[en] In this work the neutron fluence of the Varian Clinac 2100C Medical Accelerator has been evaluated using CR39 track dosimeter. The assessment of fast neutron dose to a patient for typical treatment of 200 cGy with an 18 MV photons beam is performed at surface-source distance of 100 cm with a field size of 20x20 cm². Fast neutron leakage around of the accelerator head is evaluated

[en] This paper describes the radioactive waste management in Mexico, particularly the activities that the National Institute of Nuclear Research (NINR) is undertaking in this field. Classification and annual generation of radioactive waste, together with practices and facilities relating to the management of radioactive waste are addressed. The respective national legal framework and policy are outlined. (author)

[en] 'Full text:' The main sources of electricity generation in Mexico are fossil fuels, mainly gas. At the end of 2013, nearly 50% of total electricity generated in Mexico, was generated by gas and 12% using coal. The Mexican 2012-2026 National Strategy for Energy (ENE-2012) proposes a diversification of generating sources in the electricity sector. Also states an objective indicating that by 2026, at least 35% of the total electricity produced should be by means of non-fossil fuels. Currently, Mexico has one nuclear power plant (Laguna Verde) consisting of two BWR units, with a combined capacity of de 1,610 MW. This power represents 3.08% as total installed capacity in the country, and represents 4.6% of the country's generated electrical energy on 2013. This work analyzes ENE-2012, considering different scenarios for nuclear energy in order to comply with the participation of clean energy sources by 2026. From this analysis we can conclude that nuclear energy should have more participation in the Mexican electricity generation mix for the near future. (author)

[en] The article presents results of a study using the pulsed laser heating method of thermal resistances of liquid lead contacts in gaps simulating a heat-conducting liquid metal sublayer in fuel elements of new-generation fast reactors. The paper describes the designs of measuring cells, the methodology for obtaining and processing experimental data, provides information on the thermophysical properties of parts of the measuring cells, presents the results of estimating the measurement error, and studies the dependence of thermal resistances of liquid lead contacts in model gaps on the temperature and number of meltings and crystallizations of lead. When studying the wetting of steel and ceramic surfaces with lead and lead-sodium alloy, based on the results of measuring the contact angles using the sessile drop method, the measurements were performed during heating and subsequent cooling of the contacting pair. The results of the experiments showed that in the temperature range of 400-650 °C typical for the operation of the liquid metal heat-conducting sublayer, the thermal resistance of the contacts of the lead melt with the heat-transfer surfaces reaches 1·10-5 (m²·K)/W. Such thermal resistances must be taken into account when analyzing heat transfer in fuel elements with a liquid metal sublayer. The results of the wetting study showed the potential of using a lead-sodium alloy as a heat-conducting sublayer metal.

[en] In Mexico, it is necessary to establish in a few years, a definitive repository for low and intermediate level radioactive waste in order to satisfy the necessities for the next 50 to 80 years. Therefore, it is required to estimate the radioactive waste volumes generated annually, the updated stored volumes and their projection to the medium term. The results from the simulation of four cases considering the operation for 40 and 60 years of two nuclear power reactors, the use of conventional treatment technology and solid super compaction. If the decommissioning of a nuclear power plant is done 30 years after its lifetime, then it could be saved 50% of the total volume of radioactive waste, without considering that the reactor was operated during 40 or 60 years. Also, this operation extension for the two units represented an increase of 20% for the total volume. (author)

[en] The AAPM TG-43 modified protocol was used for the calculation of the neutron dose rate of ²⁵²Cf sources for two tissue substitute materials, five normal tissues and six tumours. The ²⁵²Cf AT source model was simulated using the Monte Carlo MCNPX code in spherical geometry for the following factors: a) neutron air kerma strength conversion factor, b) dose rate constant, c) radial dose function, d) geometry factor, e) anisotropy function and f) neutron dose rate. The calculated dose rate in water at 1 cm and 90 degrees from the source long axis, using the Watt fission spectrum, was D_n(r₀, θ₀)= 1.9160 cGy/h-μg. When this value is compared with Rivard et al. calculation using MCNP4B code, 1.8730 cGy/h-μg, a difference of 2.30% is obtained. The results for the reference neutron dose rate in other media show how small variations in the elemental composition between the tissues and malignant tumours, produce variations in the neutron dose rate up to 12.25%. (Author)

[en] Full text: The AAPM TG-43 dosimetry protocol was used for the calculation of the neutron absorbed dose rate from ²⁵²Cf sources in some media for medical applications; a) Tissue substitute materials: water, A-150 plastic and b) Normal tissue: muscle (skeletal), brain. Monte Carlo MNCPX code was employed to characterize the neutron dosimetry in the vicinity of three ²⁵²Cf sources, using the Maxwell fission neutron energy spectrum. The neutron air kerma rate is obtained with the +F6 heating tally to estimate the total energy deposition per mass unit of all particles in a spherical geometry of 200 cm in radius. Each set of evaluations required transport of 10⁸ particle histories to achieve relative standard uncertainties < 0.1%. Three models of ²⁵²Cf sources, AT, Vari Source and μSelectron were simulated using the Monte Carlo MCNPX code in spherical geometry for the following factors: a) conversion coefficient of neutron air kerma strength, b) neutron dose rate constant, c) neutron radial dose function, d) geometry factor, e) anisotropy function and f) neutron absorbed dose rate. The simulations consider that neutrons were emitted isotropically and uniformly from a cylindrical active element of Pd:Cf₂O₃ ceramic-metal matrix. The reference neutron dose rate in water at 1 cm and 90 degrees from the source long axis, using the Maxwell fission spectrum have been determined for the models AT, Vari Source and μSelectron: 1.8606±0.0003 c Gy h^-1 μg^-1, 2.012±0.0003 c Gy h^-1 μg^-1 and 2.147±0.0003 c Gy h^-1 μg^-1, respectively. (Author)

[en] In this study, sand and granular activated carbon (GAC) biofilters were comparatively assessed as post-treatment technologies of secondary effluents, including the fate of 18 organic micropollutants (OMPs). To determine the contribution of adsorption and biotransformation in OMP removal, four reactors were operated (two biofilters (with biological activity) and two filters (without biological activity)). In addition, the influence of empty bed contact time (EBCT), ranging from 0.012 to 3.2 d, and type of secondary effluent (anaerobic and aerobic) were evaluated. Organic matter, ammonium and nitrate were removed in both biofilters, being their adsorption higher on GAC than on sand. According to the behaviour exhibited, OMPs were classified in three different categories: I) biotransformation and high adsorption on GAC and sand (galaxolide, tonalide, celestolide and triclosan), II) biotransformation, high adsorption on GAC but low or null adsorption on sand (ibuprofen, naproxen, fluoxetine, erythromycin, roxythromycim, sulfamethoxazole, trimethoprim, bisphenol A, estrone, 17β-estradiol and 17α-ethinylestradiol), and, III) only adsorption on GAC (carbamazepine, diazepam and diclofenac). No influence of EBCT (in the range tested) and type of secondary effluent was observed in GAC reactors, whereas saturation and kinetic limitation of biotransformation were observed in sand reactors. Taking into account that most of the organic micropollutants studied (around 60%) fell into category II, biotransformation is crucial for the elimination of OMPs in sand biofilters. - Highlights: • OMP removal was comparatively assessed in sand and GAC biofilters. • The contribution of adsorption and biotransformation in OMP removal was identified. • The filtering material did not affect the biological activities in biofilters. • There is no direct correlation between EBCT and OMP removal in biofilters. • The type of secondary effluent determines the lifespan of filtering material.

[en] The action guides for the primary task forces of the Firemen Corps and the Public Security Elements, before the presence of radiological accidents are presented. The objective of these guides is to give the recommendations for the initial response to the emergency that fulfills the following criterions: to to Apply with readiness all the reasonable measures to protect the victims and the public, in order to minimizing the radiological and non radiological effects to the health; b) to Retrieve the control of the situation and to mitigate the consequences; c) to Protect the emergency personnel during the response operations; d) to Compile and to protect the information that can be useful to treat the effects to the health, of the victims and public in general, and to prevent that emergencies seemed are repeat in the future; e) to Create and to maintain the public's trust in the response; f) to Establish a base for an action of lingering response, and to Suggest evaluation criterions of the radiological emergency, the organization criterions, operation and surrounding of the emergency area. (Author)

[en] Neutron kerma coefficients were calculated in different media: 4 malignant tumours, 5 normal tissues and 3 tissue substitute in the range 11 eV-29 MeV. The objective was to identify which is the material that better reproduces the behavior of these tumours and tissues. These tissues have clinical interest in interstitial brachytherapy applications with fast neutron source (Cf-252). The small differences of elemental composition among these tissues produce variation in the neutron kerma coefficients. The results show that the neutron kerma coefficients for malignant tumours are smaller than soft tissue from 6% to 9%. Also, the muscle is the tissue that best represents the dosimetric behavior for the tumours and tissues analyzed in this paper for neutron energies >1 keV, where this coefficients show minor variation.