Neutron pumping of active medium formed by gadolinium isotopes 155Gd and 156Gd pair: A feasibility study.

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.

Comparison of H*(10) estimations, due to neutrons and γ-rays, around FANT with two 241Am/9Be sources.

FANT (Fuente Ampliada de Neutrones Térmicos; in Spanish) is a thermal neutron irradiation facility with an extended and very uniform irradiation area, which has been developed by the Neutron Measurement Laboratory of the Energy Engineering Department at Universidad Politécnica de Madrid (LMN-UPM). In FANT, an isotopic neutron source (241Am/9Be) produces the primary neutrons. The design and facility optimization were carried out by extensive Monte Carlo calculations. In addition, Monte Carlo methods were used to evaluate the facility's performance to produce a constant and uniform thermal neutron field; these results were validated through experimental methods. FANT is designed to have two neutron sources; the objective of this work is to estimate the ambient dose equivalent due to neutrons and gamma-rays by Monte Carlo methods, and to compare these values with measured experimental doses. Thus, the performance of FANT with the two 241Am/9Be sources of LMN-UPM, with regard to the ambient dose equivalent H*(10) produced by both neutrons and photons around the facility, is analyzed in this work. The results are compared with those previously obtained in the framework of the results obtained with the LB6411 device around FANT.

Mechanical and radiation shielding characterization of W-based alloys for advanced nuclear unit.

The paper aims to investigate the mechanical and radiation shielding properties of two tungsten-based alloys manufactured by powder technology; their features when compared with two standard stainless steel grades for advanced nuclear applications. Multiple measurements were performed to characterize the alloys' structural and mechanical properties. XRD analysis and average surface roughness measurements showed the crystalline and morphological structure of the alloys. Surface microhardness added to the abrasive wear analysis showed the final wear resistance of the selected alloys. In addition, the shielding features against gamma and fast neutrons radiation were calculated. The superior characteristics of W-based alloys manufactured by powder technology make them suitable to be used in advanced nuclear power units.

Dosimetric evaluation of 123I (Iodide) and 99mTc (Pertechnetate) in the thyroid of neonates using Cristy-Eckerman and Segars anatomical representations.

Using the MIRD formalism, and the Cristy-Eckeman and Segars anthropomorphic representations, the absorbed dose in the thyroid of newborns, was calculated when 123I (iodide) and 99mTc (pertechnetate) are used during the diagnostic procedures. The dose results will allow exploring the dosimetric impact generated by the use of these radiopharmaceutical compounds and the use of two representations. Regardless the radiopharmaceutical compound and the anthropomorphic representation is the thyroid self-dose is the greatest, due to electrons emitted during the 123I and 99mTc radioisotopes. The relative difference in total dose to the newborn thyroid gland using the Cristy-Eckerman and Segars anthropomorphic representations for the compounds 123I(iodide) and 99mTc(pertechnetate) is 1.82%, and 1.33%, respectively. Regardless of the radiopharmaceutical compound, the replacement of Cristy-Eckerman by Segars phantom does not reflect significant changes in the estimated absorbed dose to the newborn thyroid. Regardless of the anthropomorphic representation, the lowest absorbed dose in newborn's thyroid is obtained when using 99mTc (pertechnetate) is used due to the residence times.

Estimation with Monte Carlo of thermal neutrons in the FANT, using 252Cf and 241Am/9Be neutron source.

The thermal neutron irradiation device (FANT), developed at the Neutron Measurements Laboratory of the Energy Engineering Department at Universidad Politécnica de Madrid, is a high-density polyethylene regular parallelepiped, with a rather uniform neutron fluence inside its irradiation chamber. It uses a Am95241/Be49 neutron source aiming to provide thermal neutron fluence rates. Neutron spectra and neutron fluences were estimated with Monte Carlo methods in the FANT irradiation chamber when a Cf98252 neutron source is used and were compared with the results obtained with the Am95241/Be49 source. Regardless of the neutron source, the largest contribution is due to thermal neutrons, producing also epithermal and fast neutrons. Per neutron emitted by the source, the use of Cf98252 produces a larger amount of neutrons.

Spectra, fluence and absorbed doses in sensitive organs due to scattered X-rays during a chest CT.

The use of ionizing radiation for the treatment and diagnosis of diseases is becoming more frequent. The technologies associated with diagnostic imaging are constantly evolving, allowing faster and cheaper diagnoses to benefit the patient. However, this has caused an increase in the exposure to ionizing radiation of patients and health professionals. One of the diagnostic techniques for obtaining high-resolution anatomical images of patients is computed tomography (CT). Due to the detail and quality of the images obtained with CT, its use is becoming more frequent. The information provided by these images allows the specialist to make better diagnoses; however, exposure to X-rays deposits a dose in the patient. CT represents approximately 20% of all X-ray examinations but it is responsible for 70% of the medical dose accumulated by the patient. During the acquisition of the images, the highest dose is deposited in the area of the body whose image is to be obtained. During the incidence of X-rays, there is dispersion of these that reach sensitive organs whose dose is not evaluated. The objective of this work was to estimate, using Monte Carlo methods, the fluence and X-ray spectra and to obtain a factor that allows knowing the absorbed dose in sensitive organs due to scattered radiation during a chest CT. With the MCNP5 code, the CT equipment and a hybrid anthropomorphic phantom, type BOMAB it was found that the absorbed dose in these organs depends on the size of the organ and the distance between the organ and the surface of the slice on the thorax where the X-rays are incident.

How much should you worry about contaminant neutrons in spatially fractionated grid radiation therapy?

Neutron contamination in radiation therapy is of concern in treatment with high-energy photons (> 10 MV). With the development of new radiotherapy modalities such as spatially fractionated grid radiation therapy (SFGRT) or briefly grid radiotherapy, more studies are required to evaluate the risks associated with neutron contamination. In 15 MV SFGRT, high-Z materials such as lead and cerrobend are used as the block on the tray of linear accelerator (linac) which can probably increase the photoneutron production. On the other hand, the high-dose fractions (10-20 Gy) used in SFGRT can induce high neutron contamination. The current study was devoted to addressing these concerns via compression of neutron fluence (Φn) and ambient dose equivalent ([Formula: see text]) at the patient table and inside the maze between SFGRT and conventional fractionated radiation therapy (CFRT). The main components of the 15 MV Siemens Primus equipped with different grids and located inside a typical radiotherapy bunker were simulated by the MCNPX® Monte Carlo code. Evidence showed that the material used for grid construction does not significantly increase neutron contamination inside the maze. However, at the end of the maze, neutron contamination in SFGRT is significantly higher than in CFRT. In this regard, a delay time of 15 minutes after SFGRT is recommended for all radiotherapy staff before entering the maze. It can be also concluded that [Formula: see text] at the patient table is at least 10 times more pronounced than inside the maze. Therefore, the patient is more at risk of neutrons compared to the staff. The [Formula: see text] at the isocenter in SFGRT with grids made of lead and cerrobend was nearly equal to CFRT. Nevertheless, it was dramatically lower than in CFRT by 30% if the brass grid is used. Accordingly, SFGRT with the brass grid is recommended, from radiation protection aspects.

Improving the properties of commercial steel alloys in the industrial/nuclear field by RF plasma carbonitriding.

The study investigates the effect of RF plasma on the commercial SS201 and Steel alloys physical, structural and nuclear properties, using an atmosphere of nitrogen (85%) plus acetylene (15%) at 500 W working for 0.5 h. The results were compared with untreated alloys and AISI304L. The study indicated the ability of RF plasma to improve both the mechanical and tribological properties of treated alloys compared to the original commercial alloys and the AISI304L, and achieve good shielding properties for industrial/nuclear applications.

Doses and spectra, of leaking-out and scattered photons from X-ray tubes used in dentistry.

Diagnosis by X-ray image are procedures widely used to improve the diagnosis or to follow the evolution of a medical procedure, also are used to support the cancer treatment with linear accelerators. The procedure involves the X-ray unit and the detection system, while the X-ray beam is projected onto the patient; along this procedure X-rays are scattered out from the patient body and X-rays leak-out from the X-ray unit. Both, the scattered and the leaking out radiation, are important during the evaluation or the design of the shielding barriers. X-ray tube features like voltage, electric current, target and filter are related to the dose due to the scattered and leakage radiation. Also, the dose due to the scattered radiation depends upon the scattering angle; while the dose due to the leaking out radiation, depend upon the X-ray tube housing. The objective of this work was to estimate, using Monte Carlo methods, the X-ray spectra and doses (air Kerma and Ambient dose equivalent) produced by scattered and leaking out radiation of 70, 80 and 90 kV X-ray units.

Effect of changing the neutron moderator in a thermal subcritical nuclear reactor.

Subcritical thermal nuclear reactor is the combination of nuclear fuel, moderator and external neutron source. Nevertheless, the amount of neutrons increases through nuclear fissions in the 235U in the fuel, the absorption and leakage of neutrons maintain the subcriticality condition. There are several subcritical reactors with different features and purposes. The open tank subcritical reactor Nuclear Chicago model 9000 is the heterogeneous combination of natural uranium, in hexagonal lattice, and light water, acting as moderator and reflector of neutrons, that uses a239PuBe neutron source. Worldwide there are several of these reactors mainly used in education, where due to the contact with water the Al tubes have corrosion. A possible solution is to use polyethylene as moderator. Aiming to evaluate the effect of changing the moderator, in this work Monte Carlo methods were used to model the reactor with polyethylene and water as moderators increasing the fuel load. For each model were estimated the keff and the µ parameters, and for Case 10a model was calculated the power, the Ambient dose equivalent and the fluence, of neutrons and γ-rays at five sites outside the reactor.

Characteristics, as a shield against ionizing photons, of concrete blocks used in the construction industry.

INEGI reports that 81% of new constructions are built with durable materials, such as concrete blocks, including rooms where ionizing radiation is used, such as clinics and hospitals. The volume ratios used in the manufacture of concrete blocks have a 1: 5: 2 ratio of: portland cement, sand and crushed gravel. The percentage mass content of each atom in the sample is obtained with the energy-dispersed X-ray fluorescence, these results are used to calculate the characteristics such as shielding of the concrete blocks, which have been partially reported, in this work several characteristics are presented for photons from 1 keV to 100 GeV, such as linear attenuation coefficients, the hemireductive layer, the effective, atomic and electronic sections, the effective atomic number (Zef). The exposure of accumulation and energy absorption factors (EBF, EABF) of 0.5-40 mfp are reported, using the Photon Shielding and Dosimetry (PSD) software. These characteristics are compared to those of NBS concrete. The concrete blocks have an attenuation capacity for energies used in radiological equipment such as mammography, 20-35 keV; dental, 50-90 keV and for conventional radiography equipment, 70-150 keV, so the block and lightened concrete block must be considered in the calculations of primary and secondary shielding walls.

Radiation shielding and dosimetric parameters of mexican artisanal bricks.

In this work, were determined some radiation shielding and dosimetric parameters of three types of bricks for photons energy from 1 keV to 100 GeV photons using the Phy-X/PSD software, and for comparison also has been calculated the same parameters for NBS concrete. The parameters calculated are the linear attenuation coefficients (LAC), effective atomic numbers (Zeff), half value layers (HVL), the energy absorption buildup factors (EABF) and the exposure buildup factors (EBF), as well as the coefficients to use the geometric progression (G-P) fitting method. Obtained results show that the three types of bricks can be used safely for the design of medical facilities housing mammography units (less than 30 keV).

Age-specific calibration for in vivo monitoring of thyroid: is it necessary?

Usually, an age-specific calibration of detectors used for in vivo monitoring of 131I thyroid radioactivity is not performed in practice. This study aimed to investigate the reduction in uncertainty that one can expect if an age-specific calibration is performed. For this, voxel and stylized computational phantoms of the thyroid, corresponding to children at different age groups, were used to simulate the calibration process of 131I detectors used for thyroid monitoring. SCK•CEN physical phantoms were also used for this purpose. Both analytical and Monte Carlo methods (MCNPX version 2.6.0) were used to estimate the counting efficiencies of the considered detectors. The results show that the uncertainties in the assessment of thyroid activity at a distance of 20 cm would be reduced from a range of +8% to +30%, to a range from - 6% to +15% when age-specific calibration was performed. Using a calibration based on thyroids of adults would result in an overestimation of the thyroid activity for children by up to 30% at a detector-neck distance of about 20 cm; a larger overestimation may be expected at closer distances. It is concluded that age-specific calibration of in vivo monitoring systems for the thyroid is important and has to be taken into consideration to improve the reliability of thyroid dose assessment for children.

Gamma-ray energy spectrum unfolding of plastic scintillators using artificial neural network.

In this study, the unfolding of the plastic scintillator spectrum was undertaken using the artificial neural networks tools of MATLAB. To this purpose, the response matrix of the plastic scintillator was generated for 145 energy groups and in 512 pulse-height channels using the MCNPX2.6 code. The results confirmed that the relative error in the gamma-ray energy unfolding with artificial neural networks is less than 3.8%.

Mathematical description of the effect of HIF inhibition on the radiobiological response of LNCaP cells.

According to the National Institute of Public Health, prostate cancer (PCa) is the leading cause of cancer death in Mexican men, highly associated with aggressiveness, resistance to treatment, and metastatic spread (Bharti et al., 2019) mediated by activation of the hypoxia-inducible factor 1 (HIF-1α). The objective of the present study was to evaluate the participation of HIF-1α activation in the radiobiological response of the human prostate adenocarcinoma cell line LNCaP, describing the phenomena with a mathematical model. Four groups were formed under different exposure conditions, including hypoxic cells treated with CoCl2 (300 µM for 22 h) with or without hypoxia-inducible factor inhibitor (150 nM chetomin for 4 h added after an incubation period of 18 h with CoCl2, just before completing the incubation period of 22 h). They were exposed to a source of 60Co in a dose range between 2 and 10 Gy to obtain survival curves that are fitted to a mathematical model. CoCl2 or chetomin treatments do not affect the viability of LNCaP cells that remained unchanged after irradiation. CoCl2 induced hypoxia reduces the survivability of LNCaP, and obstruction of HIF-1α signaling with chetomine produces a slight radioprotective effect. As others report, the genetic reprogramming induced by HIF-1α activation acts as an intrinsic agent that selects cells with more aggressive behavior (Pressley et al., 2017), while chetomin protects cells from death due to its scavenger properties. Interestingly, treatment with chetomin of cells induced to hypoxia (HIF-1 activation with CoCl2) produces a significant reduction in the radioresistance of LNCaP cells, demonstrating that the simultaneous use of chetomin and gamma radiation is an effective option for the treatment of hypoxic prostate cancer. At the molecular level, we suggest that the selective force exerted by HIF-1α depends on the production of free radicals by radiation. The proposed mathematical model showed that the rate of change in cell survival as a function of radiation dose is proportional to the product of two functions, one that describes cell death and the other that describes natural or artificial resistance to radiation.

Beam port filters in a TRIGA MARK III nuclear reactor to produce epithermal neutrons for BNCT.

Glioblastoma multiforme is the most common and aggressive brain tumor and it is difficult to treat with conventional surgery, chemotherapy, or radiation therapy. An alternative treatment is boron neutron capture therapy which requires an energy modulated beam of neutrons and a10B drug capable of adhering to the tumor. In this work, MCNP6 Monte Carlo code was used to evaluate the effect on the neutron spectrum by placing two filters along the radial beam tube of the TRIGA Mark III nuclear reactor of ININ in Mexico. Every filter was made with the same amount and type of materials: Steel and Graphite for filter 1 and Cadmium, Aluminum, and Cadmium (Cd + Al + Cd) for filter 2. Two cases were analyzed for each filter as follows: Case A for filter 1 was considering 30 cm of steel and 30 cm of graphite, while for case B, the dimensions of filter 1 were 15 cm of steel, 15 cm of graphite, 15 cm of steel and 15 cm of graphite. Cases A and B for filter 2 were analyzed considering the same dimensions and amount of materials. The work was in the aim to produce epithermal neutrons for boron neutron capture therapy. Neutron spectra were calculated at three sites along the beam tube and two sites outside the beam tube; here, the ambient dose equivalent, the personal dose equivalent, and the effective doses were also estimated. At a distance of 517 cm of core, in case B, results in an epithermal-to-thermal neutron fluence ratio of 30.39 was obtained being larger than the one recommended by the IAEA of 20.

Comparison of FANT results using the ENDF/B-VII.1, JEFF-3.3 and TENDL2017 nuclear data libraries.

FANT (Fuente Ampliada de Neutrones Térmicos; in Spanish) is a thermal neutron irradiation facility with an extended and very uniform irradiation area, that has been developed by the Neutron Measurements Laboratory of the Energy Engineering Department at Universidad Politecnica de Madrid (LMN-UPM). This device is a parallelepiped box made of high-density polyethylene (HDPE), moderator material, that uses an A95241m/B49e neutron source of 111 GBq nominal activity for irradiating materials. The facility design was previously optimized, and the neutron spectra were estimated by extensive calculations with the MCNP6.1 code and carrying out experimental measurements (Bedogni et al., 2017). The facility takes advantage of the scattering reactions of neutrons with the HDPE surfaces of the chamber, where the moderation process is effective, achieving relevant thermal neutron fluence rates. The main goal of this work has been to simulate and analyse the FANT system by Monte Carlo methods using the MCNP6.1 code, employing 3 different nuclear data libraries: ENDF/B-VII.1, JEFF-3.3 and TENDL 2017. The transport of thermal neutrons in HDPE, E < 1eV, has been calculated in all the cases taking into account the thermal S (α,ß) treatment. The results achieved in this work have been compared with those previously obtained in the former development of FANT, using the MCNP6.1 code with the ENDF/B-VII.1 nuclear data, and experimental measurements. These results have shown that the JEFF-3.3 nuclear data library is the nuclear data library that provides of the best matching between the MCNP computational results, and the experimental data collected at FANT. Hence, the JEFF-3.3 nuclear data library seems to be the most correct library to design and benchmark thermal neutron activation devices.

Feasibility of 18-MV grid therapy from radiation protection aspects: unwanted dose and fatal cancer risk caused by photoneutrons and scattered photons.

PURPOSE: Photoneutron production is a common concern when using 18-MV photon beams in radiation therapy. In Spatially Fractionated Grid Radiation Therapy (SFGRT), the grid block in the collimation system modifies the neutron production, photon scattering, and electron contamination in and out of the radiation field. Such an effect was studied with grids made of different high-Z materials by Monte Carlo simulations. The results were also used to evaluate the lifetime risk of fatal cancers. METHODS: MCNPX® code (2.7.0 extensions) was employed to simulate an 18-MV LINAC (Varian 2100 C/D). Three types of grid made of brass, cerrobend, and lead were used to study the neutron and electron fluence. Output factors for each grid with different field sizes were calculated. A revised female MIRD phantom with an 8-cm spherical tumor inside the liver was used to estimate the dose to the tumor and the critical organs. A 20-Gy SFGRT plan with Anterior Posterior (AP) - Posterior Anterior (PA) grid beams was compared with a Conventional Fractionated Radiation Therapy (CFRT) plan which delivered 40-Gy to the tumor by AP-PA open beams. Neutron equivalent dose, photon equivalent dose, as well as lifetime risks of fatal cancer were calculated in the organs at risk. RESULTS: The grid blocks reduced the fluence of contaminant electrons inside the treatment field by more than 50%. The neutron fluences per electron-history in SFGRT plans with brass, cerrobend and lead were on average 55%, 31% and 31% less than that of the CFRT plan, respectively. However, when converting to fluences per delivered dose (Gy), the cerrobend and lead grid may incur higher neutron dose for 20 × 20 cm2 field size and above. The changes in neutron mean energy, as well as the correlated radiation weighting factors, were insignificant. The total risk due to the photoneutrons in the SFGRT plans was 87% or lower than that in the CFRT plans. In both SFGRT and CFRT plans, the contribution of the primary and scattered photons to the fatal cancer risk was 2 times or more than the photoneutrons. The total risks from photons in SFGRT with brass, cerrobend, and lead blocks were 1.733, 1.374, and 1.260%, respectively, which were less than 30% of the total photon-risk in CFRT (5.827%). CONCLUSION: In the brass, cerrobend, and lead grids, the attenuation of photoneutrons outweighs its photoneutron production in 18-MV SFGRT. The total cancer risks from photons and photoneutrons in the SFGRT plans were 30% or less of the risks in the CFRT plans (5.911%). Using 18 MV photon beams with brass, cerrobend, and lead grid blocks is still a feasible option for SFGRT.

Monte Carlo evaluation of out-of-field dose in 18 MV pelvic radiotherapy using a simplified female MIRD phantom.

This study was devoted to determining the unwanted dose due to scattered photons to the out-of-field organs and subsequently estimate the risk of secondary cancers in the patients undergoing pelvic radiotherapy. A typical 18 MV Medical Linear Accelerator (Varian Clinac 2100 C/D) was modeled using MCNPX®code to simulate pelvic radiotherapy with four treatment fields: anterior-posterior, posterior-anterior, right lateral, left lateral. Dose evaluation was performed inside Medical Internal Radiation Dose (MIRD) revised female phantom. The average photon equivalent dose in out-of-field organs is 8.53 mSv Gy-1, ranging from 0.17 to 72.11 mSv Gy-1, respectively, for the organs far from the Planning Treatment Volume (Brain) and those close to the treatment field (Colon). Evidence showed that colon with 4.3049% and thyroid with 0.0020% have the highest and lowest risk of secondary cancer, respectively. Accordingly, this study introduced the colon as an organ with a high risk of secondary cancer which should be paid more attention in the follow-up of patients undergoing pelvic radiotherapy. The authors believe that this simple Monte Carlo (MC) model can be also used in other radiotherapy plans and mathematical phantoms with different ages (from childhood to adults) to estimate the out-of-field dose. The extractable information by this simple MC model can be also employed for providing libraries for user-friendly applications (e.g. '.apk') which in turn increase the public knowledge about fatal cancer risk after radiotherapy and subsequently decrease the concerns in this regard among the public.