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Title: Two-Step Uncertainty Analysis of Watts Bar Nuclear 1 Cycle 1 with SCALE/PARCS

Authors:
 [1];  [2];  [2]; ORCiD logo [3]
  1. University of Michigan, Ann Arbor
  2. University of Michigan
  3. ORNL
Publication Date:
Research Org.:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org.:
USDOE
OSTI Identifier:
1376395
DOE Contract Number:
AC05-00OR22725
Resource Type:
Conference
Resource Relation:
Conference: M&C International Conference on Mathematics & Computational Methods Applied to Nuclear Science and Engineering - Jeju Island, , South Korea - 4/16/2017 12:00:00 AM-4/20/2017 12:00:00 AM
Country of Publication:
United States
Language:
English

Citation Formats

Xu, Kevin, Downar, Thomas, Ward, Andrew, and Jessee, Matthew Anderson. Two-Step Uncertainty Analysis of Watts Bar Nuclear 1 Cycle 1 with SCALE/PARCS. United States: N. p., 2017. Web.
Xu, Kevin, Downar, Thomas, Ward, Andrew, & Jessee, Matthew Anderson. Two-Step Uncertainty Analysis of Watts Bar Nuclear 1 Cycle 1 with SCALE/PARCS. United States.
Xu, Kevin, Downar, Thomas, Ward, Andrew, and Jessee, Matthew Anderson. Sat . "Two-Step Uncertainty Analysis of Watts Bar Nuclear 1 Cycle 1 with SCALE/PARCS". United States. doi:. https://www.osti.gov/servlets/purl/1376395.
@article{osti_1376395,
title = {Two-Step Uncertainty Analysis of Watts Bar Nuclear 1 Cycle 1 with SCALE/PARCS},
author = {Xu, Kevin and Downar, Thomas and Ward, Andrew and Jessee, Matthew Anderson},
abstractNote = {},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Sat Apr 01 00:00:00 EDT 2017},
month = {Sat Apr 01 00:00:00 EDT 2017}
}

Conference:
Other availability
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  • The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a collection of methods and software products known as VERA, the Virtual Environment for Reactor Applications, including a core simulation capability called VERA-CS. A key milestone for this endeavor is to validate VERA against measurements from operating nuclear power reactors. The first step in validation against plant data is to determine the ability of VERA to accurately simulate the initial startup physics tests for Watts Bar Nuclear Power Station, Unit 1 (WBN1) cycle 1. VERA-CS calculations were performed with the Insilico code developed at ORNL using cross sectionmore » processing from the SCALE system and the transport capabilities within the Denovo transport code using the SPN method. The calculations were performed with ENDF/B-VII.0 cross sections in 252 groups (collapsed to 23 groups for the 3D transport solution). The key results of the comparison of calculations with measurements include initial criticality, control rod worth critical configurations, control rod worth, differential boron worth, and isothermal temperature reactivity coefficient (ITC). The VERA results for these parameters show good agreement with measurements, with the exception of the ITC, which requires additional investigation. Results are also compared to those obtained with Monte Carlo methods and a current industry core simulator.« less
  • Acoustic emission (AE) monitoring of selected pressure boundary areas at TVA's Watts Bar, Unit 1 Nuclear Plant in the US during hot functional preservice testing is described. Background, methodology, and results are included. The work discussed here is a major milestone in a program supported by the US NRC to develop and demonstrate application of AE monitoring for continuous surveillance of reactor pressure boundaries to detect and evaluate growing flaws. The subject work demonstrated that anticipated problem areas can be overcome. Work is continuing to AE monitoring during reactor operation. 3 refs., 6 figs.
  • Spatial and functional coupling (including human actions) of nuclear power plant systems that lead to interdependencies are called Systems Interactions. At present, the US Nuclear Regulatory Commission (NRC) is investigating ways of integrating a systems interactions study with existing Probabilistic Risk Assessment efforts. One approach is based on graph-theoretic methods utilizing matrix representations of logic diagrams called Digraph Matrix Analysis (DMA). The objective in this report is to demonstrate the capabilities of Digraph Matrix Analysis to model an accident sequence (including front-line systems, support systems and human actions) as a continuous, well-integrated logic model in order to identify and evaluatemore » functional systems interactions. The selected accident sequence, loss of high pressure safety injection during a LOCA, was modeled and qualitative and quantitative comparisons were made to the Reactor Safety Study (WASH 1400) and other studies.« less