MODERNIZATION OF THE NESTLE-CANDU REACTOR SIMULATOR AND COUPLING TO SCALE-PROCESSED CROSS SECTIONS
CNS 24th Nuclear Simulation Symposium - 2012 October 15-16


Presented at:
CNS 24th Nuclear Simulation Symposium
2012 October 15-16
Location:
Ottawa,Canada
Session Title:
Technical 1: Code Development

Authors:
Shane Hart (The University of Tennessee)
G. Ivan Maldonado (The University of Tennessee)
  

Abstract

The original version of the NESTLE computer code for CANDU applications, herein referred as the NESTLE-CANDU or NESTLE-C program, was developed under sponsorship by the CNSC as a “stand-alone” program [1].  In fact, NESTLE-C emerged from the original version of NESTLE [2], applicable to light water reactors, which was written in FORTRAN 77 to solve the few-group neutron diffusion equation utilizing the Nodal Expansion Method (NEM).  Accordingly, NESTLE-C can solve the eigenvalue (criticality); eigenvalue adjoint; external fixed-source or eigenvalue initiated transient problems for CANDU reactor fuel arrangements and geometries.  This article reports a recent conversion of the NESTLE-C code to the Fortran90 standard, in addition, we highlight other code updates carried out to modularize and modernize NESTLE-C in a manner consistent with the latest updates performed with theparent NESTLE code for light water reactor (LWR) applications [3].  Also reported herein, is a simulation of a CANDU reactor employing 37-element fuel bundles, which was carried out to highlight the SCALE to NESTLE-C coupling developed for two-group collapsed and bundle homogenized cross-section generation.  The results presented are consistent with corresponding simulations that employed HELIOS generated cross-sections.

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