scatter ratios. Before applying this procedure to field measurements, a calibration including several experimental simulations must be carried out in the laboratory. In this paper it is shown that Monte Carlo methods are a valuable tool to minimize the number of experimental measurements needed for the calibration.

Modelisation and distribution of neutron flux in radium-beryllium source (226Ra-Be)

NASA Astrophysics Data System (ADS)

Didi, Abdessamad; Dadouch, Ahmed; Jai, Otman

2017-09-01

Using the Monte Carlo N-Particle code (MCNP-6), to analyze the thermal, epithermal and fast neutron fluxes, of 3 millicuries of radium-beryllium, for determine the qualitative and quantitative of many materials, using method of neutron activation analysis. Radium-beryllium source of neutron is established to practical work and research in nuclear field. The main objective of this work was to enable us harness the profile flux of radium-beryllium irradiation, this theoretical study permits to discuss the design of the optimal irradiation and performance for increased the facility research and education of nuclear physics.

MCNP output data analysis with ROOT (MODAR)

NASA Astrophysics Data System (ADS)

Carasco, C.

2010-12-01

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

Inclusion of Scatter in HADES: Final Report

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

Aufderheide, M B

Covert nuclear attack is one of the foremost threats facing the United States and is a primary focus of the War on Terror. The Domestic Nuclear Detection Office (DNDO), within the Department of Homeland Security (DHS), is chartered to develop, and improve domestic systems to detect and interdict smuggling for the illicit use of a nuclear explosive device, fissile material or radiologica1 material. The CAARS (Cargo Advanced Automated Radiography System) program is a major part of the DHS effort to enhance US security by harnessing cutting-edge technologies to detect radiological and nuclear threats at points of entry to the Unitedmore » States. DNDO has selected vendors to develop complete radiographic systems. It is crucial that the initial design and testing concepts for the systems be validated and compared prior to the substantial efforts to build and deploy prototypes and subsequent large-scale production. An important aspect of these systems is the scatter which interferes with imaging. Monte Carlo codes, such as MCNP (X-5 Monte Carlo Team, 2005 Revision) allow scatter to be calculatied, but these calculations are very time consuming. It would be useful to have a fast scatter estimation algorithm in a fast ray tracing code. We have been extending the HADES ray-tracing radiographic simulation code to model vendor systems in a flexible and quick fashion and to use this tool to study a variety of questions involving system performance and the comparative value of surrogates. To enable this work, HADES has been linked to the BRL-CAD library (BRL-CAD Open Source Project, 2010), in order to enable the inclusion of complex CAD geometries in simulations, scanner geometries have been implemented in HADES, and the novel detector responses have been included in HADES. A major extension of HADES which has been required by this effort is the inclusion of scatter in these radiographic simulations. Ray tracing codes generally do not easily allow the inclusion of scatter, because these codes define a source and a grid of detector pixels and only compute the attenuation along rays between these points. Scatter is an extremely complex set of processes which can involve rays which change directions many times between the source and detector. Scatter from outside the field of view of the imaging system, as well as within the field of view, can have an important role in image formation. In this report, we will describe how we implemented a treatment of scatter in HADES. We begin with a discussion of how we define scatter in Section 2, followed by a description of how single Compton scatter is now included in HADES in Section 3. In Section 4 we report a set of verification tests against MCNP and tests of how the technique scales with image size, number of scatters allowed and number of processors used in the calculations. In Section 5, we describe how we plan to extend this approach to other forms of scatter and conclude in Section 6. It should be emphasized that the purpose of this report is to show that a form of scatter has been implemented in HADES and has been verified against MCNP. Validation, the process of comparing simulation and experiment, is a future task.« less

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

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

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

Translation into French of: “Changes to publication requirements made at the XVIII International Botanical Congress in Melbourne – what does e-publication mean for you?”. Translated by Christian Feuillet and Valéry Malécot Changements des conditions requises pour la publication faits au XVIII e Congrès International de Botanique à Melbourne – qu’est-ce que la publication électronique représente pour vous?

PubMed Central

Knapp, Sandra; McNeill, John; Turland, Nicholas J.

2011-01-01

Résumé Les changements au Code International de Nomenclature Botanique sont décidés tous les 6 ans aux Sections de Nomenclature associées aux Congrès Internationaux de Botanique (CIB). Le XVIIIe CIB se tenait à Melbourne, Australie; la Section de Nomenclature s’est réunie les 18-22 juillet 2011 et ses décisions ont été acceptées par le Congrès en session plénière le 30 juillet. Suite à cette réunion, plusieurs modifications importantes ont été apportées au Code et vont affecter la publication de nouveaux noms. Deux de ces changements prendront effet le 1er janvier 2012, quelques mois avant que le Code de Melbourne soit publié. Les documents électroniques publiés en ligne en ‘Portable Document Format’ (PDF) avec un ‘International Standard Serial Number’ (ISSN) ou un ‘International Standard Book Number’ (ISBN) constitueront une publication effective, et l’exigence d’une description ou d’une diagnose en latin pour les noms des nouveaux taxa sera changée en l’exigence d’une description ou d’une diagnose en latin ou en anglais. De plus, à partir du 1er janvier 2013, les noms nouveaux des organismes traités comme champignons devront, pour que la publication soit valide, inclure dans le protologue (tous ce qui est associé au nom au moment de la publication valide) la citation d’un identifiant (‘identifier’) fourni par un dépôt reconnu (tel MycoBank). Une ébauche des nouveaux articles concernant la publication électronique est fournie et des conseils de bon usage sont esquissés. Pour encourager la diffusion des changements adoptés au Code International de Nomenclature pour les algues, les champignons et les plantes, cet article sera publié dans BMC Evolutionary Biology, Botanical Journal of the Linnean Society, Brittonia, Cladistics, MycoKeys, Mycotaxon, New Phytologist, North American Fungi, Novon, Opuscula Philolichenum, PhytoKeys, Phytoneuron, Phytotaxa, Plant Diversity and Resources, Systematic Botany et Taxon. PMID:22287925

Comportement instationnaire des thermoéléments à effet Peltier multi-étages

NASA Astrophysics Data System (ADS)

Monchoux, F.; Zély, D.; Cordier, A.

1995-01-01

The analysis of thermoelectric phenomena is possible based on non-equilibrium thermodynamics. Integration of the thermal balance equation leads to an analytical solution for the non-stationnary behaviour. The influence to Thomson effect is commmented. The model, introduced in the complete software TRNSYS, permits the modelling of complex systems including such elements in their thermal regulation. La thermodynamique des processus irréversibles permet l'analyse des phénomènes thermoélectriques. Par intégration de l'équation de bilan thermique, on a obtenu une solution analytique pour le régime non stationnaire donnant la température en tous points et le flux absorbé. On a analysé l'influence de l'effet Thomson. Le modèle a été inclu dans le code plus général TRNSYS qui permet la modélisation de systèmes complexes.

MCNP6 Status

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

Goorley, John T.

2012-06-25

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

User Manual for Whisper-1.1

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

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

2017-01-26

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

Analysis of sewage sludge using an experimental prompt gamma neutron activation analysis (pgnaa) set-up with an am-be source

NASA Astrophysics Data System (ADS)

Idiri, Z.; Redjem, F.; Beloudah, N.

2016-09-01

An experimental PGNAA set-up using a 1 Ci Am-Be source has been developed and used for analysis of bulk sewage sludge samples issued from a wastewater treatment plant situated in an industrial area of Algiers. The sample dimensions were optimized using thermal neutron flux calculations carried out with the MCNP5 Monte Carlo Code. A methodology is then proposed to perform quantitative analysis using the absolute method. For this, average thermal neutron flux inside the sludge samples is deduced using average thermal neutron flux in reference water samples and thermal flux measurements with the aid of a 3He neutron detector. The average absolute gamma detection efficiency is determined using the prompt gammas emitted by chlorine dissolved in a water sample. The gamma detection efficiency is normalized for sludge samples using gamma attenuation factors calculated with the MCNP5 code for water and sludge. Wet and dehydrated sludge samples were analyzed. Nutritive elements (Ca, N, P, K) and heavy metals elements like Cr and Mn were determined. For some elements, the PGNAA values were compared to those obtained using Atomic Absorption Spectroscopy (AAS) and Inductively Coupled Plasma (ICP) methods. Good agreement is observed between the different values. Heavy element concentrations are very high compared to normal values; this is related to the fact that the wastewater treatment plant is treating not only domestic but also industrial wastewater that is probably rejected by industries without removal of pollutant elements. The detection limits for almost all elements of interest are sufficiently low for the method to be well suited for such analysis.

Monte Carlo modelling of TRIGA research reactor

NASA Astrophysics Data System (ADS)

El Bakkari, B.; Nacir, B.; El Bardouni, T.; El Younoussi, C.; Merroun, O.; Htet, A.; Boulaich, Y.; Zoubair, M.; Boukhal, H.; Chakir, M.

2010-10-01

The Moroccan 2 MW TRIGA MARK II research reactor at Centre des Etudes Nucléaires de la Maâmora (CENM) achieved initial criticality on May 2, 2007. 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 neutronic analysis of the 2-MW TRIGA MARK II research reactor at CENM and validation of the results by comparisons with the experimental, operational, and available final safety analysis report (FSAR) values. The study was prepared in collaboration between the Laboratory of Radiation and Nuclear Systems (ERSN-LMR) from Faculty of Sciences of Tetuan (Morocco) and CENM. The 3-D continuous energy Monte Carlo code MCNP (version 5) was used to develop a versatile and accurate full model of the TRIGA core. The model represents in detailed all components of the core with literally no physical approximation. Continuous energy cross-section data from the more recent nuclear data evaluations (ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1, and JENDL-3.3) as well as S( α, β) thermal neutron scattering functions distributed with the MCNP code were used. The cross-section libraries were generated by using the NJOY99 system updated to its more recent patch file "up259". The consistency and accuracy of both the Monte Carlo simulation and neutron transport physics were established by benchmarking the TRIGA experiments. Core excess reactivity, total and integral control rods worth as well as power peaking factors were used in the validation process. Results of calculations are analysed and discussed.

Calculation and comparison of xenon and samarium reactivities of the HEU, LEU core in the low power research reactor.

PubMed

Dawahra, S; Khattab, K; Saba, G

2015-07-01

Comparative studies for the conversion of the fuel from HEU to LEU in the Miniature Neutron Source Reactor (MNSR) have been performed using the MCNP4C and GETERA codes. The precise calculations of (135)Xe and (149)Sm concentrations and reactivities were carried out and compared during the MNSR operation time and after shutdown for the existing HEU fuel (UAl4-Al, 90% enriched) and the potential LEU fuels (U3Si2-Al, U3Si-Al, U9Mo-Al, 19.75% enriched and UO2, 12.6% enriched) in this paper using the MCNP4C and GETERA codes. It was found that the (135)Xe and (149)Sm reactivities did not reach their equilibrium reactivities during the daily operating time of the reactor. The (149)Sm reactivities could be neglected compared to (135)Xe reactivities during the reactor operating time and after shutdown. The calculations for the UAl4-Al produced the highest (135)Xe reactivity in all the studied fuel group during the reactor operation (0.39 mk) and after the reactor shutdown (0.735 mk), It followed by U3Si-Al (0.34 mk, 0.653 mk), U3Si2-Al (0.33 mk, 0.634 mk), U9Mo-Al (0.3 mk, 0.568 mk) and UO2 (0.24 mk, 0.448 mk) fuels, respectively. Finally, the results showed that the UO2 was the best candidate for fuel conversion to LEU in the MNSR since it gave the lowest (135)Xe reactivity during the reactor operation and after shutdown. Copyright © 2015 Elsevier Ltd. All rights reserved.

Geant4 Modifications for Accurate Fission Simulations

NASA Astrophysics Data System (ADS)

Tan, Jiawei; Bendahan, Joseph

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

Dose conversion coefficients based on the Chinese mathematical phantom and MCNP code for external photon irradiation.

PubMed

Qiu, Rui; Li, Junli; Zhang, Zhan; Liu, Liye; Bi, Lei; Ren, Li

2009-02-01

A set of conversion coefficients from kerma free-in-air to the organ-absorbed dose are presented for external monoenergetic photon beams from 10 keV to 10 MeV based on the Chinese mathematical phantom, a whole-body mathematical phantom model. The model was developed based on the methods of the Oak Ridge National Laboratory mathematical phantom series and data from the Chinese Reference Man and the Reference Asian Man. This work is carried out to obtain the conversion coefficients based on this model, which represents the characteristics of the Chinese population, as the anatomical parameters of the Chinese are different from those of Caucasians. Monte Carlo simulation with MCNP code is carried out to calculate the organ dose conversion coefficients. Before the calculation, the effects from the physics model and tally type are investigated, considering both the calculation efficiency and precision. In the calculation irradiation conditions include anterior-posterior, posterior-anterior, right lateral, left lateral, rotational and isotropic geometries. Conversion coefficients from this study are compared with those recommended in the Publication 74 of International Commission on Radiological Protection (ICRP74) since both the sets of data are calculated with mathematical phantoms. Overall, consistency between the two sets of data is observed and the difference for more than 60% of the data is below 10%. However, significant deviations are also found, mainly for the superficial organs (up to 65.9%) and bone surface (up to 66%). The big difference of the dose conversion coefficients for the superficial organs at high photon energy could be ascribed to kerma approximation for the data in ICRP74. Both anatomical variations between races and the calculation method contribute to the difference of the data for bone surface.

TREAT Transient Analysis Benchmarking for the HEU Core

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

Kontogeorgakos, D. C.; Connaway, H. M.; Wright, A. E.

2014-05-01

This work was performed to support the feasibility study on the potential conversion of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory from the use of high enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel. The analyses were performed by the GTRI Reactor Conversion staff at the Argonne National Laboratory (ANL). The objective of this study was to benchmark the transient calculations against temperature-limited transients performed in the final operating HEU TREAT core configuration. The MCNP code was used to evaluate steady-state neutronics behavior, and the point kinetics code TREKIN was used tomore » determine core power and energy during transients. The first part of the benchmarking process was to calculate with MCNP all the neutronic parameters required by TREKIN to simulate the transients: the transient rod-bank worth, the prompt neutron generation lifetime, the temperature reactivity feedback as a function of total core energy, and the core-average temperature and peak temperature as a functions of total core energy. The results of these calculations were compared against measurements or against reported values as documented in the available TREAT reports. The heating of the fuel was simulated as an adiabatic process. The reported values were extracted from ANL reports, intra-laboratory memos and experiment logsheets and in some cases it was not clear if the values were based on measurements, on calculations or a combination of both. Therefore, it was decided to use the term “reported” values when referring to such data. The methods and results from the HEU core transient analyses will be used for the potential LEU core configurations to predict the converted (LEU) core’s performance.« less

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

PubMed

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

2011-01-01

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

Correlated prompt fission data in transport simulations

DOE PAGES

Talou, P.; Vogt, R.; Randrup, J.; ...

2018-01-24

Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n -n, n - γ, and γ - γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ raysmore » from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX - PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in nonproliferation, safeguards, nuclear energy, and defense programs. Here, this review provides an overview of the topic, starting from theoretical considerations of the fission process, with a focus on correlated signatures. It then explores the status of experimental correlated fission data and current efforts to address some of the known shortcomings. Numerical simulations employing the FREYA and CGMF codes are compared to experimental data for a wide range of correlated fission quantities. The inclusion of those codes into the MCNP6.2 and MCNPX - PoliMi transport codes is described and discussed in the context of relevant applications. The accuracy of the model predictions and their sensitivity to model assumptions and input parameters are discussed. Lastly, a series of important experimental and theoretical questions that remain unanswered are presented, suggesting a renewed effort to address these shortcomings.« less

Correlated prompt fission data in transport simulations

NASA Astrophysics Data System (ADS)

Talou, P.; Vogt, R.; Randrup, J.; Rising, M. E.; Pozzi, S. A.; Verbeke, J.; Andrews, M. T.; Clarke, S. D.; Jaffke, P.; Jandel, M.; Kawano, T.; Marcath, M. J.; Meierbachtol, K.; Nakae, L.; Rusev, G.; Sood, A.; Stetcu, I.; Walker, C.

2018-01-01

Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n - n, n - γ, and γ - γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ rays from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX - PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in nonproliferation, safeguards, nuclear energy, and defense programs. This review provides an overview of the topic, starting from theoretical considerations of the fission process, with a focus on correlated signatures. It then explores the status of experimental correlated fission data and current efforts to address some of the known shortcomings. Numerical simulations employing the FREYA and CGMF codes are compared to experimental data for a wide range of correlated fission quantities. The inclusion of those codes into the MCNP6.2 and MCNPX - PoliMi transport codes is described and discussed in the context of relevant applications. The accuracy of the model predictions and their sensitivity to model assumptions and input parameters are discussed. Finally, a series of important experimental and theoretical questions that remain unanswered are presented, suggesting a renewed effort to address these shortcomings.

Correlated prompt fission data in transport simulations

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

Talou, P.; Vogt, R.; Randrup, J.

Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n -n, n - γ, and γ - γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ raysmore » from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX - PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in nonproliferation, safeguards, nuclear energy, and defense programs. Here, this review provides an overview of the topic, starting from theoretical considerations of the fission process, with a focus on correlated signatures. It then explores the status of experimental correlated fission data and current efforts to address some of the known shortcomings. Numerical simulations employing the FREYA and CGMF codes are compared to experimental data for a wide range of correlated fission quantities. The inclusion of those codes into the MCNP6.2 and MCNPX - PoliMi transport codes is described and discussed in the context of relevant applications. The accuracy of the model predictions and their sensitivity to model assumptions and input parameters are discussed. Lastly, a series of important experimental and theoretical questions that remain unanswered are presented, suggesting a renewed effort to address these shortcomings.« less

Ada Integrated Environment III Computer Program Development Specification. Volume III. Ada Optimizing Compiler.

DTIC Science & Technology

1981-12-01

file.library-unit{.subunit).SYMAP Statement Map: library-file. library-unit.subunit).SMAP Type Map: 1 ibrary.fi le. 1 ibrary-unit{.subunit). TMAP The library...generator SYMAP Symbol Map code generator SMAP Updated Statement Map code generator TMAP Type Map code generator A.3.5 The PUNIT Command The P UNIT...Core.Stmtmap) NAME Tmap (Core.Typemap) END Example A-3 Compiler Command Stream for the Code Generator Texas Instruments A-5 Ada Optimizing Compiler

Modèle tridimensionnel pour coupler les équations magnétiques et électriques dans le cas de la magnétostatique

NASA Astrophysics Data System (ADS)

Piriou, F.; Razek, A.

1991-03-01

In this paper a 3D model for coupling of magnetic and electric equations is presented. The magnetic equations are solved with the help of finite element method using the magnetic vector potential formulation. To take into account the effects of magnetic saturation we use the Newton-Raphson algorithm. We develop the analysis permitting the coupling of magnetic and electric equations to obtain a difrerential system equations which can be solved with numerical integration. As example we model an iron core coil and the validity of our model is verified by a comparison of the obtained results with an analytical solution and a 2D code calculation. Dans cet article est présenté un modèle 3D qui permet de coupler les équations magnétiques et électriques. Les équations magnétiques sont résolues à l'aide de la méthode des éléments finis en utilisant une formulation en potentiel vecteur magnétique. Dans le modèle proposé les effets de la saturation du circuit magnétique sont pris en compte en utilisant l'algorithme de Newton-Raphson. On montre comment relier les équations magnétiques avec celles du circuit électrique pour aboutir à un système d'équations différentielles que l'on résout avec une intégration numérique. A titre d'exemple on modélise une bobine à noyau ferromagnétique et pour montrer la validité du modèle on compare les résultats obtenus avec une solution analytique et un code de calcul 2D.

Benchmarking Geant4 for simulating galactic cosmic ray interactions within planetary bodies

DOE PAGES

Mesick, K. E.; Feldman, W. C.; Coupland, D. D. S.; ...

2018-06-20

Galactic cosmic rays undergo complex nuclear interactions with nuclei within planetary bodies that have little to no atmosphere. Radiation transport simulations are a key tool used in understanding the neutron and gamma-ray albedo coming from these interactions and tracing these signals back to geochemical composition of the target. In this paper, we study the validity of the code Geant4 for simulating such interactions by comparing simulation results to data from the Apollo 17 Lunar Neutron Probe Experiment. Different assumptions regarding the physics are explored to demonstrate how these impact the Geant4 simulation results. In general, all of the Geant4 resultsmore » over-predict the data, however, certain physics lists perform better than others. Finally, in addition, we show that results from the radiation transport code MCNP6 are similar to those obtained using Geant4.« less

Development of a new multi-modal Monte-Carlo radiotherapy planning system.

PubMed

Kumada, H; Nakamura, T; Komeda, M; Matsumura, A

2009-07-01

A new multi-modal Monte-Carlo radiotherapy planning system (developing code: JCDS-FX) is under development at Japan Atomic Energy Agency. This system builds on fundamental technologies of JCDS applied to actual boron neutron capture therapy (BNCT) trials in JRR-4. One of features of the JCDS-FX is that PHITS has been applied to particle transport calculation. PHITS is a multi-purpose particle Monte-Carlo transport code. Hence application of PHITS enables to evaluate total doses given to a patient by a combined modality therapy. Moreover, JCDS-FX with PHITS can be used for the study of accelerator based BNCT. To verify calculation accuracy of the JCDS-FX, dose evaluations for neutron irradiation of a cylindrical water phantom and for an actual clinical trial were performed, then the results were compared with calculations by JCDS with MCNP. The verification results demonstrated that JCDS-FX is applicable to BNCT treatment planning in practical use.

Shutdown Dose Rate Analysis for the long-pulse D-D Operation Phase in KSTAR

NASA Astrophysics Data System (ADS)

Park, Jin Hun; Han, Jung-Hoon; Kim, D. H.; Joo, K. S.; Hwang, Y. S.

2017-09-01

KSTAR is a medium size fully superconducting tokamak. The deuterium-deuterium (D-D) reaction in the KSTAR tokamak generates neutrons with a peak yield of 3.5x1016 per second through a pulse operation of 100 seconds. The effect of neutron generation from full D-D high power KSTAR operation mode to the machine, such as activation, shutdown dose rate, and nuclear heating, are estimated for an assurance of safety during operation, maintenance, and machine upgrade. The nuclear heating of the in-vessel components, and neutron activation of the surrounding materials have been investigated. The dose rates during operation and after shutdown of KSTAR have been calculated by a 3D CAD model of KSTAR with the Monte Carlo code MCNP5 (neutron flux and decay photon), the inventory code FISPACT (activation and decay photon) and the FENDL 2.1 nuclear data library.

Benchmarking Geant4 for simulating galactic cosmic ray interactions within planetary bodies

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

Mesick, K. E.; Feldman, W. C.; Coupland, D. D. S.

Galactic cosmic rays undergo complex nuclear interactions with nuclei within planetary bodies that have little to no atmosphere. Radiation transport simulations are a key tool used in understanding the neutron and gamma-ray albedo coming from these interactions and tracing these signals back to geochemical composition of the target. In this paper, we study the validity of the code Geant4 for simulating such interactions by comparing simulation results to data from the Apollo 17 Lunar Neutron Probe Experiment. Different assumptions regarding the physics are explored to demonstrate how these impact the Geant4 simulation results. In general, all of the Geant4 resultsmore » over-predict the data, however, certain physics lists perform better than others. Finally, in addition, we show that results from the radiation transport code MCNP6 are similar to those obtained using Geant4.« less

Implementation, capabilities, and benchmarking of Shift, a massively parallel Monte Carlo radiation transport code

DOE PAGES

Pandya, Tara M.; Johnson, Seth R.; Evans, Thomas M.; ...

2015-12-21

This paper discusses the implementation, capabilities, and validation of Shift, a massively parallel Monte Carlo radiation transport package developed and maintained at Oak Ridge National Laboratory. It has been developed to scale well from laptop to small computing clusters to advanced supercomputers. Special features of Shift include hybrid capabilities for variance reduction such as CADIS and FW-CADIS, and advanced parallel decomposition and tally methods optimized for scalability on supercomputing architectures. Shift has been validated and verified against various reactor physics benchmarks and compares well to other state-of-the-art Monte Carlo radiation transport codes such as MCNP5, CE KENO-VI, and OpenMC. Somemore » specific benchmarks used for verification and validation include the CASL VERA criticality test suite and several Westinghouse AP1000 ® problems. These benchmark and scaling studies show promising results.« less

Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

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

Kontogeorgakos, D.; Derstine, K.; Wright, A.

2013-06-01

The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO 2more » particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N–Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.« less

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

DTIC Science & Technology

2007-09-01

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

Modeling of Radioxenon Production and Release Pathways

DTIC Science & Technology

2010-09-01

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

Montgomery Point Lock and Dam, White River, Arkansas

DTIC Science & Technology

2016-01-01

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

AREVA Developments for an Efficient and Reliable use of Monte Carlo codes for Radiation Transport Applications

NASA Astrophysics Data System (ADS)

Chapoutier, Nicolas; Mollier, François; Nolin, Guillaume; Culioli, Matthieu; Mace, Jean-Reynald

2017-09-01

In the context of the rising of Monte Carlo transport calculations for any kind of application, AREVA recently improved its suite of engineering tools in order to produce efficient Monte Carlo workflow. Monte Carlo codes, such as MCNP or TRIPOLI, are recognized as reference codes to deal with a large range of radiation transport problems. However the inherent drawbacks of theses codes - laboring input file creation and long computation time - contrast with the maturity of the treatment of the physical phenomena. The goals of the recent AREVA developments were to reach similar efficiency as other mature engineering sciences such as finite elements analyses (e.g. structural or fluid dynamics). Among the main objectives, the creation of a graphical user interface offering CAD tools for geometry creation and other graphical features dedicated to the radiation field (source definition, tally definition) has been reached. The computations times are drastically reduced compared to few years ago thanks to the use of massive parallel runs, and above all, the implementation of hybrid variance reduction technics. From now engineering teams are capable to deliver much more prompt support to any nuclear projects dealing with reactors or fuel cycle facilities from conceptual phase to decommissioning.

Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor

NASA Technical Reports Server (NTRS)

Butler, C.; Albright, D.

2007-01-01

Highly efficient, compact nuclear reactors would provide high specific impulse spacecraft propulsion. This analysis and numerical simulation effort has focused on the technical feasibility issues related to the nuclear design characteristics of a novel reactor design. The Fissioning Plasma Core Reactor (FPCR) is a shockwave-driven gaseous-core nuclear reactor, which uses Magneto Hydrodynamic effects to generate electric power to be used for propulsion. The nuclear design of the system depends on two major calculations: core physics calculations and kinetics calculations. Presently, core physics calculations have concentrated on the use of the MCNP4C code. However, initial results from other codes such as COMBINE/VENTURE and SCALE4a. are also shown. Several significant modifications were made to the ISR-developed QCALC1 kinetics analysis code. These modifications include testing the state of the core materials, an improvement to the calculation of the material properties of the core, the addition of an adiabatic core temperature model and improvement of the first order reactivity correction model. The accuracy of these modifications has been verified, and the accuracy of the point-core kinetics model used by the QCALC1 code has also been validated. Previously calculated kinetics results for the FPCR were described in the ISR report, "QCALC1: A code for FPCR Kinetics Model Feasibility Analysis" dated June 1, 2002.

Neutron field measurement at the Experimental Advanced Superconducting Tokamak using a Bonner sphere spectrometer

NASA Astrophysics Data System (ADS)

Hu, Zhimeng; Zhong, Guoqiang; Ge, Lijian; Du, Tengfei; Peng, Xingyu; Chen, Zhongjing; Xie, Xufei; Yuan, Xi; Zhang, Yimo; Sun, Jiaqi; Fan, Tieshuan; Zhou, Ruijie; Xiao, Min; Li, Kai; Hu, Liqun; Chen, Jun; Zhang, Hui; Gorini, Giuseppe; Nocente, Massimo; Tardocchi, Marco; Li, Xiangqing; Chen, Jinxiang; Zhang, Guohui

2018-07-01

The neutron field measurement was performed in the Experimental Advanced Superconducting Tokamak (EAST) experimental hall using a Bonner sphere spectrometer (BSS) based on a 3He thermal neutron counter. The measured spectra and the corresponding integrated neutron fluence and dose values deduced from the spectra at two exposed positions were compared to the calculated results obtained by a general Monte Carlo code MCNP5, and good agreements were found. The applicability of a homemade dose survey meter installed at EAST was also verified with the comparison of the ambient dose equivalent H*(10) values measured by the meter and BSS.

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

Rezaei-Ochbelagh, D.; Salman-Nezhad, S.; Asadi, A.

External photon beam radiotherapy is carried out in a way to achieve an 'as low as possible' a dose in healthy tissues surrounding the target. One of these surroundings can be heart as a vital organ of body. As it is impossible to directly determine the absorbed dose by heart, using phantoms is one way to acquire information around it. The other way is Monte Carlo method. In this work we have presented a simulation of heart geometry by introducing of different surfaces in MCNP code. We used 14 surface equations in order to determine human heart modeling. Those surfacesmore » are borders of heart walls and contents.« less

Multigroup cross section library for GFR2400

NASA Astrophysics Data System (ADS)

Čerba, Štefan; Vrban, Branislav; Lüley, Jakub; Haščík, Ján; Nečas, Vladimír

2017-09-01

In this paper the development and optimization of the SBJ_E71 multigroup cross section library for GFR2400 applications is discussed. A cross section processing scheme, merging Monte Carlo and deterministic codes, was developed. Several fine and coarse group structures and two weighting flux options were analysed through 18 benchmark experiments selected from the handbook of ICSBEP and based on performed similarity assessments. The performance of the collapsed version of the SBJ_E71 library was compared with MCNP5 CE ENDF/B VII.1 and the Korean KAFAX-E70 library. The comparison was made based on integral parameters of calculations performed on full core homogenous models.

D-T Neutron Skyshine Experiments at JAERI/FNS

NASA Astrophysics Data System (ADS)

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

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

Optimization of Neutron Spectrum in Northwest Beam Tube of Tehran Research Reactor for BNCT, by MCNP Code

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

Zamani, M.; End of North Kargar st, Atomic Energy Organization of Iran, P.O. Box: 14155-1339, Tehran; Kasesaz, Y.

2015-07-01

In order to gain the neutron spectrum with proper components specification for BNCT, it is necessary to design a Beam Shape Assembling (BSA), include of moderator, collimator, reflector, gamma filter and thermal neutrons filter, in front of the initial radiation beam from the source. According to the result of MCNP4C simulation, the Northwest beam tube has the most optimized neuron flux between three north beam tubes of Tehran Research Reactor (TRR). So, it has been chosen for this purpose. Simulation of the BSA has been done in four above mentioned phases. In each stage, ten best configurations of materials withmore » different length and width were selected as the candidates for the next stage. The last BSA configuration includes of: 78 centimeters of air as an empty space, 40 centimeters of Iron plus 52 centimeters of heavy-water as moderator, 30 centimeters of water or 90 centimeters of Aluminum-Oxide as a reflector, 1 millimeters of lithium (Li) as thermal neutrons filter and finally 3 millimeters of Bismuth (Bi) as a filter of gamma radiation. The result of Calculations shows that if we use this BSA configuration for TRR Northwest beam tube, then the best neutron flux and spectrum will be achieved for BNCT. (authors)« less

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

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

Lessard, O.J.

1993-09-01

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

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

NASA Astrophysics Data System (ADS)

Kouznetsov, A.; Cully, C. M.

2017-12-01

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

Fast fission neutron detection using the Cherenkov effect

NASA Astrophysics Data System (ADS)

Millard, Matthew James

The Cherenkov effect in optically clear media of varying indices of refraction and composition was investigated for quantification of fast neutrons. The ultimate application of the proposed detection system is criticality monitoring. The optically clear medium, composed of select target nuclei, was coupled to a photomultiplier tube. Neutron reaction products of the target nuclei contained within the optical medium emit beta particles and gamma rays that produce Cherenkov photons within the medium which can be detected. Assessed media include quartz (SiO2), sapphire (Al2O3), spinel (MgAl2O4), and zinc sulfide (ZnS), which were irradiated with un-moderated 252Cf. Monte Carlo N-Particle (MCNP) code simulations were conducted to quantify the neutron flux incident on the media. High resolution gamma-ray spectroscopic measurements of the samples were conducted to verify the MCNP estimate. The threshold reactions of interest were 28Si (n, p) 28Al, 27 Al (n, p) 27Mg, 24Mg(n, p)24 Na, and 64Zn(n, p)64Cu which have neutron reaction cross sections in the 1 to 10 MeV range on the order of 0.1 barn. The detection system offers a unique way to measure a criticality event; it can count in place, making retrieval by emergency personnel unnecessary.

Data and methods to estimate fetal dose from fluoroscopically guided prophylactic hypogastric artery balloon occlusion

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

Solomou, G.; Stratakis, J.; Perisinakis, K.

Purpose: To provide data for estimation of fetal radiation dose (D{sub F}) from prophylactic hypogastric artery balloon occlusion (HABO) procedures. Methods: The Monte-Carlo-N-particle (MCNP) transport code and mathematical phantoms representing a pregnant patient at the ninth month of gestation were employed. PA, RAO 20° and LAO 20° fluoroscopy projections of left and right internal iliac arteries were simulated. Projection-specific normalized fetal dose (NFD) data were produced for various beam qualities. The effects of projection angle, x-ray field location relative to the fetus, field size, maternal body size, and fetal size on NFD were investigated. Presented NFD values were compared tomore » corresponding values derived using a physical anthropomorphic phantom simulating pregnancy at the third trimester and thermoluminescence dosimeters. Results: NFD did not considerably vary when projection angle was altered by ±5°, whereas it was found to markedly depend on tube voltage, filtration, x-ray field location and size, and maternal body size. Differences in NFD < 7.5% were observed for naturally expected variations in fetal size. A difference of less than 13.5% was observed between NFD values estimated by MCNP and direct measurements. Conclusions: Data and methods provided allow for reliable estimation of radiation burden to the fetus from HABO.« less

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

PubMed

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

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

Theoretical modeling of a portable x-ray tube based KXRF system to measure lead in bone

PubMed Central

Specht, Aaron J; Weisskopf, Marc G; Nie, Linda Huiling

2017-01-01

Objective K-shell x-ray fluorescence (KXRF) techniques have been used to identify health effects resulting from exposure to metals for decades, but the equipment is bulky and requires significant maintenance and licensing procedures. A portable x-ray fluorescence (XRF) device was developed to overcome these disadvantages, but introduced a measurement dependency on soft tissue thickness. With recent advances to detector technology, an XRF device utilizing the advantages of both systems should be feasible. Approach In this study, we used Monte Carlo simulations to test the feasibility of an XRF device with a high-energy x-ray tube and detector operable at room temperature. Main Results We first validated the use of Monte Carlo N-particle transport code (MCNP) for x-ray tube simulations, and found good agreement between experimental and simulated results. Then, we optimized x-ray tube settings and found the detection limit of the high-energy x-ray tube based XRF device for bone lead measurements to be 6.91 μg g−1 bone mineral using a cadmium zinc telluride detector. Significance In conclusion, this study validated the use of MCNP in simulations of x-ray tube physics and XRF applications, and demonstrated the feasibility of a high-energy x-ray tube based XRF for metal exposure assessment. PMID:28169835

Theoretical modeling of a portable x-ray tube based KXRF system to measure lead in bone.

PubMed

Specht, Aaron J; Weisskopf, Marc G; Nie, Linda Huiling

2017-03-01

K-shell x-ray fluorescence (KXRF) techniques have been used to identify health effects resulting from exposure to metals for decades, but the equipment is bulky and requires significant maintenance and licensing procedures. A portable x-ray fluorescence (XRF) device was developed to overcome these disadvantages, but introduced a measurement dependency on soft tissue thickness. With recent advances to detector technology, an XRF device utilizing the advantages of both systems should be feasible. In this study, we used Monte Carlo simulations to test the feasibility of an XRF device with a high-energy x-ray tube and detector operable at room temperature. We first validated the use of Monte Carlo N-particle transport code (MCNP) for x-ray tube simulations, and found good agreement between experimental and simulated results. Then, we optimized x-ray tube settings and found the detection limit of the high-energy x-ray tube based XRF device for bone lead measurements to be 6.91 µg g -1 bone mineral using a cadmium zinc telluride detector. In conclusion, this study validated the use of MCNP in simulations of x-ray tube physics and XRF applications, and demonstrated the feasibility of a high-energy x-ray tube based XRF for metal exposure assessment.

Effective dose in the manufacturing process of rutile covered welding electrodes.

PubMed

Herranz, M; Rozas, S; Pérez, C; Idoeta, R; Núñez-Lagos, R; Legarda, F

2013-03-01

Shielded metal arc welding using covered electrodes is the most common welding process. Sometimes the covering contains naturally occurring radioactive materials (NORMs). In Spain the most used electrodes are those covered with rutile mixed with other materials. Rutile contains some detectable natural radionuclides, so it can be considered a NORM. This paper mainly focuses on the use of MCNP (Monte Carlo N-Particle Transport Code) as a predictive tool to obtain doses in a factory which produces this type of electrode and assess the radiological impact in a specific facility after estimating the internal dose.To do this, in the facility, areas of highest radiation and positions of workers were identified, radioactive content of rutile and rutile covered electrodes was measured, and, considering a worst possible scenario, external dose at working points has been calculated using MCNP. This procedure has been validated comparing the results obtained with those from a pressurised ionisation chamber and TLD dosimeters. The internal dose has been calculated using DCAL (dose and risk calculation). The doses range between 8.8 and 394 μSv yr(-1), always lower than the effective dose limit for the public, 1 mSv yr(-1). The highest dose corresponds to the mixing area.

The pineapple AcMADS1 promoter confers high level expression in tomato and arabidopsis flowering and fruiting tissues, but AcMADS1 does not complement the tomato LeMADS-RIN (rin) mutant

USDA-ARS?s Scientific Manuscript database

A previous EST study identified a MADS box transcription factor coding sequence, AcMADS1, that is strongly induced during non-climacteric pineapple fruit ripening. Phylogenetic analyses place the AcMADS1 protein in the same superclade as LeMADS-RIN, a master regulator of fruit ripening upstream of e...

Validation of the analytical methods in the LWR code BOXER for gadolinium-loaded fuel pins

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

Paratte, J.M.; Arkuszewski, J.J.; Kamboj, B.K.

1990-01-01

Due to the very high absorption occurring in gadolinium-loaded fuel pins, calculations of lattices with such pins present are a demanding test of the analysis methods in light water reactor (LWR) cell and assembly codes. Considerable effort has, therefore, been devoted to the validation of code methods for gadolinia fuel. The goal of the work reported in this paper is to check the analysis methods in the LWR cell/assembly code BOXER and its associated cross-section processing code ETOBOX, by comparison of BOXER results with those from a very accurate Monte Carlo calculation for a gadolinium benchmark problem. Initial results ofmore » such a comparison have been previously reported. However, the Monte Carlo calculations, done with the MCNP code, were performed at Los Alamos National Laboratory using ENDF/B-V data, while the BOXER calculations were performed at the Paul Scherrer Institute using JEF-1 nuclear data. This difference in the basic nuclear data used for the two calculations, caused by the restricted nature of these evaluated data files, led to associated uncertainties in a comparison of the results for methods validation. In the joint investigations at the Georgia Institute of Technology and PSI, such uncertainty in this comparison was eliminated by using ENDF/B-V data for BOXER calculations at Georgia Tech.« less

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

Depriest, Kendall

Unsuccessful attempts by members of the radiation effects community to independently derive the Norgett-Robinson-Torrens (NRT) damage energy factors for silicon in ASTM standard E722-14 led to an investigation of the software coding and data that produced those damage energy factors. The ad hoc collaboration to discover the reason for lack of agreement revealed a coding error and resulted in a report documenting the methodology to produce the response function for the standard. The recommended changes in the NRT damage energy factors for silicon are shown to have significant impact for a narrow energy region of the 1-MeV(Si) equivalent fluence responsemore » function. However, when evaluating integral metrics over all neutrons energies in various spectra important to the SNL electronics testing community, the change in the response results in a small decrease in the total 1- MeV(Si) equivalent fluence of ~0.6% compared to the E722-14 response. Response functions based on the newly recommended NRT damage energy factors have been produced and are available for users of both the NuGET and MCNP codes.« less

Analysis of neutron and gamma-ray streaming along the maze of NRCAM thallium production target room.

PubMed

Raisali, G; Hajiloo, N; Hamidi, S; Aslani, G

2006-08-01

Study of the shield performance of a thallium-203 production target room has been investigated in this work. Neutron and gamma-ray equivalent dose rates at various points of the maze are calculated by simulating the transport of streaming neutrons, and photons using Monte Carlo method. For determination of neutron and gamma-ray source intensities and their energy spectrum, we have applied SRIM 2003 and ALICE91 computer codes to Tl target and its Cu substrate for a 145 microA of 28.5 MeV protons beam. The MCNP/4C code has been applied with neutron source term in mode n p to consider both prompt neutrons and secondary gamma-rays. Then the code is applied for the prompt gamma-rays as the source term. The neutron-flux energy spectrum and equivalent dose rates for neutron and gamma-rays in various positions in the maze have been calculated. It has been found that the deviation between calculated and measured dose values along the maze is less than 20%.

Performance and accuracy of criticality calculations performed using WARP – A framework for continuous energy Monte Carlo neutron transport in general 3D geometries on GPUs

DOE PAGES

Bergmann, Ryan M.; Rowland, Kelly L.; Radnović, Nikola; ...

2017-05-01

In this companion paper to "Algorithmic Choices in WARP - A Framework for Continuous Energy Monte Carlo Neutron Transport in General 3D Geometries on GPUs" (doi:10.1016/j.anucene.2014.10.039), the WARP Monte Carlo neutron transport framework for graphics processing units (GPUs) is benchmarked against production-level central processing unit (CPU) Monte Carlo neutron transport codes for both performance and accuracy. We compare neutron flux spectra, multiplication factors, runtimes, speedup factors, and costs of various GPU and CPU platforms running either WARP, Serpent 2.1.24, or MCNP 6.1. WARP compares well with the results of the production-level codes, and it is shown that on the newestmore » hardware considered, GPU platforms running WARP are between 0.8 to 7.6 times as fast as CPU platforms running production codes. Also, the GPU platforms running WARP were between 15% and 50% as expensive to purchase and between 80% to 90% as expensive to operate as equivalent CPU platforms performing at an equal simulation rate.« less

Performance and accuracy of criticality calculations performed using WARP – A framework for continuous energy Monte Carlo neutron transport in general 3D geometries on GPUs

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

Bergmann, Ryan M.; Rowland, Kelly L.; Radnović, Nikola

In this companion paper to "Algorithmic Choices in WARP - A Framework for Continuous Energy Monte Carlo Neutron Transport in General 3D Geometries on GPUs" (doi:10.1016/j.anucene.2014.10.039), the WARP Monte Carlo neutron transport framework for graphics processing units (GPUs) is benchmarked against production-level central processing unit (CPU) Monte Carlo neutron transport codes for both performance and accuracy. We compare neutron flux spectra, multiplication factors, runtimes, speedup factors, and costs of various GPU and CPU platforms running either WARP, Serpent 2.1.24, or MCNP 6.1. WARP compares well with the results of the production-level codes, and it is shown that on the newestmore » hardware considered, GPU platforms running WARP are between 0.8 to 7.6 times as fast as CPU platforms running production codes. Also, the GPU platforms running WARP were between 15% and 50% as expensive to purchase and between 80% to 90% as expensive to operate as equivalent CPU platforms performing at an equal simulation rate.« less

Summer 2016

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

Mendoza, Paul Michael

2016-08-31

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

Coastal Benthic Boundary Layer Special Research Program: A Review of the First Year. Volume 1.

DTIC Science & Technology

1994-04-06

also Indebted to Dr. LeBlanc for his supervision of the relaxation time numerical analyses, and Lachlan Munro, a3 graduate ONR AASERT student, for coding...R. Smith and E. Besancon Code 7174 Naval Research Laboratory Stennis Space Center, MS 39529-5004 I INTRODUCTION: This brief report outlines the

Turbomachinery Forced Response Prediction System (FREPS): User's Manual

NASA Technical Reports Server (NTRS)

Morel, M. R.; Murthy, D. V.

1994-01-01

The turbomachinery forced response prediction system (FREPS), version 1.2, is capable of predicting the aeroelastic behavior of axial-flow turbomachinery blades. This document is meant to serve as a guide in the use of the FREPS code with specific emphasis on its use at NASA Lewis Research Center (LeRC). A detailed explanation of the aeroelastic analysis and its development is beyond the scope of this document, and may be found in the references. FREPS has been developed by the NASA LeRC Structural Dynamics Branch. The manual is divided into three major parts: an introduction, the preparation of input, and the procedure to execute FREPS. Part 1 includes a brief background on the necessity of FREPS, a description of the FREPS system, the steps needed to be taken before FREPS is executed, an example input file with instructions, presentation of the geometric conventions used, and the input/output files employed and produced by FREPS. Part 2 contains a detailed description of the command names needed to create the primary input file that is required to execute the FREPS code. Also, Part 2 has an example data file to aid the user in creating their own input files. Part 3 explains the procedures required to execute the FREPS code on the Cray Y-MP, a computer system available at the NASA LeRC.

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

DTIC Science & Technology

2014-04-21

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

Simplification of an MCNP model designed for dose rate estimation

NASA Astrophysics Data System (ADS)

Laptev, Alexander; Perry, Robert

2017-09-01

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

A Deep Penetration Problem Calculation Using AETIUS:An Easy Modeling Discrete Ordinates Transport Code UsIng Unstructured Tetrahedral Mesh, Shared Memory Parallel

NASA Astrophysics Data System (ADS)

KIM, Jong Woon; LEE, Young-Ouk

2017-09-01

As computing power gets better and better, computer codes that use a deterministic method seem to be less useful than those using the Monte Carlo method. In addition, users do not like to think about space, angles, and energy discretization for deterministic codes. However, a deterministic method is still powerful in that we can obtain a solution of the flux throughout the problem, particularly as when particles can barely penetrate, such as in a deep penetration problem with small detection volumes. Recently, a new state-of-the-art discrete-ordinates code, ATTILA, was developed and has been widely used in several applications. ATTILA provides the capabilities to solve geometrically complex 3-D transport problems by using an unstructured tetrahedral mesh. Since 2009, we have been developing our own code by benchmarking ATTILA. AETIUS is a discrete ordinates code that uses an unstructured tetrahedral mesh such as ATTILA. For pre- and post- processing, Gmsh is used to generate an unstructured tetrahedral mesh by importing a CAD file (*.step) and visualizing the calculation results of AETIUS. Using a CAD tool, the geometry can be modeled very easily. In this paper, we describe a brief overview of AETIUS and provide numerical results from both AETIUS and a Monte Carlo code, MCNP5, in a deep penetration problem with small detection volumes. The results demonstrate the effectiveness and efficiency of AETIUS for such calculations.

An Object Oriented Analysis Method for Ada and Embedded Systems

DTIC Science & Technology

1989-12-01

expansion of the paradligm from the coding anld desiningactivities into the earlier activity of reurmnsalyi.Ts hpl, begins by discussing the application of...response time: 0.1 seconds.I Step le: Identify Known Restrictions on the Software.I " The cruise control system object code must fit within 16K of mem- orv...application of object-oriented techniques to the coding and desigll phases of the life cycle, as well as various approaches to requirements analysis. 3

Littoral Infrared Ship Self Defence Technology Studies (Autodefense cotiere infrarouge des navires etudes technologiques)

DTIC Science & Technology

2014-05-01

de simulation du Simulateur de Contre- mesures de la Menace Navale afin de pouvoir inclure des leurres et des autodirecteurs de missiles ; 4) Une...sur le littoral ; 2) La détection des petites cibles de surface sur le littoral ; 3) L’amélioration et la validation de la modélisation et du code...amélioration et une validation supplémentaire de la modélisation et du

Error Processing Techniques for the Modified Read Facsimile Code.

DTIC Science & Technology

1981-09-01

suppose la cr~ation d’un support commun pour tout** les informations, une vtritable " Banque do dones". repartic sur des moyens detraitement nationaux...Veisctnce dv 20 rdgions. Dos calculateurs ant #td implant~s done le pass# au main@ dans toutem lea plus important**. On trouve ainsi des machines Bull...rA, un certain nombre dlentre eux devront n~cessairement etre mc- cesaiblos, voire min & jour en temps rdel :parmi, ccs derniers le fichier commercial

Assessment of doses caused by electrons in thin layers of tissue-equivalent materials, using MCNP.

PubMed

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.

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

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

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

1995-02-01

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

Involvement of LeMRP, an ATP-binding cassette transporter, in shikonin transport and biosynthesis in Lithospermum erythrorhizon.

PubMed

Zhu, Y; Chu, S-J; Luo, Y-L; Fu, J-Y; Tang, C-Y; Lu, G-H; Pang, Y-J; Wang, X-M; Yang, R-W; Qi, J-L; Yang, Y-H

2018-03-01

Shikonin and its derivatives are important medicinal secondary metabolites accumulating in roots of Lithospermum erythrorhizon. Although some membrane proteins have been identified as transporters of secondary metabolites, the mechanisms underlying shikonin transport and accumulation in L. erythrorhizon cells still remain largely unknown. In this study, we isolated a cDNA encoding LeMRP, an ATP-binding cassette transporter from L. erythrorhizon, and further investigated its functions in the transport and biosynthesis of shikonin using the yeast transformation and transgenic hairy root methods, respectively. Real-time PCR was applied for expression analyses of LeMRP and shikonin biosynthetic enzyme genes. Functional analysis of LeMRP using the heterologous yeast cell expression system showed that LeMRP could be involved in shikonin transport. Transgenic hairy roots of L. erythrorhizon demonstrated that LeMRP overexpressing hairy roots produced more shikonin than the empty vector (EV) control. Real-time PCR results revealed that the enhanced shikonin biosynthesis in the overexpression lines was mainly caused by highly up-regulated expression of genes coding key enzymes (LePAL, HMGR, Le4CL and LePGT) involved in shikonin biosynthesis. Conversely, LeMRP RNAi decreased the accumulation of shikonin and effectively down-regulated expression level of the above genes. Typical inhibitors of ABC proteins, such as azide and buthionine sulphoximine, dramatically inhibited accumulation of shikonin in hairy roots. Our findings provide evidence for the important direct or indirect role of LeMRP in transmembrane transport and biosynthesis of shikonin. © 2017 German Society for Plant Sciences and The Royal Botanical Society of the Netherlands.

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

PubMed Central

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

2017-01-01

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

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

PubMed

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

2017-01-01

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

Plasma Theory and Simulation.

DTIC Science & Technology

1978-07-01

l l) A paper t i t led “Part icle-Fluid Hybrid Codes Applied to Beam- Plasma , Ring -Plasma Instabi l i ties ” was presented at Monterey (see Section V...ic le-Fluid Hybr id Codes Applied to Beam- Plasma , Ring -Plasma Ins tab i l i t ies”. (2) A. Peiravi and C. K. Birdsall , “Self-Heating of id Therma l

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

PubMed

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

2008-01-01

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

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

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

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

1995-01-01

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

Reactive transport modeling of uranium 238 and radium 226 in groundwater of the Königstein uranium mine, Germany

NASA Astrophysics Data System (ADS)

Nitzsche, O.; Merkel, B.

Knowledge of the transport behavior of radionuclides in groundwater is needed for both groundwater protection and remediation of abandoned uranium mines and milling sites. Dispersion, diffusion, mixing, recharge to the aquifer, and chemical interactions, as well as radioactive decay, should be taken into account to obtain reliable predictions on transport of primordial nuclides in groundwater. This paper demonstrates the need for carrying out rehabilitation strategies before closure of the Königstein in-situ leaching uranium mine near Dresden, Germany. Column experiments on drilling cores with uranium-enriched tap water provided data about the exchange behavior of uranium. Uranium breakthrough was observed after more than 20 pore volumes. This strong retardation is due to the exchange of positively charged uranium ions. The code TReAC is a 1-D, 2-D, and 3-D reactive transport code that was modified to take into account the radioactive decay of uranium and the most important daughter nuclides, and to include double-porosity flow. TReAC satisfactorily simulated the breakthrough curves of the column experiments and provided a first approximation of exchange parameters. Groundwater flow in the region of the Königstein mine was simulated using the FLOWPATH code. Reactive transport behavior was simulated with TReAC in one dimension along a 6000-m path line. Results show that uranium migration is relatively slow, but that due to decay of uranium, the concentration of radium along the flow path increases. Results are highly sensitive to the influence of double-porosity flow. Résumé La protection des eaux souterraines et la restauration des sites miniers et de prétraitement d'uranium abandonnés nécessitent de connaître le comportement des radionucléides au cours de leur transport dans les eaux souterraines. La dispersion, la diffusion, le mélange, la recharge de l'aquifère et les interactions chimiques, de même que la décroissance radioactive, doivent être prises en compte pour obtenir des prédictions fiables concernant le transport des nucléides primaires dans les eaux souterraines. Ce papier montre la nécessité d'établir des stratégies de réhabilitation avant la fermeture de la mine d'uranium de Knigstein, près de Dresde (Allemagne). Des expériences de lessivage en colonne sur des carottes avec de l'eau enrichie en uranium fournissent des données sur le comportement de l'échange de l'uranium. La restitution de l'uranium a été observée après un lessivage par un volume supérieur à 20 fois celui des pores. Ce fort retard est dûà l'échange d'ions uranium positifs. Le code TReAC est un code de transport réactif en 1D, 2D et 3D, qui a été modifié pour prendre en compte la décroissance radioactive de l'uranium et les principaux nucléides descendants, et pour introduire l'écoulement dans un milieu à double porosité. TReAC a simulé de façon satisfaisante les courbes de restitution des expériences sur colonne et a fourni une première approche des paramètres de l'échange. L'écoulement souterrain dans la région de la mine de Knigstein a été simulé au moyen du code FLOWPATH. Le comportement du transport réactif a été simulé avec TReAC en une dimension, le long d'un axe d'écoulement long de 6000 m. Les résultats montrent que la migration de l'uranium est relativement lente ; mais du fait de la décroissance radioactive de l'uranium, la concentration en radium le long de cet axe augmente. Les résultats sont très sensibles à l'influence de l'écoulement en milieu à double porosité.

The NASA research and technology program on batteries

NASA Technical Reports Server (NTRS)

Bennett, Gary L.

1990-01-01

The NASA research and technology program on batteries is being carried out within the Propulsion, Power and Energy Division (Code RP) of NASA's Office of Aeronautics, Exploration and Technology (OAET). The program includes development of high-performance, long-life, cost-effective primary and secondary (rechargeable) batteries. The NASA OAET battery program is being carried out at Lewis Research Center (LeRC) and the Jet Propulsion Laboratory (JPL). LeRC is focusing primarily on nickel-hydrogen batteries (both individual pressure vessel or IPV and bipolar). LeRC is also involved in a planned flight experiment to test a sodium-sulfur battery design. JPL is focusing primarily on lithium rechargeable batteries, having successfully transferred its lithium primary battery technology to the U.S. Air Force for use on the Centaur upper stage. Both LeRC and JPL are studying advanced battery concepts that offer even higher specific energies. The long-term goal is to achieve 100 Wh/kg.

DUKSUP: A Computer Program for High Thrust Launch Vehicle Trajectory Design and Optimization

NASA Technical Reports Server (NTRS)

Williams, C. H.; Spurlock, O. F.

2014-01-01

From the late 1960's through 1997, the leadership of NASA's Intermediate and Large class unmanned expendable launch vehicle projects resided at the NASA Lewis (now Glenn) Research Center (LeRC). One of LeRC's primary responsibilities --- trajectory design and performance analysis --- was accomplished by an internally-developed analytic three dimensional computer program called DUKSUP. Because of its Calculus of Variations-based optimization routine, this code was generally more capable of finding optimal solutions than its contemporaries. A derivation of optimal control using the Calculus of Variations is summarized including transversality, intermediate, and final conditions. The two point boundary value problem is explained. A brief summary of the code's operation is provided, including iteration via the Newton-Raphson scheme and integration of variational and motion equations via a 4th order Runge-Kutta scheme. Main subroutines are discussed. The history of the LeRC trajectory design efforts in the early 1960's is explained within the context of supporting the Centaur upper stage program. How the code was constructed based on the operation of the Atlas/Centaur launch vehicle, the limits of the computers of that era, the limits of the computer programming languages, and the missions it supported are discussed. The vehicles DUKSUP supported (Atlas/Centaur, Titan/Centaur, and Shuttle/Centaur) are briefly described. The types of missions, including Earth orbital and interplanetary, are described. The roles of flight constraints and their impact on launch operations are detailed (such as jettisoning hardware on heating, Range Safety, ground station tracking, and elliptical parking orbits). The computer main frames on which the code was hosted are described. The applications of the code are detailed, including independent check of contractor analysis, benchmarking, leading edge analysis, and vehicle performance improvement assessments. Several of DUKSUP's many major impacts on launches are discussed including Intelsat, Voyager, Pioneer Venus, HEAO, Galileo, and Cassini.

DUKSUP: A Computer Program for High Thrust Launch Vehicle Trajectory Design and Optimization

NASA Technical Reports Server (NTRS)

Spurlock, O. Frank; Williams, Craig H.

2015-01-01

From the late 1960s through 1997, the leadership of NASAs Intermediate and Large class unmanned expendable launch vehicle projects resided at the NASA Lewis (now Glenn) Research Center (LeRC). One of LeRCs primary responsibilities --- trajectory design and performance analysis --- was accomplished by an internally-developed analytic three dimensional computer program called DUKSUP. Because of its Calculus of Variations-based optimization routine, this code was generally more capable of finding optimal solutions than its contemporaries. A derivation of optimal control using the Calculus of Variations is summarized including transversality, intermediate, and final conditions. The two point boundary value problem is explained. A brief summary of the codes operation is provided, including iteration via the Newton-Raphson scheme and integration of variational and motion equations via a 4th order Runge-Kutta scheme. Main subroutines are discussed. The history of the LeRC trajectory design efforts in the early 1960s is explained within the context of supporting the Centaur upper stage program. How the code was constructed based on the operation of the AtlasCentaur launch vehicle, the limits of the computers of that era, the limits of the computer programming languages, and the missions it supported are discussed. The vehicles DUKSUP supported (AtlasCentaur, TitanCentaur, and ShuttleCentaur) are briefly described. The types of missions, including Earth orbital and interplanetary, are described. The roles of flight constraints and their impact on launch operations are detailed (such as jettisoning hardware on heating, Range Safety, ground station tracking, and elliptical parking orbits). The computer main frames on which the code was hosted are described. The applications of the code are detailed, including independent check of contractor analysis, benchmarking, leading edge analysis, and vehicle performance improvement assessments. Several of DUKSUPs many major impacts on launches are discussed including Intelsat, Voyager, Pioneer Venus, HEAO, Galileo, and Cassini.

Gamma-ray spectroscopy measurements and simulations for uranium mining

NASA Astrophysics Data System (ADS)

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

2018-01-01

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

A CT and MRI scan to MCNP input conversion program.

PubMed

Van Riper, Kenneth A

2005-01-01

We describe a new program to read a sequence of tomographic scans and prepare the geometry and material sections of an MCNP input file. Image processing techniques include contrast controls and mapping of grey scales to colour. The user interface provides several tools with which the user can associate a range of image intensities to an MCNP material. Materials are loaded from a library. A separate material assignment can be made to a pixel intensity or range of intensities when that intensity dominates the image boundaries; this material is assigned to all pixels with that intensity contiguous with the boundary. Material fractions are computed in a user-specified voxel grid overlaying the scans. New materials are defined by mixing the library materials using the fractions. The geometry can be written as an MCNP lattice or as individual cells. A combination algorithm can be used to join neighbouring cells with the same material.

DNDO Report

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

Liegey, Lauren Rene; Wilcox, Trevor; Mckinney, Gregg Walter

2015-08-07

My internship program was the Domestic Nuclear Detection Office Summer Internship Program. I worked at Los Alamos National Laboratory with Trevor A. Wilcox and Gregg W. McKinney in the NEN-5 group. My project title was “MCNP Physical Model Interoperability & Validation”. The goal of my project was to write a program to predict the solar modulation parameter for dates in the future and then implement it into MCNP6. This update to MCNP6 can be used to calculate the background more precisely, which is an important factor in being able to detect Special Nuclear Material. We will share our work inmore » a published American Nuclear Society (ANS) paper, an ANS presentation, and a LANL student poster session. Through this project, I gained skills in programming, computing, and using MCNP. I also gained experience that will help me decide on a career or perhaps obtain employment in the future.« less

HR Public meeting

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

Heuer, Rolf; Catherin, Anne-Sylvie; Vuillemin, Vin

2010-06-25

Cher(e)s collègues, En collaboration avec le Département HR, le Directeur général a le plaisir de vous convier à une réunion publique qui se tiendra le vendredi 25 juin 2010 à 9h30 dans l’Amphithéâtre principal (Bâtiment 500)*. Un café d’accueil y sera servi à partir de 9h. Cette réunion abordera les thèmes suivants : • Valeurs de l’Organisation (Directeur général) • Code de Conduite (Directeur général / Anne-Sylvie Catherin) • Création du nouveau rôle d’Ombudsperson (Vincent Vuillemin); Ces présentations seront suivies d’une séance de questions-réponses. Nous espérons vous retrouver nombreux le 25 juin ! Meilleures salutations, Anne-Sylvie Catherin Chef du Départementmore » des Ressources humaines *Cette réunion sera retransmise simultanément dans l’Amphithéâtre BE de Prévessin (Bâtiment 864) et également disponible à l’adresse suivante : http://webcast.cern.ch. Dear colleagues, In collaboration with HR Department, the Director-General would like to invite you to an information meeting which will be held on Friday 25 June 2010 at 9:30 am in the Main Auditorium (Building 500)*. A welcome coffee will be available from 9:00 am. During this meeting, information will be given about: • Organization’s values (Director-General) • Code of Conduct (Director-General / Anne-Sylvie Catherin) • New Ombudsperson role (Vincent Vuillemin) These presentations will be followed by a questions & answers session. We look forward to seeing you all on 25 June! Best regards, Anne-Sylvie Catherin Head, Human Resources Department. This meeting will be simultaneously retransmitted in BE Auditorium (Building 864) and available at the following address: http://webcast.cern.ch.« less

HR Public meeting

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

None

2010-10-12

Cher(e)s collègues, En collaboration avec le Département HR, le Directeur général a le plaisir de vous convier à une réunion publique qui se tiendra le vendredi 25 juin 2010 à 9h30 dans l’Amphithéâtre principal (Bâtiment 500)*. Un café d’accueil y sera servi à partir de 9h. Cette réunion abordera les thèmes suivants : Valeurs de l’Organisation (Directeur général); Code de Conduite (Directeur général / Anne-Sylvie Catherin); Création du nouveau rôle d’Ombudsperson (Vincent Vuillemin) Ces présentations seront suivies d’une séance de questions-réponses. Nous espérons vous retrouver nombreux le 25 juin ! Meilleures salutations, Anne-Sylvie Catherin Chef du Département des Ressources humainesmore » *Cette réunion sera retransmise simultanément dans l’Amphithéâtre BE de Prévessin (Bâtiment 864) et également disponible à l’adresse suivante : http://webcast.cern.ch [Dear colleagues, In collaboration with HR Department, the Director-General would like to invite you to an information meeting which will be held on Friday 25 June 2010 at 9:30 am in the Main Auditorium (Building 500)*. A welcome coffee will be available from 9:00 am. During this meeting, information will be given about: Organization’s values (Director-General); Code of Conduct (Director-General / Anne-Sylvie Catherin); New Ombudsperson role (Vincent Vuillemin); These presentations will be followed by a questions & answers session. We look forward to seeing you all on 25 June! Best regards, Anne-Sylvie Catherin Head, Human Resources Department *This meeting will be simultaneously retransmitted in BE Auditorium (Building 864) and available at the following address: http://webcast.cern.ch.« less

Simulation of prompt gamma-ray emission during proton radiotherapy.

PubMed

Verburg, Joost M; Shih, Helen A; Seco, Joao

2012-09-07

The measurement of prompt gamma rays emitted from proton-induced nuclear reactions has been proposed as a method to verify in vivo the range of a clinical proton radiotherapy beam. A good understanding of the prompt gamma-ray emission during proton therapy is key to develop a clinically feasible technique, as it can facilitate accurate simulations and uncertainty analysis of gamma detector designs. Also, the gamma production cross-sections may be incorporated as prior knowledge in the reconstruction of the proton range from the measurements. In this work, we performed simulations of proton-induced nuclear reactions with the main elements of human tissue, carbon-12, oxygen-16 and nitrogen-14, using the nuclear reaction models of the GEANT4 and MCNP6 Monte Carlo codes and the dedicated nuclear reaction codes TALYS and EMPIRE. For each code, we made an effort to optimize the input parameters and model selection. The results of the models were compared to available experimental data of discrete gamma line cross-sections. Overall, the dedicated nuclear reaction codes reproduced the experimental data more consistently, while the Monte Carlo codes showed larger discrepancies for a number of gamma lines. The model differences lead to a variation of the total gamma production near the end of the proton range by a factor of about 2. These results indicate a need for additional theoretical and experimental study of proton-induced gamma emission in human tissue.

Hybrid parallel code acceleration methods in full-core reactor physics calculations

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

Courau, T.; Plagne, L.; Ponicot, A.

2012-07-01

When dealing with nuclear reactor calculation schemes, the need for three dimensional (3D) transport-based reference solutions is essential for both validation and optimization purposes. Considering a benchmark problem, this work investigates the potential of discrete ordinates (Sn) transport methods applied to 3D pressurized water reactor (PWR) full-core calculations. First, the benchmark problem is described. It involves a pin-by-pin description of a 3D PWR first core, and uses a 8-group cross-section library prepared with the DRAGON cell code. Then, a convergence analysis is performed using the PENTRAN parallel Sn Cartesian code. It discusses the spatial refinement and the associated angular quadraturemore » required to properly describe the problem physics. It also shows that initializing the Sn solution with the EDF SPN solver COCAGNE reduces the number of iterations required to converge by nearly a factor of 6. Using a best estimate model, PENTRAN results are then compared to multigroup Monte Carlo results obtained with the MCNP5 code. Good consistency is observed between the two methods (Sn and Monte Carlo), with discrepancies that are less than 25 pcm for the k{sub eff}, and less than 2.1% and 1.6% for the flux at the pin-cell level and for the pin-power distribution, respectively. (authors)« less

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

NASA Astrophysics Data System (ADS)

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

2005-06-01

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

MCNPX CALCULATIONS OF SPECIFIC ABSORBED FRACTIONS IN SOME ORGANS OF THE HUMAN BODY DUE TO APPLICATION OF 133Xe, 99mTc and 81mKr RADIONUCLIDES.

PubMed

Jovanovic, Z; Krstic, D; Nikezic, D; Ros, J M Gomez; Ferrari, P

2018-03-01

Monte Carlo simulations were performed to evaluate treatment doses with wide spread used radionuclides 133Xe, 99mTc and 81mKr. These different radionuclides are used in perfusion or ventilation examinations in nuclear medicine and as indicators for cardiovascular and pulmonary diseases. The objective of this work was to estimate the specific absorbed fractions in surrounding organs and tissues, when these radionuclides are incorporated in the lungs. For this purpose a voxel thorax model has been developed and compared with the ORNL phantom. All calculations and simulations were performed by means of the MCNP5/X code.

Monte Carlo calculations of lung dose in ORNL phantom for boron neutron capture therapy.

PubMed

Krstic, D; Markovic, V M; Jovanovic, Z; Milenkovic, B; Nikezic, D; Atanackovic, J

2014-10-01

Monte Carlo simulations were performed to evaluate dose for possible treatment of cancers by boron neutron capture therapy (BNCT). The computational model of male Oak Ridge National Laboratory (ORNL) phantom was used to simulate tumours in the lung. Calculations have been performed by means of the MCNP5/X code. In this simulation, two opposite neutron beams were considered, in order to obtain uniform neutron flux distribution inside the lung. The obtained results indicate that the lung cancer could be treated by BNCT under the assumptions of calculations. © The Author 2014. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

A possible approach to 14MeV neutron moderation: A preliminary study case.

PubMed

Flammini, D; Pilotti, R; Pietropaolo, A

2017-07-01

Deuterium-Tritium (D-T) interactions produce almost monochromatic neutrons with about 14MeV energy. These neutrons are used in benchmark experiments as well as for neutron cross sections assessment in fusion reactors technology. The possibility to moderate 14MeV neutrons for purposes beyond fusion is worth to be studied in relation to projects of intense D-T sources. In this preliminary study, carried out using the MCNP Monte Carlo code, the moderation of 14MeV neutrons is approached foreseeing the use of combination of metallic materials as pre-moderator and reflectors coupled to standard water moderators. Copyright © 2017 Elsevier Ltd. All rights reserved.

Neutronic Calculation Analysis for CN HCCB TBM-Set

NASA Astrophysics Data System (ADS)

Cao, Qixiang; Zhao, Fengchao; Zhao, Zhou; Wu, Xinghua; Li, Zaixin; Wang, Xiaoyu; Feng, Kaiming

2015-07-01

Using the Monte Carlo transport code MCNP, neutronic calculation analysis for China helium cooled ceramic breeder test blanket module (CN HCCB TBM) and the associated shield block (together called TBM-set) has been carried out based on the latest design of HCCB TBM-set and C-lite model. Key nuclear responses of HCCB TBM-set, such as the neutron flux, tritium production rate, nuclear heating and radiation damage, have been obtained and discussed. These nuclear performance data can be used as the basic input data for other analyses of HCCB TBM-set, such as thermal-hydraulics, thermal-mechanics and safety analysis. supported by the Major State Basic Research Development Program of China (973 Program) (No. 2013GB108000)

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

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

Lockhart, Madeline Louise; McMath, Garrett Earl

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

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

DOE PAGES

Lockhart, Madeline Louise; McMath, Garrett Earl

2017-10-26

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

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

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

Solomon, Jr., Clell J.

2014-05-05

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

Development and application of a hybrid transport methodology for active interrogation systems

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

Royston, K.; Walters, W.; Haghighat, A.

A hybrid Monte Carlo and deterministic methodology has been developed for application to active interrogation systems. The methodology consists of four steps: i) neutron flux distribution due to neutron source transport and subcritical multiplication; ii) generation of gamma source distribution from (n, 7) interactions; iii) determination of gamma current at a detector window; iv) detection of gammas by the detector. This paper discusses the theory and results of the first three steps for the case of a cargo container with a sphere of HEU in third-density water cargo. To complete the first step, a response-function formulation has been developed tomore » calculate the subcritical multiplication and neutron flux distribution. Response coefficients are pre-calculated using the MCNP5 Monte Carlo code. The second step uses the calculated neutron flux distribution and Bugle-96 (n, 7) cross sections to find the resulting gamma source distribution. In the third step the gamma source distribution is coupled with a pre-calculated adjoint function to determine the gamma current at a detector window. The AIMS (Active Interrogation for Monitoring Special-Nuclear-Materials) software has been written to output the gamma current for a source-detector assembly scanning across a cargo container using the pre-calculated values and taking significantly less time than a reference MCNP5 calculation. (authors)« less

Optimization of the Efficiency of a Neutron Detector to Measure (α, n) Reaction Cross-Section

NASA Astrophysics Data System (ADS)

Perello, Jesus; Montes, Fernando; Ahn, Tony; Meisel, Zach; Joint InstituteNuclear Astrophysics Team

2015-04-01

Nucleosynthesis, the origin of elements, is one of the greatest mysteries in physics. A recent particular nucleosynthesis process of interest is the charge-particle process (cpp). In the cpp, elements form by nuclear fusion reactions during supernovae. This process of nuclear fusion, (α,n), will be studied by colliding beam elements produced and accelerated at the National Superconducting Cyclotron Laboratory (NSCL) to a helium-filled cell target. The elements will fuse with α (helium nuclei) and emit neutrons during the reaction. The neutrons will be detected for a count of fused-elements, thus providing us the probability of such reactions. The neutrons will be detected using the Neutron Emission Ratio Observer (NERO). Currently, NERO's efficiency varies for neutrons at the expected energy range (0-12 MeV). To study (α,n), NERO's efficiency must be near-constant at these energies. Monte-Carlo N-Particle Transport Code (MCNP6), a software package that simulates nuclear processes, was used to optimize NERO configuration for the experiment. MCNP6 was used to simulate neutron interaction with different NERO configurations at the expected neutron energies. By adding additional 3He detectors and polyethylene, a near-constant efficiency at these energies was obtained in the simulations. With the new NERO configuration, study of the (α,n) reactions can begin, which may explain how elements are formed in the cpp. SROP MSU, NSF, JINA, McNair Society.

Detection of radioactive particles offshore by γ-ray spectrometry Part I: Monte Carlo assessment of detection depth limits

NASA Astrophysics Data System (ADS)

Maučec, M.; de Meijer, R. J.; Rigollet, C.; Hendriks, P. H. G. M.; Jones, D. G.

2004-06-01

A joint research project between the British Geological Survey and Nuclear Geophysics Division of the Kernfysisch Versneller Instituut, Groningen, the Netherlands, was commissioned by the United Kingdom Atomic Energy Authority to establish the efficiency of a towed seabed γ-ray spectrometer for the detection of 137Cs-containing radioactive particles offshore Dounreay, Scotland. Using the MCNP code, a comprehensive Monte Carlo feasibility study was carried out to model various combinations of geological matrices, particle burial depth and lateral displacement, source activity and detector material. To validate the sampling and absolute normalisation procedures of MCNP for geometries including multiple (natural and induced) heterogeneous sources in environmental monitoring, a benchmark experiment was conducted. The study demonstrates the ability of seabed γ-ray spectrometry to locate radioactive particles offshore and to distinguish between γ count rate increases due to particles from those due to enhanced natural radioactivity. The information presented in this study will be beneficial for estimation of the inventory of 137Cs particles and their activity distribution and for the recovery of particles from the sea floor. In this paper, the Monte Carlo assessment of the detection limits is presented. The estimation of the required towing speed and acquisition times and their application to radioactive particle detection and discrimination offshore formed a supplementary part of this study.

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

DOE PAGES

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

2017-03-01

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

Consistent criticality and radiation studies of Swiss spent nuclear fuel: The CS2M approach.

PubMed

Rochman, D; Vasiliev, A; Ferroukhi, H; Pecchia, M

2018-06-15

In this paper, a new method is proposed to systematically calculate at the same time canister loading curves and radiation sources, based on the inventory information from an in-core fuel management system. As a demonstration, the isotopic contents of the assemblies come from a Swiss PWR, considering more than 6000 cases from 34 reactor cycles. The CS 2 M approach consists in combining four codes: CASMO and SIMULATE to extract the assembly characteristics (based on validated models), the SNF code for source emission and MCNP for criticality calculations for specific canister loadings. The considered cases cover enrichments from 1.9 to 5.0% for the UO 2 assemblies and 4.8% for the MOX, with assembly burnup values from 7 to 74 MWd/kgU. Because such a study is based on the individual fuel assembly history, it opens the possibility to optimize canister loadings from the point-of-view of criticality, decay heat and emission sources. Copyright © 2018 Elsevier B.V. All rights reserved.

Cross section of the 197Au(n,2n)196Au reaction

NASA Astrophysics Data System (ADS)

Kalamara, A.; Vlastou, R.; Kokkoris, M.; Diakaki, M.; Serris, M.; Patronis, N.; Axiotis, M.; Lagoyannis, A.

2017-09-01

The 197Au(n,2n)196Au reaction cross section has been measured at two energies, namely at 17.1 MeV and 20.9 MeV, by means of the activation technique, relative to the 27Al(n,α)24Na reference reaction cross section. Quasi-monoenergetic neutron beams were produced at the 5.5 MV Tandem T11/25 accelerator laboratory of NCSR "Demokritos", by means of the 3H(d,n)4He reaction, implementing a new Ti-tritiated target of ˜ 400 GBq activity. The induced γ-ray activity at the targets and reference foils has been measured with HPGe detectors. The cross section for the population of the second isomeric (12-) state m2 of 196Au was independently determined. Auxiliary Monte Carlo simulations were performed using the MCNP code. The present results are in agreement with previous experimental data and with theoretical calculations of the measured reaction cross sections, which were carried out with the use of the EMPIRE code.

Neutron Capture gamma ENDF libraries for modeling and identification of neutron sources

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

Sleaford, B

2007-10-29

There are a number of inaccuracies and data omissions with respect to gammas from neutron capture in the ENDF libraries used as field reference information and by modeling codes used in JTOT. As the use of Active Neutron interrogation methods is expanded, these shortfalls become more acute. A new, more accurate and complete evaluated experimental database of gamma rays (over 35,000 lines for 262 isotopes up to U so far) from thermal neutron capture has recently become available from the IAEA. To my knowledge, none of this new data has been installed in ENDF libraries and disseminated. I propose tomore » upgrade libraries of {sup 184,186}W, {sup 56}Fe, {sup 204,206,207}Pb, {sup 104}Pd, and {sup 19}F the 1st year. This will involve collaboration with Richard Firestone at LBL in evaluating the data and installing it in the libraries. I will test them with the transport code MCNP5.« less

Comparison of UWCC MOX fuel measurements to MCNP-REN calculations

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

Abhold, M.; Baker, M.; Jie, R.

1998-12-31

The development of neutron coincidence counting has greatly improved the accuracy and versatility of neutron-based techniques to assay fissile materials. Today, the shift register analyzer connected to either a passive or active neutron detector is widely used by both domestic and international safeguards organizations. The continued development of these techniques and 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, as it is currently used, fails to accurately predict detector response in highly multiplying mediums such as mixed-oxide (MOX) lightmore » water reactor fuel assemblies. For this reason, efforts have been made to modify the currently used Monte Carlo codes and to develop new analytical methods so that this model is not required to predict detector response. The authors describe their efforts to modify a widely used Monte Carlo code for this purpose and also compare calculational results with experimental measurements.« less

SABRINA - an interactive geometry modeler for MCNP

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

West, J.T.; Murphy, J.

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

Analysis of a Neutronic Experiment on a Simulated Mercury Spallation Neutron Target Assembly Bombarded by Giga-Electron-Volt Protons

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

Maekawa, Fujio; Meigo, Shin-ichiro; Kasugai, Yoshimi

2005-05-15

A neutronic benchmark experiment on a simulated spallation neutron target assembly was conducted by using the Alternating Gradient Synchrotron at Brookhaven National Laboratory and was analyzed to investigate the prediction capability of Monte Carlo simulation codes used in neutronic designs of spallation neutron sources. The target assembly consisting of a mercury target, a light water moderator, and a lead reflector was bombarded by 1.94-, 12-, and 24-GeV protons, and the fast neutron flux distributions around the target and the spectra of thermal neutrons leaking from the moderator were measured in the experiment. In this study, the Monte Carlo particle transportmore » simulation codes NMTC/JAM, MCNPX, and MCNP-4A with associated cross-section data in JENDL and LA-150 were verified based on benchmark analysis of the experiment. As a result, all the calculations predicted the measured quantities adequately; calculated integral fluxes of fast and thermal neutrons agreed approximately within {+-}40% with the experiments although the overall energy range encompassed more than 12 orders of magnitude. Accordingly, it was concluded that these simulation codes and cross-section data were adequate for neutronics designs of spallation neutron sources.« less

SU-F-T-111: Investigation of the Attila Deterministic Solver as a Supplement to Monte Carlo for Calculating Out-Of-Field Radiotherapy Dose

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

Mille, M; Lee, C; Failla, G

Purpose: To use the Attila deterministic solver as a supplement to Monte Carlo for calculating out-of-field organ dose in support of epidemiological studies looking at the risks of second cancers. Supplemental dosimetry tools are needed to speed up dose calculations for studies involving large-scale patient cohorts. Methods: Attila is a multi-group discrete ordinates code which can solve the 3D photon-electron coupled linear Boltzmann radiation transport equation on a finite-element mesh. Dose is computed by multiplying the calculated particle flux in each mesh element by a medium-specific energy deposition cross-section. The out-of-field dosimetry capability of Attila is investigated by comparing averagemore » organ dose to that which is calculated by Monte Carlo simulation. The test scenario consists of a 6 MV external beam treatment of a female patient with a tumor in the left breast. The patient is simulated by a whole-body adult reference female computational phantom. Monte Carlo simulations were performed using MCNP6 and XVMC. Attila can export a tetrahedral mesh for MCNP6, allowing for a direct comparison between the two codes. The Attila and Monte Carlo methods were also compared in terms of calculation speed and complexity of simulation setup. A key perquisite for this work was the modeling of a Varian Clinac 2100 linear accelerator. Results: The solid mesh of the torso part of the adult female phantom for the Attila calculation was prepared using the CAD software SpaceClaim. Preliminary calculations suggest that Attila is a user-friendly software which shows great promise for our intended application. Computational performance is related to the number of tetrahedral elements included in the Attila calculation. Conclusion: Attila is being explored as a supplement to the conventional Monte Carlo radiation transport approach for performing retrospective patient dosimetry. The goal is for the dosimetry to be sufficiently accurate for use in retrospective epidemiological investigations.« less

Voltage-gated potassium channel-associated limbic encephalitis in the West of Scotland: case reports and literature review.

PubMed

Reid, J M; Foley, P; Willison, H J

2009-11-01

The syndrome of limbic encephalitis (LE) associated with antibodies against voltage-gated potassium channels (VGKC-LE) has recently been described. The number of published cases is however small. We therefore aimed to review all cases seen at our centre and compare with published cases. Retrospective cases of VGKC-LE were identified using a questionnaire to Neurologists at the Southern General hospital, Glasgow, and by reviewing patients with a positive VGKC antibody test (2002-2007). Case-note review of identified cases and a literature review of all published cases of VGKC-LE were performed. Seven cases were identified (four female, age range 51-81). Patients presented sub-acutely with seizures and anterograde memory loss. Five patients had medial temporal lobe change on cranial imaging. No paraneoplastic cases were identified. 5/7 patients made some improvement with immunotherapy. In 2006, 3/18 (17%) patients with a coded discharge of encephalitis were diagnosed with VGKC-LE. The literature review revealed 40 patients with VGKC-LE. Age, gender or VGKC level did not predict likelihood for a significant recovery. Patients treated < or =5 months of symptom onset with immunotherapy were more likely to make a significant recovery (83% vs. 45%, p=0.04). VGKC-LE is being increasingly diagnosed and is best identified early and treated with immunotherapy to offer the greatest chance of recovery. This series and literature review expands the current published evidence in VGKC-LE.

Extended Range Underwater Loudhailer for Port Security Applications

DTIC Science & Technology

2006-06-01

and Subtitle Extended Range Underwater Loudhailer for Port Security Applications 6. Performing Organization Code Project No. 5903 7. Author( s ...used in the audio market . The name ‘RCA’ derives from the Radio Corporation of America, which introduced the design, by the early 1940s, to allow...Test June 2005 Test Range (yds) S pe ct ru m B an d Le ve l ( dB re 1 µ P a) S pe ct ru m B an d Le ve l ( dB re 1 µ P a) Figure 9

Measurement of Compression Factor and Error Sensitivity Factor of the Modified READ Facsimile Coding Technique.

DTIC Science & Technology

1980-08-01

Compression factor and error sensitivity together with statistical data have also been tabulated. This TIB is a companion drcument to NCS TIB’s 79-7...vu donner la priorit6 pour lour r~alisation. Chaque application est conf ice A un " chef do projet", responsable successivoment do sa conception. de son...pilote depend des r~sultats obtenus et fait I’objet d’une d~cision- de ’.a Direction Gdnerale. Ndanmoins, le chef do projet doit dOs le d~part consid~rer

Adjoint-Based Uncertainty Quantification with MCNP

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

Seifried, Jeffrey E.

2011-09-01

This work serves to quantify the instantaneous uncertainties in neutron transport simulations born from nuclear data and statistical counting uncertainties. Perturbation and adjoint theories are used to derive implicit sensitivity expressions. These expressions are transformed into forms that are convenient for construction with MCNP6, creating the ability to perform adjoint-based uncertainty quantification with MCNP6. These new tools are exercised on the depleted-uranium hybrid LIFE blanket, quantifying its sensitivities and uncertainties to important figures of merit. Overall, these uncertainty estimates are small (< 2%). Having quantified the sensitivities and uncertainties, physical understanding of the system is gained and some confidence inmore » the simulation is acquired.« less

NASA Lewis Research Center Workshop on Forced Response in Turbomachinery

NASA Technical Reports Server (NTRS)

Stefko, George L. (Compiler); Murthy, Durbha V. (Compiler); Morel, Michael (Compiler); Hoyniak, Dan (Compiler); Gauntner, Jim W. (Compiler)

1994-01-01

A summary of the NASA Lewis Research Center (LeRC) Workshop on Forced Response in Turbomachinery in August, 1993 is presented. It was sponsored by the following NASA organizations: Structures, Space Propulsion Technology, and Propulsion Systems Divisions of NASA LeRC and the Aeronautics and Advanced Concepts & Technology Offices of NASA Headquarters. In addition, the workshop was held in conjunction with the GUIde (Government/Industry/Universities) Consortium on Forced Response. The workshop was specifically designed to receive suggestions and comments from industry on current research at NASA LeRC in the area of forced vibratory response of turbomachinery blades which includes both computational and experimental approaches. There were eight presentations and a code demonstration. Major areas of research included aeroelastic response, steady and unsteady fluid dynamics, mistuning, and corresponding experimental work.

Deep learning for galaxy surface brightness profile fitting

NASA Astrophysics Data System (ADS)

Tuccillo, D.; Huertas-Company, M.; Decencière, E.; Velasco-Forero, S.; Domínguez Sánchez, H.; Dimauro, P.

2018-03-01

Numerous ongoing and future large area surveys (e.g. Dark Energy Survey, EUCLID, Large Synoptic Survey Telescope, Wide Field Infrared Survey Telescope) will increase by several orders of magnitude the volume of data that can be exploited for galaxy morphology studies. The full potential of these surveys can be unlocked only with the development of automated, fast, and reliable analysis methods. In this paper, we present DeepLeGATo, a new method for 2-D photometric galaxy profile modelling, based on convolutional neural networks. Our code is trained and validated on analytic profiles (HST/CANDELS F160W filter) and it is able to retrieve the full set of parameters of one-component Sérsic models: total magnitude, effective radius, Sérsic index, and axis ratio. We show detailed comparisons between our code and GALFIT. On simulated data, our method is more accurate than GALFIT and ˜3000 time faster on GPU (˜50 times when running on the same CPU). On real data, DeepLeGATo trained on simulations behaves similarly to GALFIT on isolated galaxies. With a fast domain adaptation step made with the 0.1-0.8 per cent the size of the training set, our code is easily capable to reproduce the results obtained with GALFIT even on crowded regions. DeepLeGATo does not require any human intervention beyond the training step, rendering it much automated than traditional profiling methods. The development of this method for more complex models (two-component galaxies, variable point spread function, dense sky regions) could constitute a fundamental tool in the era of big data in astronomy.

Monte Carlo source simulation technique for solution of interference reactions in INAA experiments: a preliminary report

NASA Astrophysics Data System (ADS)

Allaf, M. Athari; Shahriari, M.; Sohrabpour, M.

2004-04-01

A new method using Monte Carlo source simulation of interference reactions in neutron activation analysis experiments has been developed. The neutron spectrum at the sample location has been simulated using the Monte Carlo code MCNP and the contributions of different elements to produce a specified gamma line have been determined. The produced response matrix has been used to measure peak areas and the sample masses of the elements of interest. A number of benchmark experiments have been performed and the calculated results verified against known values. The good agreement obtained between the calculated and known values suggests that this technique may be useful for the elimination of interference reactions in neutron activation analysis.

Background studies for the MINER Coherent Neutrino Scattering reactor experiment

NASA Astrophysics Data System (ADS)

Agnolet, G.; Baker, W.; Barker, D.; Beck, R.; Carroll, T. J.; Cesar, J.; Cushman, P.; Dent, J. B.; De Rijck, S.; Dutta, B.; Flanagan, W.; Fritts, M.; Gao, Y.; Harris, H. R.; Hays, C. C.; Iyer, V.; Jastram, A.; Kadribasic, F.; Kennedy, A.; Kubik, A.; Lang, K.; Mahapatra, R.; Mandic, V.; Marianno, C.; Martin, R. D.; Mast, N.; McDeavitt, S.; Mirabolfathi, N.; Mohanty, B.; Nakajima, K.; Newhouse, J.; Newstead, J. L.; Ogawa, I.; Phan, D.; Proga, M.; Rajput, A.; Roberts, A.; Rogachev, G.; Salazar, R.; Sander, J.; Senapati, K.; Shimada, M.; Soubasis, B.; Strigari, L.; Tamagawa, Y.; Teizer, W.; Vermaak, J. I. C.; Villano, A. N.; Walker, J.; Webb, B.; Wetzel, Z.; Yadavalli, S. A.

2017-05-01

The proposed Mitchell Institute Neutrino Experiment at Reactor (MINER) experiment at the Nuclear Science Center at Texas A&M University will search for coherent elastic neutrino-nucleus scattering within close proximity (about 2 m) of a 1 MW TRIGA nuclear reactor core using low threshold, cryogenic germanium and silicon detectors. Given the Standard Model cross section of the scattering process and the proposed experimental proximity to the reactor, as many as 5-20 events/kg/day are expected. We discuss the status of preliminary measurements to characterize the main backgrounds for the proposed experiment. Both in situ measurements at the experimental site and simulations using the MCNP and GEANT4 codes are described. A strategy for monitoring backgrounds during data taking is briefly discussed.

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

Van Der Marck, S. C.

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

A Monte Carlo simulation and setup optimization of output efficiency to PGNAA thermal neutron using 252Cf neutrons

NASA Astrophysics Data System (ADS)

Zhang, Jin-Zhao; Tuo, Xian-Guo

2014-07-01

We present the design and optimization of a prompt γ-ray neutron activation analysis (PGNAA) thermal neutron output setup based on Monte Carlo simulations using MCNP5 computer code. In these simulations, the moderator materials, reflective materials, and structure of the PGNAA 252Cf neutrons of thermal neutron output setup are optimized. The simulation results reveal that the thin layer paraffin and the thick layer of heavy water moderating effect work best for the 252Cf neutron spectrum. Our new design shows a significantly improved performance of the thermal neutron flux and flux rate, that are increased by 3.02 times and 3.27 times, respectively, compared with the conventional neutron source design.

Case Studies of Navigation Channel and Port Sedimentation

DTIC Science & Technology

2011-01-14

Engineering Supplement to Limited Reevaluation Report, Vol. II (1995) 1/14/2011 18 MCNP  Hypotheses Why is channel shoaling more than anticipated? (1...activities have modified shoaling patterns and magnitudes (trawling, dredging and disposal, subsidence/fluid withdrawal). MCNP Hypotheses Why is channel

ARES: A System for Real-Time Operational and Tactical Decision Support

DTIC Science & Technology

1986-12-01

In B]LE LCLGf. 9 NAVAL POSTGRADUATE SCHOOL Monterey, California Vi,-. %*.. THESIS - ’ A RE S A SYSTEM -OR REAL- 1I I .-.. --- OPERATIONAL AND...able) aval Postgraduate School 54 Naval Postgraduate School NN DRESS (City,. State,. and ZIP Code) 7b ADDRESS (City,. State,. and ZIP Code...SUBJECT TERMS (Continue on reverse if necessaty and identify by block number) LD GROUP SUB-GROUP Decision Support System, Logistics Model, Operational

Theory and methods for measuring the effective multiplication constant in ADS

NASA Astrophysics Data System (ADS)

Rugama Saez, Yolanda

2001-10-01

In the thesis an absolute measurements technique for the subcriticality determination is presented. The ADS is a hybrid system where a subcritical system is fed by a proton accelerator. There are different proposals to define an ADS, one is to use plutonium and minor actinides from power plants waste as fuel to be transmuted into non radioactive isotopes (transmuter/burner, ATW). Another proposal is to use a Th232-U233 cycle (Energy Amplifier), being that thorium is an interesting and abundant fertile isotope. The development of accelerator driven systems (ADS) requires the development of methods to monitor and control the subcriticality of this kind of system without interfering with its normal operation mode. With this finality, we have applied noise analysis techniques that allow us to characterise the system when it is operating. The method presented in this thesis is based on the stochastic neutron and photon transport theory that can be implemented by presently available neutron/photon transport codes. In this work, first we analyse the stochastic transport theory which has been applied to define a parameter to determine the subcritical reactivity monitoring measurements. Finally we give the main limitations and recommendations for these subcritical monitoring methodology. As a result of the theoretical methodology, done in the first part of this thesis, a monitoring measurement technique has been developed and verified using two coupled Monte Carlo programs. The first one, LAHET, simulates the spallation collisions and the high energy transport and the other, MCNP-DSP, is used to estimate the counting statistics from a neutron/photon ray counter in a fissile system, as well as the transport for neutron with energies less than 20 MeV. From the coupling of both codes we developed the LAHET/MCNP-DSP code which, has the capability to simulate the total process in the ADS from the proton interaction to the signal detector processing. In these simulations, we compute the cross power spectral densities between pairs of detectors located inside the system which, is defined as the measured parameter. From the comparison of the theoretical predictions with the Monte Carlo simulations, we obtain some practical and simple methods to determine the system multiplication constant. (Abstract shortened by UMI.)

HR Public meeting

ScienceCinema

None

2018-05-24

Cher(e)s collègues, En collaboration avec le Département HR, le Directeur général a le plaisir de vous convier à une réunion publique qui se tiendra le vendredi 25 juin 2010 à 9h30 dans l’Amphithéâtre principal (Bâtiment 500)*. Un café d’accueil y sera servi à partir de 9h. Cette réunion abordera les thèmes suivants : Valeurs de l’Organisation (Directeur général); Code de Conduite (Directeur général / Anne-Sylvie Catherin); Création du nouveau rôle d’Ombudsperson (Vincent Vuillemin) Ces présentations seront suivies d’une séance de questions-réponses. Nous espérons vous retrouver nombreux le 25 juin ! Meilleures salutations, Anne-Sylvie Catherin Chef du Département des Ressources humaines *Cette réunion sera retransmise simultanément dans l’Amphithéâtre BE de Prévessin (Bâtiment 864) et également disponible à l’adresse suivante : http://webcast.cern.ch [Dear colleagues, In collaboration with HR Department, the Director-General would like to invite you to an information meeting which will be held on Friday 25 June 2010 at 9:30 am in the Main Auditorium (Building 500)*. A welcome coffee will be available from 9:00 am. During this meeting, information will be given about: Organization’s values (Director-General); Code of Conduct (Director-General / Anne-Sylvie Catherin); New Ombudsperson role (Vincent Vuillemin); These presentations will be followed by a questions & answers session. We look forward to seeing you all on 25 June! Best regards, Anne-Sylvie Catherin Head, Human Resources Department *This meeting will be simultaneously retransmitted in BE Auditorium (Building 864) and available at the following address: http://webcast.cern.ch.

A New Source Biasing Approach in ADVANTG

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

Bevill, Aaron M; Mosher, Scott W

2012-01-01

The ADVANTG code has been developed at Oak Ridge National Laboratory to generate biased sources and weight window maps for MCNP using the CADIS and FW-CADIS methods. In preparation for an upcoming RSICC release, a new approach for generating a biased source has been developed. This improvement streamlines user input and improves reliability. Previous versions of ADVANTG generated the biased source from ADVANTG input, writing an entirely new general fixed-source definition (SDEF). Because volumetric sources were translated into SDEF-format as a finite set of points, the user had to perform a convergence study to determine whether the number of sourcemore » points used accurately represented the source region. Further, the large number of points that must be written in SDEF-format made the MCNP input and output files excessively long and difficult to debug. ADVANTG now reads SDEF-format distributions and generates corresponding source biasing cards, eliminating the need for a convergence study. Many problems of interest use complicated source regions that are defined using cell rejection. In cell rejection, the source distribution in space is defined using an arbitrarily complex cell and a simple bounding region. Source positions are sampled within the bounding region but accepted only if they fall within the cell; otherwise, the position is resampled entirely. When biasing in space is applied to sources that use rejection sampling, current versions of MCNP do not account for the rejection in setting the source weight of histories, resulting in an 'unfair game'. This problem was circumvented in previous versions of ADVANTG by translating volumetric sources into a finite set of points, which does not alter the mean history weight ({bar w}). To use biasing parameters without otherwise modifying the original cell-rejection SDEF-format source, ADVANTG users now apply a correction factor for {bar w} in post-processing. A stratified-random sampling approach in ADVANTG is under development to automatically report the correction factor with estimated uncertainty. This study demonstrates the use of ADVANTG's new source biasing method, including the application of {bar w}.« less

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

DTIC Science & Technology

2008-09-01

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

Neutronics Analyses of the Minimum Original HEU TREAT Core

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

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

2014-04-01

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

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

NASA Astrophysics Data System (ADS)

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

2017-08-01

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

Deaths Due to Use of Lethal Force by Law Enforcement: Findings From the National Violent Death Reporting System, 17 U.S. States, 2009-2012.

PubMed

DeGue, Sarah; Fowler, Katherine A; Calkins, Cynthia

2016-11-01

Several high-profile cases in the U.S. have drawn public attention to the use of lethal force by law enforcement (LE), yet research on such fatalities is limited. Using data from a public health surveillance system, this study examined the characteristics and circumstances of these violent deaths to inform prevention. All fatalities (N=812) resulting from use of lethal force by on-duty LE from 2009 to 2012 in 17 U.S. states were examined using National Violent Death Reporting System data. Case narratives were coded for additional incident circumstances. Victims were majority white (52%) but disproportionately black (32%) with a fatality rate 2.8 times higher among blacks than whites. Most victims were reported to be armed (83%); however, black victims were more likely to be unarmed (14.8%) than white (9.4%) or Hispanic (5.8%) victims. Fatality rates among military veterans/active duty service members were 1.4 times greater than among their civilian counterparts. Four case subtypes were examined based on themes that emerged in incident narratives: about 22% of cases were mental health related; 18% were suspected "suicide by cop" incidents, with white victims more likely than black or Hispanic victims to die in these circumstances; 14% involved intimate partner violence; and about 6% were unintentional deaths due to LE action. Another 53% of cases were unclassified and did not fall into a coded subtype. Regression analyses identified victim and incident characteristics associated with each case subtype and unclassified cases. Knowledge about circumstances of deaths due to the use of lethal force can inform the development of prevention strategies, improve risk assessment, and modify LE response to increase the safety of communities and officers and prevent fatalities associated with LE intervention. Copyright © 2016. Published by Elsevier Inc.

Calculation of Absorbed Dose in Target Tissue and Equivalent Dose in Sensitive Tissues of Patients Treated by BNCT Using MCNP4C

NASA Astrophysics Data System (ADS)

Zamani, M.; Kasesaz, Y.; Khalafi, H.; Pooya, S. M. Hosseini

Boron Neutron Capture Therapy (BNCT) is used for treatment of many diseases, including brain tumors, in many medical centers. In this method, a target area (e.g., head of patient) is irradiated by some optimized and suitable neutron fields such as research nuclear reactors. Aiming at protection of healthy tissues which are located in the vicinity of irradiated tissue, and based on the ALARA principle, it is required to prevent unnecessary exposure of these vital organs. In this study, by using numerical simulation method (MCNP4C Code), the absorbed dose in target tissue and the equiavalent dose in different sensitive tissues of a patiant treated by BNCT, are calculated. For this purpose, we have used the parameters of MIRD Standard Phantom. Equiavelent dose in 11 sensitive organs, located in the vicinity of target, and total equivalent dose in whole body, have been calculated. The results show that the absorbed dose in tumor and normal tissue of brain equal to 30.35 Gy and 0.19 Gy, respectively. Also, total equivalent dose in 11 sensitive organs, other than tumor and normal tissue of brain, is equal to 14 mGy. The maximum equivalent doses in organs, other than brain and tumor, appear to the tissues of lungs and thyroid and are equal to 7.35 mSv and 3.00 mSv, respectively.

Efficient Geometry and Data Handling for Large-Scale Monte Carlo - Thermal-Hydraulics Coupling

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard

2014-06-01

Detailed coupling of thermal-hydraulics calculations to Monte Carlo reactor criticality calculations requires each axial layer of each fuel pin to be defined separately in the input to the Monte Carlo code in order to assign to each volume the temperature according to the result of the TH calculation, and if the volume contains coolant, also the density of the coolant. This leads to huge input files for even small systems. In this paper a methodology for dynamical assignment of temperatures with respect to cross section data is demonstrated to overcome this problem. The method is implemented in MCNP5. The method is verified for an infinite lattice with 3x3 BWR-type fuel pins with fuel, cladding and moderator/coolant explicitly modeled. For each pin 60 axial zones are considered with different temperatures and coolant densities. The results of the axial power distribution per fuel pin are compared to a standard MCNP5 run in which all 9x60 cells for fuel, cladding and coolant are explicitly defined and their respective temperatures determined from the TH calculation. Full agreement is obtained. For large-scale application the method is demonstrated for an infinite lattice with 17x17 PWR-type fuel assemblies with 25 rods replaced by guide tubes. Again all geometrical detailed is retained. The method was used in a procedure for coupled Monte Carlo and thermal-hydraulics iterations. Using an optimised iteration technique, convergence was obtained in 11 iteration steps.

Material Assessment for ITER's Collective Thomson Scattering first mirror

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

Santos, R.; Policarpo, H.; Goncalves, B.

2015-07-01

The International Thermonuclear Energy Reactor (ITER) Collective Thomson Scattering (CTS) system is a diagnostic instrument that measures plasma density and velocity through Thomson scattering of microwave radiation. Some of the key components of the CTS are quasi-optical mirrors that are used to produce astigmatic beam patterns, which have impact on the strength and spatial resolution of the diagnostic signal. The mirrors are exposed to neutron radiation, which may alter the quality of the signal received. In this work, three different materials (molybdenum (Mo), stainless steel 316 (SS-316) and tungsten (W)) are considered for the first mirror of the CTS. Themore » objective is to access which of the material studied are best suited for this mirror, considering different neutron radiation loads simulated scenarios defined by ITER, based on the resultant stresses and temperature distributions. For it, the neutron irradiation, and subsequent heat-load on the mirrors are simulated using the Monte Carlo N-Particle (MCNP) code. Based on the MCNP heat-load results, transient thermal-structural Finite Element Analysis (FEA) of the mirror over a 400 s discharge, with and without cooling on the rear side, are conducted using in commercial FEA software ANSYS. Results show that of the tested materials Mo and W are the most suitable material for this application. Even though, this study does not yet consider the variation of the material properties with temperature, it presents a quick initial satisfactory assessment that may be considered in subsequent and more complex analysis. (authors)« less

Rapid Acute Dose Assessment Using MCNP6

NASA Astrophysics Data System (ADS)

Owens, Andrew Steven

Acute radiation doses due to physical contact with a high-activity radioactive source have proven to be an occupational hazard. Multiple radiation injuries have been reported due to manipulating a radioactive source with bare hands or by placing a radioactive source inside a shirt or pants pocket. An effort to reconstruct the radiation dose must be performed to properly assess and medically manage the potential biological effects from such doses. Using the reference computational phantoms defined by the International Commission on Radiological Protection (ICRP) and the Monte Carlo N-Particle transport code (MCNP6), dose rate coefficients are calculated to assess doses for common acute doses due to beta and photon radiation sources. The research investigates doses due to having a radioactive source in either a breast pocket or pants back pocket. The dose rate coefficients are calculated for discrete energies and can be used to interpolate for any given energy of photon or beta emission. The dose rate coefficients allow for quick calculation of whole-body dose, organ dose, and/or skin dose if the source, activity, and time of exposure are known. Doses are calculated with the dose rate coefficients and compared to results from the International Atomic Energy Agency (IAEA) reports from accidents that occurred in Gilan, Iran and Yanango, Peru. Skin and organ doses calculated with the dose rate coefficients appear to agree, but there is a large discrepancy when comparing whole-body doses assessed using biodosimetry and whole-body doses assessed using the dose rate coefficients.

Hypersonic Shock Interactions About a 25 deg/65 deg Sharp Double Cone

NASA Technical Reports Server (NTRS)

Moss, James N.; LeBeau, Gerald J.; Glass, Christopher E.

2002-01-01

This paper presents the results of a numerical study of shock interactions resulting from Mach 10 air flow about a sharp double cone. Computations are made with the direct simulation Monte Carlo (DSMC) method by using two different codes: the G2 code of Bird and the DAC (DSMC Analysis Code) code of LeBeau. The flow conditions are the pretest nominal free-stream conditions specified for the ONERA R5Ch low-density wind tunnel. The focus is on the sensitivity of the interactions to grid resolution while providing information concerning the flow structure and surface results for the extent of separation, heating, pressure, and skin friction.

The NASA-LeRC wind turbine sound prediction code

NASA Technical Reports Server (NTRS)

Viterna, L. A.

1981-01-01

Development of the wind turbine sound prediction code began as part of an effort understand and reduce the noise generated by Mod-1. Tone sound levels predicted with this code are in good agreement with measured data taken in the vicinity Mod-1 wind turbine (less than 2 rotor diameters). Comparison in the far field indicates that propagation effects due to terrain and atmospheric conditions may amplify the actual sound levels by 6 dB. Parametric analysis using the code shows that the predominant contributors to Mod-1 rotor noise are (1) the velocity deficit in the wake of the support tower, (2) the high rotor speed, and (3) off-optimum operation.

Analysis of the Temporal Response of Coupled Asymmetrical Zero-Power Subcritical Bare Metal Reactor Systems

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

Klain, Kimberly L.

The behavior of symmetrical coupled-core systems has been extensively studied, yet there is a dearth of research on asymmetrical systems due to the increased complexity of the analysis of such systems. In this research, the multipoint kinetics method is applied to asymmetrical zeropower, subcritical, bare metal reactor systems. Existing research on asymmetrical reactor systems assumes symmetry in the neutronic coupling; however, it will be shown that this cannot always be assumed. Deep subcriticality adds another layer of complexity and requires modification of the multipoint kinetics equations to account for the effect of the external neutron source. A modified set ofmore » multipoint kinetics equations is derived with this in mind. Subsequently, the Rossi-alpha equations are derived for a two-region asymmetrical reactor system. The predictive capabilities of the radiation transport code MCNP6 for neutron noise experiments are shown in a comparison to the results of a series of Rossi-alpha measurements performed by J. Mihalczo utilizing a coupled set of symmetrical bare highly-enriched uranium (HEU) cylinders. The ptrac option within MCNP6 can generate time-tagged counts in a cell (list-mode data). The list-mode data can then be processed similarly to measured data to obtain values for system parameters such as the dual prompt neutron decay constants observable in a coupled system. The results from the ptrac simulations agree well with the historical measured values. A series of case studies are conducted to study the effects of geometrical asymmetry in the coupling between two bare metal HEU cylinders. While the coupling behavior of symmetrical systems has been reported on extensively, that of asymmetrical systems remains sparse. In particular, it appears that there has been no previous research in obtaining the coupling time constants for asymmetrically-coupled systems. The difficulty in observing such systems is due in part to the inability to determine the individual coupling coefficients from measurement: unlike the symmetrical cases, only the product of the values can be obtained. A method is proposed utilizing MCNP6 tally ratios to separate the coupling coefficients for such systems. This work provides insight into the behavior of asymmetrically-coupled systems as the separation distance between the two cores is changed and also as the asymmetry is increased. As the asymmetry increases, both the slower and the faster observable prompt neutron decay constants increase in magnitude. The coupling time constants are determined from the measured decay constants. As the separation distance increases, both coupling coefficients decrease as expected. Based on these findings, an effective computational method utilizing MCNP6 and the Rossialpha technique can be applied to the prediction of asymmetrical coupled system measurements.« less

A Selection of Experimental Test Cases for the Validation of CFD Codes (Recueil de cas d’essai experimentaux pour la validation des codes de l’aerodynamique numerique). Volume 1

DTIC Science & Technology

1994-08-01

volume H1. Le rapport ext accompagnt5 doun jeo die disqoettex contenant les donn~es appropri~es Li bous let cas d’essai. (’es disqoettes sont disponibles ...GERMANY PURPL’Sb OF THE TESi The tests are part of a larger effort to establish a database of experimental measurements for missile configurations

Identification d'une loi thermo-élasto-viscoplastique en vue de la modélisation du laminage à chaud du cuivre

NASA Astrophysics Data System (ADS)

Moureaux, P.; Moto Mpong, S.; Remy, M.; Bouffioux, C.; Lecomte-Beckers, J.; Habraken, A. M.

2002-12-01

la mise au point d'un modèle de simulation de la dernière passe du laminage à chaud du cuivre ne présente à priori pas de problème du point de vue numérique pour un code d'éléments finis non linéaire. La collecte d'informations précises tant sur le procédé industriel que sur le comportement du matériau est par contre une opération non triviale. Cet article présente les diverses méthodes expérimentales mises en œuvre pour caractériser le matériau : essais de compression à chaud, mesures d'analyse thermique différentielle, essais de dilatométrie et de diffusivité. Les méthodes permettant d'identifier les paramètres de la loi élasto-visco-plastique de type Norton-Hoff à partir des essais sont présentées et une analyse bibliographique investigue le problème de la détermination du module de Young à haute température. Tant les hypothèses supplémentaires relatives au procédé et nécessaires au modèle que les résultats finaux sont résumés.

Small Engine Technology (Set) Task 8 Aeroelastic Prediction Methods

NASA Technical Reports Server (NTRS)

Eick, Chris D.; Liu, Jong-Shang

1998-01-01

AlliedSignal Engines, in cooperation with NASA LeRC, completed an evaluation of recently developed aeroelastic computer codes using test cases from the AlliedSignal Engines fan blisk database. Test data for this task includes strain gage, light probe, performance, and steady-state pressure information obtained for conditions where synchronous or flutter vibratory conditions were found to occur. Aeroelastic codes evaluated include the quasi 3-D UNSFLO (developed at MIT and modified to include blade motion by AlliedSignal), the 2-D FREPS (developed by NASA LeRC), and the 3-D TURBO-AE (under development at NASA LeRC). Six test cases each where flutter and synchronous vibrations were found to occur were used for evaluation of UNSFLO and FREPS. In addition, one of the flutter cases was evaluated using TURBO-AE. The UNSFLO flutter evaluations were completed for 75 percent radial span and provided good agreement with the experimental test data. Synchronous evaluations were completed for UNSFLO but further enhancement needs to be added to the code before the unsteady pressures can be used to predict forced response vibratory stresses. The FREPS evaluations were hindered as the steady flow solver (SFLOW) was unable to converge to a solution for the transonic flow conditions in the fan blisk. This situation resulted in all FREPS test cases being attempted but no results were obtained during the present program. Currently, AlliedSignal is evaluating integrating FREPS with our existing steady flow solvers to bypass the SFLOW difficulties. ne TURBO-AE steady flow solution provided an excellent match with the AlliedSignal Engines calibrated DAWES 3-D viscous solver. Finally, the TURBO-AE unsteady analyses also matched experimental observations by predicting flutter for the single test case evaluated.

Overview and Current Status of Analyses of Potential LEU Design Concepts for TREAT

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

Connaway, H. M.; Kontogeorgakos, D. C.; Papadias, D. D.

2015-10-01

Neutronic and thermal-hydraulic analyses have been performed to evaluate the performance of different low-enriched uranium (LEU) fuel design concepts for the conversion of the Transient Reactor Test Facility (TREAT) from its current high-enriched uranium (HEU) fuel. TREAT is an experimental reactor developed to generate high neutron flux transients for the testing of nuclear fuels. The goal of this work was to identify an LEU design which can maintain the performance of the existing HEU core while continuing to operate safely. A wide variety of design options were considered, with a focus on minimizing peak fuel temperatures and optimizing the powermore » coupling between the TREAT core and test samples. Designs were also evaluated to ensure that they provide sufficient reactivity and shutdown margin for each control rod bank. Analyses were performed using the core loading and experiment configuration of historic M8 Power Calibration experiments (M8CAL). The Monte Carlo code MCNP was utilized for steady-state analyses, and transient calculations were performed with the point kinetics code TREKIN. Thermal analyses were performed with the COMSOL multi-physics code. Using the results of this study, a new LEU Baseline design concept is being established, which will be evaluated in detail in a future report.« less

Development of a technique using MCNPX code for determination of nitrogen content of explosive materials using prompt gamma neutron activation analysis method

NASA Astrophysics Data System (ADS)

Nasrabadi, M. N.; Bakhshi, F.; Jalali, M.; Mohammadi, A.

2011-12-01

Nuclear-based explosive detection methods can detect explosives by identifying their elemental components, especially nitrogen. Thermal neutron capture reactions have been used for detecting prompt gamma 10.8 MeV following radioactive neutron capture by 14N nuclei. We aimed to study the feasibility of using field-portable prompt gamma neutron activation analysis (PGNAA) along with improved nuclear equipment to detect and identify explosives, illicit substances or landmines. A 252Cf radio-isotopic source was embedded in a cylinder made of high-density polyethylene (HDPE) and the cylinder was then placed in another cylindrical container filled with water. Measurements were performed on high nitrogen content compounds such as melamine (C3H6N6). Melamine powder in a HDPE bottle was placed underneath the vessel containing water and the neutron source. Gamma rays were detected using two NaI(Tl) crystals. The results were simulated with MCNP4c code calculations. The theoretical calculations and experimental measurements were in good agreement indicating that this method can be used for detection of explosives and illicit drugs.

Methodology for the nuclear design validation of an Alternate Emergency Management Centre (CAGE)

NASA Astrophysics Data System (ADS)

Hueso, César; Fabbri, Marco; de la Fuente, Cristina; Janés, Albert; Massuet, Joan; Zamora, Imanol; Gasca, Cristina; Hernández, Héctor; Vega, J. Ángel

2017-09-01

The methodology is devised by coupling different codes. The study of weather conditions as part of the data of the site will determine the relative concentrations of radionuclides in the air using ARCON96. The activity in the air is characterized depending on the source and release sequence specified in NUREG-1465 by RADTRAD code, which provides results of the inner cloud source term contribution. Known activities, energy spectra are inferred using ORIGEN-S, which are used as input for the models of the outer cloud, filters and containment generated with MCNP5. The sum of the different contributions must meet the conditions of habitability specified by the CSN (Spanish Nuclear Regulatory Body) (TEDE <50 mSv and equivalent dose to the thyroid <500 mSv within 30 days following the accident doses) so that the dose is optimized by varying parameters such as CAGE location, flow filtering need for recirculation, thicknesses and compositions of the walls, etc. The results for the most penalizing area meet the established criteria, and therefore the CAGE building design based on the methodology presented is radiologically validated.

Control rod calibration and reactivity effects at the IPEN/MB-01 reactor

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

Pinto, Letícia Negrão; Gonnelli, Eduardo; Santos, Adimir dos

2014-11-11

Researches that aim to improve the performance of neutron transport codes and quality of nuclear cross section databases are very important to increase the accuracy of simulations and the quality of the analysis and prediction of phenomena in the nuclear field. In this context, relevant experimental data such as reactivity worth measurements are needed. Control rods may be made of several neutron absorbing materials that are used to adjust the reactivity of the core. For the reactor operation, these experimental data are also extremely important: with them it is possible to estimate the reactivity worth by the movement of themore » control rod, understand the reactor response at each rod position and to operate the reactor safely. This work presents a temperature correction approach for the control rod calibration problem. It is shown the control rod calibration data of the IPEN/MB-01 reactor, the integral and differential reactivity curves and a theoretical analysis, performed by the MCNP-5 reactor physics code, developed and maintained by Los Alamos National Laboratory, using the ENDF/B-VII.0 nuclear data library.« less

Visualization of nuclear particle trajectories in nuclear oil-well logging

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

Case, C.R.; Chiaramonte, J.M.

Nuclear oil-well logging measures specific properties of subsurface geological formations as a function of depth in the well. The knowledge gained is used to evaluate the hydrocarbon potential of the surrounding oil field. The measurements are made by lowering an instrument package into an oil well and slowly extracting it at a constant speed. During the extraction phase, neutrons or gamma rays are emitted from the tool, interact with the formation, and scatter back to the detectors located within the tool. Even though only a small percentage of the emitted particles ever reach the detectors, mathematical modeling has been verymore » successful in the accurate prediction of these detector responses. The two dominant methods used to model these devices have been the two-dimensional discrete ordinates method and the three-dimensional Monte Carlo method has routinely been used to investigate the response characteristics of nuclear tools. A special Los Alamos National Laboratory version of their standard MCNP Monte carlo code retains the details of each particle history of later viewing within SABRINA, a companion three-dimensional geometry modeling and debugging code.« less

Radiological Shielding Design for the Neutron High-Resolution Backscattering Spectrometer EMU at the OPAL Reactor

NASA Astrophysics Data System (ADS)

Ersez, Tunay; Esposto, Fernando; Souza, Nicolas R. de

2017-09-01

The shielding for the neutron high-resolution backscattering spectrometer (EMU) located at the OPAL reactor (ANSTO) was designed using the Monte Carlo code MCNP 5-1.60. The proposed shielding design has produced compact shielding assemblies, such as the neutron pre-monochromator bunker with sliding cylindrical block shields to accommodate a range of neutron take-off angles, and in the experimental area - shielding of neutron focusing guides, choppers, flight tube, backscattering monochromator, and additional shielding elements inside the Scattering Tank. These shielding assemblies meet safety and engineering requirements and cost constraints. The neutron dose rates around the EMU instrument were reduced to < 0.5 µSv/h and the gamma dose rates to a safe working level of ≤ 3 µSv/h.

Direct Discrete Method for Neutronic Calculations

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

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

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

Evaluation Of Shielding Efficacy Of A Ferrite Containing Ceramic Material

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

Verst, C.

2015-10-12

The shielding evaluation of the ferrite based Mitsuishi ceramic material has produced for several radiation sources and possible shielding sizes comparative dose attenuation measurements and simulated projections. High resolution gamma spectroscopy provided uncollided and scattered photon spectra at three energies, confirming theoretical estimates of the ceramic’s mass attenuation coefficient, μ/ρ. High level irradiation experiments were performed using Co-60, Cs-137, and Cf-252 sources to measure penetrating dose rates through steel, lead, concrete, and the provided ceramic slabs. The results were used to validate the radiation transport code MCNP6 which was then used to generate dose rate attenuation curves as a functionmore » of shielding material, thickness, and mass for photons and neutrons ranging in energy from 200 keV to 2 MeV.« less

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

Asadi, Somayeh; Masoudi, Seyed Farhad, E-mail: masoudi@kntu.ac.ir; Shahriari, Majid

In ophthalmic brachytherapy dosimetry, it is common to consider the water phantom as human eye anatomy. However, for better clinical analysis, there is a need for the dose determination in different parts of the eye. In this work, a full human eye is simulated with MCNP-4C code by considering all parts of the eye, i.e., the lens, cornea, retina, choroid, sclera, anterior chamber, optic nerve, and bulk of the eye comprising vitreous body and tumor. The average dose in different parts of this full model of the human eye is determined and the results are compared with the dose calculatedmore » in water phantom. The central axes depth dose and the dose in whole of the tumor for these 2 simulated eye models are calculated as well, and the results are compared.« less

Design of an epithermal column for BNCT based on D D fusion neutron facility

NASA Astrophysics Data System (ADS)

Durisi, E.; Zanini, A.; Manfredotti, C.; Palamara, F.; Sarotto, M.; Visca, L.; Nastasi, U.

2007-05-01

Boron Neutron Capture Therapy (BNCT) is currently performed on patients at nuclear reactors. At the same time the international BNCT community is engaged in the development of alternative facilities for in-hospital treatments. This paper investigates the potential of a novel high-output D-D neutron generator, developed at Lawrence Berkeley National Laboratory (CA, USA), for BNCT. The simulation code MCNP-4C is used to realize an accurate study of the epithermal column in view of the treatment of deep tumours. Different materials and Beam Shaping Assemblies (BSA) are investigated and an optimized configuration is proposed. The neutron beam quality is defined by the standard free beam parameters, calculated averaging over the collimator aperture. The results are discussed and compared with the performances of other facilities.

Use of a Si(Li) detector as β spectrometer.

PubMed

Dryák, P; Kovář, P

2014-05-01

The aim of this work is to demonstrate the capability of a Si(Li) detector for the measurement of β spectra, despite the energy absorption in air and in the Be window. A simple source holder fixes the source on the symmetry axis at 3mm from the detector window. The β-sources are produced by evaporation on a plastic backing plate. Absorbing materials between the source and the sensitive volume of the detector are 3 mm of air, a Be window, 0.1 μm Si and 20 nm of gold. A model of the detector was created for β spectra simulation using the MCNP 4A code. Experimental spectra of (14)C, (147)Pm, (204)Tl, (90)Sr/(90)Y were compared with simulated spectra. © 2013 Published by Elsevier Ltd.

Evaluation of the effective dose during BNCT at TRR thermal column epithermal facility.

PubMed

Jarahi, Hossein; Kasesaz, Yaser; Saleh-Koutahi, Seyed Mohsen

2016-04-01

An epithermal neutron beam has been designed for Boron neutron Capture Therapy (BNCT) at the thermal column of Tehran Research Reactor (TRR) recently. In this paper the whole body effective dose, as well as the equivalent doses of several organs have been calculated in this facility using MCNP4C Monte Carlo code. The effective dose has been calculated by using the absorbed doses determined for each individual organ, taking into account the radiation and tissue weighting factors. The ICRP 110 whole body male phantom has been used as a patient model. It was found that the effective dose during BNCT of a brain tumor is equal to 0.90Sv. This effective dose may induce a 4% secondary cancer risk. Copyright © 2016 Elsevier Ltd. All rights reserved.

Investigation on the reflector/moderator geometry and its effect on the neutron beam design in BNCT.

PubMed

Kasesaz, Y; Rahmani, F; Khalafi, H

2015-12-01

In order to provide an appropriate neutron beam for Boron Neutron Capture Therapy (BNCT), a special Beam Shaping Assembly (BSA) must be designed based on the neutron source specifications. A typical BSA includes moderator, reflector, collimator, thermal neutron filter, and gamma filter. In common BSA, the reflector is considered as a layer which covers the sides of the moderator materials. In this paper, new reflector/moderator geometries including multi-layer and hexagonal lattice have been suggested and the effect of them has been investigated by MCNP4C Monte Carlo code. It was found that the proposed configurations have a significant effect to improve the thermal to epithermal neutron flux ratio which is an important neutron beam parameter. Copyright © 2015 Elsevier Ltd. All rights reserved.

Characterization of the radiation environment at the UNLV accelerator facility during operation of the Varian M6 linac

NASA Astrophysics Data System (ADS)

Hodges, M.; Barzilov, A.; Chen, Y.; Lowe, D.

2016-10-01

The bremsstrahlung photon flux from the UNLV particle accelerator (Varian M6 model) was determined using MCNP5 code for 3 MeV and 6 MeV incident electrons. Human biological equivalent dose rates due to accelerator operation were evaluated using the photon flux with the flux-to-dose conversion factors. Dose rates were computed for the accelerator facility for M6 linac use under different operating conditions. The results showed that the use of collimators and linac internal shielding significantly reduced the dose rates throughout the facility. It was shown that the walls of the facility, in addition to the earthen berm enveloping the building, provide equivalent shielding to reduce dose rates outside to below the 2 mrem/h limit.

Thermal neutron radiative capture on cadmium as a counting technique at the INES beam line at ISIS: A preliminary investigation of detector cross-talk.

PubMed

Festa, G; Grazzi, F; Pietropaolo, A; Scherillo, A; Schooneveld, E M

2017-12-01

Experimental tests are presented that assess the cross-talk level among three scintillation detectors used as neutron counters exploiting the thermal neutron radiative capture on Cd. The measurements were done at the INES diffractometer operating at the ISIS spallation neutron source (Rutherford Appleton Laboratory, UK). These tests follow a preliminary set of measurements performed on the same instrument to study the effectiveness of this thermal neutron counting strategy in neutron diffraction measurements, typically performed on INES using squashed 3 He filled gas tubes. The experimental data were collected in two different geometrical configurations of the detectors and compared to results of Monte Carlo simulations, performed using the MCNP code. Copyright © 2017 Elsevier Ltd. All rights reserved.

Larvae of the genus Eleodes (Coleoptera, Tenebrionidae): matrix-based descriptions, cladistic analysis, and key to late instars

PubMed Central

Smith, Aaron D.; Dornburg, Rebecca; Wheeler, Quentin D.

2014-01-01

Abstract Darkling beetle larvae (Coleoptera, Tenebrionidae) are collectively referred to as false wireworms. Larvae from several species in the genus Eleodes are considered to be agricultural pests, though relatively little work has been done to associate larvae with adults of the same species and only a handful of species have been characterized in their larval state. Morphological characters from late instar larvae were examined and coded to produce a matrix in the server-based content management system mx. The resulting morphology matrix was used to produce larval species descriptions, reconstruct a phylogeny, and build a key to the species included in the matrix. Larvae are described for the first time for the following 12 species: Eleodes anthracinus Blaisdell, Eleodes carbonarius (Say), Eleodes caudiferus LeConte, Eleodes extricatus (Say), Eleodes goryi Solier, Eleodes hispilabris (Say), Eleodes nigropilosus LeConte, Eleodes pilosus Horn, Eleodes subnitens LeConte, Eleodes tenuipes Casey, Eleodes tribulus Thomas, and Eleodes wheeleri Aalbu, Smith & Triplehorn. The larval stage of Eleodes armatus LeConte is redescribed with additional characters to differentiate it from the newly described congeneric larvae. PMID:25009429

Review of Hybrid (Deterministic/Monte Carlo) Radiation Transport Methods, Codes, and Applications at Oak Ridge National Laboratory

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

Wagner, John C; Peplow, Douglas E.; Mosher, Scott W

2011-01-01

This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or moremore » localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(102-4), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications.« less

CAD-Based Shielding Analysis for ITER Port Diagnostics

NASA Astrophysics Data System (ADS)

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

2017-09-01

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

Metal Matrix Laminate Tailoring (MMLT) code: User's manual

NASA Technical Reports Server (NTRS)

Murthy, P. L. N.; Morel, M. R.; Saravanos, D. A.

1993-01-01

The User's Manual for the Metal Matrix Laminate Tailoring (MMLT) program is presented. The code is capable of tailoring the fabrication process, constituent characteristics, and laminate parameters (individually or concurrently) for a wide variety of metal matrix composite (MMC) materials, to improve the performance and identify trends or behavior of MMC's under different thermo-mechanical loading conditions. This document is meant to serve as a guide in the use of the MMLT code. Detailed explanations of the composite mechanics and tailoring analysis are beyond the scope of this document, and may be found in the references. MMLT was developed by the Structural Mechanics Branch at NASA Lewis Research Center (LeRC).

Ensemble des troubles causés par l'alcoolisation fœtale : lignes directrices canadiennes concernant le diagnostic

PubMed Central

Chudley, Albert E.; Conry, Julianne; Cook, Jocelynn L.; Loock, Christine; Rosales, Ted; LeBlanc, Nicole

2005-01-01

Résumé LE DIAGNOSTIC DE L'ENSEMBLE DES TROUBLES CAUSÉS PAR L'ALCOOLISATION FÉTALE (ETCAF) est complexe et l'élaboration de lignes directrices concernant le diagnostic est justifiée. Un sous-comité du Comité consultatif national de l'Agence de santé publique du Canada sur l'Ensemble des troubles causés par l'alcoolisation fœtale a examiné, analysé et intégré les méthodes de diagnostic actuelles afin de parvenir à une méthode de diagnostic normalisée faisant le consensus au Canada. L'objet du présent document est d'examiner et de clarifier l'utilisation des systèmes de diagnostic actuels et de formuler des recommandations quant à leur application pour le diagnostic des déficiences liées à l'ETCAF chez des individus de tous les âges. Les lignes directrices sont fondées sur un vaste consensus de praticiens et d'autres intervenants spécialisés dans le domaine. Ces lignes directrices ont été organisées en sept catégories, soit le dépistage et l'orientation vers les spécialistes, l'examen physique et le diagnostic différentiel, l'évaluation du comportement neurologique, le traitement et le suivi, les antécédents de consommation d'alcool de la mère pendant la grossesse, les critères du diagnostic pour le syndrome d'alcoolisation fœtale (SAF), SAF partiel et troubles neurologiques du développement liés à l'alcool, l'harmonisation de l'Institute of Medicine (IOM) et des approches du Code diagnostique à 4 chiffres. Le diagnostic exige une évaluation complète des antécédents, ainsi qu'un examen physique et du comportement neurologique, tout en recourant à une approche multidisciplinaire. Les présentes lignes directrices pour le diagnostic du syndrome d'alcoolisation fœtale et des déficiences qui y sont associées sont les premières à avoir été élaborées au Canada et elles sont fondées sur la consultation d'un grand éventail de spécialistes du diagnostic.

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

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

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

Second NASA Workshop on Wiring for Space Applications

NASA Technical Reports Server (NTRS)

1994-01-01

This document contains the proceedings of the Second NASA Workshop on Wiring for Space Applications held at NASA LeRC in Cleveland, OH, 6-7 Oct. 1993. The workshop was sponsored by NASA Headquarters Code QW Office of Safety and Mission Quality, Technical Standards Division and hosted by NASA LeRC, Power Technology Division, Electrical Components and Systems Branch. The workshop addressed key technology issues in the field of electrical power wiring for space applications. Speakers from government, industry, and academia presented and discussed topics on arc tracking phenomena, wiring system design, insulation constructions, and system protection. Presentation materials provided by the various speakers are included in this document.

Methodologie de conception numerique d'un ventilateur helico-centrifuge basee sur l'emploi du calcul meridien

NASA Astrophysics Data System (ADS)

Lallier-Daniels, Dominic

La conception de ventilateurs est souvent basée sur une méthodologie « essais/erreurs » d'amélioration de géométries existantes ainsi que sur l'expérience de design et les résultats expérimentaux cumulés par les entreprises. Cependant, cette méthodologie peut se révéler coûteuse en cas d'échec; même en cas de succès, des améliorations significatives en performance sont souvent difficiles, voire impossibles à obtenir. Le projet présent propose le développement et la validation d'une méthodologie de conception basée sur l'emploi du calcul méridien pour la conception préliminaire de turbomachines hélico-centrifuges (ou flux-mixte) et l'utilisation du calcul numérique d'écoulement fluides (CFD) pour la conception détaillée. La méthode de calcul méridien à la base du processus de conception proposé est d'abord présentée. Dans un premier temps, le cadre théorique est développé. Le calcul méridien demeurant fondamentalement un processus itératif, le processus de calcul est également présenté, incluant les méthodes numériques de calcul employée pour la résolution des équations fondamentales. Une validation du code méridien écrit dans le cadre du projet de maîtrise face à un algorithme de calcul méridien développé par l'auteur de la méthode ainsi qu'à des résultats de simulation numérique sur un code commercial est également réalisée. La méthodologie de conception de turbomachines développée dans le cadre de l'étude est ensuite présentée sous la forme d'une étude de cas pour un ventilateur hélico-centrifuge basé sur des spécifications fournies par le partenaire industriel Venmar. La méthodologie se divise en trois étapes: le calcul méridien est employé pour le pré-dimensionnement, suivi de simulations 2D de grilles d'aubes pour la conception détaillée des pales et finalement d'une analyse numérique 3D pour la validation et l'optimisation fine de la géométrie. Les résultats de calcul méridien sont en outre comparés aux résultats de simulation pour la géométrie 3D afin de valider l'emploi du calcul méridien comme outil de prédimensionnement. Mots-clés Turbomachines hélico-centrifuges, Calcul méridien, Simulation numérique d'écoulement fluide, Conception numérique de turbomachines.

NASA. Lewis Research Center Advanced Modulation and Coding Project: Introduction and overview

NASA Technical Reports Server (NTRS)

Budinger, James M.

1992-01-01

The Advanced Modulation and Coding Project at LeRC is sponsored by the Office of Space Science and Applications, Communications Division, Code EC, at NASA Headquarters and conducted by the Digital Systems Technology Branch of the Space Electronics Division. Advanced Modulation and Coding is one of three focused technology development projects within the branch's overall Processing and Switching Program. The program consists of industry contracts for developing proof-of-concept (POC) and demonstration model hardware, university grants for analyzing advanced techniques, and in-house integration and testing of performance verification and systems evaluation. The Advanced Modulation and Coding Project is broken into five elements: (1) bandwidth- and power-efficient modems; (2) high-speed codecs; (3) digital modems; (4) multichannel demodulators; and (5) very high-data-rate modems. At least one contract and one grant were awarded for each element.

The legumin gene family: structure of a B type gene of Vicia faba and a possible legumin gene specific regulatory element.

PubMed Central

Bäumlein, H; Wobus, U; Pustell, J; Kafatos, F C

1986-01-01

The field bean, Vicia faba L. var. minor, possesses two sub-families of 11 S legumin genes named A and B. We isolated from a genomic library a B-type gene (LeB4) and determined its primary DNA sequence. Gene LeB4 codes for a 484 amino acid residue prepropolypeptide, encompassing a signal peptide of 22 amino acid residues, an acidic, very hydrophilic alpha-chain of 281 residues and a basic, somewhat hydrophobic beta-chain of 181 residues. The latter two coding regions are immediately contiguous, but each is interrupted by a short intron. Type A legumin genes from soybean and pea are known to have introns in the same two positions, in addition to an extra intron (within the alpha-coding sequence). Sequence comparisons of legumin genes from these three plants revealed a highly conserved sequence element of at least 28 bp, centered at approximately 100 bp upstream of each cap site. The element is absent from the equivalent position of all non-legumin and other plant and fungal genes examined. We tentatively name this element "legumin box" and suggest that it may have a function in the regulation of legumin gene expression. PMID:3960730

HR Public meeting

ScienceCinema

Heuer, Rolf; Catherin, Anne-Sylvie; Vuillemin, Vin

2018-05-25

Cher(e)s collègues, En collaboration avec le Département HR, le Directeur général a le plaisir de vous convier à une réunion publique qui se tiendra le vendredi 25 juin 2010 à 9h30 dans l’Amphithéâtre principal (Bâtiment 500)*. Un café d’accueil y sera servi à partir de 9h. Cette réunion abordera les thèmes suivants : • Valeurs de l’Organisation (Directeur général) • Code de Conduite (Directeur général / Anne-Sylvie Catherin) • Création du nouveau rôle d’Ombudsperson (Vincent Vuillemin); Ces présentations seront suivies d’une séance de questions-réponses. Nous espérons vous retrouver nombreux le 25 juin ! Meilleures salutations, Anne-Sylvie Catherin Chef du Département des Ressources humaines *Cette réunion sera retransmise simultanément dans l’Amphithéâtre BE de Prévessin (Bâtiment 864) et également disponible à l’adresse suivante : http://webcast.cern.ch. Dear colleagues, In collaboration with HR Department, the Director-General would like to invite you to an information meeting which will be held on Friday 25 June 2010 at 9:30 am in the Main Auditorium (Building 500)*. A welcome coffee will be available from 9:00 am. During this meeting, information will be given about: • Organization’s values (Director-General) • Code of Conduct (Director-General / Anne-Sylvie Catherin) • New Ombudsperson role (Vincent Vuillemin) These presentations will be followed by a questions & answers session. We look forward to seeing you all on 25 June! Best regards, Anne-Sylvie Catherin Head, Human Resources Department. This meeting will be simultaneously retransmitted in BE Auditorium (Building 864) and available at the following address: http://webcast.cern.ch.

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

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

Rolison, L; Samant, S; Baciak, J

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

Words-in-Freedom and the Oral Tradition.

ERIC Educational Resources Information Center

Webster, Michael

1989-01-01

Explores how oral and print characteristics mesh or clash in "words-in-freedom," a form of visual poetry invented by Filippo Tommaso Marinetti. Analyzes Marinetti's poster-poem "Apres la Marne, Joffre visita le front en auto," highlighting the different natures of the two media and the coding difficulties occasioned by…

A D-D/D-T fusion reaction based neutron generator system for liver tumor BNCT

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

Koivunoro, H.; Lou, T.P.; Leung, K. N.

2003-04-02

Boron-neutron capture therapy (BNCT) is an experimental radiation treatment modality used for highly malignant tumor treatments. Prior to irradiation with low energetic neutrons, a 10B compound is located selectively in the tumor cells. The effect of the treatment is based on the high LET radiation released in the {sup 10}B(n,{alpha}){sup 7}Li reaction with thermal neutrons. BNCT has been used experimentally for brain tumor and melanoma treatments. Lately applications of other severe tumor type treatments have been introduced. Results have shown that liver tumors can also be treated by BNCT. At Lawrence Berkeley National Laboratory, various compact neutron generators based onmore » D-D or D-T fusion reactions are being developed. The earlier theoretical studies of the D-D or D-T fusion reaction based neutron generators have shown that the optimal moderator and reflector configuration for brain tumor BNCT can be created. In this work, the applicability of 2.5 MeV neutrons for liver tumor BNCT application was studied. The optimal neutron energy for external liver treatments is not known. Neutron beams of different energies (1eV < E < 100 keV) were simulated and the dose distribution in the liver was calculated with the MCNP simulation code. In order to obtain the optimal neutron energy spectrum with the D-D neutrons, various moderator designs were performed using MCNP simulations. In this article the neutron spectrum and the optimized beam shaping assembly for liver tumor treatments is presented.« less

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

DOE PAGES

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

2015-10-28

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

Novel paint design based on nanopowder to protection against X and gamma rays

PubMed Central

Movahedi, Mohammad Mehdi; Abdi, Adibe; Mehdizadeh, Alireza; Dehghan, Naser; Heidari, Emad; Masumi, Yusef; Abbaszadeh, Mojtaba

2014-01-01

Background: Lead-based shields are the standard method of intraoperative radiation protection in the radiology and nuclear medicine department. Human lead toxicity is well documented. The lead used is heavy, lacks durability, is difficult to launder, and its disposal is associated with environmental hazards. The aim of this study was to design a lead free paint for protection against X and gamma rays. Materials and Methods: In this pilot st we evaluated several types of nano metal powder that seemed to have good absorption. The Monte Carlo code, MCNP4C, was used to model the attenuation of X-ray photons in paints with different designs. Experimental measurements were carried out to assess the attenuation properties of each paint design. Results: Among the different nano metal powder, nano tungsten trioxide and nano tin dioxide were the two most appropriate candidates for making paint in diagnostic photon energy range. Nano tungsten trioxide (15%) and nano tin dioxide (85%) provided the best protection in both simulation and experiments. After this step, attempts were made to produce appropriate nano tungsten trioxide-nano tin dioxide paints. The density of this nano tungsten trioxide-nano tin dioxide paint was 4.2 g/cm3. The MCNP simulation and experimental measurements for HVL (Half-Value Layer) values of this shield at 100 kVp were 0.25 and 0.23 mm, respectively. Conclusions: The results showed the cost-effective lead-free paint can be a great power in absorbing the X-rays and gamma rays and it can be used instead of lead. PMID:24591777

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

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

Chandler, David; Ellis, Ronald James

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

The effect of tandem-ovoid titanium applicator on points A, B, bladder, and rectum doses in gynecological brachytherapy using 192Ir.

PubMed

Sadeghi, Mohammad Hosein; Sina, Sedigheh; Mehdizadeh, Amir; Faghihi, Reza; Moharramzadeh, Vahed; Meigooni, Ali Soleimani

2018-02-01

The dosimetry procedure by simple superposition accounts only for the self-shielding of the source and does not take into account the attenuation of photons by the applicators. The purpose of this investigation is an estimation of the effects of the tandem and ovoid applicator on dose distribution inside the phantom by MCNP5 Monte Carlo simulations. In this study, the superposition method is used for obtaining the dose distribution in the phantom without using the applicator for a typical gynecological brachytherapy (superposition-1). Then, the sources are simulated inside the tandem and ovoid applicator to identify the effect of applicator attenuation (superposition-2), and the dose at points A, B, bladder, and rectum were compared with the results of superposition. The exact dwell positions, times of the source, and positions of the dosimetry points were determined in images of a patient and treatment data of an adult woman patient from a cancer center. The MCNP5 Monte Carlo (MC) code was used for simulation of the phantoms, applicators, and the sources. The results of this study showed no significant differences between the results of superposition method and the MC simulations for different dosimetry points. The difference in all important dosimetry points was found to be less than 5%. According to the results, applicator attenuation has no significant effect on the calculated points dose, the superposition method, adding the dose of each source obtained by the MC simulation, can estimate the dose to points A, B, bladder, and rectum with good accuracy.

Using the MCNP Taylor series perturbation feature (efficiently) for shielding problems

NASA Astrophysics Data System (ADS)

Favorite, Jeffrey

2017-09-01

The Taylor series or differential operator perturbation method, implemented in MCNP and invoked using the PERT card, can be used for efficient parameter studies in shielding problems. This paper shows how only two PERT cards are needed to generate an entire parameter study, including statistical uncertainty estimates (an additional three PERT cards can be used to give exact statistical uncertainties). One realistic example problem involves a detailed helium-3 neutron detector model and its efficiency as a function of the density of its high-density polyethylene moderator. The MCNP differential operator perturbation capability is extremely accurate for this problem. A second problem involves the density of the polyethylene reflector of the BeRP ball and is an example of first-order sensitivity analysis using the PERT capability. A third problem is an analytic verification of the PERT capability.

An investigation of MCNP6.1 beryllium oxide S(α, β) cross sections

DOE PAGES

Sartor, Raymond F.; Glazener, Natasha N.

2016-03-08

In MCNP6.1, materials are constructed by identifying the constituent isotopes (or elements in a few cases) individually. This list selects the corresponding microscopic cross sections calculated from the free-gas model to create the material macroscopic cross sections. Furthermore, the free-gas model and the corresponding material macroscopic cross sections assume that the interactions of atoms do not affect the nuclear cross sections.

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

PubMed

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

2017-08-01

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

Evaluation of the Pool Critical Assembly Benchmark with Explicitly-Modeled Geometry using MCNP6

DOE PAGES

Kulesza, Joel A.; Martz, Roger Lee

2017-03-01

Despite being one of the most widely used benchmarks for qualifying light water reactor (LWR) radiation transport methods and data, no benchmark calculation of the Oak Ridge National Laboratory (ORNL) Pool Critical Assembly (PCA) pressure vessel wall benchmark facility (PVWBF) using MCNP6 with explicitly modeled core geometry exists. As such, this paper provides results for such an analysis. First, a criticality calculation is used to construct the fixed source term. Next, ADVANTG-generated variance reduction parameters are used within the final MCNP6 fixed source calculations. These calculations provide unadjusted dosimetry results using three sets of dosimetry reaction cross sections of varyingmore » ages (those packaged with MCNP6, from the IRDF-2002 multi-group library, and from the ACE-formatted IRDFF v1.05 library). These results are then compared to two different sets of measured reaction rates. The comparison agrees in an overall sense within 2% and on a specific reaction- and dosimetry location-basis within 5%. Except for the neptunium dosimetry, the individual foil raw calculation-to-experiment comparisons usually agree within 10% but is typically greater than unity. Finally, in the course of developing these calculations, geometry that has previously not been completely specified is provided herein for the convenience of future analysts.« less

Interdisciplinary Research in Viscoelasticity and Rheology

DTIC Science & Technology

1989-09-01

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Monte Carlo based dosimetry for neutron capture therapy of brain tumors

NASA Astrophysics Data System (ADS)

Zaidi, Lilia; Belgaid, Mohamed; Khelifi, Rachid

2016-11-01

Boron Neutron Capture Therapy (BNCT) is a biologically targeted, radiation therapy for cancer which combines neutron irradiation with a tumor targeting agent labeled with a boron10 having a high thermal neutron capture cross section. The tumor area is subjected to the neutron irradiation. After a thermal neutron capture, the excited 11B nucleus fissions into an alpha particle and lithium recoil nucleus. The high Linear Energy Transfer (LET) emitted particles deposit their energy in a range of about 10μm, which is of the same order of cell diameter [1], at the same time other reactions due to neutron activation with body component are produced. In-phantom measurement of physical dose distribution is very important for BNCT planning validation. Determination of total absorbed dose requires complex calculations which were carried out using the Monte Carlo MCNP code [2].

Final Stage in the Design of a Boron Neutron Capture Therapy facility at CEADEN, Cuba

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

Cabal, F. Padilla; Martin, G.

A neutron beam simulation study is carried out to determine the most suitable neutron energy for treatment of shallow and deep-seated brain tumors in the context of Boron Neutron Capture Therapy (BNCT). Two figures-of-merit, the therapeutic gain and the neutron fluence are utilized as beam assessment parameters. An irradiation cavity is used instead of a parallel beam port for the therapy. Calculations are performed using the MCNP5 code. After the optimization of our beam-shaper a study of the dose distribution in the head, neck, tyroids, lungs and upper and middle spine had been made. The therapeutic gain is increased whilemore » the current required for one hour treatment is decreased in comparison with the trading prototypes of NG used for BNCT.« less

Masters Thesis- Criticality Alarm System Design Guide with Accompanying Alarm System Development for the Radioisotope Production Laboratory in Richland, Washington

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

Greenfield, Bryce A.

2009-12-01

A detailed instructional manual was created to guide criticality safety engineers through the process of designing a criticality alarm system (CAS) for Department of Energy (DOE) hazard class 1 and 2 facilities. Regulatory and technical requirements were both addressed. A list of design tasks and technical subtasks are thoroughly analyzed to provide concise direction for how to complete the analysis. An example of the application of the design methodology, the Criticality Alarm System developed for the Radioisotope Production Laboratory (RPL) of Richland, Washington is also included. The analysis for RPL utilizes the Monte Carlo code MCNP5 for establishing detector coveragemore » in the facility. Significant improvements to the existing CAS were made that increase the reliability, transparency, and coverage of the system.« less

The MCNP Simulation of a PGNAA System at TRR-1/M1

NASA Astrophysics Data System (ADS)

Sangaroon, S.; Ratanatongchai, W.; Picha, R.; Khaweerat, S.; Channuie, J.

2017-06-01

The prompt-gamma neutron activation analysis system (PGNAA) has been installed at Thai Research Reactor-1/Modified 1 (TRR-1/M1) since 1999. The purpose of the system is for elemental and isotopic analyses. The system mainly consists of a series of the moderator and collimator, neutron and gamma-ray shielding and the HPGe detector. In this work, the condition of the system is carried out based on the Monte Carlo method using Monte Carlo N-Particle transport code and the experiment. The flux ratios (Φthermal/Φepithermal and Φthermal/Φfast) and thermal neutron flux have been obtained. The simulated prompt gamma rays of the Portland cement sample have been carried out. The simulation provides significant contribution in upgrading the PGNAA station to be available in various applications.

Determining organ dose conversion coefficients for external neutron irradiation by using a voxel mouse model

PubMed Central

Zhang, Xiaomin; Xie, Xiangdong; Qu, Decheng; Ning, Jing; Zhou, Hongmei; Pan, Jie; Yang, Guoshan

2016-01-01

A set of fluence-to-dose conversion coefficients has been calculated for neutrons with energies <20 MeV using a developed voxel mouse model and Monte Carlo N-particle code (MCNP), for the purpose of neutron radiation effect evaluation. The calculation used 37 monodirectional monoenergetic neutron beams in the energy range 10−9 MeV to 20 MeV, under five different source irradiation configurations: left lateral, right lateral, dorsal–ventral, ventral–dorsal, and isotropic. Neutron fluence-to-dose conversion coefficients for selected organs of the body were presented in the paper, and the effect of irradiation geometry conditions, neutron energy and the organ location on the organ dose was discussed. The results indicated that neutron dose conversion coefficients clearly show sensitivity to irradiation geometry at neutron energy below 1 MeV. PMID:26661852

SESAME: a software tool for the numerical dosimetric reconstruction of radiological accidents involving external sources and its application to the accident in Chile in December 2005.

PubMed

Huet, C; Lemosquet, A; Clairand, I; Rioual, J B; Franck, D; de Carlan, L; Aubineau-Lanièce, I; Bottollier-Depois, J F

2009-01-01

Estimating the dose distribution in a victim's body is a relevant indicator in assessing biological damage from exposure in the event of a radiological accident caused by an external source. This dose distribution can be assessed by physical dosimetric reconstruction methods. Physical dosimetric reconstruction can be achieved using experimental or numerical techniques. This article presents the laboratory-developed SESAME--Simulation of External Source Accident with MEdical images--tool specific to dosimetric reconstruction of radiological accidents through numerical simulations which combine voxel geometry and the radiation-material interaction MCNP(X) Monte Carlo computer code. The experimental validation of the tool using a photon field and its application to a radiological accident in Chile in December 2005 are also described.

Maintenance method and its critical issues for a fast-ignition laser fusion reactor based on a dry wall chamber

NASA Astrophysics Data System (ADS)

Someya, Y.; Matsumoto, T.; Okano, K.; Asaoka, Y.; Hiwatari, R.; Goto, T.; Ogawa, Y.

2008-05-01

The neutronics analysis has been carried out for feasibility study of the FALCON-D concept by Monte Carlo N-paticle transport code (MCNP), in order to inspect the cooling performance of in-vessel and ex-vessel components, and a connection pipe between Vacuum Vessel and reactor room. The nuclear heating rate in the Vacuum Vessel was at the same level as that of NBI duct of the ITER. The temperature of the connection pipe was found to be 345·, ·which was smaller than the melting point of structure materials (F82H). Moreover, the radiation damage of the final optics was also investigated. We propose a sliding changer concept for replacement. This method could be adapted for the replacement of one FPY cycle in the final optics system.

Urinary Phthalate Metabolite Concentrations and Reproductive Outcomes among Women Undergoing in Vitro Fertilization: Results from the EARTH Study.

PubMed

Hauser, Russ; Gaskins, Audrey J; Souter, Irene; Smith, Kristen W; Dodge, Laura E; Ehrlich, Shelley; Meeker, John D; Calafat, Antonia M; Williams, Paige L

2016-06-01

Evidence from both animal and human studies suggests that exposure to phthalates may be associated with adverse female reproductive outcomes. We evaluated the associations between urinary concentrations of phthalate metabolites and outcomes of assisted reproductive technologies (ART). This analysis included 256 women enrolled in the Environment and Reproductive Health (EARTH) prospective cohort study (2004-2012) who provided one to two urine samples per cycle before oocyte retrieval. We measured 11 urinary phthalate metabolites [mono(2-ethylhexyl) phthalate (MEHP), mono(2-ethyl-5-hydroxyhexyl) phthalate (MEHHP), mono(2-ethyl-5-oxohexyl) phthalate (MEOHP), mono(2-ethyl-5-carboxypentyl) phthalate (MECPP), mono-isobutyl phthalate (MiBP), mono-n-butyl phthalate (MBP), monobenzyl phthalate (MBzP), monoethyl phthalate (MEP), monocarboxyisooctyl phthalate (MCOP), monocarboxyisononyl phthalate (MCNP), and mono(3-carboxypropyl) phthalate (MCPP)]. We used generalized linear mixed models to evaluate the association of urinary phthalate metabolites with in vitro fertilization (IVF) outcomes, accounting for multiple IVF cycles per woman. In multivariate models, women in the highest as compared with lowest quartile of MEHP, MEHHP, MEOHP, MECPP, ΣDEHP (MEHP + MEHHP + MEOHP + MECPP), and MCNP had lower oocyte yield. Similarly, the number of mature (MII) oocytes retrieved was lower in the highest versus lowest quartile for these same phthalate metabolites. The adjusted differences (95% CI) in proportion of cycles resulting in clinical pregnancy and live birth between women in the fourth versus first quartile of ΣDEHP were -0.19 (-0.29, -0.08) and -0.19 (-0.28, -0.08), respectively, and there was also a lower proportion of cycles resulting in clinical pregnancy and live birth for individual DEHP metabolites. Urinary concentrations of DEHP metabolites were inversely associated with oocyte yield, clinical pregnancy, and live birth following ART. Hauser R, Gaskins AJ, Souter I, Smith KW, Dodge LE, Ehrlich S, Meeker JD, Calafat AM, Williams PL, for the EARTH Study Team. 2016. Urinary phthalate metabolite concentrations and reproductive outcomes among women undergoing in vitro fertilization: results from the EARTH study. Environ Health Perspect 124:831-839; http://dx.doi.org/10.1289/ehp.1509760.

Investigation of Natural and Man-Made Radiation Effects on Crews on Long Duration Space Missions

NASA Technical Reports Server (NTRS)

Bolch, Wesley E.; Parlos, Alexander

1996-01-01

Over the past several years, NASA has studied a variety of mission scenarios designed to establish a permanent human presence on the surface of Mars. Nuclear electric propulsion (NEP) is one of the possible elements in this program. During the initial stages of vehicle design work, careful consideration must be given to not only the shielding requirements of natural space radiation, but to the shielding and configuration requirements of the on-board reactors. In this work, the radiation transport code MCNP has been used to make initial estimates of crew exposures to reactor radiation fields for a specific manned NEP vehicle design. In this design, three 25 MW(sub th), scaled SP-100-class reactors are shielded by three identical shields. Each shield has layers of beryllium, tungsten, and lithium hydride between the reactor and the crew compartment. Separate calculations are made of both the exiting neutron and gamma fluxes from the reactors during beginning-of-life, full-power operation. This data is then used as the source terms for particle transport in MCNP. The total gamma and neutron fluxes exiting the reactor shields are recorded and separate transport calculations are then performed for a 10 g/sq cm crew compartment aluminum thickness. Estimates of crew exposures have been assessed for various thicknesses of the shield tungsten and lithium hydride layers. A minimal tungsten thickness of 20 cm is required to shield the reactor photons below the 0.05 Sv/y man-made radiation limit. In addition to a 20-cm thick tungsten layer, a 40-cm thick lithium hydride layer is required to shield the reactor neutrons below the annual limit. If the tungsten layer is 30-cm thick, the lithium hydride layer should be at least 30-cm thick. These estimates do not take into account the photons generated by neutron interactions inside the shield because the MCNP neutron cross sections did not allow reliable estimates of photon production in these materials. These results, along with natural space radiation shielding estimates calculated by NASA Langley Research Center, have been used to provide preliminary input data into a new Macintosh-based software tool. A skeletal version of this tool being developed will allow rapid radiation exposure and risk analyses to be performed on a variety of Lunar and Mars missions utilizing nuclear-powered vehicles.

Analyzing Aeroelasticity in Turbomachines

NASA Technical Reports Server (NTRS)

Reddy, T. S. R.; Srivastava, R.

2003-01-01

ASTROP2-LE is a computer program that predicts flutter and forced responses of blades, vanes, and other components of such turbomachines as fans, compressors, and turbines. ASTROP2-LE is based on the ASTROP2 program, developed previously for analysis of stability of turbomachinery components. In developing ASTROP2- LE, ASTROP2 was modified to include a capability for modeling forced responses. The program was also modified to add a capability for analysis of aeroelasticity with mistuning and unsteady aerodynamic solutions from another program, LINFLX2D, that solves the linearized Euler equations of unsteady two-dimensional flow. Using LINFLX2D to calculate unsteady aerodynamic loads, it is possible to analyze effects of transonic flow on flutter and forced response. ASTROP2-LE can be used to analyze subsonic, transonic, and supersonic aerodynamics and structural mistuning for rotors with blades of differing structural properties. It calculates the aerodynamic damping of a blade system operating in airflow so that stability can be assessed. The code also predicts the magnitudes and frequencies of the unsteady aerodynamic forces on the airfoils of a blade row from incoming wakes. This information can be used in high-cycle fatigue analysis to predict the fatigue lives of the blades.

Some Notes on Neutron Up-Scattering and the Doppler-Broadening of High-Z Scattering Resonances

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

Parsons, Donald Kent

When neutrons are scattered by target nuclei at elevated temperatures, it is entirely possible that the neutron will actually gain energy (i.e., up-scatter) from the interaction. This phenomenon is in addition to the more usual case of the neutron losing energy (i.e., down-scatter). Furthermore, the motion of the target nuclei can also cause extended neutron down-scattering, i.e., the neutrons can and do scatter to energies lower than predicted by the simple asymptotic models. In recent years, more attention has been given to temperature-dependent scattering cross sections for materials in neutron multiplying systems. This has led to the inclusion of neutronmore » up-scatter in deterministic codes like Partisn and to free gas scattering models for material temperature effects in Monte Carlo codes like MCNP and cross section processing codes like NJOY. The free gas scattering models have the effect of Doppler Broadening the scattering cross section output spectra in energy and angle. The current state of Doppler-Broadening numerical techniques used at Los Alamos for scattering resonances will be reviewed, and suggestions will be made for further developments. The focus will be on the free gas scattering models currently in use and the development of new models to include high-Z resonance scattering effects. These models change the neutron up-scattering behavior.« less

Monte Carlo simulations for angular and spatial distributions in therapeutic-energy proton beams

NASA Astrophysics Data System (ADS)

Lin, Yi-Chun; Pan, C. Y.; Chiang, K. J.; Yuan, M. C.; Chu, C. H.; Tsai, Y. W.; Teng, P. K.; Lin, C. H.; Chao, T. C.; Lee, C. C.; Tung, C. J.; Chen, A. E.

2017-11-01

The purpose of this study is to compare the angular and spatial distributions of therapeutic-energy proton beams obtained from the FLUKA, GEANT4 and MCNP6 Monte Carlo codes. The Monte Carlo simulations of proton beams passing through two thin targets and a water phantom were investigated to compare the primary and secondary proton fluence distributions and dosimetric differences among these codes. The angular fluence distributions, central axis depth-dose profiles, and lateral distributions of the Bragg peak cross-field were calculated to compare the proton angular and spatial distributions and energy deposition. Benchmark verifications from three different Monte Carlo simulations could be used to evaluate the residual proton fluence for the mean range and to estimate the depth and lateral dose distributions and the characteristic depths and lengths along the central axis as the physical indices corresponding to the evaluation of treatment effectiveness. The results showed a general agreement among codes, except that some deviations were found in the penumbra region. These calculated results are also particularly helpful for understanding primary and secondary proton components for stray radiation calculation and reference proton standard determination, as well as for determining lateral dose distribution performance in proton small-field dosimetry. By demonstrating these calculations, this work could serve as a guide to the recent field of Monte Carlo methods for therapeutic-energy protons.

MCMEG: Simulations of both PDD and TPR for 6 MV LINAC photon beam using different MC codes

NASA Astrophysics Data System (ADS)

Fonseca, T. C. F.; Mendes, B. M.; Lacerda, M. A. S.; Silva, L. A. C.; Paixão, L.; Bastos, F. M.; Ramirez, J. V.; Junior, J. P. R.

2017-11-01

The Monte Carlo Modelling Expert Group (MCMEG) is an expert network specializing in Monte Carlo radiation transport and the modelling and simulation applied to the radiation protection and dosimetry research field. For the first inter-comparison task the group launched an exercise to model and simulate a 6 MV LINAC photon beam using the Monte Carlo codes available within their laboratories and validate their simulated results by comparing them with experimental measurements carried out in the National Cancer Institute (INCA) in Rio de Janeiro, Brazil. The experimental measurements were performed using an ionization chamber with calibration traceable to a Secondary Standard Dosimetry Laboratory (SSDL). The detector was immersed in a water phantom at different depths and was irradiated with a radiation field size of 10×10 cm2. This exposure setup was used to determine the dosimetric parameters Percentage Depth Dose (PDD) and Tissue Phantom Ratio (TPR). The validation process compares the MC calculated results to the experimental measured PDD20,10 and TPR20,10. Simulations were performed reproducing the experimental TPR20,10 quality index which provides a satisfactory description of both the PDD curve and the transverse profiles at the two depths measured. This paper reports in detail the modelling process using MCNPx, MCNP6, EGSnrc and Penelope Monte Carlo codes, the source and tally descriptions, the validation processes and the results.

MAX: Multiplatform Applications for XAFS

NASA Astrophysics Data System (ADS)

Alain, Michalowicz; Jacques, Moscovici; Diane, Muller-Bouvet; Karine, Provost

2009-11-01

MAX is a new EXAFS and XANES analysis package, replacing our old "EXAFS pour le Mac" software suite. The major improvement is the ability to work with strictly the same code, compiled at once for Microsoft Windows, Apple MacOSX and LINUX systems, justifying the title "Multiplatform Applications for XAFS". It is organized as four modules: ABSORBIX (X-ray absorbance and fluorescence self-absorption calculations), CHEROKEE (EXAFS and XANES data treatment), ROUNDMIDNIGHT (EXAFS modeling and fit) and CRYSTALFFREV (from crystal structures and molecular modeling to FEFF EXAFS and XANES theoretical calculations). Most features developed in "EXAFS pour le Mac" are still available, but with much improvements in the user's interface, data treatment algorithms and new functionalities.

Lighting Control Systems Handbook.

DTIC Science & Technology

1985-06-01

cost, both initial and operating. Initially, the control system designer must collect in- formation and then study and weigh several areas including...8217odLe 045. Pearl Harbor. III: Code 11 Pearl Harbor ar ho I ir I L ’ odk 402. R IYI& [’. Plearl II arbor I II: Li bra ry. Pearl HaIitrbor. I ai

A One Dimensional, Time Dependent Inlet/Engine Numerical Simulation for Aircraft Propulsion Systems

NASA Technical Reports Server (NTRS)

Garrard, Doug; Davis, Milt, Jr.; Cole, Gary

1999-01-01

The NASA Lewis Research Center (LeRC) and the Arnold Engineering Development Center (AEDC) have developed a closely coupled computer simulation system that provides a one dimensional, high frequency inlet/engine numerical simulation for aircraft propulsion systems. The simulation system, operating under the LeRC-developed Application Portable Parallel Library (APPL), closely coupled a supersonic inlet with a gas turbine engine. The supersonic inlet was modeled using the Large Perturbation Inlet (LAPIN) computer code, and the gas turbine engine was modeled using the Aerodynamic Turbine Engine Code (ATEC). Both LAPIN and ATEC provide a one dimensional, compressible, time dependent flow solution by solving the one dimensional Euler equations for the conservation of mass, momentum, and energy. Source terms are used to model features such as bleed flows, turbomachinery component characteristics, and inlet subsonic spillage while unstarted. High frequency events, such as compressor surge and inlet unstart, can be simulated with a high degree of fidelity. The simulation system was exercised using a supersonic inlet with sixty percent of the supersonic area contraction occurring internally, and a GE J85-13 turbojet engine.

Core–Shell Nanoparticle-Based Peptide Therapeutics and Combined Hyperthermia for Enhanced Cancer Cell Apoptosis

PubMed Central

2015-01-01

Mitochondria-targeting peptides have garnered immense interest as potential chemotherapeutics in recent years. However, there is a clear need to develop strategies to overcome the critical limitations of peptides, such as poor solubility and the lack of target specificity, which impede their clinical applications. To this end, we report magnetic core–shell nanoparticle (MCNP)-mediated delivery of a mitochondria-targeting pro-apoptotic amphipathic tail-anchoring peptide (ATAP) to malignant brain and metastatic breast cancer cells. Conjugation of ATAP to the MCNPs significantly enhanced the chemotherapeutic efficacy of ATAP, while the presence of targeting ligands afforded selective delivery to cancer cells. Induction of MCNP-mediated hyperthermia further potentiated the efficacy of ATAP. In summary, a combination of MCNP-mediated ATAP delivery and subsequent hyperthermia resulted in an enhanced effect on mitochondrial dysfunction, thus resulting in increased cancer cell apoptosis. PMID:25133971

MCNP simulation to optimise in-pile and shielding parts of the Portuguese SANS instrument.

PubMed

Gonçalves, I F; Salgado, J; Falcão, A; Margaça, F M A; Carvalho, F G

2005-01-01

A Small Angle Neutron Scattering instrument is being installed at one end of the tangential beam tube of the Portuguese Research Reactor. The instrument is fed using a neutron scatterer positioned in the middle of the beam tube. The scatterer consists of circulating H2O contained in a hollow disc of Al. The in-pile shielding components and the shielding installed around the neutron selector have been the object of an MCNP simulation study. The quantities calculated were the neutron and gamma-ray fluxes in different positions, the energy deposited in the material by the neutron and gamma-ray fields, the material activation resulting from the neutron field and radiation doses at the exit wall of the shutter and around the shielding. The MCNP results are presented and compared with results of an analytical approach and with experimental data collected after installation.

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

Ortiz-Ramírez, Pablo, E-mail: rapeitor@ug.uchile.cl; Ruiz, Andrés

The Monte Carlo simulation of the gamma spectroscopy systems is a common practice in these days. The most popular softwares to do this are MCNP and Geant4 codes. The intrinsic spatial efficiency method is a general and absolute method to determine the absolute efficiency of a spectroscopy system for any extended sources, but this was only demonstrated experimentally for cylindrical sources. Due to the difficulty that the preparation of sources with any shape represents, the simplest way to do this is by the simulation of the spectroscopy system and the source. In this work we present the validation of themore » intrinsic spatial efficiency method for sources with different geometries and for photons with an energy of 661.65 keV. In the simulation the matrix effects (the auto-attenuation effect) are not considered, therefore these results are only preliminaries. The MC simulation is carried out using the FLUKA code and the absolute efficiency of the detector is determined using two methods: the statistical count of Full Energy Peak (FEP) area (traditional method) and the intrinsic spatial efficiency method. The obtained results show total agreement between the absolute efficiencies determined by the traditional method and the intrinsic spatial efficiency method. The relative bias is lesser than 1% in all cases.« less