Neutron self-shielding, cover reactions and burnup corrections for reactor dosimetry applications

Neutron spectrum adjustments for reactor dosimetry applications can be made using a least squares computer code such as STAY'SL, in conjunction with the measured reaction rates, neutron cross-sections, and their uncertainties and covariances, as contained in IRDF-2002. However, prior to the spectral adjustment, corrections to the neutron cross-sections must be applied for neutron self-shielding or cover reactions. Such corrections are critical and may produce significant changes in the calculated reaction rate, especially for reactions that have large thermal or resonance cross-sections. If reactor measurements are made with highly dilute monitors, neutron self-shielding corrections may not be required. However, non-dilute monitors will always show significantly reduced reaction rates compared with dilute monitors, since thermal and resonance neutrons may be absorbed in the outer layers of a foil or wire, thereby reducing the activation rate in the interior of the material. Cover materials such as boron, cadmium or gadolinium are frequently used to suppress thermal neutrons, and cadmium ratios are used as an indicator of the ratio of thermal to epithermal or fast neutrons. Ideally, neutron transport computer codes should be used to determine the neutron self-shielding and cover corrections that are to be applied to a given material. The reason for this requirement is that neutron scattering will result in higher than expected neutron fluxes in those energy groups that correspond to large thermal or resonance neutron cross-sections. Failure to include such neutron scattering effects will result in an overestimation of the neutron self-shielding corrections. Such neutron transport calculations require the use of neutron scattering cross-sections, usually taken as the total neutron cross-sections, in addition to the neutron activation cross-sections. The total neutron cross-sections are included in IRDF-2002 for target materials that have a dosimetry quality (n,γ) reaction. Fortunately, lengthy neutron transport calculations may not be required for relatively thin samples or simple covers, where the neutron mean free paths for neutron scattering tend to be larger than the dimensions of the sample. Approximations have been developed that may allow sufficiently accurate neutron self-shielding and cover calculations. Some of the approximation formulas are described. Such approximate calculations are not necessarily applicable to all material types used for a specific reactor dosimetry application. The adequacy of the approximation may also be tested by using samples of different thickness, or comparing the results from dilute and non-dilute types of material. There are also a number of other effects that can have a significant impact on the calculation of activation rates, such as flux depression or the partial shielding of one sample by an adjacent sample. Such effects may not add linearly, especially when neutron scattering effects are significant (author)