[en] The Sixth International Conference on Simulation Methods in Nuclear Engineering was held in Montreal, Quebec, Canada on October 12-15, 2004. The conference attracted over 100 specialists from 11 countries to share approaches and techniques in the . Most of the papers and presentations focused on the analysis of difficult engineering problems associated with nuclear reactors with the aid of complex computer programs. Many included comparison of calculations with observations from experiments or actual plant operations. A measure of the advances that have been made in simulation methods was the good agreement between calculation and measurement, a fact that has led to increasing reliance on simulation methods. The titles of the various sessions give some indication of the scope of topics addressed: codes and modelling; safety analysis; fuel channels; simulator; operations support; reactor physics; thermalhydraulics; neutronics methods; and, containment

[en] The channel of a nuclear power reactor contains the fuel bundles which are made up of fuel elements distributed in a manner that creates a series of interconnected subchannels through which the coolant flows. Therefore, the 'Subchannel' modeling is a quite well technique used to predict local flow variables in nuclear fuel channels. In general, these models require the use of correlations for calculating mixing effect across the lateral interconnected subchannels. However, there is no unique set of correlations that can be used to predict a complete rage of experimental conditions. In order to avoid this major drawback, in this paper the coupling of a Genetic Algorithm (GA) to a subchannel model is presented. The use of a GA in conjunction with an appropriate objective function allows the subchannel model to internally determine the optimal values of the coefficients required by the lateral mixing correlations, without user intervention. The predictions obtained by using the coupled model were compared with experimental data collected in two interconnected subchannels. The model was able to predict the experimental trends without using external tuning coefficients. (author)

[en] Incorporation of full three-dimensional (3D) models of the reactor core into system transient codes allows for a 'best-estimate' calculation of interactions between the core behavior and plant dynamics. Recent progress in computer technology has made the development of coupled system thermal-hydraulic (T-H) and neutron kinetics code systems feasible. Considerable efforts have been made in various countries and organizations in this direction. (author)

[en] The mass distributions according to major locations and removal mechanisms for CsI are analyzed, using the ISAAC 2.0 code, to evaluate the fission product transport and deposition behavior under severe accident conditions. As for the analysis results, the mechanisms of gravitational settlement and diffusiophoresis affect evenly in an early release while the mechanism of gravitational settlement dominates in a late release. The mechanism of thermophoresis always appeared to occur but its effect on the late release was one order of a magnitude less than the diffusiophoresis while an impaction appeared to hardly have an effect during the whole period. (author)

[en] The fuel management calculation of CANDU reactor is commonly carried out every three days. It requires that the fuel management calculation should be finished within two or three hours. The selection of channels for refuelling is a challenging task for the fuelling engineer at a CANDU station. This activity is fairly time consuming and requires a considerable amount of experience, coupled to good judgement and intuition. Moreover, operating experience shows that it is very difficult to fully translate human decision processes into algorithms for a complete automation of the selection process. Therefore, an efficient refueling channel selection method is necessary. (author)

[en] The objective of this paper is to identify standardized ways of data exchange between scientific codes compared to the current ad-hoc case by case approach. The Hierarchical Data Format (HDF) developed by the National Center for Supercomputing Applications (NCSA) is recommended. To demonstrate feasibility, a prototype application of HDF for data exchange between two Canadian Industry Standard Toolset (IST) codes, SMART (fission product behaviour in containment) and ADDAM (atmospheric dispersion and public dose calculation), was successfully developed and tested. (author)

[en] The paper presents the activity performed at University of Pisa in the framework of the participation to the Phase II of the BEMUSE (Best Estimate Methods - Uncertainty and Sensitivity Evaluation) Programme. This activity has been promoted by the Working Group on Accident Management and Analysis (GAMA) and endorsed by the Committee on the Safety of Nuclear Installations (CSNI). The scope of the Programme is to perform Large Break Loss-Of-Coolant Accident (LBLOCA) analyses making reference to experimental data and to a Nuclear Power Plant (NPP) in order to address the issue of 'the capabilities of computational tools' including scaling/uncertainty analysis. The justification for such an activity comes from the consideration that a wide spectrum of uncertainty methods applied to Best Estimate codes exist and are used in research laboratories, but their practicability and/or validity is not sufficiently established to support general use of the codes and acceptance by industry and safety authorities. The consideration of the Best Estimate codes and uncertainty evaluation for Design Basis Accident (DBA), by itself, shows the safety significance of the proposed activity. (author)

[en] NB Power is planning to conduct an 18-month maintenance outage of the Point Lepreau Generating Station (PLGS) beginning in April 2008. The major activity would be the replacement of all 380 Fuel Channel and Calandria Tube Assemblies and the connecting feeder pipes. This activity is referred to as Retube. NB Power would also take advantage of this outage to conduct a number of repairs, replacements, inspections and upgrades (such as rewinding or replacing the generator, replacement of shutdown system trip computers, replacement of certain valves and expansion joints, inspection of systems not normally accessible, etc). These collective activities are referred to as Refurbishment. This would allow the station to operate for an additional 25 to 30 years. The scope of the project was determined from the outcome of a two-year study involving a detailed condition assessment of the station that examined issues relating to ageing and obsolescence. The majority of the plant components were found to be capable of supporting extended operation without needing replacement or changes. In addition to the condition assessment, a detailed review of Safety and Licensing issues associated with extended operation was performed. This included a review of known regulatory and safety issues, comparison of the station against current codes and standards, and comparison of the station against safety related modifications made to more recent CANDU 6 units. Benefit cost analyses (BCA) were performed to assist the utility in determining which changes were appropriate to include in the project scope. As a Probabilistic Safety Assessment (PSA) for PLGS did not exist at the time, a risk baseline for the station had to be determined for use in the BCA. Extensive dialogue with the Canadian Nuclear Safety Commission staff was also undertaken during this phase. A comprehensive Licensing Framework was produced upon which the CNSC provided feedback to NB Power. This feedback was important in terms of achieving clarity of the regulatory position and thus to minimize the financial risk associated with regulatory uncertainty. The Refurbishment outage is preceded by a detailed Engineering Project Phase that includes: Finalizing details of the Retube process including modeling, tooling development, site facilities and training of personnel; Perform Engineering activities related to design modifications, safety analysis and level II PSA; Construction of new waste storage structures to house Retube Waste and other additional waste storage structures for the extended life of the station; Setup necessary temporary construction facilities (offices, storage areas, change rooms, decontamination and maintenance areas); Perform detailed outage planning; Initiate development of detailed layup, commissioning and return to service procedures; Procure equipment and components. Although final project approval is still pending, NB Power has been carrying on a limited scope of activities that are important in reducing overall project financial risk. A number of these up-front activities relate to safety analysis and licensing issues related to life extension. In particular, a level II PSA along with additional safety analyses are being performed to complement that which currently support the existing Operating License for the station. This paper discusses the Safety and Licensing activities that were involved in defining the project scope and outlines the safety analysis related activities that will be performed in support of the Refurbishment project and extended operation. (author)

[en] The evaluation of uncertainty constitutes the necessary supplement of Best Estimate (BE) calculations performed to understand accident scenarios in water cooled nuclear reactors. The needs come from the imperfection of computational tools on the one side and from the interest in using such tool to get more precise evaluation of safety margins. In the present paper the approaches to uncertainty are outlined and the CIAU (Code with capability of Internal Assessment of Uncertainty) method proposed by the University of Pisa is described including ideas at the basis and results from applications. An activity in progress at the International Atomic Energy Agency (IAEA) is considered. Two approaches are distinguished that are characterized as 'propagation of code input uncertainty' and 'propagation of code output errors'. For both methods, the thermal-hydraulic code is at the centre of the process of uncertainty evaluation: in the former case the code itself is adopted to compute the error bands and to propagate the input errors, in the latter case the errors in code application to relevant measurements are used to derive the error bands. The CIAU method exploits the idea of the 'status approach' for identifying the thermalhydraulic conditions of an accident in any Nuclear Power Plant (NPP). Errors in predicting such status are derived from the comparison between predicted and measured quantities and, in the stage of the application of the method, are used to compute the uncertainty. (author)

[en] NUCIRC is a steady-state thermalhydraulic code that was developed to examine the behaviour of the heat transport system of a CANDU nuclear reactor over a wide range of single- and two-phase operating conditions. One of the principal parameters calculated by NUCIRC, which is used to assess the performance of CANDU reactors, is the critical channel power (CCP) used to determine regional overpower protection (ROP) trip set points. This paper reviews the evolution of NUCIRC plant and thermalhydraulic model development for reactor conditions that change with increasing Effective Full Power Days (EFPD) of operation. The paper first reviews the initial models used for calculating fuel channel pressure drop and critical heat flux (CHF), and the methodology for preparing site-specific (defined through a user-input file) plant models. Model development, including pressure drop models to address the manifold effect and the phenomenon of pressure tube creep, is then reviewed and discussed. Improvements in CHF and onset of significant void correlations are also presented in the context of providing accurate site-specific models for aging plants. Finally, the current status of calculating accurate and reliable CCPs for a nuclear plant is summarised. The paper concludes by making recommendations for further development to improve plant monitoring capability and to accurately assess plant performance. (author)