M&C 2017 Workshop Program (Sunday, April 16, 2016)

Transcription

1 M&C 2017 Workshop Program (Sunday, April 16, 2016) WS-ID Title Organizer Lectures Room Time Ivo Kodeli SUSD3D Cross-Section (Jozef Stefan WS-1 Sensitivity and Ivo Kodeli (JSI) 303A Institute, Uncertainty Code Slovenia) WS-2 WS-3 WS-4 WS-5 WS-6 MOOSE: Enabling Multiphysics Simulation Advanced features and the tutorial of SuperMC Uncertainty Quantification in SCALE6.2 for Depletion and Decay Problems McCARD Applications for the Research Reactor Analysis and the S/U Analysis High Fidelity Multiphysics Power Reactor Simulation Workshop (Hosted by KHNP - Description provided below Derek Gaston (MIT) Shengpeng Yu (Institute of Nuclear Energy Safety Technology, CAS) Germina Ilas (ORNL) Hyung Jin Shim (SNU) Deokjung Lee (UNIST) Derek Gaston (MIT) Jing Song (INEST) Shengpeng Yu (INEST) Bin Wu (INEST) Guangyao Sun (INEST) Will Wieselquist (ORNL) Benjamin Betzler (ORNL) Chang Hyo Kim (SNU) Hyun Chul Lee (PNU) Byungchul Lee (KAERI) Chang Je Park (Sejong University) Hyung Jin Shim (SNU) Jae Yong Lee (KHNP- Shane Stimpson (ORNL) Ben Collins (ORNL) Jinyoung Cho (KAERI) Brendan Kochunas (University of Michigan) Sooyoung Choi and Hyunsuk Lee (UNIST) Hwan Soo Lee (KHNP 303B :00-12:30 13:30-17:30

2 WS-1. SUSD3D Cross-Section Sensitivity and Uncertainty Code as Part of the XSUN Windows Interface Environment for Deterministic Radiation Transport Calculations 09:00-12:30, April 16 (Sun), 303A Organizer: Kodeli (IJS), SUSD3D code uses the first-order perturbation theory to calculate the sensitivity coefficients and standard deviation in the calculated detector responses or design parameters of interest (such as reaction rates, doses, multiplication factor - keff, effective delayed neutron fraction - eff) due to input cross sections and their uncertainties. Complex one-, two- and threedimensional shielding and criticality problems can be studied. The first version of the SUSD3D code was developed in the 1990-ies in the scope of the European Fusion Programme. Several types of uncertainties can be considered, i.e. those due to: (1) neutron/gamma multi-group cross sections, (2) energy-dependent response functions, (3) secondary angular distribution (SAD) or secondary energy distribution (SED) uncertainties. SUSD3D is now available as part of the XSUN-2017 Windows interface developed to facilitate the deterministic radiation transport and cross-section sensitivity-uncertainty calculations. The package assists the users in the preparation of input cards, rapid modification and execution of the complete chain of codes including TRANSX, PARTISN and SUSD3D, all available from the OECD/NEA Data Bank and RSICC. The objective was to make the input and output handling for these codes as user friendly as possible, passing information among the codes internally. XSUN-2017 allows an interactive viewing of results obtained from the PARTISN and SUSD3D programs. Plotting utilities include 2- dimensional (2D) color schemes of PARTISN geometries, 3D plots of the neutron flux distribution and sensitivity profile plots. The first version of the Windows interface XSUN was developed in An updated version, XSUN-2017, will be released through OECD/NEA Data Bank Computer Code Collection and RSICC early The description of the XSUN-2017 and SUSD3D code system, the recent improvements and updates will be presented (such as eff, SAD, SED S/U analysis). Examples of the use and validation will be demonstrated, including the S/U inter-comparison exercise using the SNEAK-7A and -7B benchmark experiments involving the XSUN-2017 code system comparison with the codes such as TSUNAMI-3D, SERPENT2, MCNP6, ERANOS and XSUSA, several fusion applications and the sensitivity and uncertainty analysis of the keff and eff parameters for the MYRRHA accelerator driven system (ADS).

3 WS-2. MOOSE: Enabling Multiphysics Simulation 09:00-12:30, April 16 (Sun), 303B Organizer: Derek Gaston (MIT), The open source Multiphysics Object Oriented Simulation Environment (MOOSE) software package was originally developed at Idaho National Laboratory for the purpose of supporting nuclear reactor simulation. Since that time it has grown to support scientists and engineers studying a myriad of physics including: heat conduction, solid mechanics, neutronics, chemistry, fluid flow, material evolution, porous flow and much more. One capability that sets MOOSE apart is the ability to utilize massively parallel computers and clusters. MOOSE is completely free and available at It is developed by a growing community of scientists and engineers from around the globe on GitHub ( The purpose of this tutorial is to familiarize the participants with working with MOOSE, preparing them to utilize MOOSE to study their own physics systems. In addition, there will be demonstrations of the use of MOOSE in reactor analysis. Topics covered in this tutorial will be: an overview of the design of MOOSE, an introduction to adding your own physics, the pre-existing physics modules and multi-scale simulation. Some advanced topics such as mesh adaptivity and adaptive timestepping will also be introduced. Each section will be accompanied by example applications that can be built and run on the workshop participant s laptops (in addition to being demonstrated by the speaker). All workshop materials will be made freely available before the workshop begins. WS-3. Advanced features and the tutorial of SuperMC 09:00-12:30, April 16 (Sun), 303B Organizer: Shengpeng Yu (INEST), shengpeng.yu@fds.org.cn SuperMC, a general, intelligent, accurate and precise simulation software system for the nuclear design and safety evaluation of nuclear systems, is designed to support the comprehensive neutronics calculation. The main usability features of SuperMC include automatic modeling of geometry and physics, visualization and virtual simulation and cloud computing services. The latest version of SuperMC can accomplish the transport calculation of n, γ, burnup and activation calculation, and can be applied for criticality and shield design of reactors, medical physics analysis, etc. SuperMC has been verified and validated by more than 2000 benchmark models and experiments, such as ICSBEP, SINBAD, IRPhEP, ITER, IAEA-BN600, BEAVRS, HM etc. SuperMC has been used in 50+ nations and more than 30 major nuclear engineering projects. It has been publicly distributed by OECD/NEA Data Bank. SuperMC has passed ITER benchmarking and supported to create a series of ITER neutronics reference models. The SuperMC workshop consists of two presentations. One presentation is for the advance features and development progress of SuperMC and the other presentation would give exhaustive introduction of the usage of SuperMC. Your participation will be among the highlights of the workshop.

4 WS-4. Uncertainty Quantification in SCALE6.2 for Depletion and Decay Problems 13:30-17:30, April 16 (Sun), 301 Organizer: William Wieselquist (ORNL), SCALE 6.2 contains extensive capabilities for uncertainty quantification (UQ) in depletion and decay problems. The general strategy is to wrap an existing nominal problem in the new Sampler sequence which defines two categories of uncertain input data: nuclear data and input file parameters. With nuclear data, uncertainty is available for three broad data types: decay data, fission yield data, and cross section data. With input file parameters, a distribution may be assigned to each variable by name and variables may be composed of other variables including standard mathematical operations (+, -,/,*, sin, cos, etc.). Supported distributions for input file parameters are: normal, truncated normal, uniform, and beta. Nuclear data has been assumed to have a truncated normal distribution with samples pre-calculated for speed of execution. Input file parameters are sampled on-the-fly. Workshop attendees will be given a short introductory lecture on the Sampler methodology followed by a demonstration of the following three applications: Sampler+Polaris for nodal cross-section data UQ, Sampler+ORIGEN for neutron, gamma, and decay heat source term UQ, Sampler+Polaris+ORIGEN for destructive assay isotopic UQ. Input files for the demonstrations will be provided at the beginning of the workshop. The tutorial is open to all participants at the conference. Participants wishing to follow along with the tutorial should bring their own computer, have a valid license for SCALE or the most recent version, and have this SCALE version installed on their computer.

5 WS-5. McCARD Applications for the Research Reactor Analysis and the S/U Analysis 13:30-17:30, April 16 (Sun), 302 Organizer: Hyungjin Shim (SNU), McCARD is a Monte Carlo (MC) neutron-photon transport simulation code designed exclusively for neutronics analyses of various nuclear reactor and fuel systems. McCARD estimates neutronics design parameters such as effective multiplication factor, neutron flux and current, fission power, etc. by using continuous-energy cross section libraries and detailed geometrical data of the system. Since its predecessor MCNAP was first introduced in 1999 as a MC burnup analysis tool with an ORIGEN2-type fuel depletion equation solver, it has evolved to a versatile MC tool which is capable of performing the whole-core neutronics calculations, the reactor fuel burnup analysis, the few group diffusion theory constant generation, sensitivity and uncertainty (S/U) analysis, and uncertainty propagation analysis. This workshop will focus on two subjects McCARD applications for the research reactor design and analysis and its capability of the nuclear data S/U analysis. Core design efforts using McCARD are presented for newly-deployed and -designed research reactors such as Jordan Research and Training Reactor and Kijang Research Reactor. And the McCARD S/U analysis lecture covers the MC perturbation techniques including the k-adjoint and the generalized adjoint function estimations, sensitivity estimation algorithms in the MC Wielandt method, and its applications for the nuclear data S/U analysis. For a practice of the McCARD S/U analysis, its executable and library files in the Windows system will be distributed during the workshop. WS-6. High Fidelity Multiphysics Power Reactor Simulation Workshop 13:30-17:30, April 16 (Sun), 303 Organizer: Deokjung Lee (UNIST), deokjung@unist.ac.kr This workshop will present the researches and simulation results of the high fidelity multiphysics power reactor simulation. For high fidelity light water reactor simulations, four different disciplines need to be coupled together: neutronics, thermal hydraulics, fuel behavior, and water chemistry. The presentations will include multiphysics simulation results from various state-of-the-art reactor analysis computer codes such as VERA-CS (ORNL), ntracer (SNU), MPACT (University of Michigan), BISON (INL), MAMBA (CASL), COBRA-TF (NCSU) STREAM (UNIST), and MCS (UNIST). The speakers will be experts from ORNL, University of Michigan, NCSU, SNU, and UNIST.