MELCOR Integrated Severe Accident Code for Non-LWR applications
Abstract not provided.
Publications
Modeling Heat Pipe Reactors in the MELCOR Severe Accident Code
Abstract not provided.
MELCOR Integrated Severe Accident Code for Non-LWR applications
Abstract not provided.
MELCOR Code Change History (Revision 14959 to 18019)
This document summarily provides brief descriptions of the MELCOR code enhancement made between code revision number 14959and 18019. Revision 14959 represents the previous official code release; therefore, the modeling features described within this document are provided to assist users that update to the newest official MELCOR code release, 18019. Along with the newly updated MELCOR Users Guide and Reference Manual, users are aware and able to assess the new capabilities for their modeling and analysis applications.
MELCOR Code Change History: Revision 11932 to 14959 Patch Release Addendum
This document summarily provides brief descriptions of the MELCOR code enhancement made between code revision number 11932 and 14959. Revision 11932 represents the last official code release; therefore, the modeling features described within this document are provided to assist users that update to the newest official MELCOR code release, 14959. Along with the newly updated MELCOR Users' Guide [2] and Reference Manual [3], users will be aware and able to assess the new capabilities for their modeling and analysis applications. Following the official release an addendum section has been added to this report detailing modifications made to the official release which support the accompanying patch release. The addendums address user reported issues and previously known issues within the official code release which extends the original Quick look document to also support the patch release. Furthermore, the addendums section documents the recent changes to input records in the Users' Guide applicable to the patch release and corrects a few issues in the revision 14959 release as well. This page left blank.
Preliminary Results of ASTEC-MELCOR Crosswalk
Abstract not provided.
Informing Fukushima Decommissioning with Extended MELCOR Calculations
Abstract not provided.
BSAF Phase 2: MELCOR Analysis Activities
Abstract not provided.
Sensitivity Analysis of OECD Benchmark Tests in BISON
This report summarizes a NEAMS (Nuclear Energy Advanced Modeling and Simulation) project focused on sensitivity analysis of a fuels performance benchmark problem. The benchmark problem was defined by the Uncertainty Analysis in Modeling working group of the Nuclear Science Committee, part of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD ). The benchmark problem involv ed steady - state behavior of a fuel pin in a Pressurized Water Reactor (PWR). The problem was created in the BISON Fuels Performance code. Dakota was used to generate and analyze 300 samples of 17 input parameters defining core boundary conditions, manuf acturing tolerances , and fuel properties. There were 24 responses of interest, including fuel centerline temperatures at a variety of locations and burnup levels, fission gas released, axial elongation of the fuel pin, etc. Pearson and Spearman correlatio n coefficients and Sobol' variance - based indices were used to perform the sensitivity analysis. This report summarizes the process and presents results from this study.
Optimization and parallelization of the thermal-hydraulic subchannel code CTF for high-fidelity multi-physics applications
Annals of Nuclear Energy
This paper describes major improvements to the computational infrastructure of the CTF subchannel code so that full-core, pincell-resolved (i.e., one computational subchannel per real bundle flow channel) simulations can now be performed in much shorter run-times, either in stand-alone mode or as part of coupled-code multi-physics calculations. These improvements support the goals of the Department Of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL) Energy Innovation Hub to develop high fidelity multi-physics simulation tools for nuclear energy design and analysis. A set of serial code optimizations - including fixing computational inefficiencies, optimizing the numerical approach, and making smarter data storage choices - are first described and shown to reduce both execution time and memory usage by about a factor of ten. Next, a "single program multiple data" parallelization strategy targeting distributed memory "multiple instruction multiple data" platforms utilizing domain decomposition is presented. In this approach, data communication between processors is accomplished by inserting standard Message-Passing Interface (MPI) calls at strategic points in the code. The domain decomposition approach implemented assigns one MPI process to each fuel assembly, with each domain being represented by its own CTF input file. The creation of CTF input files, both for serial and parallel runs, is also fully automated through use of a pressurized water reactor (PWR) pre-processor utility that uses a greatly simplified set of user input compared with the traditional CTF input. To run CTF in parallel, two additional libraries are currently needed: MPI, for inter-processor message passing, and the Parallel Extensible Toolkit for Scientific Computation (PETSc), which is used to solve the global pressure matrix in parallel. Results presented include a set of testing and verification calculations and performance tests assessing parallel scaling characteristics up to a full-core, pincell-resolved model of a PWR core containing 193 17 × 17 assemblies under hot full-power conditions. This model, representative of Watts Bar Unit 1 and containing about 56,000 pins, was modeled with roughly 59,000 subchannels, leading to about 2.8 million thermal-hydraulic control volumes in total. Results demonstrate that CTF can now perform full-core analysis of a PWR (not previously possible owing to excessively long runtimes and memory requirements) on the order of 20 min. This new capability not only is useful to stand-alone CTF users, but also is being leveraged in support of coupled code multi-physics calculations being done in the CASL program.
Sensitivity Analysis of the Gap Heat Transfer Model in BISON
This report summarizes the result of a NEAMS project focused on sensitivity analysis of the heat transfer model in the gap between the fuel rod and the cladding used in the BISON fuel performance code of Idaho National Laboratory. Using the gap heat transfer models in BISON, the sensitivity of the modeling parameters and the associated responses is investigated. The study results in a quantitative assessment of the role of various parameters in the analysis of gap heat transfer in nuclear fuel.
Virtual Nuclear Reactor Modeling Using LIME
Abstract not provided.
Coupling for a Virtual Nuclear Reactor
Abstract not provided.
LIME: Software for Robust and Flexible Multiphysics Coupling
Abstract not provided.
Sodium fast reactor safety and licensing research plan. Volume II
Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.
VERA MultiphysicsCoupling with LIME: Code Requirements
Abstract not provided.
An introduction to LIME 1.0 and its use in coupling codes for multiphysics simulations
LIME is a small software package for creating multiphysics simulation codes. The name was formed as an acronym denoting 'Lightweight Integrating Multiphysics Environment for coupling codes.' LIME is intended to be especially useful when separate computer codes (which may be written in any standard computer language) already exist to solve different parts of a multiphysics problem. LIME provides the key high-level software (written in C++), a well defined approach (with example templates), and interface requirements to enable the assembly of multiple physics codes into a single coupled-multiphysics simulation code. In this report we introduce important software design characteristics of LIME, describe key components of a typical multiphysics application that might be created using LIME, and provide basic examples of its use - including the customized software that must be written by a user. We also describe the types of modifications that may be needed to individual physics codes in order for them to be incorporated into a LIME-based multiphysics application.
Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety
This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the experienced user-base and the experimental validation base was decaying away quickly.
ODTLES simulations of wall-bounded turbulent flows
Physics of Fluids
Abstract not provided.
Multiphysics coupling via LIME :
Abstract not provided.
A theory manual for multi-physics code coupling in LIME
The Lightweight Integrating Multi-physics Environment (LIME) is a software package for creating multi-physics simulation codes. Its primary application space is when computer codes are currently available to solve different parts of a multi-physics problem and now need to be coupled with other such codes. In this report we define a common domain language for discussing multi-physics coupling and describe the basic theory associated with multiphysics coupling algorithms that are to be supported in LIME. We provide an assessment of coupling techniques for both steady-state and time dependent coupled systems. Example couplings are also demonstrated.
LIME : lightweight integrated multiphysics environment
Abstract not provided.
Mathematical and computational tools for predictive simulation of complex coupled systems under uncertainty
Abstract not provided.
Foundational development of an advanced nuclear reactor integrated safety code
This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.
Development of a lightweight multi-scale multi-physics coupling strategy for nuclear reactor safety simulations
Abstract not provided.
Results 1–25 of 40