Verification of the Parallel Pin-Wise Core Simulator pCTF/PARCSv3.2 in Operational Control Rod Drop Transient Scenarios. (2nd September 2017)

• Record Type: Journal Article
• Authors: Ramos, Enrique; Roman, Jose E.; Abarca, Agustín; Miró, Rafael; Bermejo, Juan A.; Ortego, Alberto; Posada, Jose M.
• Is Part Of: Nuclear science and engineering. Volume 187:Number 3(2017)
• Page Start: 254
• Page End: 267
• Publication Date: 2017-09-02
• Subjects: Subchannel simulation -- parallel computing -- COBRA-TF; Nuclear energy -- Periodicals; Nuclear engineering -- Periodicals; Nuclear energy; Nuclear engineering; Periodicals; 539.705
• DOI: 10.1080/00295639.2017.1320892
• Languages: English
• ISSNs: 0029-5639

Abstract

Abstract: Thanks to advances in computer technology, it is feasible to obtain detailed reactor core descriptions for safety analysis of the light water reactor (LWR), in order to represent realistically the fuel elements design, as is the case for three-dimensional coupled simulations for local neutron kinetics and thermal hydraulics. This scenario requires an efficient thermal-hydraulic code that can produce a response in a reasonable time for large-scale, detailed models. In two-fluid codes, such as the thermal-hydraulic subchannel code COBRA-TF, the time restriction is even more important, since the set of equations to be solved is more complex. We have developed a message passing interface parallel version of COBRA-TF, called pCTF. The parallel code is based on a cell-oriented domain decomposition approach, and performs well in models that consist of many cells. The Jacobian matrix is computed in parallel, with each processor in charge of calculating the coefficients related to a subset of the cells. Furthermore, the resulting system of linear equations is also solved in parallel, by exploiting solvers and preconditioners from PETSc. The goal of this study is to demonstrate the capability of the recently developed pCTF/PARCS coupled code to simulate large cores with a pin-by-pin level of detail in an acceptable computational time, using for this purpose two control rod drop operational transients that took place in the core of a three-loop pressurized water reactor. As a result, the main safety parameters of the core hot channel have been calculated by the coupled code in a pin level of detail, obtaining best estimate results for this transient.