[en] The problem of estimating the time-varying net reactivity from flux measurements is solved for a point reactor kinetics model using a linear filtering technique in an H_∞ settings. In order to sue this technique, an appropriate dynamical model of the reactivity is constructed that can be embedded into the reactor model as one of its variables. A filter, which minimizes the H_∞ norm of the estimation error power spectrum, operates on neutron density measurements corrupted by noise and provides an estimate of the dynamic net reactivity. Computer simulations are performed to reveal the basic characteristics of the H_∞ optimal filter. The results of the simulation indicate that the filter can be used to determine the time-varying reactivity from neutron density measurements that have been corrupted by noise

[en] Some imprecisions are pointed out in the derivation by a previous author of a correction for the neutron flux in a heterogeneous lattice cell. In particular, the equation which is basic to the mathematical formulation does not properly take into account the lethargy and energy variables. It is shown that this and other errors in the derivation lead to an overestimation of the proposed correction. (UK)

[en] Requirements on detector and cable are defined, tests are described, and rules for the physical design of the measuring system are stated. The standard has to be equivalently applied to other instrumentations if used in place of the neutron fluence rate for determination of the power density or the power density distribution. (orig./HP)

No abstract available

[en] We analyze the one-component, one-dimensional, reaction-diffusion equation through a simple inverse method. We confine the system and fix the boundary conditions as to induce pattern formation. We analyze the stability of those patterns. Our goal is to get information about the reaction term out of the preknowledgment of the pattern. (author). 5 refs

[en] The choice of reflector material is crucial for fusion and hybrid reactors as it was for the fission reactors. Multiple reflector materials were studied for pure fusion blanket design. The purpose of reflector in fusion blanket is to enhance the tritium breeding ratio (TBR). In fusion fission hybrid blanket the roll of reflector is slightly changed as it include the fission core and the performance of fission core also needs to be optimized and evaluated with the choice of reflector material, along with the enhancement of TBR. The performance parameters of Hyb-WT are significantly influenced by the choice of reflector material. TiC is best for TRU transmutation, TBR and reduced the neutron wall loading and graphite is best for FP transmutation. Strategy of multi reflector materials gives the best TRU and FP transmutation performance and also enhanced TBR with reduced neutron wall loading and it is a better choice for Hyb-WT reflector. The neutron flux is primarily dominated by the fission neutrons

[en] We have developed a computer code that employs the nodal flux expansion method (FEM) in order to calculate the eigenvalue and the neutron flux for two energy groups for bidimensional problems. This microcomputer-oriented code employs also the two-parameter Chebyshev acceleration method for improving the iterative process convergence due to the discretization approach adopted. The results obtained so far demonstrate the applicability of the acceleration technique to the FLUX Expansion Method. (author)

No abstract available

[en] A radically new approach is presented for the estimation of flux at a point by Monte Carlo methods. A mathematical review of the method is given followed by examples of isotropic and non-isotropic sources in spherical media which could be related to the interpretation of shielding benchmark experiments. (author)

[en] A method to estimate the accuracy of approximating neutron flux with a functional system and restoring the flux using discrete measurements is proposed. The method is based on a limited character of variations of non-controllable parameters governing the neutron field. The estimations obtained may be used in both error analysis and deciding an approximating functional system as well as number and arrangement of neutron detectors. As a paragon, the one-group neutron diffusion equation of an infinite cylindrical reactor in a steady state is considered