OECD/NEA benchmark on 3-D deterministic transport calculations
without spatial homogenization (3-D C5G7 MOX)
 
  Byung-Chan Na e-mail
 

OECD Nuclear Energy Agency


Voice:
+33 1 45 24 10 91
web: http: //www.nea.fr

 
description
 
A direct calculation for whole core heterogeneous geometries was not feasible in the past due to the limited capability of computers. One had to rely upon homogenization techniques to collapse the spatial heterogeneities into a tractable homogenous description. These homogenization techniques can introduce substantial error into the flux distribution and consequently reaction rates in the homogenized zone can be significantly in error. With modern computational abilities, direct whole core heterogeneous calculations are becoming ever increasingly more feasible. Given the trend in computational ability observed thus far, it is not unreasonable to assume that, with time, such whole core calculations will eventually become common place.

An OECD/NEA benchmark problem was therefore created to test the ability of modern deterministic transport methods and codes to treat such reactor core problems without spatial homogenization. Participants in the benchmark are invited to present details of their results and calculation methods at this session.

 
 
 
     

 

The Changing RÙle of Analytical Transport Methods
in NuclearScience and Engineering and Related Fields
 
  B. Ganapol e-mail
 
 

Professor
University of Arizona
Department of Aerospace and Mechanical Engineering
Tucson, AZ


web: personal page

 
 
description
 
With advances in computational platform efficiency and the development of more powerful numerical methods, the rÙle of analytical neutron transport theory as a predictive reactor design tool has given way to direct computation. In the mid '60s, the zenith of analytical transport theory, there was hardly an issue of NS&E that did not contain at least one article on a neutron transport theoretic investigation. In today's computationally intensive reactor physics environment, no one would seriously apply a purely theoretical transport study to predict realistic reactor behavior. Simultaneously, increased computational power has enabled more sophisticated and comprehensive analytical transport studies. While the existence of analytical neutron transport theory has certainly not disappeared, its nature has changed. For example, semi-analytical benchmark quality solutions in the multidimensional multigroup approximation have emerged for both steady state and transient transport scenarios. The concept of embedding benchmarks to enable online real-time diagnostics is seriously being studied. Fusing highly accurate 2- and 3- D solutions with Monte Carlo to reduce variance and permit problems, which currently are not possible because of low signal to noise, to become reality. The intent of this session is to encourage transport theorists to showcase novel applications of analytical transport theory methods in nuclear science and engineering and related fields.
 
 
 
Resonance self-shielding models and algorithms
for reactor calculations
 
  M. Ouisloumen e-mail
 
  Westinghouse Elec. Co.  
 
  H. Matsumoto e-mail
 
  Mitsubishi Heavy Industries Ltd  
 
description
 
The isotopic resonance self-shielding calculation is an essential part in the process of fuel assembly calculations. It is also a very complex problem that involves physics models as well as numerical algorithms.
Although the self-shielding of the resonance cross-section is a problem that was dealt with at the beginning of the reactor physics codes development, it is not yet solved in a satisfactory manner. With increasing demand in core calculation accuracy, the approximations used in the old lattice code generation to treat the energy and space mutual isotopic resonance interferences are no longer sufficient. The complexity in both the geometry and fuel material of the new generation of reactors makes this a very challenging problem.
In this session we expect papers on the latest advances and developments in resonance self-shielding methods. This includes models and numerical algorithms developed to solve this problem.
 
 
 
Monte Carlo Simulation: recent trends in modeling and applications
 
  Mohamed Eid e-mail
 
 

CEA/DEN/DM2S/SERMA
CEA de Saclay, F-91191 Gif sur Yvette cedex


voice:†+33 1 6908 3175
Fax:†+33 1 6908 9935

 
 
description
 
The session aims at monitoring recent trends in Monte-Carlo Simulation modeling and application. Monte-Carlo Simulation methods receive an increasing interest from the Mathematics & Computing community as well as engineers and industry. Progress in numerical methods and technical development of powerful computers sustain this regain of interest in Monte-Carlo Simulation. Engineers and industry are getting more familiar with Monte-Carlo Simulation, as well. The session will monitor the most recent R&D trends both in modeling and applications.
 
 
 
iterative methods for transport problems
 
  Marvin L. Adams e-mail
 
 

Professor
Department of Nuclear Engineering
Texas A&M, USA


web: personal page

 
 
description
 
Deterministic particle-transport problems of practical interest usually involve large numbers of equations and unknowns. Iterative methods for solving these large systems have a significant impact on the kinds of problems that can be solved in practice. In this session we expect papers on the latest advances in efficient, rapidly convergent iterative methods for transport problems. This includes methods that are well suited for use on massively parallel architectures.
 
 
 
Multi-physics coupled code systems
for nuclear reactor design and safety
 
  Kostadin N. Ivanov e-mail
 
 

Professor
Nuclear Engineering Program
Department of Mechanical and Nuclear Engineering
The Pennsylvania State University
230 Reber Building, PA 16802, USA


voice:†+1 (814) 865-0040
Fax:†+1 (814) 865-8499
web: to get to personal page

 
 
description
 
Current trends in nuclear power generation and regulation as well as the design of next generation reactor concepts along with the continuing computer technology progress stimulate the development, qualification and application of multi-physics coupled code systems. These efforts have been focused on both extending the analysis capabilities by coupling models, which simulate different phenomena and processes or different components of a nuclear power plant as well as refining the scale and level of detail of the coupling. The proposed special technical session will cover: coupling of models of different plant components; coupled neutronics/thermal-hydraulics on both global (assembly-wise) and local (pin-wise) levels; modeling of coupled neutronics/thermal- hydraulics/fuel performance/mechanical phenomena; improving coupled calculations by utilizing CFD calculations in coarse-mesh thermal-hydraulic models; uncertainty analysis methods for coupled calculations; and advances in numerical and computation techniques for coupled code simulations.
 
 
 
Numerical modelling of underground radioactive waste repository:
a focus on radionuclides transfer through geological medium
 
  C. Serres e-mail
 
 

IRSN


web: IRSN page

 
 
  B. amaziane e-mail
 
  Université de Pau
CNRS UMR 5142


web: Université de Pau

 
 
description
 
A solution proposed to close the nuclear fuel cycle is to dispose radioactive waste in deep geological formations chosen for their ability to attenuate any possible release of radionuclides in the geosphere. Release and transport simulations of radionuclides from the repository to the biosphere is an important issue of the safety assessment process in order to quantify the confinement capability of design options and host rock. The main migration mechanism is transport in groundwater under unsaturated/saturated conditions. A mathematical formulation leads to a coupled system of partial differential equations that includes an elliptic pressure-velocity equation and a diffusion-convection concentration equation. Diffusive and advective terms are of various importances with respect of each others, depending on the characteristics of the medium. Because of the different characteristics of the geological medium, associated hydrogeological input data may differ from several orders of magnitude. When the concentration equation is convection dominated a special care should be taken in discretization but the numerical scheme has also to be able to account for the diffusion-dispersion term that cannot be neglected in several cases. Possible topics of interest for this session are (not exclusive):
fleche.gif Advances in 2D/3D robust schemes applicable for unstructured grids
   and guarantying that the approximate solution has various interesting properties
   which correspond to the properties of the physical solution,
fleche.gif Advances in homogenization methods to upscale physical properties
   or transport mechanisms at the large scale of the repository,
fleche.gif Advances in 3D meshing to capture geological settings but also verifying
   good properties for FE/FV discretization,
fleche.gif A posteriori error estimators and adaptive meshes
 
 
 
     
high temperature reactors physics and methods
 
  Abderrafi M. Ougouag e-mail
 
  Frederik Reitsma e-mail
     
 
description
 
In recent years, the two main variants of the gas-cooled High Temperature Reactor (HTR), the Pebble Bed (PBR) and the Prismatic concepts, have received renewed great emphasis in the nuclear reactor research and development community. The efforts have led to the construction of HTTR in Japan and of HTR-10 in China. Other efforts around the world, notably in South Africa (PBMR), in China (high power PBR), in the United States (NGNP), and in France and Korea portend of great advances to come. All of the HTR plans share the use of the TRISO fuel design or variants thereof and all utilize graphite as a moderator and helium as a coolant. The methods for analyzing these reactors must therefore also share many common features. Research and development (R&D) work on methods for analyzing these reactors have proceeded alongside the design efforts. Some of these R&D efforts include the development of improved and advanced codes and the development of mathematical methods for the planning and design of such reactors. It is the intent for this session to bring together the most current advances in the methods for design, analysis, optimization, and safety evaluation of the new HTR concepts. In addition, the session will incorporate reviews with a historical perspective. Papers and presentations are planned on neutronics methods, mathematical models, core fuel cycle advances as well as other mathematical developments, for both the pebble bed and the prismatic reactor concepts. Submittals on these topics are sought and encouraged.
 
 
 
Transport calculation methods for unstructured meshes
 
  Tamara Poveschenko e-mail
 
  Kurchatov Institute
Moscow, Russia
 
 
  Cassiano de Oliveira e-mail
 
  Nuclear and Radiological Engineering Program
The George W. Woodruff School of Technology
Georgia Institute of Technology
Atlanta, Georgia

web: web page

 
 
description
 
Transport calculation methods in unstructured meshes have become mature and used, in particular, as a standard tool for assembly and core design. The session addresses recent developments in this area, including improved spatial approximation, treatment of boundary conditions, linear and nonlinear acceleration techniques and 3D methodology.

Contributions are also sought in the use of spectral calculations for the generation of multigroup cross sections with particular emphasis on anisotropic (PN) cross sections.

 
 
 
Recent Developments in Stochastic and Deterministic Methods
for Charged Particle Transport
 
  Anil K. Prinja e-mail
 
 

Professor and Associate Chair
Chemical and Nuclear Engineering Department
209 Farris Engineering Center
The University of New Mexico, USA


voice: (505) 277-4600
Fax: (505) 277-5433

 
 
description
 

Charged particle elastic nuclear and inelastic electronic interactions are characterized by extremely short mean free paths and near singular differential cross sections that are highly peaked about forward directions and small energy transfers. Numerical solution of the underlying linear transport equation to investigate the transverse spatial spreading and energy straggling of pencil beams of charged radiation presents a formidable challenge even with presently available computational hardware and software.
In particular, it is clear that traditional methods of solution developed and matured for comparatively low energy neutral particle transport with weakly anisotropic scattering cannot be readily adapted to treat charged particles.
The field calls for creative modeling and computationally efficient approaches to this hallenging transport problem with widespread applications
.

Papers are invited in theoretical and mathermatical modeling, numerical methods development and computational implementation issues, for the interaction and transport of charged particles (electrons, light and heavy ions, exotic particles) in amorphous media, specifically addressing the handling of highly forward peaked scattering and energy-loss straggling. Work based on deterministic numerical schemes and Monte Carlo simulation for analog, nonanalog and condensed history formulations is strongly encouraged, as are theoretical approaches that complement such numerical work.

 
 
 
Stochastic Transport
 
  Philippe Humbert e-mail
 
  CEA/DAM, France  
 
  Edward W. Larsen e-mail
 
 

Professor University of Michigan
Department of Nuclear Engineering
Ann Arbor, MI


web: personal page

 
 
description
 
This special session will cover theoretical and computational aspects of stochastic
particle transport, including neutron noise analysis, the development of fission chains
in super-critical systems (the extinction/initiation problem), and particle transport in
stochastic media. The session will involve mathematical modeling, deterministic and
Monte Carlo numerical methods, and experimental applications that relate to
the mathematical/numerical models. These applications can include stochastic aspects
of such physical problems as subcritical multiplication measurements, monitoring
of accelerator-driven systems, the behavior of pulsed reactors, criticality accident
analysis, and particle transport in pebble-bed reactors.
 
 
 
Advances in System Modeling
for Probabilistic Risk Assessment (PRA)
 
  Tunc Aldemir
e-mail
 
 

Professor
The Ohio State University
Department of Mechanical Engineering
Building 1, Room 130B
Suite 255
650 Ackerman Road
Columbus, OH 43202 U.S.A.

Voice:†(614) 292-4627
Fax: (614) 292-3163
Web: personal page

 
 
description
 
The conventional fault-tree/event-tree approach can only represent the sequencing of events in system evolution. Explicit representation of the time element in system evolution may be needed in PRA if the system has more than one failure mode, control loops and/or hardware/process variable/software/human interaction, in general, with resulting computational complexity. Accounting for uncertainties in the probabilistic system model can be also computationally challenging. The session targets techniques
that have been proposed to address these issues.
 
 
 
Advanced Methods in 3D Deterministic Transport
 
  Christian Aussourd e-mail
 
  CEA/DAM, France  
 
description
 
Since the emergence of massively parallel processor architectures using off-the-shelf hardware components, the cost of large 3D deterministic transport calculations is becoming steadily smaller. The growing interest in these methods contributed to improve their accuracy and efficiency.

For this special session, we are seeking for outstanding lectures covering the area of 3D transport calculations, such as surveys of mature numerical schemes or parallel techniques along with insights into brand-new methods. Special attention will be paid to algorithms aimed at significantly reducing the cost of the calculations while preserving accuracy.

 
 
 
Advanced Computational Methods for Research Reactors
 
  Guy Willermoz e-mail
 
 

CEA, France

Voice: +33 4 42 25 70 82

 
 
description
 
Research reactors are essential to the nuclear development: MTR for the material and fuel improvements and radio-isotope production, neutron sources for basic and applied physics and BNCT, Test reactors for global safety qualification, critical mock ups for code and nuclear data validation.

The typically small size and heterogeneous nature of these reactors' cores lead to interesting modelling challenges where classical approximations, as Fick's law or homogenisation theory, are not applicable. This type of reactor leads to a wide diversity of physical problems: nuclear heating, material damage, tiny response volumes, extreme neutron spectra, ä . On the other hand, their limited scale makes them the ideal test bed for advanced and innovative computation methods, prior to the deployment of these techniques in power reactor calculation.

The session will offer a forum for the presentation of leading-edge developments in the computational methods employed for the design and the analysis of research reactor cores, which are going to be the foundation of the next generation simulations tools.

 
 
 

details & contacts for invited technical sessions

 

 

 

 

General Technical Sessions
fleche.gif  Deterministic Transport Theory Methods
fleche.gif  Radiative Transfer and Biomedical Applications
fleche.gif  Reactor Core Methods
fleche.gif  Monte Carlo Methods
fleche.gif  Computational Methods on Advanced Computers
fleche.gif  Physics and Computational Methods
for Advanced Reactor Concepts
fleche.gif  Perturbation Theory and Variational Methods
fleche.gif  Reactor Analysis
fleche.gif  Lattice Physics Methods
fleche.gif  Neutron Kinetics and Dynamics
fleche.gif  Criticality and Safety Analysis
fleche.gif  Accelerators and Subcritical
Systems
fleche.gif  Fuel loading optimization
and fuel design
fleche.gif  Radiation Protection and Shielding
fleche.gif  Nuclear Data Analysis and Methods
fleche.gif  Computational Fluid Dynamics
fleche.gif  Models and Methods for Material Sciences
fleche.gif  Environmental and Bioscience
fleche.gif  Computer Codes and Benchmarks
(poster session)
 
invited Technical Sessions
fleche.gif  Advanced Computational Methods for
Research Reactors   details & contacts
fleche.gif  Advanced Methods in 3D Deterministic
Transport   details & contacts
fleche.gif  Advances in System Modeling
for Probabilistic Risk Assessment (PRA)   details & contacts
fleche.gif  Stochastic Transport   details & contacts
fleche.gif  Recent Developments in Stochastic and Deterministic Methods for Charged Particle Transport   details & contacts
fleche.gif  Transport calculation methods for unstructured meshes   details & contacts
fleche.gif  High Temperature Reactors Physics and Methods details & contacts
fleche.gif  Multi-physics coupled code systems
for nuclear reactor design and safety
details & contacts
fleche.gif  Iterative methods for transport problems   details & contacts
fleche.gif  Monte Carlo Simulation: recent trends in modeling and applications  
details & contacts
fleche.gif  Resonance self-shielding models and algorithms for reactor calculations  
details & contacts
fleche.gif  The Changing Rôle of Analytical Transport Methods in Nuclear Science and Engineering and Related Fields   details & contacts
fleche.gif  Numerical modelling of underground radioactive waste repository: a focus
on radionuclides transfer through
geological medium details & contacts
fleche.gif  OECD/NEA benchmark on 3-D deterministic transport calculations without spatial homogenization
(3-D C5G7 MOX)
details & contacts

dom_f
Technical sessions