Study of the Applicability of the TRACE Code to CANDU Safety Analysis by Modeling the RD-14M Experimental Facility

Abstract

<p>A modern thermal-hydraulics system code is used to model the behaviour of a nuclear power plant's heat transport system in best-estimate safety analyses. The U.S. Nuclear Regulatory Commission's newly developed TRAC/RELAP Advanced Computational Engine (TRACE) code represents the state of the art in the field. Before a code can be accepted for use in safety analyses, however, it must first be demonstrated that it can produce meaningful results through a process of verification and validation. To demonstrate the applicability of the TRACE code to modeling CANDU type reactors, a model of the RD-14M integral test facility has thus been created using TRACE V5.0.</p> <p>The RD-14M facility is an electrically heated, scaled representation of a typical CANDU primary heat transport system. Tests at the facility are done to simulate various postulated accident scenarios, including loss of coolant accidents. A specific test (B9401) is of note for serving as benchmark data in an International Atomic Energy Agency (IAEA) code intercomparison and validation exercise, and has been modeled with TRACE.</p> <p>TRACE was demonstrated to reproduce the general trends of the test transient with fidelity comparable to other codes included in the IAEA exercise. Differences between the measured and predicted peak fuel sheath temperatures and break discharge rate were found to partially result from limitations in TRACE's constituent thermal-hydraulic models' ability to account for some horizontal flow stratification effects and break discharge phenomena common in CANDU accident analysis. Methods are proposed within the modeling methodology to account for these limitations. Ultimately, TRACE predicted the important test parameters as well as or better than some of the established codes used in the IAEA exercise, attesting to its maturity as an analysis tool.</p>