Two passive neutron spectrometers were designed using Monte Carlo methods. Each spectrometer utilizes six 20.5 cm-diameter cylindrical moderators of varying thicknesses (0.5, 4.5, 8.5, 12.5, 16.5, and 20.5 cm). One spectrometer employs polyethylene moderators, while the other uses polymethylmethacrylate. Thermal neutrons are measured using pairs of thermoluminescent chips with different 6Li content (TLD600 and TLD700). Responses to gamma-rays of these TLDs and are approximately the same because have the same elemental composition, the same effective atomic number and similar densities. For both spectrometers, fluence response functions and absorbed doses from neutrons and secondary gamma-rays were estimated for 45 monoenergetic neutron energies ranging from 10-9 to 20 MeV. As monoenergetic neutrons are moderated to thermal energies, they reach the TLDs with an energy distribution, inducing 6Li (n, alpha) reactions that define the fluence response function. Energy deposition from these neutrons and gamma-rays within the TLDs allows for absorbed dose calculation. The fluence response matrices are similar for both moderator materials; however, neutron absorbed dose is consistently higher in TLD600 than in TLD700, independent of the moderator. Conversely, gamma-ray absorbed dose is similar across TLD types and moderators. These spectrometers offer a smaller size and lower weight compared to a passive Bonner sphere spectrometer, while also being easier to construct and more portable.

Living organisms are continuously exposed to both natural and artificial sources of radiation. Among artificial sources, those associated with radiotherapy and radiodiagnostic procedures represent the most significant contributors to ionizing radiation exposure. Given the potential biological risks associated with such exposure, radiological equipment must be installed in specially designed facilities constructed with appropriate shielding materials to ensure that radiation levels remain within regulatory limits and to protect occupational workers and the general public. The objective was to estimate, through Monte Carlo simulations, the linear attenuation coefficient, half-value layer, ambient dose equivalent, air kerma, and transmission factors of a concrete block. The MCNP5 code was employed to model an X-ray tube by varying the metallic target material and the tube voltage within the range of 30-120 kV. The generated X-ray spectra were subsequently used as the source term in a narrow beam transmission geometry simulation. The results indicate that an increase in X-ray energy leads to a reduction in the attenuation capacity of the concrete block, but it remains a viable option for shielding. These findings contribute to the limited of knowledge regarding the radiological characterization of construction materials in Mexico and provide relevant data for shielding design in diagnostic radiology facilities.

La radiación está presente en varias formas en nuestro entorno, se agrupa en fuentes naturales y artificiales. En las artificiales se utiliza en diferentes campos como industrial, investigación, medico, entre otros. En este último se utiliza para radioterapia y radiodiagnóstico, los equipos de radiación ionizante empleados con este fin, es necesario que estén contenidos en instalaciones con un blindaje adecuado del tal modo que proteja y limite la exposición a la radiación a un nivel aceptable. Existen varios materiales que se utilizan para blindaje como el plomo, el acero y el hormigón; pero dependiendo de la región suelen utilizarse materiales locales de construcción, pero no todos estos materiales están caracterizados. El objetivo de este trabajo fue estimar mediante métodos Monte Carlo (código MCNP5), el coeficiente de atenuación lineal y el valor de la capa hemirreductora del block de concreto. Se modelo un tubo de rayos X con un blanco metálico de tungsteno y se utilizó un rango de energía desde 30 a 120 keV. Posteriormente se modelo un experimento de transmisión de geometría angosta, donde se utilizó como termino fuente los espectros de rayos X obtenidos. Los resultados mostraron que ambos parámetros solo dependen de la energía de los rayos X. Para este rango de energías es seguro utilizar el block de concreto como barrera de blindaje de salas de diagnóstico médico.

Seven shape memory alloys, together with AISI 321 stainless steel and polyethylene, were investigated as moderators for a241Am-Be neutron source. Monte Carlo simulations were performed to estimate the neutron spectra at 100 cm from the source, with the aim of producing realistic neutron radiation fields comprising thermal, epithermal, and fast neutrons suitable for calibration of neutron radiation protection instruments. In addition, gamma-ray spectra induced by neutron interactions with, the source, alloy, steel, and polyethylene nuclei were evaluated. Using the calculated neutron spectra, the ambient dose equivalent, isotropic effective dose, and ambient dose were determined, while the ambient dose equivalent and ambient dose were also calculated for the gamma-ray spectra. The neutron spectra of alloy-moderated 241Am-Be sources, including configurations with thin and thick polyethylene shells, exhibited neutron absorption features associated with 55Mn present in all alloys. Neutron spectra produced by thin and thick alloy shell-polyethylene moderated 241Am-Be sources showed welldefined thermal, epithermal, and fast neutron components, with mean neutron energies ranging from 1.53 to 2.48 MeV. Among the investigated configurations, the thick ASH4P/241Am-Be source produced the highest neutron fluence, with a mean energy of approximately 2.16 MeV and a balanced contribution of thermal, epithermal, and fast neutrons. These results demonstrate that shape memory alloys, in combination with polyethylene, can be effectively used to tailor 241Am-Be neutron fields, enabling the generation of realistic spectral and dosimetric characteristics suitable for extending the calibration capabilities of neutron calibration facilities.

La tomografía computarizada (TC) es una técnica de diagnóstico médico que en las últimas décadas se ha convertido en una técnica de referencia. Debido a esto es responsable de la mayor dosis (70%) acumulada por la población debida a fuentes artificiales de radiación. La dosis que se administra al paciente es 50-500 veces mayor que la dosis en un examen radiológico convencional. Actualmente, existen una variedad de exámenes de TC, donde alguno o varios órganos radiosensibles pueden quedar expuestos a la radiación de manera directa o indirecta, uno de los exámenes que se ha incrementado es la TC de cabeza y cerebro. Para conocer la dosis que reciben estos órganos se puede medir de manera experimental o estimar mediante simulaciones computacionales. El objetivo de este trabajo fue estimar mediante métodos Monte Carlo la dosis absorbida por órganos radiosensibles debida a la radiación que se dispersa durante una TC de cabeza. Se modelo un fantoma antropomórfico hibrido tipo BOMAB, así como, el equipo de tomografía computarizada y se simulo el examen. Se encontró que la mayor dosis absorbida la recibe la tiroides por su cercanía a la zona de exploración, mientras que la menor dosis corresponde a los órganos reproductores masculinos por su lejanía. Las dosis que reciben son pequeñas pero son acumulativas y pueden representar un riesgo para el paciente debido a que hay quienes se realizan más de un examen en su vida.

Clay-based bricks are used worldwide in the construction industry due to their mechanical properties and durability. To design or evaluate the shielding of a room having a X-ray radiation source, is necessary to determine the walls thickness (barriers) necessary to reduce the dose due to primary, scattered and leakage radiation. For these calculations is required to know the dose transmission of room walls materials. In this work, Monte Carlo methods were used to estimate the Ka, H*(10) and H* transmission values of the red brick for X-rays produced when 70, 90 and 120 keV electrons collide with a tungsten target and pass through a 0.28 cm-thick aluminum filter. Dose transmission values in function of shielding thickness were fitted to a semiempirical function. Curves of dose transmission, and the parameters obtained by the fitting process, can be used to evaluate or to design a shielding based on clay bricks.

The Bonner sphere spectrometer is widely used to measure the neutron spectrum; however, it is bulky and heavy. Variants of Bonner sphere spectrometer aiming to reduce weight and volume have been reported where moderating spheres have been changed by multi or nested cylinders and cubes. In these variants passive and active thermal neutron detectors have been included to habilitate the spectrometer to different experimental situations. In this work Monte Carlo methods were used to design a novel passive neutron spectrometer with six 20.5 x 20.5 cm2 surface area regular parallelepipeds made of polyethylene moderators with 0.5, 4.5, 8.5, 12.5, 16.5 and 20.5 cm-thickness. Thermal neutrons are detected using the technique of pairs of thermoluminescent dosimeters (TLD600 and TLD700). For each moderator the response function was estimated; in each TLD, the response is the amount of 6Li(n, alpha) nuclear reactions occurring in the TLD600 and TLD700 for 1E(-9) to 20 MeV neutrons. In addition, the absorbed dose, due to neutrons and gamma-rays, in the TLDs were estimated. The total responses for the TLD600 and TLD700 inserted in the Regular Parallelepiped Spectrometer were calculated and compared with the total response of Bonner sphere spectrometer with TLD600. The fabrication of this novel spectrometer is simple, it is portable and compact whose weight and volume is 3.39 times smaller than weight and volume of the passive Bonner sphere spectrometer.

This computational study using MCNP5 evaluated the feasibility of replacing 6061-T6 aluminum with 316L stainless steel (SS-316L) for the tubes hosting the uranium slugs in the subcritical nuclear reactor Nuclear Chicago model 9000, thereby contributing to its preservation as a key resource for nuclear research and education in Mexico. Simulations and dosimetric analyses (ICRP/ICRU) confirmed subcriticality in both configurations. Notably, SS-316L demonstrated an effective attenuation of peripheral gamma radiation and a reduction in the ambient neutron dose, indicating a considerable improvement in radiological safety. Although a reduction in thermal and epithermal neutron fluence was observed, the similarity in the gamma spectrum suggests no significant alteration for gamma spectroscopic experiments. In conclusion, SS-316L presents a promising alternative that enhances radiological safety and reactor longevity, making it a worthy consideration as a replacement material. Further experimental investigation is recommended to assess material activation and the gamma dose in the vicinity of the fuel.

The present study focuses on the development of datasets characterizing the neutron field generated by the collimation channel of the Prizm-AN irradiation stand, which utilizes an internal bounded neutron (IBN) radioisotope capsule as a fast neutron source. It is aimed at supporting the prototyping of the proposed Thermal Neutron Imaging System (TENIS) diagnostics framework, currently in the feasibility assessment phase. To this end, comprehensive calculations of the radiation fields and dose characteristics associated with the Prizm-AN system-tailored to the operational specifics of the TENIS equipment-are presented. These insights will be valuable for developers in the preparatory stages leading to the creation of a demonstration version of the system. Additionally, the study explores the potential application of detection technologies for the real-time monitoring and control of parameters related to leaking deuterium-tritium (d-t) plasma.

In this study, four r-PET-based microcomposites (P1C1, P1C2, P1C3, and P1C4) were synthesized via solution casting, incorporating barite (BaSO4) and different metal oxide microparticles (WO3, Bi2O3, Fe2O3, and TiO2) to enhance radiation attenuation. The linear attenuation coefficient calculated by XCOM and Phy-X/PSD, the mean value layer calculation, as well as a comparative analysis with other materials reported in the literature and used in construction for ionizing energy shielding, show that the alternative of using r-PET as a matrix offers many advantages in the manufacture of X-ray and gamma-ray shielding materials. The reported materials were evaluated for an energy range between 0.015 and 10 MeV, covering the energy ranges used in medical and industrial applications. These results demonstrate that r-PET-based composites combine high shielding efficiency with reduced mass and a sustainable composition, representing a promising, environmentally friendly, and ergonomically favorable alternative to conventional lead-based shielding.

The selection of fuel holder tube material in subcritical reactors is critical for operational safety and long-term efficiency. In this study, a Monte Carlo analysis using the MCNP5 code was conducted to evaluate the impact of replacing Al-6061-T6 with its nitrided variant (Al-6061-T6-N), treated via Plasma Immersion Ion Implantation (PIII), as a structural material for the fuel holder tubes in a low-power subcritical reactor. The results showed that the surface modification does not induce significant changes in the effective multiplication factor or in the neutron and gamma-ray spectra. Likewise, the Ambient dose equivalents measured at various positions around the reactor remained virtually unchanged. These findings suggest that the PIII technique preserves the essential nuclear interaction properties of the material without compromising radiological safety. However, complementary experimental studies are recommended to validate its long-term performance under real operating conditions and radiation-induced aging.

To ensure long-term and trouble-free operation, the reactor fuel is modified by introducing various homogeneous and heterogeneous additives. This modified fuels exhibit satisfactory performances under irradiation at elevated temperatures and burnup levels. However, the challenges related to radiation safety during the handling of fresh and spent nuclear fuel remain unresolved. In this study, neutron emission spectra and effective doses for VVER1200 reactor fuel containing a heterogenous distribution of natGd2O3 and Am2O3, microspheres were calculated. Unlike in the case of homogenous variants, this design did not reduce the thermal conductivity of the fuel and positively affected the core neutronics and thermophysics. Comparing the radiation characteristics of fresh FAs revealed a considerable increase in the neutron (more than 3.7 center dot 102 times at a distance of 10 cm from the FA) and photonic (about 104 times) components of Am-containing fuel compared to U and U-Gd fuels. The yield and effective dose of neutrons emitted from spent Am-containing fuel is two times higher than that of uranium-based fuels. When calculating the dose, it is important to consider the energy spectrum of (alpha, n) neutrons in Am2O2 microcapsules. The findings of this study provide insights into the development of procedures and regulations for manufacturing and post-irradiation handling of new fuel in reactors.

Metallic alloys of different compositions are basic structures for building different types of nuclear reactors. This study evaluates the nuclear properties for three medium entropy alloys against incident neutrons and gamma radiation. The alloys had different chemical compositions prepared by powder technology and were compared with two stainless steel alloys for use in constructing different parts of nuclear power plant units. The shielding parameters were calculated: linear attenuation coefficient, half-value layer, tenth-value layer, mean free path, effective atomic number (Z(eff)), effective electronic number, and neutron removal cross-section. The Z(eff) of all investigated alloys had a range of 25.46-25.93.Sample 1 medium entropy alloy had the lowest neutron absorption feature and the greatest density (7.890 +/- 0.323 g/cm(3)) and Sample 3 medium entropy alloy had the largest neutron absorption feature. The study indicates that medium entropy alloys have potential for enhancing efficiency and safety of nuclear reactors.

Gamma rays are part of the electromagnetic spectrum and are known for their high concentration of energy; that is, they are classified as a type of ionizing radiation with the highest energy and are generated due to the decay of an excited atomic nucleus of unstable atoms, radioactive elements, particle acceleration, particle-antiparticle annihilation, among others. In this chapter, the understanding of gamma rays from their generation, interaction processes, as well as a comparison of different materials used to attenuate this kind of ionizing radiation will be deepened; we will also focus on how to characterize new materials, where the most important equations for it and the calculations for distributed radiation sources are included. The chapter will end with the expectations on materials, which, based on the theory, possibly exist for attenuating gamma rays in the short and long term.

Uranium fission fragments, as well as the products of He-3(n,p)H-3 and B-10(n,alpha)Li-7 nuclear reactions were utilized in the nuclear reactor for gas ionization and excitation. However, the Li-6(n,alpha)H-3 nuclear reaction was less examined. The use of lithium-6 as a surface source of excitation of the gas medium, due to the long path length of tritium nuclei in the gas, allows to excite large volumes of gas as opposed to using U-235 or B-10. While investigating the luminescence of noble gases in the core of the IVG.1M research reactor, we noted an appearance of alkali metal lines and a sharp increase in the intensity of these lines at temperatures above 570 K. It was determined that the population of levels of lithium atoms has practically no effect on the population of the 2p-levels of atoms of noble gases. The selectivity of p- and s-levels deactivation by lithium atoms implies the possibility of creating inversion of population at 2p-1s transitions of noble gas atoms. Successful experiments to study the luminescence of gases upon excitation by Li-6(n,alpha)H-3 nuclear reaction products allow us to proceed to experiments to achieve the laser action threshold and study the lasing characteristics of gas mixtures at the IGR pulsed nuclear reactor with thermal neutron flux density up to 7 center dot 10(16) n/cm(2)s. For this purpose, an experimental device designs were proposed to perform experiments on the IGR reactor. A step-by-step procedure of fabrication of a nuclear-excited source for excitation of gas mixtures is provided. The results of reactor experiments aimed at determining the spectral and temporal characteristics of optical radiation during excitation of gas mixtures by Li-6(n,alpha)H-3 nuclear reaction products are presented.

Air Kerma was measured and compared with calculated air Kerma aiming to validate the use of air Kerma gamma factors. The air Kerma was measured in radiology departments in three hospitals at the Lambayeque region in Peru. In each department were obtained three times different radiographs, for each radiograph the X-ray operational parameters were recorded such us: voltage, current, and triggering time; the average values of these parameters, and the measured air Kerma were calculated. In addition was recorded the focal spot-to r distance, and the filter thickness. Monte Carlo methods were used to calculate the spectra, air Kerma, and Ambient dose equivalent for an X-ray tube (W target, Al2O3 filter) working to 40, 50, and 60 kV; these values and those previously reported were used to calculate the spectra mean energy, the air Kerma and the Ambient dose equivalent gamma factors in function of voltage. Mean values of measured voltage were used to calculate the gamma factors for air kerma and this with the current, and triggering time mean values, in addition to the focal spot-to-r distance, and the filter thickness were used to calculate the air Kerma. The relative difference between calculated air Kerma and the measured air Kerma varies from -6.84 to 10.82% at the Belen hospital, -2.22 to 9.64% at Las Mercedes hospital and from 0.41% to 6.30% at the Ferreñafe hospital; thus, calculated air Kerma values using the gamma factors were validated with measured air Kerma values. Air Kerma gamma factors can be used for dosimetric purposes using the operational features of the X-ray tube with W target and Al2O3 filter; in addition can be used during the design of primary barriers of radiology hall.

The possibility of creating technical means for controlling the processes of accumulation and conversion of the energies of thermal and epithermal neutrons into the energy of monoenergetic photons due to neutron pumping of an active medium consisting of nuclei with long-lived isomeric states was studied in this work. The system under study consisted of an external pulse-periodic source of deuterium -tritium neutrons (PSN) and a subcritical blanket, which included a variable neutron-collimation beam-shaping assembly (vBSA) and an active medium. The vBSA was composed of moderating blocks and selective plates designed to trap and shape a pulsed neutron flux with subsequent conversion of a millisecond signature into monoenergetic photon emission. Gadolinium oxide enriched in 155Gd isotope was used as the active medium, where the heavier one could be at different excited states, the de-excitations of which were accompanied by photon emission. In this research, the possibility of using the conjugate system (i.e., blanket - PSN - vBSA) for converting excess neutron energy accumulated in the inverse state of 156Gd nuclei into photon emission was demonstrated in detail.

Neutron moderation characteristics and manufacturing viability of ABS (Acrylonitrile Butadiene Styrene) spheres were studied using a 6LiI(Eu) scintillator detector and a commercial 3D printer with FDM-type extruder. Response functions (RFs) for the detector with ABS spheres of varying diameters were calculated with the Monte Carlo (MC) code MCNPX and compared with RFs obtained for high-density polyethylene (HDPE) spheres. Designed ABS spheres were produced and irradiated with 241Am–Be neutron beams to validate MC simulations.

Neutron spectrum unfolding is a crucial process in radiation protection and dosimetry. Unfolding codes using iterative algorithms require a criterion to stop the iterations. One approach often relies on the Root Mean Square Error (RMSE) criterion to assess the convergence of iterative algorithms. The aim of this work is to present a new criteria: Average Ratio Scaled (AVGS) and Relative Change in AVGS (dAVGS) to address specific challenges associated with RMSE. Extensive validation tests were conducted, covering a range of scenarios with results showing high level of agreement between the unfolded spectra and the reference.

Ionizing radiations are tools in diagnosis and treatment of diseases. Leukopenia from exposure to ionizing radiation has been reported. Due to their radiosensitivity, leukocytes are a biological model to analyze cell damage. Therefore, cell viability, DNA damage, and Hsp70 and p53 expression in human leukocytes exposed to low-dose gamma radiation fields from a 137Cs source were evaluated. A decrease in cell viability, DNA damage and an increase in the expression of Hsp70 and p53 proportional to the radiation dose received was found, which was 0.2, 0.4, 0.6, 0.8 and 1.0 mGy.