8th International Conference on Simulation Methods in Nuclear Science and Engineering

Workshop Details

DRAGON Workshop

DRAGON is a software for nuclear reactor lattice simulation developed and maintained by Polytechnique Montréal. Designed around two deterministic solvers for the 3D neutron transport equation (CP and MoC), it includes all the modeling tools required to generate burnup as well as local and global parameters dependent database for finite reactor core calculations. The topics that will be covered in this workshop are
  1. Overview of DRAGON and the GANLIB utilities. (G. Marleau and A. Hébert)
  2. Programming a DRAGON input deck with CLE-2000. (G. Marleau)
  3. Using DRAGON for burnup dependent lattice calculations. (A. Hébert or G. Marleau)
  4. Selecting the adequate DRAGON geometry for transport problems. (G. Marleau)
  5. Mixture composition for resonance self-shielding calculation. (A. Hébert)
  6. Creating a multi-parameter reactor database using DRAGON. (A. Hébert)
  7. Questions and conclusions. (G. Marleau and A. Hébert)

SCALE Workshop

SCALE is a comprehensive modeling and simulation suite for nuclear safety analysis and design developed and maintained by Oak Ridge National Laboratory under contract with the U.S. Nuclear Regulatory Commission, U.S. Department of Energy, and the National Nuclear Security Administration to perform reactor physics, criticality safety, radiation shielding, and spent fuel characterization for nuclear facilities and transportation/storage package designs.
In this workshop we will present:
  1. a brief overview of the capabilities in the latest SCALE 6.2.3 release,
  2. an introduction to the TSUNAMI 3D code for assessing cross section sensitivity and uncertainty as well as system similarity,
  3. an introduction to the Sampler uncertainty quantification code for quantifying uncertainty in arbitrary input parameters and nuclear data, and
  4. a summary of ongoing activities for the next SCALE 6.3.0 release including new ENDF/B-VIII nuclear data libraries and the SHIFT Monte Carlo code.

SuperMC Workshop

Super Multi-functional Calculation Program for Nuclear Design and Safety Evaluation (SuperMC) is the large-scaled integrated software system for neutronics design. Taking radiation transport calculation as the core, SuperMC supports the whole process neutronics calculation containing depletion, radiation source term/dose/biohazard, and material activation, and is also in support of the multi-physics coupling calculation of thermo-hydraulics, structural mechanics, chemistry, biology, etc. Besides, based on the cloud computing framework, it integrates accurate modeling, visualized analyses and virtual simulations and comprehensive data libraries as a whole. SuperMC can be used for the design and safety evaluation of nuclear energy systems, as well as nuclear technology application field including radiation medicine, nuclear detection and so on. SuperMC has been widely applied in over 60 nations and more than 30 mega nuclear engineering projects. It has been publicly distributed by the OECD/NEA.

The workshop of SuperMC will contain following sessions:
  1. Introduction to SuperMC: The development history, the main functions and the advanced methods of SuperMC
  2. SuperMC Tutorial:This part of the workshop will give the introduction to the usage of the automatic modeling, calculation and visualization modules.
  3. Advanced features of SuperMC:This part of the workshop will give detailed introduction of the advanced methods, such as the shutdown dose rate calculation, GWWG, etc.
  4. Discussion

COBRA-TF Workshop

1. What is CTF?
CTF is a shortened name given to the version of COBRA-TF being developed and improved by the Reactor Dynamics and Fuel Modeling Group (RDFMG) at the North Carolina State University (NCSU) in cooperation with Oak Ridge National Laboratory (ORNL). COBRA-TF is a thermal-hydraulic simulation code designed for LWR vessel analysis. It uses a two-fluid (hence the “TF” designation), three-field modeling approach. The original COBRA-TF code was developed as a thermal-hydraulic rod-bundle analysis code in 1980 by Pacific Northwest National Laboratory (PNNL) under sponsorship of the Nuclear Regulatory Commission (NRC). It was subsequently implemented in the COBRA-TRAC code system and further validated and refined as part of the FLECHT-SEASET 163-Rod Blocked Bundle Test and analysis program. Over the past several decades, the COBRA series of codes has been used extensively throughout the nuclear industry, resulting in many variants of the code being created and validated.

In the last decade, CTF has been extensively validated for Pressurized Water Reactor (PWR), Boiling Water Reactor (BWR), VVER, Small Modular Reactor (SMR), and research reactor applications. Improvements have included development of models, enhancing computational efficiency, as well as improving software quality and associated Quality Assurance (Q&A) procedures and documentation of CTF. Modifications and validation of CTF to analyze CANDU and advanced reactors is underway. As a result, CTF has become state-of-the-art sub-channel code for reactor thermal-hydraulics bundle and core analysis.

CTF have been distributed under code and collaboration agreements to different organizations, which resulted in further improvements, modifications, verification & validation activities and applications. The CTF has been included in two large projects – U.S. Department of Energy (DOE) Consortium for Advanced LWR Simulation (CASL) and European Commission (EC) NUclear REactor SAFEty simulation platform (NURESAFE). Within CASL, CTF has become an important component of VERA-CS, a “Virtual Environment for Reactor Applications”, Core Simulator (CS).

In order to leverage and combine all non-proprietary developments, improvements, modifications and error fixes as well as the available verification and validation database and application experience of CTF from different organizations and activities, a CTF User Group (UG) has been established in order to provide and maintain the so-called “gold-standard” of CTF. RDFMG/NCSU is the keeper of the gold-standard CTF and taking on the responsibility of maintaining and merging all developments and modifications. The gold-standard version of CTF uses GIT source control and is hosted on GITHUB to be accessible by all members of CTF UG. The code can be run in serial or parallel modes and is being distributed via a code agreement/license to interesting parties.

2. Topics to be covered at the workshop
‒ Overview of CTF models and correlations;
‒ Introduction to the fuel rod model of CTF - CTFFuel;
‒ CTF input and output options: pre- and post-processing tools;
‒ Coding guidelines, software quality assurance requirements, CTF versions, GITHUB access, source control, maintenance, testing, etc.;
‒ Verification and validation matrix, coverage matrix, and UQ of CTF;
‒ Demonstration of CTF applications;
‒ High-to-low model information in CTF;
‒ Multi-physics and multi-scale modeling and simulations involving CTF.

3. Who are expected audiences - it is expected that CTF users at different levels, as well as non-users would all get benefit from the workshop.