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[en] The neutronics of reactors in operating (critical) states is generally modeled as a mathematical eigenvalue problem, with the multiplication constant, the eigenvalue of the solution, being the neutron 'balancing factor' that brings equality between neutron production and loss terms. In sub-critical assemblies, however, the mathematical problem is no longer an eigenvalue one, but rather an exactly determined linear algebra problem 'driven' by a fixed-source term. In Canada, the Refuelling Simulation Program (RFSP) is the industry-standard two-group neutron diffusion code for pressurized heavy water reactors. This paper describes a modification to RFSP to allow fixed-source calculations, and hence to allow the modelling of sub-critical systems. The new capability in RFSP was applied to a proposed new guaranteed shutdown system ('rod-based' GSS) for the Pickering B nuclear generating station. This system was modelled with both RFSP and a fixed-source Monte- Carlo model using MCNP. Comparisons are made between both the fixed-source and eigenvalue solutions for each code (RFSP and MCNP), and also between the RFSP fixed-source solution and the MCNP fixed-source solution. (authors)
[en] Highlights: • A flux reconstruction method is presented that uses a 3D transport theory form factor. • 3D form factor is a 2D xy-plane component times an approximate 1D z-axis component. • Method is used to simulate travelling flux detector scan (TFD scan) readings. - Abstract: Even with current computing capabilities, detailed full core three-dimensional (3-D) transport calculations are still not practical. However, if we are satisfied with knowing only the average values of spatial flux distributions, the 3-D diffusion solution will constitute the final solution. On the other hand, in reactor design and safety analysis, direct information about the local flux distribution for the heterogeneous assemblies is required to assess the design and determine the safety margins. For this reason, after having solved the full-reactor-core problem, we have to look into the possibilities of recovering in a second step the information on local properties of single heterogeneous assemblies. In particular, the detector readings at detector locations are derived using these global homogenized parameters by applying appropriate numerical methods such as advanced interpolations. In this paper, we propose a method based on flux reconstruction to calculate the simulated detector readings in three-dimensions with high fidelity. Data from detector readings are very important in ensuring optimal reactor operations as well as in detecting any deviations from normal operations. Thus, calculating the detector readings with high fidelity will allow improvements to operating and safety margins. To validate this method, comparisons between detector reading simulation results and measurements from an operating CANDU reactor will be conducted and results will be presented.