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Thin 33S samples for the study of the 33S(n, α)30Si cross-section at the n_TOF facility at CERN were made by thermal evaporation of 33S powder onto a dedicated substrate made of kapton covered with thin layers of copper, chromium and titanium. This method has provided for the first time bare sulfur samples a few centimeters in diameter. The samples have shown an excellent adherence with no mass loss after few years and no sublimation in vacuum at room temperature. The determination of the mass thickness of 33S has been performed by means of Rutherford backscattering spectrometry. The samples have been successfully tested under neutron irradiation.

The resonance neutron self-shielding factor, G res, is required in neutron metrology and activation data analysis. In a previous paper, the authors have shown that a dimensionless variable can be introduced which converts the dependence of G res on the physical and nuclear properties of the material samples into an universal curve, valid for the isolated resonances of any nuclide. This work presents a methodology based on the universal curve, which enables to calculate G res for a group of isolated resonances by weighting its individual contributions. A good agreement was reached with results calculated by the MCNP code and with experimental values for Mo foils and wires.

The interpretation of the sample activation in a nuclear reactor requires the knowledge of two corrective parameters: the thermal neutron self-shielding factor, G th, and the resonance neutron self-shielding factor, G res. The authors established a universal curve of G res for isolated resonances and various geometries. The present paper deals with the description of G th in foils, wires, spheres and cylinders by means of a universal curve on the basis of a dimensionless variable which includes the physical, nuclear and geometrical properties of the sample. The universal curve is in good agreement with the experimental and calculated values obtained from the literature.

The presence of a nuclide sample in an epithermal neutron field of a nuclear reactor creates a perturbation of the local neutron flux. This effect can be very important, especially if the nuclide cross-section exhibits a prominent resonance peak. To take into account the effect of the neutron flux perturbation in the sample activation, a resonance neutron selfshielding factor (Gres) must be considered. This factor depends on the geometry and dimension of the sample, as well as on the physical and nuclear properties of the nuclide. On the basis of a dimensionless variable which includes the relevant characteristics of the sample, an universal curve is proposed, which enables the determination of the factor Gres for isolated resonances of any nuclide and samples of various geometries (foils, wires and spheres). The proposed universal curve is in very good agreement with experimental and calculated values obtained from the literature. r 2003 Elsevier Science Ltd. All rights reserved.

Epithermal neutron resonance self-shielding factors in foils of materials used as activation detectors or as targets for radionuclide production have been calculated using the MCNP code. Two irradiation conditions have been considered: (a) foils immersed into an isotropic neutron flux and (b) foils submitted to a collimated neutron beam. The self-shielding factors, Gres(t), for gold, indium, manganese and cobalt have been compared with available published values. The self-shielding factor depends on various physical and nuclear parameters. However, it is shown that for the isotropic case and for high absorber elements submitted to a collimated beam, a dimensionless variable could be adopted that describes the self-shielding factors of different materials by quasi "universal curves". Gres(t) for the collimated beam are always higher than those for the isotropic case.

In a previous work, a simulation study to investigate the enhancement of the fast neutron flux inside a hollow lead cylinder located near the core of a pool type reactor has been carried out. In this complementary work, the calculation of gamma radiation spectra inside the same irradiation device was performed using the MCNP code. Two radiation spectra were considered as input sources: a typical fission gamma spectrum in a nuclear reactor, for the calculation of the socalled fission γ spectrum inside the device; and a fission neutron spectrum, for the calculation of the (n, γ) spectrum. The calculation was made considering two diffusing media adjacent to the reactor core: water and beryllium + water. The lead wall thickness, t, adopted for the hollow lead cylinder ranges from 2 to 10 cm. In order to compare the results with and without the irradiation device, reference cases (t = 0) have also been considered. The values of the attenuation factors F att (fission γ fluence) and F att [(n, γ) fluence] are presented. These results, together with those from the previous work, allow an optimisation of the irradiation conditions in order to maximise the absorbed dose due to fast neutrons.

A simulation study was carried out in order to investigate the enhancement of the fast neutron spectra in a sample irradiated near the core qf a pool type nuclear reactor. The irradiation device consisted of a lead hollow cylinder where a thin aluminium sample holder containing the sample to be irradiated is placed. Calculations were performed considering water or beryllium + water between the simulated fission neutron source and the irradiation device; the thickness of the wall of the lead cylinder was varied up to 10 cm. The fast neutron spectra in the sample position were calculated and the average fast neutron energy, the total neutron fluence. Φ(E>0.1 MeV), the fast neutron fluence, Φ(E>1 MeV), and the conversion factor, C = Φ/Φ, were determined. It was found that the lead cylinder surrounding the sample induces an appreciable enhancement of the fast neutron spectra. This effect can be described using a gain factor defined, at each point, as the ratio between the fast neutron fluences with and without the irradiation device, G =Φ/Φ(t =0). The gain of the neutron fluence increases from 2 to 4.5 when the lead thickness varies between 2 and 10 cm.

The MCNP-4A code, running on a Pentium PC, was used to simulate a conveyor belt system for on-line cement raw material analysis, namely the determination of calcium, silicon, iron and aluminium contents. The influence of different system and sample parameters (overall system geometry, neutron moderator/reflector composition and thickness, neutron absorber lining of detector shields, sample bulk density and water content) on the background count rate originating from structural materials was studied. A configuration has been obtained which corresponds to a lower background due to structural materials and a relative higher contribution from γ-rays produced in (n, γ) reactions in the bulk raw material. In this configuration, the dependence of the count rate on the sample density and water content is different in the lower and upper detectors. In the upper detectors, the count rate from inelastic reactions in lead surrounding the neutron source is shown to decrease exponentially with increasing density values, thus enabling density measurements to be made.

The introduction of an appropriate dimensionless variable, which takes into account, simultaneously, the material absorbing and scattering properties, can express the thermal neutron self-shielding factor of spherical samples by a unique curve. It is shown that the differences between the values calculated by Blaauw and those of the proposed curve are

Resonance neutron self-shielding factors for cylindrical samples of nuclides used as activation detectors or as targets for radionuclide production have been calculated using the MCNP code. These factors depend on the sample dimensions, as well as on the physical and nuclear properties of the nuclides. However, defining a dimensionless variable, which includes the relevant characteristics of the samples, it is possible to extend to cylinders a previously deduced universal curve for isolated resonances of any nuclide and samples of other geometries (foils, wires and spheres).

This paper describes a complete Monte Carlo study of the irradiator geometry of the Portuguese Gamma Irradiation Facility, using the MCNP code. In this work, eight different source rack configurations were simulated. Routine and non-routine irradiations have been studied. For routine irradiations, both the dose per source activity and the dose uniformity remain constant for all source configurations studied. For non-routine irradiations some dependence on source configuration has been found concerning the dose uniformity in the inner rows.

In this work, the resonance self-shielding factor is calculated by means of the Monte Carlo technique for different materials (Au, Co, Mn), geometries (circular foils and wires) and incidence of neutrons (isotropic field and collimated beam). The results are compared with the values obtained by other authors.