bluuebunny/crossref_metadata_2025 huggingface dataset:
Spatial Correlations in Monte Carlo Criticality Simulations
Authors: Dumonteil E. , Malvagi F. , Zoia A. , Mazzolo A. , Artusio D. , Dieudonné C. , De Mulatier C.
Published: June 2014

Temporal correlations arising in Monte Carlo criticality codes have focused the attention of both developers and practitioners for a long time. Those correlations affects the evaluation of tallies of loosely coupled systems, where the system’s typical size is very large compared to the diffusion/absorption length scale of the neutrons. These time correlations are closely related to spatial correlations, both variables being linked by the transport equation. Therefore this paper addresses the question of diagnosing spatial correlations in Monte Carlo criticality simulations. In that aim, we will propose a spatial correlation function well suited to Monte Carlo simulations, and show its use while simulating a fuel pin-cell. The results will be discussed, modeled and interpreted using the tools of branching processes of statistical mechanics. A mechanism called “neutron clustering”, affecting simulations, will be discussed in this frame.
Application of new information entropy to source convergence diagnostics in Monte Carlo criticality calculations
Authors: Naito Yoshitaka , Namekawa Masakazu
Published: June 2014

In criticality safety evaluation, effective multiplication factor and its corresponding source distribution are often evaluated by the Monte Carlo method. For source convergence diagnostics a kind of information entropy (Shannon entropy) has, sometimes, been used. In the last conference (2009), we proposed new type information entropy in which the source distribution is expressed on an eigen-function space in order to evaluate the convergence not only of the effective multiplication factor but also of source distribution without increasing computing resources requirements. However, there arose questions concerning our entropy. Here, we answer some of the questions : (1) how to estimate expansion coefficients, (2) how to estimate the difference between the effective multiplication factor and the corresponding maximum eigenvalue, and (3) how to estimate the deviations of expansion coefficients due to the power iteration method using source distribution calculated by Monte Carlo methods. By applying our entropy to power iteration method, more effective source convergence diagnostics method combined with the “Sandwich Method” is attained.
Experimental validation of VESTA 2.1
Authors: Haeck W. , Cochet B. , Bernard F. , Tymen A.
Published: June 2014

Depletion codes such as VESTA are used to calculate the evolution of a material subjected to radiation (be it neutrons or another type of particle) for a wide variety of applications in the fields of nuclear safety, radiation protection and environmental health safety. For these applications, experimental validation is paramount. In this paper we will describe the experimental validation of the latest version of VESTA using a set of 76 samples consisting of radiochemical assay data and decay heat measurements. We will describe the general calculation procedure that has been applied to determine the uncertainty on each individual nuclide measurement as well as the general tendencies and detailed results.
Calculation of ex-core physical quantities using the 3D importance functions
Authors: Trakas Christos , De Laubiere Xavier
Published: June 2014

Diverse physical quantities are calculated in engineering studies with penalizing hypotheses to assure the required operation margins for each reactor. Today, these physical quantities are obtained by direct calculations from deterministic or Monte Carlo codes. The related states are critical or sub-critical.

The current physical quantities are for example: the SRD counting rates (source range detector) in the sub-critical state, the IRD (intermediary range detector) and PRD (power range detector) counting rates (neutron particles only), the deposited energy in the reflector (neutron + photon particles), the fluence or the DPA (displacement per atom) in the reactor vessel (neutron particles only).

The reliability of the proposed methodology is tested in the EPR reactor. The main advantage of the new methodology is the simplicity to obtain the physical quantities by an easy matrix calculation importance linked to nuclear power sources for all the cycles of the reactor. This method also allows to by-pass the direct calculations of the physical quantity of irradiated cores by Monte Carlo Codes, these calculations being impossible today (too many isotopic concentrations / MCNP5 limit).

This paper presents the first feasibility study for the physical quantities calculation outside of the core by the importance method instead of the direct calculations used currently by AREVA.
MINARET: Towards a time-dependent neutron transport parallel solver
Authors: Baudron A.-M. , Lautard J.J. , Maday Y. , Mula O.
Published: June 2014

We present the newly developed time-dependent 3D multigroup discrete ordinates neutron transport solver that has recently been implemented in the MINARET code. The solver is the support for a study about computing acceleration techniques that involve parallel architectures. In this work, we will focus on the parallelization of two of the variables involved in our equation: the angular directions and the time. This last variable has been parallelized by a (time) domain decomposition method called the parareal in time algorithm.
Shared Memory Parallelism for 3D Cartesian Discrete Ordinates Solver
Authors: Moustafa Salli , Dutka-Malen Ivan , Plagne Laurent , Ponçot Angélique , Ramet Pierre
Published: June 2014

This paper describes the design and the performance of DOMINO, a 3D Cartesian SN solver that implements two nested levels of parallelism (multicore+SIMD) on shared memory computation nodes. DOMINO is written in C++, a multi-paradigm programming language that enables the use of powerful and generic parallel programming tools such as Intel TBB and Eigen. These two libraries allow us to combine multi-thread parallelism with vector operations in an efficient and yet portable way. As a result, DOMINO can exploit the full power of modern multi-core processors and is able to tackle very large simulations, that usually require large HPC clusters, using a single computing node. For example, DOMINO solves a 3D full core PWR eigenvalue problem involving 26 energy groups, 288 angular directions (S16), 46 × 106 spatial cells and 1 × 1012 DoFs within 11 hours on a single 32-core SMP node. This represents a sustained performance of 235 GFlops and 40:74% of the SMP node peak performance for the DOMINO sweep implementation. The very high Flops/Watt ratio of DOMINO makes it a very interesting building block for a future many-nodes nuclear simulation tool.
Hybrid Parallel Programming Models for AMR Neutron Monte-Carlo Transport
Authors: Dureau David , Poëtte Gaël
Published: June 2014

This paper deals with High Performance Computing (HPC) applied to neutron transport theory on complex geometries, thanks to both an Adaptive Mesh Refinement (AMR) algorithm and a Monte-Carlo (MC) solver. Several Parallelism models are presented and analyzed in this context, among them shared memory and distributed memory ones such as Domain Replication and Domain Decomposition, together with Hybrid strategies. The study is illustrated by weak and strong scalability tests on complex benchmarks on several thousands of cores thanks to the petaflopic supercomputer Tera100.
Thread Divergence and Photon Transport on the GPU (U)
Authors: Aulwes Rob T. , Zukaitis Anthony
Published: June 2014

Monte Carlo methods are commonly used to solve numerically the particle transport problems. A major disadvantage to Monte Carlo methods is the time required to obtain accurate solutions. Graphical Processing Units (GPUs) have increased in use as accelerators for improving performance in high-performance computing. Extracting the best performance from GPUs places requires careful consideration on code execution and data movement. In particular, performance can be reduced if threads diverge due to branching, and Monte Carlo codes are susceptible to branching penalties. We explore different schemes to reduce thread divergence in photonics transport and report on our performance findings.
A CUDA Monte Carlo simulator for radiation therapy dosimetry based on Geant4
Authors: Henderson N. , Murakami K. , Amako K. , Asai M. , Aso T. , Dotti A. , Kimura A. , Gerritsen M. , Kurashige H. , Perl J. , Sasaki T.
Published: June 2014

Geant4 is a large-scale particle physics package that facilitates every aspect of particle transport simulation. This includes, but is not limited to, geometry description, material definition, tracking of particles passing through and interacting with matter, storage of event data, and visualization. As more detailed and complex simulations are required in different application domains, there is much interest in adapting the code for parallel and multi-core architectures. Parallelism can be achieved by tracking many particles at the same time. The complexity in the context of a GPU/CUDA adaptation is the highly serialized nature of the Geant4 package and the presence of large lookup tables that guide the simulation. This work presents G4CU, a CUDA implementation of the core Geant4 algorithm adapted for dose calculations in radiation therapy. For these applications the geometry is a block of voxels and the physics is limited to low energy electromagnetic physics. These features allow efficient tracking of many particles in parallel on the GPU. Experiments with radiotherapy simulations in G4CU demonstrate about 40 times speedups over Geant4.
Comparison of Two Accelerators for Monte Carlo Radiation Transport Calculations, NVIDIA Tesla M2090 GPU and Intel Xeon Phi 5110p Coprocessor: A Case Study for X-ray CT Imaging Dose Calculation
Authors: Liu Tianyu , George Xu X , Carothers Christopher D.
Published: June 2014

Hardware accelerators are currently becoming increasingly important in boosting high performance computing sys- tems. In this study, we tested the performance of two accelerator models, NVIDIA Tesla M2090 GPU and Intel Xeon Phi 5110p coprocessor, using a new Monte Carlo photon transport package called ARCHER-CT we have developed for fast CT imaging dose calculation. The package contains three code variants, ARCHER − CTCPU, ARCHER – CTGPU and ARCHER − CTCOP to run in parallel on the multi-core CPU, GPU and coprocessor architectures respectively. A detailed GE LightSpeed Multi-Detector Computed Tomography (MDCT) scanner model and a family of voxel patient phantoms were included in the code to calculate absorbed dose to radiosensitive organs under specified scan protocols. The results from ARCHER agreed well with those from the production code Monte Carlo N-Particle eXtended (MCNPX). It was found that all the code variants were significantly faster than the parallel MCNPX running on 12 MPI processes, and that the GPU and coprocessor performed equally well, being 2.89~4.49 and 3.01~3.23 times faster than the parallel ARCHER − CTCPU running with 12 hyperthreads.
A comparative study of history-based versus vectorized Monte Carlo methods in the GPU/CUDA environment for a simple neutron eigenvalue problem
Authors: Liu Tianyu , Du Xining , Ji Wei , Xu X. George , Brown Forrest B.
Published: June 2014

For nuclear reactor analysis such as the neutron eigenvalue calculations, the time consuming Monte Carlo (MC) simulations can be accelerated by using graphics processing units (GPUs). However, traditional MC methods are often history-based, and their performance on GPUs is affected significantly by the thread divergence problem. In this paper we describe the development of a newly designed event-based vectorized MC algorithm for solving the neutron eigenvalue problem. The code was implemented using NVIDIA’s Compute Unified Device Architecture (CUDA), and tested on a NVIDIA Tesla M2090 GPU card. We found that although the vectorized MC algorithm greatly reduces the occurrence of thread divergence thus enhancing the warp execution efficiency, the overall simulation speed is roughly ten times slower than the history-based MC code on GPUs. Profiling results suggest that the slow speed is probably due to the memory access latency caused by the large amount of global memory transactions. Possible solutions to improve the code efficiency are discussed.
Memory Bottlenecks and Memory Contention in Multi-Core Monte Carlo Transport Codes
Authors: Tramm John R. , Siegel Andrew R.
Published: June 2014

We have extracted a kernel that executes only the most computationally expensive steps of the Monte Carlo particle transport algorithm - the calculation of macroscopic cross sections - in an effort to expose bottlenecks within multi-core, shared memory architectures.
Preliminary Studies on the Resiliency of Stochastic Linear Solvers
Authors: Forget Benoit
Published: June 2014

With the advent of exascale computing and the realization that memory errors will be an ever important part of the high performance computing landscape, this paper proposes the reconsideration of stochastic linear solvers for their inherent scalability and resiliency capabilities. This paper addresses the latter by analyzing the resiliency of stochastic solvers to randomly occurring memory errors that go undetected. The premise is that, in stochastic solvers, undetected errors can be considered as part of the random process while detectable errors can be filtered using basic statistics. Thus, the goal is not to detect all memory errors, but only those that matter and quantifying their frequency which will impact efficiency. A simple iterative stochastic solver was implemented and all double-precision input variables were imposed a bit-flip to determine their impact on the final solution. Accepted batches contribute to the sample mean and variance that is used to determine whether or not to accept or reject the following batch. The test case indicated that 3-6 % of all soft memory errors in the double-precision variables were detected as exceeding the normal noise of the method. Since only a small fraction of memory errors truly matter, stochastic solvers offer a path that could potentially avoid error correction code all together. Multiple consecutive bit errors were also analyzed for the specific case of five neighboring bit flips. The test on one variable produced a soft error detection rate 1 % higher than the single bit-flip.
A Multiple-Set Overlapping-Domain Decomposed Monte Carlo Synthetic Acceleration Method for Linear Systems
Authors: Slattery Stuart R. , Evans Thomas M. , Wilson Paul P.H.
Published: June 2014

We present a multiple-set overlapping-domain decomposed strategy for parallelizing the Monte Carlo Synthetic Acceleration method. Monte Carlo Synthetic Acceleration methods use the Neumann-Ulam class of Monte Carlo solvers for linear systems to accelerate a fixed-point iteration sequence. Effective parallel algorithms for these methods require the parallelization of the underlying Neumann-Ulam solvers. To do this in a domain decomposed environment, we borrow strategies traditionally implemented in Monte Carlo particle transport to parallelize the problem. The parallel Neumann-Ulam and multiple-set overlapping-domain decomposition algorithms are presented along with parallel scaling data for the resulting implementation using the Titan Cray XK7 machine at Oak Ridge National Laboratory.
Integration of the DRAGON5/DONJON5 codes in the SALOME platform for performing multi-physics calculations in nuclear engineering
Authors: Hébert Alain
Published: June 2014

We are presenting the computer science techniques involved in the integration of codes DRAGON5 and DONJON5 in the SALOME platform. This integration brings new capabilities in designing multi-physics computational schemes, with the possibility to couple our reactor physics codes with thermal-hydraulics or thermo-mechanics codes from other organizations. A demonstration is presented where two code components are coupled using the YACS module of SALOME, based on the CORBA protocol. The first component is a full-core 3D steady-state neuronic calculation in a PWR performed using DONJON5. The second component implement a set of 1D thermal-hydraulics calculations, each performed over a single assembly.
Benchmarking and parallel scalability of MANCINTAP, a Parallel High-Performance Tool For Neutron Activation Analysis in Complex 4D Scenarios
Authors: Firpo G. , Frambati S. , Frignani M. , Gerra G.
Published: June 2014

MANCINTAP is a parallel computational tool developed by Ansaldo Nucleare to perform 4D neutron transport, activation and time-resolved dose-rate calculations in very complex geometries for CPU-intensive fission and fusion applications. MANCINTAP creates an automated link between the 3D radiation transport code MCNP5—which is used to evaluate both the neutron fluxes for activation calculations and the resulting secondary gamma dose rates—and the zero-dimensional activation code Anita2000 by handling crucial processes such as data exchange, determination of material mixtures and generation of cumulative probability distributions. A brief description of the computational tool is given here, with particular emphasis on the key technical choices underlying the project. Benchmarking of MANCINTAP has been performed in three steps: (i) against a very simplified model, where an analytical solution is available for comparison; (ii) against the well-established deterministic transport and activation code ATTILA and (iii) against experimental data obtained at the Frascati Neutron Generator (FNG) facility. An analysis of MANCINTAP scalability performances is proposed to demonstrate the robustness of its parallel structure, tailored for HPC applications, which makes it—to the best of our knowledge—a novel tool.
Application of MCNP for neutronic calculations at VR-1 training reactor
Authors: Huml Ondřej , Rataj Jan , Bílý Tomáš
Published: June 2014

The paper presents utilization of Monte Carlo MCNP transport code for neutronic calculations of training reactor VR-1. Results of calculations are compared with results of measurements realized during last few critical experiments with various reactor core configurations. Very good agreement between calculations and measurements is observed.
Improvements in the simulation of a main steam line break with steam generator tube rupture
Authors: Gallardo Sergio , Querol Andrea , Verdú Gumersindo
Published: June 2014

The result of simultaneous Main Steam Line Break (MSLB) and a Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR) is a depressurization in the secondary and primary system because both systems are connected through the SGTR. The OECD/NEA ROSA-2 Test 5 performed in the Large Scale Test Facility (LSTF) reproduces these simultaneous breaks in a Pressurized Water Reactor (PWR). A simulation of this Test 5 was made with the thermal-hydraulic code TRACE5. Some discrepancies found, such as an underestimation of SG-A secondary pressure during the depressurization and overestimation of the primary pressure drop after the first Power Operated Relief Valve (PORV) opening can be improved increasing the nodalization of the Upper Head in the pressure vessel and meeting the actual fluid conditions of Upper Head during the transient.
Withdrawn article – Performance Assessment of the Commercial CFD Software for the Prediction of the Reactor Internal Flow
Authors: Hee Lee Gong , Seok Bang Young , Woong Woo Sweng , Kim Do Hyeong , Ku Kang Min
Published: June 2014

This paper published in SNA + MC 2013 has been formally withdrawn. It should not be cited or referred to in the future.

A corrected version replaces this article, see SNA + MC 2013, 05128 (2014), DOI: 10.1051/snamc/201405128.

Request approved by the Editors and the Publisher on July 22, 2014.
Atomistic simulation of track formation in nuclear fuels due to heavy ion irradiation
Authors: Starikov S. , Stegailov V. , Pisarev V.
Published: June 2014

At moving of the swift heavy ion (e.g. Xe ion - the typical fission fragment), the track formation takes place in nuclear materials. There is a large interest in understanding of the mechanisms of this process. In this work, the atomistic simulation of track formation due to the swift heavy ion irradiation is performed for various nuclear materials (U, UO2). The two-temperature atomistic model with explicit account of electron pressure and electron thermal conductivity is used. This two-temperature model describes ionic subsystem by means of molecular dynamics while the electron subsystem is considered in the continuum approach. The various mechanisms of track formation are examined.
Feasibility study for a realistic training dedicated to radiological protection improvement
Authors: Courageot Estelle , Reinald Kutschera , Gaillard-Lecanu Emmanuelle , Sylvie Jahan , Riedel Alexandre , Therache Benjamin
Published: June 2014

Any personnel involved in activities within the controlled area of a nuclear facility must be provided with appropriate radiological protection training. An evident purpose of this training is to know the regulation dedicated to workplaces where ionizing radiation may be present, in order to properly carry out the radiation monitoring, to use suitable protective equipments and to behave correctly if unexpected working conditions happen. A major difficulty of this training consist in having the most realistic reading from the monitoring devices for a given exposure situation, but without using real radioactive sources. A new approach is developed at EDF R&D for radiological protection training. This approach combines different technologies, in an environment representative of the workplace but geographically separated from the nuclear power plant: a training area representative of a workplace, a Man Machine Interface used by the trainer to define the source configuration and the training scenario, a geolocalization system, fictive radiation monitoring devices and a particle transport code able to calculate in real time the dose map due to the virtual sources. In a first approach, our real-time particles transport code, called Moderato, used only an attenuation low in straight line. To improve the realism further, we would like to switch a code based on the Monte Carlo transport of particles method like Geant 4 or MCNPX instead of Moderato. The aim of our study is the evaluation of the code in our application, in particular, the possibility to keep a real time response of our architecture.
Sensitivity Analyses in Small Break LOCA with HPI-Failure: Effect of Break-Size in Secondary-Side Depressurization
Authors: Kinoshita Ikuo , Torige Toshihide , Yamada Minoru
Published: June 2014

In the case of total failure of the high pressure injection (HPI) system following small break loss of coolant accident (SBLOCA) in pressurized water reactor (PWR), the break size is so small that the primary system does not depressurize to the accumulator (ACC) injection pressure before the core is uncovered extensively. Therefore, steam generator (SG) secondary-side depressurization is necessary as an accident management in order to grant accumulator system actuation and core reflood. A thermal-hydraulic analysis using RELAP5/MOD3 was made on SBLOCA with HPI-failure for Oi Units 3/4 operated by Kansai Electoric Power Co., which are conventional 4 loop PWR plants. The effectiveness of SG secondary-side depressurization procedure was investigated for the real plant design and operational characteristics. The sensitivity analyses using RELAP5/MOD3.2 showed that the accident management was effective for a wide range of break sizes, various orientations and positions. The critical break can be 3 inch cold-leg bottom break.
Contribution to arc plasma modeling for welding TIG application
Authors: Borel Damien , Delalondre Clarisse , Carpreau Jean-Michel , Chéron B.G. , Boubert J.-P.
Published: June 2014

In this paper we present a numerical model that simulates transferred energy by a welding thermal plasma to the weld pool. This energy transfer allows materials melting. The originality of our model is to include the modeling of transition zones and the vaporization of the anode. The cathodic and anodic areas are taken into account in the model by means of heat balance at the gas-solid interfaces. We report the heating and cooling effects they induce on the solid (cathode, anode) and plasma. Code_Saturne® the CFD software developed at EDF R&D is used for this work Comparisons between simulations and measurements of temperature and electron density confirm the model assumptions for TIG welding.
Theory and Performance of AIMS for Active Interrogation
Authors: Walters William J. , Royston Katherine E. K. , Haghighat Alireza
Published: June 2014

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.
Usage of Simulated Response Matrices for the Scintillation Spectra Unfolding
Authors: Klusoň J. , Urban T.
Published: June 2014

Response of the scintillation spectrometer to incident photons has complex character due to complex character of photon interaction processes. Experimental spectrum thus provide qualitative information given by the identifiable peaks positions, while extraction of the full, quantitative information (i.e. photon flux energy distribution) require spectra unfolding based on the spectrometer response (represented by the response matrix) knowledge. As experimental determination of response matrix is difficult or impossible in most cases, the Monte Carlo simulation is the proved solution. Usage of the unfolding for experimental spectra processing enable to determine dosimetric characteristics of the incident photon fields and/or characteristics of the sources, creating those photon fields.

The response matrices for scintillation spectrometers with different detectors (commonly used NaI, BGO and new LaBr) were calculated using Monte Carlo method. The response to the internal activity of the LaBr detector was also simulated and considered in spectra processing. Calculated matrices were used for unfolding of the experimental spectra from different applications to determine desired dosimetric quantities. Typical applications are environmental monitoring, monitoring of the working environment, accidental monitoring and contamination measurement, etc. Method is suitable also for the airborne monitoring or security applications. Results for individual detection systems are compared and discussed with aim to analyze potential advantages of the LaBr detector for considered applications.
Comparison of DNA strand-break simulated with different DNA models
Authors: Xie Wenzhang , Li Junli , Li Chunyan , Qiu Rui , Yan Congchong , Zeng Zhi
Published: June 2014

In Monte Carlo simulation of DNA damage, the geometric model of DNA is of great importance. To study the influence of DNA model on the simulation of DNA damage, three DNA models were created in this paper. They were a volume model and two atomic models with different parameters. Direct DNA strand-break induced by low-energy electrons were simulated respectively with the three models. The results show that most of the energy depositions in the DNA segments do not lead to strand-breaks. The simple single strand-break (SSB) tends to be the predominant damage type, and the contribution of complex double strand-break (DSB) to the total DSB cannot be neglected. Among the yields of all the three DNA target models applied here, the yields of the volume model are the highest, the yields of the atomic model with double van der Waals radii (r) take the second place, whereas the yields of the atomic model with single r come last. On average, the ratios of SSB yields are approximately equivalent to the corresponding ratios of the models’ volume. However, there seems to be no clear relationship between the DSB yields and the models’ volume.
Cartesian Meshing Impacts for PWR Assemblies in Multigroup Monte Carlo and Sn Transport
Authors: Manalo K. , Chin M. , Sjoden and G.
Published: June 2014

Hybrid methods of neutron transport have increased greatly in use, for example, in applications of using both Monte Carlo and deterministic transport to calculate quantities of interest, such as flux and eigenvalue in a nuclear reactor. Many 3D parallel Sn codes apply a Cartesian mesh, and thus for nuclear reactors the representation of curved fuels (cylinder, sphere, etc.) are impacted in the representation of proper fuel inventory (both in deviation of mass and exact geometry representation). For a PWR assembly eigenvalue problem, we explore the errors associated with this Cartesian discrete mesh representation, and perform an analysis to calculate a slope parameter that relates the pcm to the percent areal/volumetric deviation (areal corresponds to 2D and volumetric to 3D, respectively). Our initial analysis demonstrates a linear relationship between pcm change and areal/volumetric deviation using Multigroup MCNP on a PWR assembly compared to a reference exact combinatorial MCNP geometry calculation. For the same multigroup problems, we also intend to characterize this linear relationship in discrete ordinates (3D PENTRAN) and discuss issues related to transport cross-comparison. In addition, we discuss auto-conversion techniques with our 3D Cartesian mesh generation tools to allow for full generation of MCNP5 inputs (Cartesian mesh and Multigroup XS) from a basis PENTRAN Sn model.
Development of Safety Analysis Code System of Beam Transport and Core for Accelerator Driven System
Authors: Aizawa Naoto , Iwasaki Tomohiko
Published: June 2014

Safety analysis code system of beam transport and core for accelerator driven system (ADS) is developed for the analyses of beam transients such as the change of the shape and position of incident beam. The code system consists of the beam transport analysis part and the core analysis part. TRACE 3-D is employed in the beam transport analysis part, and the shape and incident position of beam at the target are calculated. In the core analysis part, the neutronics, thermo-hydraulics and cladding failure analyses are performed by the use of ADS dynamic calculation code ADSE on the basis of the external source database calculated by PHITS and the cross section database calculated by SRAC, and the programs of the cladding failure analysis for thermoelastic and creep.

By the use of the code system, beam transient analyses are performed for the ADS proposed by Japan Atomic Energy Agency. As a result, the rapid increase of the cladding temperature happens and the plastic deformation is caused in several seconds. In addition, the cladding is evaluated to be failed by creep within a hundred seconds. These results have shown that the beam transients have caused a cladding failure.
Transport benchmarks for one-dimensional binary markovian mixtures revisited
Authors: Malvagi Fausto
Published: June 2014

The classic benchmarks for transport through a binary markovian mixure are revisited to look at the probability distribution function of the chosen “results”: reflection, transmission and scalar flux. We argue that knowledge of the ensemble averaged results is not sufficient for reliable predictions: a measure of the dispersion must also be obtained. An algorithm to estimate this dispersion is tested.
Improvements and New Findings in Monte Carlo Method with Complex-valued Weights for Neutron Leakage-corrected Assembly Calculations
Authors: Yamamoto Toshihiro
Published: June 2014

The author of this paper recently proposed a Monte Carlo calculation algorithm to solve a complex transport equation with complex-valued weights. The algorithm enables one to generate neutron leakage-corrected group constants and anisotropic diffusion coefficients for a unit fuel pin cell or assembly. The group constants are subsequently used for multi-group deterministic core calculations. The technique, however, had some limitations in applying itself to general problems. Some improvements have been done in this paper. The reflective boundary condition has newly become available. It has been found that a cumbersome weight cancellation of fission sources with positive and negative weights can be omitted in general fuel assembly geometries. A homogenization method of diffusion coefficients for a fuel assembly has been proposed.
Propagation of Uncertainty in System Parameters of a LWR Model by Sampling MCNPX Calculations – Burnup Analysis
Authors: Campolina Daniel de A. M. , Lima Claubia P. B. , Veloso Maria Auxiliadora F.
Published: June 2014

For all the physical components that comprise a nuclear system there is an uncertainty. Assessing the impact of uncertainties in the simulation of fissionable material systems is essential for a best estimate calculation that has been replacing the conservative model calculations as the computational power increases. The propagation of uncertainty in a simulation using a Monte Carlo code by sampling the input parameters is recent because of the huge computational effort required. In this work a sample space of MCNPX calculations was used to propagate the uncertainty. The sample size was optimized using the Wilks formula for a 95th percentile and a two-sided statistical tolerance interval of 95%. Uncertainties in input parameters of the reactor considered included geometry dimensions and densities. It was showed the capacity of the sampling-based method for burnup when the calculations sample size is optimized and many parameter uncertainties are investigated together, in the same input.
Microstructure engineering of Pt-Al alloy thin films through Monte Carlo simulations
Authors: Harris R.A. , Terblans J.J. , Swart H.C.
Published: June 2014

A kinetic algorithm, based on the regular solution model, was used in conjunction with the Monte Carlo method to simulate the evolution of a micro-scaled thin film system during exposure to a high temperature environment. Pt-Al thin films were prepared via electron beam physical vapor deposition (EB-PVD) with an atomic concentration ratio of Pt63:Al37. These films were heat treated at an annealing temperature of 400 °C for 16 and 49 minutes. Scanning Auger Microscopy (SAM) (PHI 700) was used to obtain elemental maps while sputtering through the thin films. Simulations were run for the same annealing temperatures and thin-film composition. From these simulations theoretical depth profiles and simulated microstructures were obtained. These were compared to the experimentally measured depth profiles and elemental maps.
Efficient parametric studies computations on grids and clouds with Civa software
Authors: Rougeron G. , Leberre S. , Gaine J. , Lonné S.
Published: June 2014

This poster describes the implementation of distributed computations of parametric studies in Civa NDT simulation software. It discusses the optimal use of remote resources in order to optimize time computation and minimize the costs when using a commercial cloud computing system.
DNS of MHD turbulent flow via the HELIOS supercomputer system at IFERC-CSC
Authors: Satake Shin-ichi , Kimura Masato , Yoshimori Hajime , Kunugi Tomoaki , Takase Kazuyuki
Published: June 2014

The simulation plays an important role to estimate characteristics of cooling in a blanket for such high heating plasma in ITER-BA. An objective of this study is to perform large -scale direct numerical simulation (DNS) on heat transfer of magneto hydro dynamic (MHD) turbulent flow on coolant materials assumed from Flibe to lithium. The coolant flow conditions in ITER-BA are assumed to be Reynolds number and Hartmann number of a higher order. The maximum target of the DNS assumed by this study based on the result of the benchmark of Helios at IFERC-CSC for Project cycle 1 is 116 TB (2048 nodes). Moreover, we tested visualization by ParaView to visualize directly the large-scale computational result. If this large-scale DNS becomes possible, an essential understanding and modelling of a MHD turbulent flow and a design of nuclear fusion reactor contributes greatly.
Parallel computing in the PRIZMA code
Authors: Kandiev Y.Z. , Mukhamadiev R.F. , Orlov V.G.
Published: June 2014

The paper describes parallelization and load balance algorithms implemented in the PRIZMA code.
ARCHER, a New Monte Carlo Software Tool for Emerging Heterogeneous Computing Environments
Authors: George Xu X. , Liu Tianyu , Su Lin , Du Xining , Riblett Matthew , Ji Wei , Gu Deyang , Carothers Christopher D. , Shephard Mark S. , Brown Forrest B. , Kalra Mannudeep K. , Liu Bob
Published: June 2014

The Monte Carlo radiation transport community faces a number of challenges associated with peta- and exa-scale computing systems that rely increasingly on heterogeneous architectures involving hardware accelerators such as GPUs. Existing Monte Carlo codes and methods must be strategically upgraded to meet emerging hardware and software needs. In this paper, we describe the development of a software, called ARCHER (Accelerated Radiation-transport Computations in Heterogeneous EnviRonments), which is designed as a versatile testbed for future Monte Carlo codes. Preliminary results from five projects in nuclear engineering and medical physics are presented.
COG11.1 Description, New Features, and Development Activities
Authors: Buck R. M. , Heinrichs D. P. , Lee C. K. , Lent E. M.
Published: June 2014

The latest COG code version, COG11.1BETA2, is described along with a summary of application features and ongoing and future development activities. This version is available from the Radiation Shielding Information Computer Center and the Nuclear Energy Agency Data Bank.
Overview of the FLUKA code
Authors: Battistoni Giuseppe , Boehlen Till , Cerutti Francesco , Wai Chin Pik , Salvatore Esposito Luigi , Fassò Alberto , Ferrari Alfredo , Mereghetti Alessio , Garcia Ortega Pablo , Ranft Johannes , Roesler Stefan , Sala Paola R. , Vlachoudis Vasilis
Published: June 2014

The capabilities and physics model implemented inside the FLUKA code are briefly described, with emphasis on hadronic interaction. Examples of the performances of the code are presented including basic (thin target) and complex benchmarks, and radiation detector specific applications. In particular the ability of FLUKA in describing existing calorimeter performances and in predicting those of future ones, as well as the use of the code for neutron and mixed field radiation detectors will be demonstrated with several examples.
Recent Developments in Geant4
Authors:   The Geant4 Collaboration
Published: June 2014

During the last two to three years extensive development of the Geant4 simulation toolkit has occurred, which will culminate in a major release at the end of 2013. This development includes the adoption of multi-threading, the extension and improvement of physics models, improvements to the geometry modeler with a revised implementation of most geometrical primitives, advances in visualization and graphical user interfaces (GUI), the move from GNU make to CMake and the extension and automation of the Geant4 testing suite. The reasons for and implementation of multi-threading will be discussed. Certain electromagnetic and hadronic model extensions will also be discussed along with their effects on calorimeter results. Finally, visualization and GUI improvements will be highlighted along with the new configuration, build and testing systems.
MC21 v.6.0 – A Continuous-Energy Monte Carlo Particle Transport Code with Integrated Reactor Feedback Capabilities
Authors: Griesheimer D. P. , Gill D. F. , Nease B. R. , Sutton T. M. , Stedry M. H. , Dobreff P. S. , Carpenter D. C. , Trumbull T. H. , Caro E. , Joo H. , Millman D. L.
Published: June 2014

MC21 is a continuous-energy Monte Carlo radiation transport code for the calculation of the steady-state spatial distributions of reaction rates in three-dimensional models. The code supports neutron and photon transport in fixed source problems, as well as iterated-fission-source (eigenvalue) neutron transport problems. MC21 has been designed and optimized to support large-scale problems in reactor physics, shielding, and criticality analysis applications. The code also supports many in-line reactor feedback effects, including depletion, thermal feedback, xenon feedback, eigenvalue search, and neutron and photon heating. MC21 uses continuous-energy neutron/nucleus interaction physics over the range from 10−5 eV to 20 MeV. The code treats all common neutron scattering mechanisms, including fast-range elastic and non-elastic scattering, and thermal- and epithermal-range scattering from molecules and crystalline materials. For photon transport, MC21 uses continuous-energy interaction physics over the energy range from 1 keV to 100 GeV. The code treats all common photon interaction mechanisms, including Compton scattering, pair production, and photoelectric interactions. All of the nuclear data required by MC21 is provided by the NDEX system of codes, which extracts and processes data from EPDL-, ENDF-, and ACE-formatted source files. For geometry representation, MC21 employs a flexible constructive solid geometry system that allows users to create spatial cells from first- and second-order surfaces. The system also allows models to be built up as hierarchical collections of previously defined spatial cells, with interior detail provided by grids and template overlays. Results are collected by a generalized tally capability which allows users to edit integral flux and reaction rate information. Results can be collected over the entire problem or within specific regions of interest through the use of phase filters that control which particles are allowed to score each tally. The tally system has been optimized to maintain a high level of efficiency, even as the number of edit regions becomes very large.
Features of MCNP6
Authors: Goorley T. , James M. , Booth T. , Brown F. , Bull J. , Cox L.J. , Durkee J. , Elson J. , Fensin M. , Forster R.A. , Hendricks J. , Hughes H.G. , Johns R. , Kiedrowski B. , Martz R. , Mashnik S. , McKinney G. , Pelowitz D. , Prael R. , Sweezy J. , Waters L. , Wilcox T. , Zukaitis T.
Published: June 2014

MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, but it is much more than the sum of these two computer codes. MCNP6 is the result of six years of effort by the MCNP5 and MCNPX code development teams. These groups of people, residing in Los Alamos National Laboratory’s X Computational Physics Division, Monte Carlo Codes Group (XCP-3) and Nuclear Engineering and Nonproliferation Division, Radiation Transport Modeling Team (NEN-5) respectively, have combined their code development efforts to produce the next evolution of MCNP. While maintenance and major bug fixes will continue for MCNP5 1.60 and MCNPX 2.7.0 for upcoming years, new code development capabilities only will be developed and released in MCNP6. In fact, the initial release of MCNP6 contains numerous new features not previously found in either code. These new features are summarized in this document. Packaged with MCNP6 is also the new production release of the ENDF/B-VII.1 nuclear data files usable by MCNP. The high quality of the overall merged code, usefulness of these new features, along with the desire in the user community to start using the merged code, have led us to make the first MCNP6 production release: MCNP6 version 1. High confidence in the MCNP6 code is based on its performance with the verification and validation test suites, comparisons to its predecessor codes, our automated nightly software debugger tests, the underlying high quality nuclear and atomic databases, and significant testing by many beta testers.
Overview of the MCU Monte Carlo Software Package
Authors: Kalugin M. A , Oleynik D. S. , Shkarovsky D. A.
Published: June 2014

MCU (Monte Carlo Universal) is a project on development and practical use of a universal computer code for simulation of particle transport (neutrons, photons, electrons, positrons) in three-dimensional systems by means of the Monte Carlo method. This paper provides the information on the current state of the project. The developed libraries of constants are briefly described, and the potentialities of the MCU-5 package modules and the executable codes compiled from them are characterized. Examples of important problems of reactor physics solved with the code are presented.
Recent Developments of JAEA’s Monte Carlo Code MVP for Reactor Physics Applications
Authors: Nagaya Yasunobu , Okumura Keisuke , Mori Takamasa
Published: June 2014

This paper describes the recent development status of a Monte Carlo code MVP developed at Japan Atomic Energy Agency. The basic features and capabilities of MVP are overviewed. In addition, new capabilities useful for reactor analysis are also described.
Overview of Particle and Heavy Ion Transport Code System PHITS
Authors: Sato Tatsuhiko , Niita Koji , Matsuda Norihiro , Hashimoto Shintaro , Iwamoto Yosuke , Furuta Takuya , Noda Shusaku , Ogawa Tatsuhiko , Iwase Hiroshi , Nakashima Hiroshi , Fukahori Tokio , Okumura Keisuke , Kai Tetsuya , Chiba Satoshi , Sihver Lembit
Published: June 2014

A general purpose Monte Carlo Particle and Heavy Ion Transport code System, PHITS, is being developed through the collaboration of several institutes in Japan and Europe. The Japan Atomic Energy Agency is responsible for managing the entire project. PHITS can deal with the transport of nearly all particles, including neutrons, protons, heavy ions, photons, and electrons, over wide energy ranges using various nuclear reaction models and data libraries. It is written in Fortran language and can be executed on almost all computers. All components of PHITS such as its source, executable and data-library files are assembled in one package and then distributed to many countries via the Research organization for Information Science and Technology, the Data Bank of the Organization for Economic Co-operation and Development’s Nuclear Energy Agency, and the Radiation Safety Information Computational Center. More than 1,000 researchers have been registered as PHITS users, and they apply the code to various research and development fields such as nuclear technology, accelerator design, medical physics, and cosmic-ray research. This paper briefly summarizes the physics models implemented in PHITS, and introduces some important functions useful for specific applications, such as an event generator mode and beam transport functions.
The Serpent Monte Carlo Code: Status, Development and Applications in 2013
Authors: Leppänen Jaakko , Pusa Maria , Viitanen Tuomas , Valtavirta Ville , Kaltiaisenaho Toni
Published: June 2014

The Serpent Monte Carlo reactor physics burnup calculation code has been developed at VTT Technical Research Centre of Finland since 2004, and is currently used in 100 universities and research organizations around the world. This paper presents the brief history of the project, together with the currently available methods and capabilities and plans for future work. Typical user applications are introduced in the form of a summary review on Serpent-related publications over the past few years.