[en] 'Progress in Simulation Tools and Methods' was the theme of the Canadian Nuclear Society's 24th Nuclear Simulation Symposium, held in Ottawa, Ontario, Canada on October 14-16, 2012. The objective of the symposium was to provide a form for discussion and exchange of information, results and views amongst scientists, engineers and academics working in various fields of nuclear engineering. The scope of the Symposium covered all aspects of nuclear modelling and simulation, including, but not limited to: Reactor Physics; Thermalhydraulics; Safety Analysis; Fuel and Fuel Channels; and, Computer Codes and Modelling.

[en] The nuclear power industry is facing a very real challenge that affects its day-to-day activities: a rapidly aging workforce. For New Nuclear Build (NNB) countries, the challenge is even greater, having to develop a completely new workforce with little to no prior experience or exposure to nuclear power. The workforce replacement introduces workers of a new generation with different backgrounds and affinities than its predecessors. Major lifestyle differences between the new and the old generation of workers result, amongst other things, in different learning habits and needs for this new breed of learners. Interactivity, high visual content and quick access to information are now necessary to achieve high level of retention. (author)

[en] DRAGON is a deterministic code that can be used to perform lattice cell calculations based on numerical solutions of neutron transport equation. DRAGON can also be used for full core 2-D and 3-D simulations in transport. One alternative to the use of such a deterministic code consist in following the history of neutrons in the core based on statistical Monte Carlo simulation with codes like MCNP and SERPENT. This second calculation approach has been used successfully for SLOWPOKE-2 simulation in the past. Here we present a comparison between DRAGON and SERPENT calculations for the SLOWPOKE-2 reactor. We also compare the flux distribution obtained using both codes for a copper sample placed inside a small irradiation site. (author)

[en] Within the framework of the ERCOSAM-SAMARA project, co-funded by the European Union and the Russian State Atomic Energy Corporation, planning and pre-test calculations are performed to examine sensitivity parameters that can affect the break-up (erosion) of a helium (substitute for hydrogen) layer by mitigation devices (i.e., cooler, spray, or Passive Autocatalytic Recombiner - PAR). This paper reports the GOTHIC analysis results for the spray tests to be performed in the PANDA facility. The effects of spray flow rate, temperature and injection height on depressurization, erosion of helium cloud and gas transport behavior are studied. This analysis is valuable because only a limited number of conditions will be examined in the planned experiments. The study provides a useful understanding of the interaction of spray with a stratified atmosphere. (author)

[en] Reactivity device manoeuvres were completed at 30%FP reactor power, on May 18th 2012, during startup after a 2012 long outage. The main objective of these tests was to confirm that the new SDS1 and SDS2 NOP detectors were positioned at the correct locations in the reactor. Reactor physics pre-simulations were executed using a combination of the stand-alone RFSP code and the coupled RFSP/RRS Emulator code suite. The intent of these simulations was to calculate the response of the NOP detectors signals and predict the impact of the reactivity device manoeuvres on operational metrics, such as flux tilt, in order to ensure that these metrics are within acceptable operational limits. Good agreement was found between measurements and the pre-simulations. (author)

[en] A proposed modification to the OPG Pickering Nuclear Generation Station Emergency Water Supply (EWS) system was analyzed using the Industry Standard Toolset code GOTHIC to determine the acceptability of the proposed system configuration during pump start-up. The new configuration of the system included a vertical dead-ended pipe, initially filled with air. The simulation demonstrated that no significant water hammer effects were predicted and tests performed with the new configuration confirmed the analysis results. (author)

[en] Atomic Energy of Canada (AECL) has initiated a program to develop Computational Fluid Dynamics capability (CFD) to simulate single-phase and two-phase flow conditions in rod bundles. This paper presents a comparative analysis of turbulence models available in commercial CFD programs for their use in nuclear thermalhydraulic applications. The objective was accomplished using ANSYS FLUENT 6.3.26 for two geometries: a 5x5 bundle with split-type spacers and a 28-element bundle with appendages. Apart from the analysis of the two- and seven-equation models, superior predictive capabilities of the DES (Detached Eddy Simulation) and LES (Large Eddy Simulations) approaches were demonstrated for the 28-element uncrept CANDU fuel channel with appendages. (author)

[en] For the CANDU industry, analysis software is an important tool for scientists and engineers to examine issues related to safety, operation, and design. However, the software quality assurance approach currently used for these tools assumes the software is the delivered product. In this paper, we present a model that shifts the emphasis from software being the end-product to software being support for the end-product, the science. We describe a novel software development paradigm that supports this shift and provides the groundwork for re-examining the quality assurance practices used for analysis software. (author)

[en] Piping installed in nuclear power plants is affected by various degradation mechanisms and may be ruptured after being gradually thinned. The degradation mechanisms such as flow-accelerated corrosion (FAC), cavitation, liquid droplet impingement erosion (LDIE), etc., can lead to costly outages and repairs, resulting in plant reliability. In August 2008, the header pipe in the high pressure feedwater heater vent system was leaked at a Korean CANDU nuclear power plant. As the result of inspection after cutting the pipe, it was identified that the leak was evoked due to LDIE. This paper presents the numerical analysis result using ANSYS FLUENT for the purpose of identifying the cause of the leak, the comparison results of the wear rate based on the both the existing evaluation models and the measured thickness data. (author)

[en] Use of coupled codes for the safety analysis of nuclear power plants is highly desirable, as it permits multi-disciplinary studies of complex reactor behaviors and, in particular, accident simulations. The present work demonstrates the potential of the SALOME platform as an interface for creating integrated, multi-disciplinary simulations of reactor scenarios. For this purpose two codes currently in use within the Canadian nuclear industry, CATHENA and ELOCA, were coupled by means of SALOME. The coupled codes were used to model the Power Burst Facility (PBF)-CANDU Test, which was to test the thermal-mechanical behavior of PHWR (pressurized heavy water reactor) fuel during a simulated Large Loss-Of-Coolant Accident (LLOCA). The results of the SALOME-coupled simulations are compared with a previous analysis in which the two codes were coupled using a package of scripts. (author)