[en] A natural extension of the polynomic development programmed in RHENO code is presented, which adds to the variable order one-dimensional functions sum, a number of terms that represent functions of production. These new terms, which provide a direct determination of transverse escapes, are calculated from the new variables coupling among nodes: the 4 fluxes in rectangle vortices (bidimensional Cartesian geometry) or the 12 fluxes half-way through the parallelepiped edges (tridimensional Cartesian geometry). (Author)

[en] An extension to nodal diffusion theory was developed to deal with time-dependent boundary source terms. It is shown that this extension is easily introduced in advanced, full-functional diffusion theory by means of Dirac's delta spatially-dependent functions at the nodal boundaries, and allows the evaluation of reactor transients showing the propagation of neutron waves or the effects of pulsed neutron sources. A model problem with exact solutions for both the diffusion and the P₁ (telegrapher's) equations was developed to test the capabilities of the theoretical extension. The larger discrepancies occur at the earliest times computed showing, at t=50 μs, mean-square deviations between the exact diffusion solution and the numerical approximations of 2.95, 1.72, 0.32, and 0.16% for the fourth, sixth, eighth, and tenth polynomial expansion order, respectively. The availability of the exact telegrapher's solution, however, demonstrates that improved accuracy is meaningless since the mean-square deviation between the exact diffusion and exact telegrapher's solution is very similar to the mean-square deviation between the exact diffusion and the poorer (fourth-order) diffusion approximation

[en] An error estimator application was evaluated for the analysis and interpretation of nodal solutions of the neutron diffusion problem extended to reactor core level. This error estimator application relies on the residual of the equation and gives a global and local characterisation of approximate solutions by the determination of unbalances in the phase-space discretisation. It has been provided a physical meaning in terms of spurious or strange neutron reactions per cubic centimetre and second. The error estimator was applied to a wide scenario of benchmark problems, twelve of which are here reported, with the purpose of comparison between classical quality criteria and the new quality criterion based on the error estimator. The results show that the new observation criterion is able to qualify the approximate solutions in coincidence with the classical criteria but more simply and directly, concluding that approximate solutions producing a global unbalance of about (1(10)), (1(100)), (1(1000)) spurious reactions per neutron removed are respectively unacceptable, reasonably acceptable or very precise. This is achieved systematically and rigorously in the global sense because it is not possible to get an absolute local error for the diffusion problem without the help of a complementary solution; however, the observation of local errors provides valuable, practical insight into the nature of the optimal solution obtainable

[en] The success of the point-kinetics approximation to represent the time behavior of research reactors relies on the fact that research reactor cores are small enough to be neutronically tightly coupled; the core is small when measured in diffusion lengths. This fact implies that a certain change in a part of the core is immediately observed by the whole system. The propagation of changes is so fast that the core exhibits a shape function that is practically unchanged during the transient; the amplitude function, the only unknown of the problem, represents the full knowledge of the core response. One is immediately warned to look for the truth of this assumption. How small should a research reactor core be to be sure that point kinetics is a valid assumption? This question is becoming increasingly important because the tendency is to increase the size of research reactor cores to make them capable of various simultaneous uses (multipurpose characteristics), with powers in the range of tens of thermal megawatts. One of the lines of investigation at the Department of Reactor Physics is related to scenarios of Materials Test Reactor (MTR)-type research reactor transients for which space-time kinetics would bring a more profound insight than point kinetics

[en] Research reactors usually require the periodic replacement of control rods because of the burnup of the neutron absorber material, even if devices have been mainly withdrawn. The purpose of this paper is the study of the reactivity worth of hafnium as a neutron absorber in the multi-purpose-type research reactor (MPTRR). The MPTRR design develops 20 MW, and it is configured by five rows of six low-enriched uranium (200% enrichment) materials test reactor fuels inside a Zircaloy chimney. The standard control plates are given by two rows of three plates of silver-indium-cadmium alloy (85% silver, 10% indium, and 5% cadmium, in weight percents; 0.26 cm of maximum absorber width), separating the central three rows of fuel elements from the peripheral first and fifth rows. The burnup of control plates decreases the reactivity worth below the licensing criteria in ∼6 months

[en] The Multi-Purpose Type Research Reactor (MPTRR) control and first shutdown systems consist of six control plates of a Ag-In-Cd alloy. The second shutdown system is given by a liquid neutron absorber, a solution of boric acid, filling the inside of the chimney separating the core from the reflectors. The chimney is composed of four sides (four independent systems), each of them formed by two Zircaloy walls of 0.6 cm, and the empty central space inside (1.8 cm) filled with pure nitrogen during normal operation. The licensing criterion for the second shutdown system (the chimney, for brevity) requires a reactivity worth of ∼5000 pcm. The purpose of the work is the analysis of alternative liquid absorbers and the maximum reactivity worth reachable by the second shut-down system

[en] A synthesis technique is formulated as an alternative for the standard solution of the low order Interface Partial Current Equations, common in the within-group iteration schemes of Nodal Expansion and Response Matrix methods. This non-linear technique represents each node (mesh) in the iterations by its total partial in-current (jⁱⁿ), instead of the K partial in-currents at the nodal faces. At the same time, the distribution vector f is introduced, whose elements are the fractions of jⁱⁿ entering through each face, which are assumed known from the previous fission-source iteration. This simple scheme, with no other non-linear feedback required, gives a nodal formulation of the inner problem with one unknown per node and energy group, while the problem still retains the original number of unknowns. The method has been tested on the IAEA-2D and -3D, and a 2D pressurized water reactor (Biblis) benchmarks. The results show savings of about 15% to 25% of the CPU time, identical convergence despite the irregular nodal discretizations, and compatability to the rest of the iteration strategy. (author)

[en] A simple one-dimensional model for vibrational behaviour of Uf6 enrichment ultracentrifuges was developed. The equation obtained a Sturm-Liouville problem, was solved exactly and numerically for five test cases, and the results showed critical speeds and resonance modes of vibration. (author). 6 refs, 2 figs, 3 tabs

[en] Reactor physics problems have been mainly analyzed and solved by procedure-oriented methodologies. Object-oriented methodologies are a potentially powerful frame instead. In this work, object-oriented analysis for reactor physics problems is described, recognizing classes and highlighting their mathematical advantages. As a real application, a method for the solution of the neutron diffusion equation was designed within this frame, showing the simplicity and efficiency of this new focus. (author). 7 refs, 6 figs

[en] The nuclear power plants in Switzerland have followed since long the strategy to upgrade burnups, aiming fundamentally at front-end incentives to reduce the assembly production and intermediate storage costs. Some of them implemented a rather aggressive high burnup strategy, like the Gösgen nuclear power plangits core design and fuel management needed to be adapted and the safety of high burnup fuel had to be consequently demonstrated through long-term experimental programs. In particular, new zirconium-based alloys were tested and fuel behaviour during normal and transient conditions required extensive analytical and experimental validation. The paper shows a summary of the basic results of these programs for Gösgen, leading to the current licensed burnup limits that represent record commercial values for this reactor type (70, 75 and 82 GWd/tHM for maximum fuel-assembly average, maximum fuel rod average and maximum local burnups, respectively). Additionally to the technical and nuclear safety challenges that this effort required, the economic aspects of the introduction of high enrichment, high burnup core loadings on the front- and back-end fronts are briefly summarised. (author)