bluuebunny/crossref_metadata_2025 huggingface dataset:
The application of “Helios” supercomputer in radiation safety studies for the IFMIF
Authors: Kondo Keitaro , Fischer Ulrich , Gröschel Friedrich , Heinzel Volker , Leichtle Dieter , Serikov Arkady
Published: June 2014

The HELIOS supercomputer system at International Fusion Energy Research Centre, Aomori, Japan has been extensively utilized in radiation safety studies for the International Fusion Materials Irradiation Facility (IFMIF). This paper is focusing on the neutronic analysis to support the layout of the high energy beam transport (HEBT) section of IFMIF. The McDeLicious-11 Monte Carlo code, which is an enhancement to MCNP5, was utilized in order to simulate the neutron generation in the IFMIF lithium target through d-Li(d,xn) reactions and the R2Smesh approach was utilized to evaluate the ambient dose distribution after shutdown. The necessary thickness of the biological shielding and additional local shielding for the HEBT section has been evaluated. The accessibility of HEBT rooms and a necessary cooling time is discussed based on the result of shutdown dose analysis.
Core Disruptive Accident Analysis using ASTERIA-FBR
Authors: Ishizu Tomoko , Endo Hiroshi , Yamamoto Toshihisa , Tatewaki Isao
Published: June 2014

JNES is developing a core disruptive accident analysis code, ASTERIA-FBR, which tightly couples the thermal-hydraulics and the neutronics to simulate the core behavior during core disruptive accidents of fast breeder reactors (FBRs). ASTERIA-FBR consists of the three-dimensional thermal-hydraulics calculation module: CONCORD, the fuel pin behavior calculation module: FEMAXI-FBR, and the space-time neutronics module: Dynamic-GMVP or PARTISN/RKIN. This paper describes a comparison between characteristics of GMVP and PARTISN and summarizes the challenging issues on applying Dynamic-GMVP to the calculation against unprotected loss-of-flow (ULOF) event which is a typical initiator of core disruptive accident of FBR. The statistical error included in the calculation results may affect the super-prompt criticality during ULOF event and thus the amount of released energy.
Monte Carlo Modeling of Minor Actinide Burning in Fissile Spallation Targets
Authors: Malyshkin Yury , Pshenichnov Igor , Mishustin Igor , Greiner Walter
Published: June 2014

Minor actinides (MA) present a harmful part of spent nuclear fuel due to their long half-lives and high radio-toxicity. Neutrons produced in spallation targets of Accelerator Driven Systems (ADS) can be used to transmute and burn MA. Non-fissile targets are commonly considered in ADS design. However, additional neutrons from fission reactions can be used in targets made of fissile materials. We developed a Geant4-based code MCADS (Monte Carlo model for Accelerator Driven Systems) for simulating neutron production and transport in different spallation targets. MCADS is suitable for calculating spatial distributions of neutron flux and energy deposition, neutron multiplication factors and other characteristics of produced neutrons and residual nuclei. Several modifications of the Geant4 source code described in this work were made in order to simulate targets containing MA. Results of MCADS simulations are reported for several cylindrical targets made of U+Am, Am or Am2O3 including more complicated design options with a neutron booster and a reflector. Estimations of Am burning rates are given for the considered cases.
Computational Challenges of Fusion Neutronics for ITER Ports
Authors: Serikov A. , Fischer U. , Pitcher C.S. , Suarez A. , Weinhorst B.
Published: June 2014

This paper elaborates computational challenges tackled for providing neutronics service supplied for developing the design of the Diagnostics Equatorial and Upper Port Plugs (EPP and UPP). The aim was to guide and assist the EPP and UPP design developers with optimal shielding solutions which are characterised of maintain the diagnostics purposes of the systems together with adequate radiation shielding performance. The target parameter for the shielding optimization was the minimum of Shut-Down Dose Rate (SDDR) inside the interspace between the port back-side and ITER bioshield. This aim was reached by parametric neutronic analyses of the shielding geometry and material composition, mitigating direct streaming of neutrons from the plasma by arranging the labyrinths and horizontal rails. Variation of many geometrical parameters of the labyrinths was possible only by applying the high performance parallel computations with MCNP5 using pure MPI and hybrid OpenMP/MPI parallelization techniques on several available supercomputers. MCNP5 parallel performance assessments were carried out to find an efficient way to run the code in a parallel regime. It was found a strong scaling (up to 4096 cores) performance of the MCNP5 jobs running with analogue Monte Carlo sampling and weak scaling for the tasks with biased sampling as a variance reduction technique, such as the MCNP5 intrinsic weight window generator. Deep penetrating radiation in the complex ITER tokamak geometry combined blocks of strong attenuation of the radiation together with the void gaps along which the particles are streamed freely contributes to computation challenges of radiation transport.
Interpolation Method for Calculation of Computed Tomography Dose from Angular Varying Tube Current
Authors: Caracappa Peter F. , Gao Yiming , George Xu X.
Published: June 2014

The scope and magnitude of radiation dose from computed tomography (CT) examination has led to increased scrutiny and focus on accurate dose tracking. The use of tube current modulation (TCM) results complicates dose tracking by generating unique scans that are specific to the patient. Three methods of estimating the radiation dose from a CT examination that uses TCM are compared: using the average current for an entire scan, using the average current for each slice in the scan, and using an estimation of the angular variation of the dose contribution. To determine the impact of TCM on the radiation dose received, a set of angular weighting functions for each tissue of the body are derived by fitting a function to the relative dose contributions tabulated for the four principle exposure projections. This weighting function is applied to the angular tube current function to determine the organ dose contributions from a single rotation. Since the angular tube current function is not typically known, a method for estimating that function is also presented. The organ doses calculated using these three methods are compared to simulations that explicitly include the estimated TCM function.
Modeling of the radiation measurement device FALCON 5000 by MCNPX: simulated efficiency for the on-site measurement
Authors: Courageot Estelle , Duong Emilie , Gaillard-Lecanu Emmanuelle , Jahan Sylvie
Published: June 2014

In order to evaluate the efficiency of the new spectrometry detector available in the STEP unit of the R&D from EDF, a numerical model of the FALCON 5000 radiation measurement system, based on the Ge technology, has been constructed thanks to the computer calculation code MCNPX. Due to the ignorance of the source term and the geometry of irradiation in our facilities, some cases of irradiation induce by a radioactive source has been planed to be simulated with different parameters which has been studied. For different energies and different distances and angles of incidence, simulations have been performed in order to assess an average the response of the modeled detector by MCNPX. Thanks to all these results, a global curve for the efficiency has been established taking into account of the energy, the distance between the source and the detector and the angle of incidence of the photon. The modeling of the detector, its validation and the efficiency curve have been presented in this paper.
Monte Carlo parametric studies of neutron interrogation with the Associated Particle Technique for cargo container inspections
Authors: Deyglun Clément , Carasco Cédric , Pérot Bertrand
Published: June 2014

The detection of Special Nuclear Materials (SNM) by neutron interrogation is extensively studied by Monte Carlo simulation at the Nuclear Measurement Laboratory of CEA Cadarache (French Alternative Energies and Atomic Energy Commission). The active inspection system is based on the Associated Particle Technique (APT). Fissions induced by tagged neutrons (i.e. correlated to an alpha particle in the DT neutron generator) in SNM produce high multiplicity coincidences which are detected with fast plastic scintillators. At least three particles are detected in a short time window following the alpha detection, whereas nonnuclear materials mainly produce single events, or pairs due to (n,2n) and (n,n’γ) reactions. To study the performances of an industrial cargo container inspection system, Monte Carlo simulations are performed with the MCNP-PoliMi transport code, which records for each neutron history the relevant information: reaction types, position and time of interactions, energy deposits, secondary particles, etc. The output files are post-processed with a specific tool developed with ROOT data analysis software. Particles not correlated with an alpha particle (random background), counting statistics, and time-energy resolutions of the data acquisition system are taken into account in the numerical model. Various matrix compositions, suspicious items, SNM shielding and positions inside the container, are simulated to assess the performances and limitations of an industrial system.
Calculation of Gamma-ray Responses for HPGe Detectors with TRIPOLI-4 Monte Carlo Code
Authors: Lee Yi-Kang , Garg Ruchi
Published: June 2014

The gamma-ray response calculation of HPGe (High Purity Germanium) detector is one of the most important topics of the Monte Carlo transport codes for nuclear instrumentation applications. In this study the new options of TRIPOLI-4 Monte Carlo transport code for gamma-ray spectrometry were investigated. Recent improvements include the gamma-rays modeling of the electron-position annihilation, the low energy electron transport modeling, and the low energy characteristic X-ray production. The impact of these improvements on the detector efficiency of the gamma-ray spectrometry calculations was verified. Four models of HPGe detectors and sample sources were studied. The germanium crystal, the dead layer of the crystal, the central hole, the beryllium window, and the metal housing are the essential parts in detector modeling. A point source, a disc source, and a cylindrical extended source containing a liquid radioactive solution were used to study the TRIPOLI-4 calculations for the gamma-ray energy deposition and the gamma-ray self-shielding. The calculations of full-energy-peak and total detector efficiencies for different sample-detector geometries were performed. Using TRIPOLI-4 code, different gamma-ray energies were applied in order to establish the efficiency curves of the HPGe gamma-ray detectors.
Neutron/Photon Transport Simulations and Fissile Mass Estimation
Authors: Humbert Ph. , Méchitoua B.
Published: June 2014

Detection and mass estimation of fissile material are important tasks. The aim of this paper is to show how one can try to solve such an inverse problem with an iterative forward approach.
Performance of the New WCOS Technique for the TITAN SPECT Formulation
Authors: Royston Katherine E. K. , Haghighat Alireza
Published: June 2014

The TITAN code is a parallel hybrid deterministic transport code that includes novel algorithms for simulation of a Single Photon Emission Computed Tomography (SPECT) device. Monte Carlo methods are generally used for SPECT simulation, but can be time consuming and are subject to statistical uncertainties. Representation of the collimator in SPECT is difficult in a deterministic code and a new method is presented to improve the accuracy of the existing hybrid formulation, a combination of SN with fictitious quadrature and ray-tracing. In the new technique, weighted circular ordinate splitting (WCOS), the angular fluxes along the fictitious directions are weighted by the detector surface area projected along each direction to the front of the collimator. For a simple water phantom with a Tc-99m source, the behavior of the WCOS technique for varying parameters is examined for a low energy, general purpose collimator and a low energy, high sensitivity collimator. The TITAN detector flux distribution is benchmarked against the MCNP5 solution and shown to require significantly less computation time. Finally, the parallel computation time of the TITAN SPECT simulation is analyzed.
Application of new nuclear de-excitation model of PHITS for prediction of isomer yield and prompt gamma-ray production
Authors: Ogawa Tatsuhiko , Hashimoto Shintaro , Sato Tatsuhiko , Niita Koji
Published: June 2014

A new nuclear de-excitation model, intended for accurate simulation of isomeric transition of excited nuclei, was incorporated into PHITS and applied to various situations to clarify the impact of the model. The case studies show that precise treatment of gamma de-excitation and consideration for isomer production are important for various applications such as detector performance prediction, radiation shielding calculations and the estimation of radioactive inventory including isomers.
EASY-II: a system for modelling of n, d, p, γ and α activation and transmutation processes
Authors: Sublet Jean-Christophe , Eastwood James , Morgan Guy , Koning Arjan , Rochman Dimitri
Published: June 2014

EASY-II is designed as a functional replacement for the previous European Activation System, EASY-2010. It has extended nuclear data and new software, FISPACT-II, written in object-style Fortran to provide new capabilities for predictions of activation, transmutation, depletion and burnup. The new FISPACT-II code has allowed us to implement many more features in terms of energy range, up to GeV; incident particles: alpha, gamma, proton, deuteron and neutron; and neutron physics: self-shielding effects, temperature dependence, pathways analysis, sensitivity and error estimation using covariance data. These capabilities cover most application needs: nuclear fission and fusion, accelerator physics, isotope production, waste management and many more. In parallel, the maturity of modern general-purpose libraries such as TENDL-2012 encompassing thousands of target nuclides, the evolution of the ENDF format and the capabilities of the latest generation of processing codes PREPRO-2012, NJOY2012 and CALENDF-2010 have allowed the FISPACT-II code to be fed with more robust, complete and appropriate data: cross-sections with covariance, probability tables in the resonance ranges, kerma, dpa, gas and radionuclide production and 24 decay types. All such data for the five most important incident particles are placed in evaluated data files up to an incident energy of 200 MeV. The resulting code and data system, EASY-II, includes many new features and enhancements. It has been extensively tested, and also benefits from the feedback from wide-ranging validation and verification activities performed with its predecessor
An accurate method for energy spectrum reconstruction of Linac beams based on EPID measurements of scatter radiation
Authors: Juste B. , Miró R. , Verdú G. , Santos A.
Published: June 2014

This work presents a methodology to reconstruct a Linac high energy photon spectrum beam. The method is based on EPID scatter images generated when the incident photon beam impinges onto a plastic block. The distribution of scatter radiation produced by this scattering object placed on the external EPID surface and centered at the beam field size was measured. The scatter distribution was also simulated for a series of monoenergetic identical geometry photon beams. Monte Carlo simulations were used to predict the scattered photons for monoenergetic photon beams at 92 different locations, with 0.5 cm increments and at 8.5 cm from the centre of the scattering material. Measurements were performed with the same geometry using a 6 MeV photon beam produced by the linear accelerator. A system of linear equations was generated to combine the polyenergetic EPID measurements with the monoenergetic simulation results. Regularization techniques were applied to solve the system for the incident photon spectrum. A linear matrix system, A×S=E, was developed to describe the scattering interactions and their relationship to the primary spectrum (S). A is the monoenergetic scatter matrix determined from the Monte Carlo simulations, S is the incident photon spectrum, and E represents the scatter distribution characterized by EPID measurement. Direct matrix inversion methods produce results that are not physically consistent due to errors inherent in the system, therefore Tikhonov regularization methods were applied to address the effects of these errors and to solve the system for obtaining a consistent bremsstrahlung spectrum.
A Hybrid Ant Colony Algorithm for Loading Pattern Optimization
Authors: Hoareau F.
Published: June 2014

Electricité de France (EDF) operates 58 nuclear power plant (NPP), of the Pressurized Water Reactor (PWR) type. The loading pattern (LP) optimization of these NPP is currently done by EDF expert engineers. Within this framework, EDF R&D has developed automatic optimization tools that assist the experts. The latter can resort, for instance, to a loading pattern optimization software based on ant colony algorithm.

This paper presents an analysis of the search space of a few realistic loading pattern optimization problems. This analysis leads us to introduce a hybrid algorithm based on ant colony and a local search method. We then show that this new algorithm is able to generate loading patterns of good quality.
A New Method for the Calculation of Diffusion Coefficients with Monte Carlo
Authors: Dorval Eric
Published: June 2014

This paper presents a new Monte Carlo-based method for the calculation of diffusion coefficients. One distinctive feature of this method is that it does not resort to the computation of transport cross sections directly, although their functional form is retained. Instead, a special type of tally derived from a deterministic estimate of Fick’s Law is used for tallying the total cross section, which is then combined with a set of other standard Monte Carlo tallies. Some properties of this method are presented by means of numerical examples for a multi-group 1-D implementation. Calculated diffusion coefficients are in general good agreement with values obtained by other methods.
P1 adaptation of TRIPOLI-4® code for the use of 3D realistic core multigroup cross section generation
Authors: Cai Li , Pénéliau Yannick , Diop Cheikh M. , Malvagi Fausto
Published: June 2014

In this paper, we discuss some improvements we recently implemented in the Monte-Carlo code TRIPOLI-4® associated with the homogenization and collapsing of subassemblies cross sections. The improvement offered us another approach to get critical multigroup cross sections with Monte-Carlo method. The new calculation method in TRIPOLI-4® tries to ensure the neutronic balances, the multiplicative factors and the critical flux spectra for some realistic geometries. We make it by at first improving the treatment of the energy transfer probability, the neutron excess weight and the neutron fission spectrum. This step is necessary for infinite geometries. The second step which will be enlarged in this paper is aimed at better dealing with the multigroup anisotropy distribution law for finite geometries. Usually, Monte-Carlo homogenized multi-group cross sections are validated within a core calculation by a deterministic code. Here, the validation of multigroup constants will also be carried out by Monte-Carlo core calculation code. Different subassemblies are tested with the new collapsing method, especially for the fast neutron reactors subassemblies.
AN Core Analysis
Authors: Barbarino Andrea , Tomatis Daniele
Published: June 2014

Several alternative approximations of neutron transport have been proposed in years to move around the known limitations imposed by neutron diffusion in the modeling of nuclear cores. However, only a few complied with the industrial requirements of fast numerical computation, concentrating more on physical accuracy. In this work, the AN transport methodology is discussed with particular interest in core performance calculations. The implementation of the methodology in full core codes is discussed with particular attention to numerical issues and to the integration within the entire simulation process. Finally, first results from core studies in AN transport are analyzed in detail and compared to standard results of neutron diffusion.
Tracking strategies in 3D axial geometries for a MOC solver
Authors: Sciannandrone Daniele , Santandrea Simone
Published: June 2014

The solution of the neutron transport equation in 3D domains is a high resource demanding task. One of the most successful method of solution of this equation is the Method of Characteristics (MOC). However, the huge number of trajectories constitutes a serious issue in terms of memory needs and in terms of number of floating point operations. In this paper, we propose some techniques to compress tracking data and to speed up the MOC solver. These methods take advantage of regularities of 3D prismatic geometries. The new algorithms have been implemented in the 3D MOC solver TDT and applied to a typical UOx/MOx colorset.
Generation and Testing of XS Libraries for VVER Using APOLLO2 and TRIPOLI4
Authors: Zheleva Nonka , Petrov Nikolay , Todorova Galina , Kolev Nikola
Published: June 2014

MOC based calculation schemes with APOLLO2 were used to generate few-group cross-section libraries for VVER-1000 at the nodal and pin level. This paper presents an overview of the testing of the schemes and the libraries, as well as the computational aspects. Two major ameliorations are considered: application of new developments in APOLLO2 and multicore computation for an acceptable trade-off between accuracy and efficiency. Two-level Pij-MOC industrial calculation schemes were tested against TRIPOLI4 reference results. Benchmarking of the schemes shows that the higher-order linear surface method of characteristics (LS MOC) is an efficient option for cross-section library generation. There is a significant potential for further refinement of the MOC energy mesh and the MOC parameters with the progress in distributed computing.

A multi-parameter cross-section library for MSLB analysis with homogenized nodes was tested in 2D core simulation with COBAYA3 vs. whole-core TRIPOLI4 solutions on the CEA CCRT HPC system. Pin-by-pin cross-sections and interface discontinuity factors of Black Box Homogenization type were tested in diffusion calculations with COBAYA3 pin-by-pin against transport reference solutions. Good agreement is displayed.
High-Fidelity Coupled Monte-Carlo/Thermal-Hydraulics Calculations
Authors: Ivanov Aleksandar , Sanchez Victor , Ivanov Kostadin
Published: June 2014

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.
Python-based framework for coupled MC-TH reactor calculations
Authors: Travleev Anton A. , Molitor Richard , Sanchez Victor
Published: June 2014

We develop a set of Python packages to provide a modern programming interface to codes used for analysis of nuclear reactors. Currently implemented interfaces to the Monte Carlo (MC) neutronics code MCNP and thermo-hydraulic (TH) code SCF allow efficient description of calculation models and provide a framework for coupled calculations. In this paper we illustrate how these interfaces can be used to describe a pin model, and report results of coupled MCNP-SCF calculations performed for a PWR fuel assembly, organized by means of the interfaces.
Uncertainty analysis of Lead cross sections on reactor safety for ELECTRA
Authors: Alhassan Erwin , Sjöstrand Henrik , Duan Junfeng , Gustavsson Cecilia , Pomp Stephan , Österlund Michael , Rochman Dimitri , Koning Arjan J.
Published: June 2014

The Total Monte Carlo (TMC) method was used in this study to assess the impact of Pb-206, 207 and 208 nuclear data uncertainties on keff , βeff, coolant temperature coefficient, the coolant void worth for the ELECTRA reactor. Relatively large uncertainties were observed in the keff and the coolant void worth for all the isotopes with significant contribution coming from Pb-208 nuclear data. The large Pb-208 nuclear data uncertainty observed was further investigated by studying the impact of partial channels on the keff and the βeff. Various sections of ENDF file: elastic scattering (n, el), inelastic scattering (n, inl), neutron capture (n, γ), (n, 2n), resonance parameters and the angular distribution were varied randomly and distributions in keff and the βeff obtained. The dominant contributions to the uncertainty in the keff from Pb-208 came from uncertainties in the resonance parameters; however, elastic scattering cross section and the angular distribution also had significant impact. The impact of nuclear data uncertainties on the βeff was observed to be small.
Sensitivity and Uncertainty Analysis of the keff for VHTR fuel
Authors: Young Han Tae , Chul Lee Hyun , Man Noh Jae
Published: June 2014

For the uncertainty and sensitivity analysis of PMR200 designed as a VHTR in KAERI, MUSAD was implemented based on the deterministic method in the connection with DeCART/CAPP code system. The sensitivity of the multiplication factor was derived using the classical perturbation theory and the sensitivity coefficients for the individual cross sections were obtained by the adjoint method within the framework of the transport equation. Then, the uncertainty of the multiplication factor was calculated from the product of the covariance matrix and the sensitivity. For the verification calculation of the implemented code, the uncertainty analysis on GODIVA benchmark and PMR200 pin cell problem were carried out and the results were compared with the reference codes, TSUNAMI and McCARD. As a result, they are in a good agreement except the uncertainty by the scattering cross section which was calculated using the different scattering moment.
Sensitivity Analysis for Activation Problems
Authors: Arter Wayne , Morgan Guy
Published: June 2014

A study has been made as to how to develop further the techniques for sensitivity analysis used by Fispact-II. Fispact-II is a software suite for the analysis of nuclear activation and transmutation problems, developed for all nuclear applications. The software already permits sensitivity analysis to be performed by Monte Carlo sampling, and a faster uncertainty analysis is made possible by a powerful graph-based approach which generates a reduced set of nuclides on pathways leading to significant contributions to radiological quantities. The peculiar aspects of the sensitivity analysis problem for activation are the large number, typically thousands, of rate equation parameters(decay rates and reaction cross-sections) which all have some degree of associated error, and the fact that activity as a function of time varies as a sum of exponentials, so appears discontinuous as rate parameters are varied unless the sampling frequency is impractically fast. Nevertheless, Monte Carlo sampling is a generic approach and it is therefore conceivable that techniques more targeted to the activation problem might be beneficial. Moreover, recent theoretical developments have highlighted the importance of a two-stage approach to mathematically similar problems, where in the first stage, information is collected about the global behaviour of the problem, such as the identification of the rate parameters which cause the greatest variation in dose or nuclear activity, before a second stage examines a problem with its scope restricted by the information from the first. In the second stage, for example, Quasi-Monte Carlo sampling may be used in a restricted parameter space. The current work concentrates on the first stage and consists of a review of possible techniques with a detailed examination of the most promising pathways reduction approach, examined directly using Fispact-II. All the evidence obtained demonstrates the strong potential of this approach.
Processing of the GALILEOTM fuel rod code model uncertainties within the AREVA LWR realistic thermal-mechanical analysis methodology
Authors: Mailhe P. , Barbier B. , Garnier Ch. , Landskron H. , Sedlacek R. , Arimescu I. , Smith M. , Bellanger Ph.
Published: June 2014

The availability of reliable tools and associated methodology able to accurately predict the LWR fuel behavior in all conditions is of great importance for safe and economic fuel usage. For that purpose, AREVA has developed its new global fuel rod performance code GALILEOTM along with its associated realistic thermal-mechanical analysis methodology. This realistic methodology is based on a Monte Carlo type random sampling of all relevant input variables. After having outlined the AREVA realistic methodology, this paper will be focused on the GALILEOTM code benchmarking process on its extended experimental database and the GALILEOTM model uncertainties assessment. The propagation of these model uncertainties through the AREVA realistic methodology is also presented. This GALILEOTM model uncertainties processing is of the utmost importance for accurate fuel design margin evaluation as illustrated on some application examples.

With the submittal of Topical Report for GALILEOTM to the U.S. NRC in 2013, GALILEOTM and its methodology are on the way to be industrially used in a wide range of irradiation conditions.
Unified Monte Carlo: Evaluation, Uncertainty Quantification and Propagation of the Prompt Fission Neutron Spectrum
Authors: Rising Michael E. , Talou Patrick , Prinja Anil K. , White Morgan C.
Published: June 2014

The unified Monte Carlo (UMC) method is used for the quantification of uncertainties associated with the evaluation of the prompt fission neutron spectrum (PFNS) for the n(0.5 MeV)+239Pu fission reaction and compared to the Kalman filter. Ultimately, the UMC and Kalman filter approaches lead to very similar evaluated PFNS while UMC is also capable of capturing the nonlinearities present in the Los Alamos (LA) model used to calculate the PFNS. Next, the unified Monte Carlo + total Monte Carlo (UMC+TMC) method is implemented to propagate uncertainties from the prior LA model parameters through the Flattop critical assemblies. Due to the fact that cross-experiment correlations are neglected in the present evaluation work, the UMC+TMC method predicts uncertainties in the integral quantities smaller by an order of magnitude or more compared to direct sampling from the posterior LA model parameters. Last, the UMC method is proposed for use as an evaluation tool that can be used with the new prompt fission Monte Carlo Hauser-Feshbach codes that are currently under development.
Probabilistic approach for decay heat uncertainty estimation using URANIE platform and MENDEL depletion code
Authors: Tsilanizara A. , Gilardi N. , Huynh T.D. , Jouanne C. , Lahaye S. , Martinez J.M. , Diop C.M.
Published: June 2014

The knowledge of the decay heat quantity and the associated uncertainties are important issues for the safety of nuclear facilities. Many codes are available to estimate the decay heat. ORIGEN, FISPACT, DARWIN/PEPIN2 are part of them. MENDEL is a new depletion code developed at CEA, with new software architecture, devoted to the calculation of physical quantities related to fuel cycle studies, in particular decay heat.

The purpose of this paper is to present a probabilistic approach to assess decay heat uncertainty due to the decay data uncertainties from nuclear data evaluation like JEFF-3.1.1 or ENDF/B-VII.1. This probabilistic approach is based both on MENDEL code and URANIE software which is a CEA uncertainty analysis platform. As preliminary applications, single thermal fission of uranium 235, plutonium 239 and PWR UOx spent fuel cell are investigated.
Numerically robust geometry engine for compound solid geometries
Authors: Vlachoudis V. , Sinuela-Pastor D.
Published: June 2014

Monte Carlo programs heavily rely on a fast and numerically robust solid geometry engines. However the success of solid modeling, depends on facilities for specifying and editing parameterized models through a user-friendly graphical front-end. Such a user interface has to be fast enough in order to be interactive for 2D and/or 3D displays, but at the same time numerically robust in order to display possible modeling errors at real time that could be critical for the simulation. The graphical user interface “flair”(1) for FLUKA(2,3) currently employs such an engine where special emphasis has been given on being fast and numerically robust. The numerically robustness is achieved by a novel method of estimating the floating precision of the operations, which dynamically adapts all the decision operations accordingly. Moreover a predictive caching mechanism is ensuring that logical errors in the geometry description are found online, without compromising the processing time by checking all regions.
Computing Numerically-Optimal Bounding Boxes for Constructive Solid Geometry (CSG) Components in Monte Carlo Particle Transport Calculations
Authors: Millman David L. , Griesheimer David P. , Nease Brian R. , Snoeyink Jack
Published: June 2014

For large, highly detailed models, Monte Carlo simulations may spend a large fraction of their run-time performing simple point location and distance to surface calculations for every geometric component in a model. In such cases, the use of bounding boxes (axis-aligned boxes that bound each geometric component) can improve particle tracking efficiency and decrease overall simulation run time significantly. In this paper we present a robust and efficient algorithm for generating the numerically-optimal bounding box (optimal to within a user-specified tolerance) for an arbitrary Constructive Solid Geometry (CSG) object defined by quadratic surfaces. The new algorithm uses an iterative refinement to tighten an initial, conservatively large, bounding box into the numerically-optimal bounding box. At each stage of refinement, the algorithm subdivides the candidate bounding box into smaller boxes, which are classified as inside, outside, or intersecting the boundary of the component. In cases where the algorithm cannot unambiguously classify a box, the box is refined further. This process continues until the refinement near the component’s extremal points reach the user-selected tolerance level. This refinement/classification approach is more efficient and practical than methods that rely on computing actual boundary representations or sampling to determine the extent of an arbitrary CSG component. A complete description of the bounding box algorithm is presented, along with a proof that the algorithm is guaranteed to converge to within specified tolerance of the true optimal bounding box. The paper also provides a discussion of practical implementation details for the algorithm as well as numerical results highlighting performance and accuracy for several representative CSG components.
Geant4 electromagnetic physics: improving simulation performance and accuracy
Authors: Ivanchenko V. N. , Incerti S. , Allison J. , Bagulya A. , Brown J. M. C. , Champion C. , Elles S. , Francis Z. , Grichine V. , Ivantchenko A. , Jacquemier J. , Karamitros M. , Maire M. , Mantero A. , Pandola L. , Raine M. , Reis M. A. , Santin G. , Sawkey D. , Schaelicke A. , Schenk M. , Taborda A. , Urban L. , Yamashita T.
Published: June 2014

The most recent upgrades of the electromagnetic (EM) physics “standard” and “low energy” sub-libraries of the general purpose Geant4 Monte Carlo simulation toolkit are described. These upgrades are relevant to different application domains including high energy physics, medical physics and space science. Validation results are presented and discussed.
Evaluation of Monte Carlo Electron-Transport Algorithms in the Integrated Tiger Series Codes for Stochastic-Media Simulations
Authors: Franke Brian C. , Kensek Ronald P. , Prinja Anil K.
Published: June 2014

Stochastic-media simulations require numerous boundary crossings. We consider two Monte Carlo electron transport approaches and evaluate accuracy with numerous material boundaries. In the condensed-history method, approximations are made based on infinite-medium solutions for multiple scattering over some track length. Typically, further approximations are employed for material-boundary crossings where infinite-medium solutions become invalid. We have previously explored an alternative "condensed transport" formulation, a Generalized Boltzmann-Fokker-Planck GBFP method, which requires no special boundary treatment but instead uses approximations to the electron-scattering cross sections. Some limited capabilities for analog transport and a GBFP method have been implemented in the Integrated Tiger Series (ITS) codes. Improvements have been made to the condensed history algorithm. The performance of the ITS condensed-history and condensed-transport algorithms are assessed for material-boundary crossings. These assessments are made both by introducing artificial material boundaries and by comparison to analog Monte Carlo simulations.
Enhanced Electron-Photon Transport in MCNP6
Authors: Grady Hughes H.
Published: June 2014

Recently a variety of improved and enhanced methods for low-energy photon/electron transport have been developed for the Monte Carlo particle transport code MCNP6. Aspects of this development include a significant reworking of the MCNP coding to allow for consideration of much more detail in atomic relaxation processes, new algorithms for reading and processing the Evaluated-Nuclear-Data-File photon, electron, and relaxation data capable of supporting such detailed models, and extension of the electron/photon transport energy range below the traditional 1-kilovolt limit in MCNP, with the goal of performing transport of electrons and photons down to energies in the few-electron-volt range. In this paper we provide an overview of these developments.
A Particle Population Control Method for Dynamic Monte Carlo
Authors: Sweezy Jeremy , Nolen Steve , Adams Terry , Zukaitis Anthony
Published: June 2014

A general particle population control method has been derived from splitting and Russian Roulette for dynamic Monte Carlo particle transport. A well-known particle population control method, known as the particle population comb, has been shown to be a special case of this general method. This general method has been incorporated in Los Alamos National Laboratory’s Monte Carlo Application Toolkit (MCATK) and examples of it’s use are shown for both super-critical and sub-critical systems.
Time-Dependent Tracking in the Monte Carlo Application Toolkit
Authors: Nolen Steven , Adams Terry , Sweezy Jeremy , Zukaitis Anthony
Published: June 2014

The Monte Carlo Application Toolkit (MCATK) provides solution options for problems with time varying properties in mesh geometries and simple solid bodies. The paper describes the toolkit’s mechanism for handling time varying problems focusing on managing the particle population. The included results show the robustness of this approach for systems with varying degrees of criticality without user intervention.
An Adaptive Sampling Scheme for Deep-Penetration Calculation
Authors: Wang Ruihong , Ji Zhicheng , Pei Lucheng
Published: June 2014

In this paper, a kind of particle transport random walk system under the emission point as a sampling station is built. Then, an adaptive sampling scheme which aims to get better solution with the achieved information is derived. The numerical results show that the adaptive scheme could overcome the difficulty of underestimation to the result in some degree, and the adaptive importance sampling method gets satisfactory results as well.
Depletion Calculations Based on Perturbations. Application to the Study of a Rep-Like Assembly at Beginning of Cycle with TRIPOLI-4®.
Authors: Dieudonne Cyril , Dumonteil Eric , Malvagi Fausto , Diop Cheikh M’Backé
Published: June 2014

For several years, Monte Carlo burnup/depletion codes have appeared, which couple Monte Carlo codes to simulate the neutron transport to deterministic methods, which handle the medium depletion due to the neutron flux. Solving Boltzmann and Bateman equations in such a way allows to track fine 3-dimensional effects and to get rid of multi-group hypotheses done by deterministic solvers. The counterpart is the prohibitive calculation time due to the Monte Carlo solver called at each time step.

In this paper we present a methodology to avoid the repetitive and time-expensive Monte Carlo simulations, and to replace them by perturbation calculations: indeed the different burnup steps may be seen as perturbations of the isotopic concentration of an initial Monte Carlo simulation. In a first time we will present this method, and provide details on the perturbative technique used, namely the correlated sampling. In a second time the implementation of this method in the TRIPOLI-4®code will be discussed, as well as the precise calculation scheme a meme to bring important speed-up of the depletion calculation. Finally, this technique will be used to calculate the depletion of a REP-like assembly, studied at beginning of its cycle. After having validated the method with a reference calculation we will show that it can speed-up by nearly an order of magnitude standard Monte-Carlo depletion codes.
Monte Carlo Approaches for Uncertainty Quantification of Criticality for System Dimensions
Authors: Kiedrowski Brian C. , Brown Forrest B.
Published: June 2014

MCNP6 is used to estimate uncertainties from geometric tolerances using forward and adjoint methods. Results are obtained and the forward and adjoint approaches appear to agree in some cases where the responses are not non-linearly correlated. In other cases, the uncertainties in k disagree for reasons not yet known.
Optimised Iteration in Coupled Monte Carlo – Thermal-Hydraulics Calculations
Authors: Eduard Hoogenboom J. , Dufek Jan
Published: June 2014

This paper describes an optimised iteration scheme for the number of neutron histories and the relaxation factor in successive iterations of coupled Monte Carlo and thermal-hydraulic reactor calculations based on the stochastic iteration method. The scheme results in an increasing number of neutron histories for the Monte Carlo calculation in successive iteration steps and a decreasing relaxation factor for the spatial power distribution to be used as input to the thermal-hydraulics calculation. The theoretical basis is discussed in detail and practical consequences of the scheme are shown, among which a nearly linear increase per iteration of the number of cycles in the Monte Carlo calculation. The scheme is demonstrated for a full PWR type fuel assembly. Results are shown for the axial power distribution during several iteration steps. A few alternative iteration method are also tested and it is concluded that the presented iteration method is near optimal.
Applications of Adjoint-Based Techniques in Continuous-Energy Monte Carlo Criticality Calculations
Authors: Kiedrowski Brian C. , Brown Forrest B.
Published: June 2014

The method for performing adjoint-weighted tallies in MCNP is discussed along with the applications of point kinetics, perturbation theory, and nuclear data sensitivities. Future applications and extensions such as localized adjoint-weighted fluxes, more sophisticated kinetics models, higher-order perturbation theory, and temperature coefficients.
Evaluation of Higher Order Mode Components of Fission Source Distribution in Mote-Carlo Calculation
Authors: Sakata Kotaro , Endo Tomohiro , Yamamoto Akio
Published: June 2014

In order to accurately estimate uncertainty of fission source distribution in a Monte Carlo calculation, a theoretical model predicting behavior of higher order modes in fission source distribution at each cycle is proposed considering the inter-cycle correlation. In a Monte Carlo calculation, statistical noise stimulates higher order modes in fission source distribution at each cycle thus the fission source distribution may not be fully converged to a fundamental mode even if number of power iterations is sufficiently large. This is especially true for a large geometry with high dominance ratio and is one of the root cause of underestimation of statistical uncertainty. In the present model, a fission source distribution is expanded by a series of higher order modes and their behaviors are estimated by variation of expansion coefficients at each cycle. The expansion coefficients show fluctuation due to statistical noise during a Monte Carlo calculation. By observing decay ratio of higher order modes in PWR single assembly geometry, validity of the present theoretical model is confirmed. Furthermore, behaviors of higher order mode components in PWR single or multi fuel assemblies geometry are investigated.
Continuous-Energy Adjoint Flux and Perturbation Calculation using the Iterated Fission Probability Method in Monte Carlo Code TRIPOLI-4® and Underlying Applications
Authors: Truchet G. , Leconte P. , Peneliau Y. , Santamarina A. , Malvagi F.
Published: June 2014

Pile-oscillation experiments are performed in the MINERVE reactor at the CEA Cadarache to improve nuclear data accuracy. In order to precisely calculate small reactivity variations (<10 pcm) obtained in these experiments, a reference calculation need to be achieved. This calculation may be accomplished using the continuous-energy Monte Carlo code TRIPOLI-4® by using the eigenvalue difference method. This “direct” method has shown limitations in the evaluation of very small reactivity effects because it needs to reach a very small variance associated to the reactivity in both states. To answer this problem, it has been decided to implement the exact perturbation theory in TRIPOLI-4® and, consequently, to calculate a continuous-energy adjoint flux. The Iterated Fission Probability (IFP) method was chosen because it has shown great results in some other Monte Carlo codes. The IFP method uses a forward calculation to compute the adjoint flux, and consequently, it does not rely on complex code modifications but on the physical definition of the adjoint flux as a phase-space neutron importance. In the first part of this paper, the IFP method implemented in TRIPOLI-4® is described. To illustrate the effciency of the method, several adjoint fluxes are calculated and compared with their equivalent obtained by the deterministic code APOLLO-2. The new implementation can calculate angular adjoint flux. In the second part, a procedure to carry out an exact perturbation calculation is described. A single cell benchmark has been used to test the accuracy of the method, compared with the “direct” estimation of the perturbation. Once again the method based on the IFP shows good agreement for a calculation time far more inferior to the “direct” method. The main advantage of the method is that the relative accuracy of the reactivity variation does not depend on the magnitude of the variation itself, which allows us to calculate very small reactivity perturbations with high precision. Other applications of this perturbation method are presented and tested like the calculation of exact kinetic parameters (βeff, Λeff) or sensitivity parameters.