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Abstract The primary X-ray spectrum depends on different parameters such as high voltage, filament current, high voltage ripple, anode angle and thickness of filter material. The objective of this work is to determine whether the unfolding technique based on the Tikhonov regularization method is accurate enough to estimate the X-ray spectrum when slight changes in the operation variables are considered. In this frame, several X-ray spectra are considered (extracted from the IPEM78 Catalogue Report) varying the main operation variables of the X-ray tube (high voltage, voltage ripple, filter thickness and filter material). With those spectra, the corresponding absorbed dose curves are obtained by simulation with a MCNP5 model reproducing a flat panel detector and a PMMA wedge. Once the absorbed dose curves are simulated and applying the unfolding Tikhonov regularization method, the unfolded spectrum is obtained, which is finally compared with the theoretical one (IPEM78 Catalogue Report). Discrepancies between unfolded and primary X-ray spectra can be attributed to the fact that this is an ill-posed problem, and the unfolding of the spectrum is strongly affected by the method used to improve the conditioning of the response function (response matrix). Highlights • Unfolding of X-ray spectra in the radiodiagnostic energy range. • Application of the Monte Carlo code MCNP5 to obtain dose curves. • Unfolding using Tikhonov regularization technique. • Accuracy of Tikhonov method. [ABSTRACT FROM AUTHOR]

All materials exposed to some neutron flux can be activated independently of the kind of the neutron source. In this study, a nuclear reactor has been considered as neutron source. In particular, the activation of control rods in a BWR is studied to obtain the doses produced around the storage pool for irradiated fuel of the plant when control rods are withdrawn from the reactor and installed into this pool. It is very important to calculate these doses because they can affect to plant workers in the area. The MCNP code based on the Monte Carlo method has been applied to simulate activation reactions produced in the control rods inserted into the reactor. Obtained activities are introduced as input into another MC model to estimate doses produced by them. The comparison of simulation results with experimental measurements allows the validation of developed models. The developed MC models have been also applied to simulate the activation of other materials, such as components of a stainless steel sample introduced into a training reactors. These models, once validated, can be applied to other situations and materials where a neutron flux can be found, not only nuclear reactors. For instance, activation analysis with an Am-Be source, neutrography techniques in both medical applications and non-destructive analysis of materials, civil engineering applications using a Troxler, analysis of materials in decommissioning of nuclear power plants, etc. [ABSTRACT FROM AUTHOR]

All materials present in the core of a nuclear reactor are activated by neutron irradiation. The activity so generated produces a dose around the material. This dose is a potential risk for workers in the surrounding area when materials are withdrawn from the reactor. Therefore, it is necessary to assess the activity generated and the dose produced. In previous works, neutron activation of control rods and doses around the storage pool where they are placed have been calculated for a Boiling Water Reactor using the MCNP5 code based on the Monte Carlo method. Most of the activation is produced indeed in stainless steel components of the nuclear reactor core not only control rods. In this work, a stainless steel sample is irradiated in the Training Reactor AKR-2 of the Technical University Dresden. Dose measurements around the sample have been performed for different times after the irradiation. Experimental dosimetric values are compared with results of Monte Carlo simulation of the irradiation. Comparison shows a good agreement. Hence, the activation Monte Carlo model can be considered as validated. [ABSTRACT FROM AUTHOR]

Nuclear reactions can generate radionuclides and it is necessary to know the activity produced in the reaction. Most of the activation reactions are induced by neutrons since the addition of one neutron produces normally an instability in the nucleus, but other particles like protons can also be used to produce radionuclides. Nuclear reactors and accelerators are some of the most used devices to host nuclear reactions to produce radionuclides for various applications. The MCNP5 code based on the Monte Carlo (MC) method can be used to estimate the activity generated in the reaction. A summary of the simulation of neutron reactions is presented in the paper. The comparison of simulation results with experimental measurements allows the validation of developed models. • Neutron activation of materials in the core of a nuclear reactor. • Application of the Monte Carlo method to simulate neutron activation. • Irradiation of a stainless steel sample in a nuclear reactor. • Estimation of production of radionuclides in an accelerator. [ABSTRACT FROM AUTHOR]

In environmental radioactivity measurements, High Purity Germanium (HPGe) detectors are commonly used due to their excellent resolution. Efficiency calibration of detectors is essential to determine activity of radionuclides. The Monte Carlo method has been proved to be a powerful tool to complement efficiency calculations. In aged detectors, efficiency is partially deteriorated due to the dead layer increasing and consequently, the active volume decreasing. The characterization of the radiation transport in the dead layer is essential for a realistic HPGe simulation. In this work, the MCNP5 code is used to calculate the detector efficiency. The F4MESH tally is used to determine the photon and electron fluence in the dead layer and the active volume. The energy deposited in the Ge has been analyzed using the ⁎F8 tally. The F8 tally is used to obtain spectra and to calculate the detector efficiency. When the photon fluence and the energy deposition in the crystal are known, some unfolding methods can be used to estimate the activity of a given source. In this way, the efficiency is obtained and serves to verify the value obtained by other methods. [ABSTRACT FROM AUTHOR]

High Purity Germanium (HPGe) detectors are widely used for environmental radioactivity measurements due to their excellent energy resolution. Monte Carlo (MC) codes are a useful tool to complement experimental measurements in calibration procedures at the laboratory. However, the efficiency curve of the detector can vary due to uncertainties associated with measurements. These uncertainties can be classified into some categories: geometrical parameters of the measurement (distance source–detector, volume of the source), properties of the radiation source (radionuclide activity, branching ratio), and detector characteristics (Ge dead layer, active volume, end cap thickness). The Monte Carlo simulation can be also affected by other kind of uncertainties mainly related to cross sections and to the calculation itself. Normally, all these uncertainties are not well known and it required a deep analysis to determine their effect on the detector efficiency. In this work, the Noether–Wilks formula is used to carry out the uncertainty analysis. A Probability Density Function (PDF) is assigned to each variable involved in the sampling process. The size of the sampling is determined from the characteristics of the tolerance intervals by applying the Noether–Wilks formula. Results of the analysis transform the efficiency curve into a region of possible values into the tolerance intervals. Results show a good agreement between experimental measurements and simulations for two different matrices (water and sand). [ABSTRACT FROM AUTHOR]

Abstract: Obtaining primary X-ray spectra is a complex task, mainly due to the high fluence of X-rays. In order to avoid the pile up effect in the detector, an indirect method based on the Compton scattering interactions in a low density rod material can be successfully applied in the diagnostic energy range. In this work, the Monte Carlo codes PENELOPE and MCNP5 have been used to simulate the indirect method based on a simplified Compton spectrometry technique. Both models include the X-ray focus, a poly(mehtylmethacrylate) (PMMA) rod and an HPGe detector. Because the probability that primary photons scattered in the PMMA will be emitted towards the detector is small, it is necessary to consider appropriate values for cut-offs and other simulation parameters. With these models, a response function can be determined, relating the response of the detector to the primary X-ray spectrum. This function can be normally expressed as a matrix, which can be calculated by simulating the response detector to several monochromatic X-ray beams. The main goal of this work is to test the capability of the Monte Carlo codes PENELOPE and MCNP5, together with unfolding methods to estimate the primary spectrum when the response matrix and the response of the detector for a given conditions are known. The reliability of unfolded X-ray spectra is studied by comparing them with theoretical spectra obtained from the IPEM 78 catalog and calculating the Root Mean Squared (RMS) and Quality parameters. [Copyright &y& Elsevier]