A Coupled Calculation Suite for Atucha II Operational Transients Analysis

Abstract

While more than a decade ago reactor and thermal hydraulic calculations were tedious and often needed a lot of approximations and simplifications that forced the designers to take a very conservative approach, computational resources available nowadays allow engineers to cope with increasingly complex problems in a reasonable time. The use of best-estimate calculations provides tools to justify convenient engineering margins, reduces costs, and maximises economic benefits. In this direction, a suite of coupled best-estimate specific calculation codes was developed to analyse the behaviour of the Atucha II nuclear power plant in Argentina. The developed tool includes three-dimensional spatial neutron kinetics, a channel-level model of the core thermal hydraulics with subcooled boiling correlations, a one-dimensional model of the primary and secondary circuits including pumps, steam generators, heat exchangers, and the turbine with all their associated control loops, and a complete simulation of the reactor control, limitation, and protection system working in closed-loop conditions as a faithful representation of the real power plant. In the present paper, a description of the coupling scheme between the codes involved is given, and some examples of their application to Atucha II are shown.

1/3D modeling of the core coolant circuit of a PHWR nuclear power plant

“…All the CC outlets are subcooled with differences of more than 5 °C between the center and the periphery. These results seem to be in good agreement with literature data (Mazzantini et al, 2010). Fig.…”

“…Araneo et al (2010) performed 3D CFD simulations (CFX) in order to investigate the thermal stratification and mechanical stresses under core re-flood during a LOCA event. Mazzantini et al (2010) used DYNETZ (an specific code developed by KWU for ATUCHA II) for modeling the whole NPP coupled with RELAP5 for modeling each of the 451 CC and PCE code (developed from PUMA code) to solve the local spatial neutron kinetic. The effect of subcooled boiling on the transient power distribution and vise versa was studied.…”

Unstructured Grids and the Multigroup Neutron Diffusion Equation

“…If, as is the general case, the cross sections depend on x through the flux (x) itself-such as by means of the xenon concentration or by local temperature distributions-then (1) is nonlinear. Nevertheless, this general case can be solved by successive (a) Coarse structured grid commonly used in diffusion codes such as in [9]. Each lattice cell is divided into a 2 × 2 grid for solving the neutron diffusion equation linear iterations so the basic problem can be regarded as being purely linear.…”

Negative power coefficient on PHWRs with CARA fuel

“…And the latter could be the cause for some short period oscillations 0029-5493/$ -see front matter © 2014 Elsevier B.V. All rights reserved. http://dx.doi.org/10.1016/j.nucengdes.2013.12.056 (not related to Xe oscillations) that state of the art calculations (Mazzantini et al, 2011) anticipate for Atucha II.…”