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Title: Assessment of SFR Wire Wrap Simulation Uncertainties

Abstract

Predictive modeling and simulation of nuclear reactor performance and fuel are challenging due to the large number of coupled physical phenomena that must be addressed. Models that will be used for design or operational decisions must be analyzed for uncertainty to ascertain impacts to safety or performance. Rigorous, structured uncertainty analyses are performed by characterizing the model’s input uncertainties and then propagating the uncertainties through the model to estimate output uncertainty. This project is part of the ongoing effort to assess modeling uncertainty in Nek5000 simulations of flow configurations relevant to the advanced reactor applications of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. Three geometries are under investigation in these preliminary assessments: a 3-D pipe, a 3-D 7-pin bundle, and a single pin from the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility. Initial efforts have focused on gaining an understanding of Nek5000 modeling options and integrating Nek5000 with Dakota. These tasks are being accomplished by demonstrating the use of Dakota to assess parametric uncertainties in a simple pipe flow problem. This problem is used to optimize performance of the uncertainty quantification strategy and to estimate computational requirements for assessments of complex geometries. A sensitivity analysis to three turbulent models wasmore » conducted for a turbulent flow in a single wire wrapped pin (THOR) geometry. Section 2 briefly describes the software tools used in this study and provides appropriate references. Section 3 presents the coupling interface between Dakota and a computational fluid dynamic (CFD) code (Nek5000 or STARCCM+), with details on the workflow, the scripts used for setting up the run, and the scripts used for post-processing the output files. In Section 4, the meshing methods used to generate the THORS and 7-pin bundle meshes are explained. Sections 5, 6 and 7 present numerical results for the 3-D pipe, the single pin THORS mesh, and the 7-pin bundle mesh, respectively.« less

Authors:
 [1];  [1];  [1];  [2]
  1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division
  2. Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
Publication Date:
Research Org.:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org.:
USDOE
OSTI Identifier:
1333666
Report Number(s):
ORNL/TM-2016/540
NT0503000; NENT026; TRN: US1700769
DOE Contract Number:
AC05-00OR22725
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; SODIUM COOLED REACTORS; REACTOR SAFETY; COMPUTERIZED SIMULATION; THERMAL HYDRAULICS; PERFORMANCE; WIRES; THREE-DIMENSIONAL CALCULATIONS; CONFIGURATION; DESIGN; SENSITIVITY ANALYSIS; MATHEMATICAL MODELS; PIPES; FUEL PINS; FUEL ELEMENT CLUSTERS; TURBULENT FLOW; N CODES

Citation Formats

Delchini, Marc-Olivier G., Popov, Emilian L., Pointer, William David, and Swiler, Laura P. Assessment of SFR Wire Wrap Simulation Uncertainties. United States: N. p., 2016. Web. doi:10.2172/1333666.
Delchini, Marc-Olivier G., Popov, Emilian L., Pointer, William David, & Swiler, Laura P. Assessment of SFR Wire Wrap Simulation Uncertainties. United States. doi:10.2172/1333666.
Delchini, Marc-Olivier G., Popov, Emilian L., Pointer, William David, and Swiler, Laura P. 2016. "Assessment of SFR Wire Wrap Simulation Uncertainties". United States. doi:10.2172/1333666. https://www.osti.gov/servlets/purl/1333666.
@article{osti_1333666,
title = {Assessment of SFR Wire Wrap Simulation Uncertainties},
author = {Delchini, Marc-Olivier G. and Popov, Emilian L. and Pointer, William David and Swiler, Laura P.},
abstractNote = {Predictive modeling and simulation of nuclear reactor performance and fuel are challenging due to the large number of coupled physical phenomena that must be addressed. Models that will be used for design or operational decisions must be analyzed for uncertainty to ascertain impacts to safety or performance. Rigorous, structured uncertainty analyses are performed by characterizing the model’s input uncertainties and then propagating the uncertainties through the model to estimate output uncertainty. This project is part of the ongoing effort to assess modeling uncertainty in Nek5000 simulations of flow configurations relevant to the advanced reactor applications of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. Three geometries are under investigation in these preliminary assessments: a 3-D pipe, a 3-D 7-pin bundle, and a single pin from the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility. Initial efforts have focused on gaining an understanding of Nek5000 modeling options and integrating Nek5000 with Dakota. These tasks are being accomplished by demonstrating the use of Dakota to assess parametric uncertainties in a simple pipe flow problem. This problem is used to optimize performance of the uncertainty quantification strategy and to estimate computational requirements for assessments of complex geometries. A sensitivity analysis to three turbulent models was conducted for a turbulent flow in a single wire wrapped pin (THOR) geometry. Section 2 briefly describes the software tools used in this study and provides appropriate references. Section 3 presents the coupling interface between Dakota and a computational fluid dynamic (CFD) code (Nek5000 or STARCCM+), with details on the workflow, the scripts used for setting up the run, and the scripts used for post-processing the output files. In Section 4, the meshing methods used to generate the THORS and 7-pin bundle meshes are explained. Sections 5, 6 and 7 present numerical results for the 3-D pipe, the single pin THORS mesh, and the 7-pin bundle mesh, respectively.},
doi = {10.2172/1333666},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2016,
month = 9
}

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