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.

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.

Benchmarking of stray neutron fields produced by synchrocyclotrons and synchrotrons in compact protontherapy centers (CPTC) using MCNP6 Monte Carlo code.

Proton therapy is an external radiotherapy using proton beams with energies between 70 and 230 MeV to treat some type of tumours with outstanding benefits, due to its energy transfer plot. There is a growing demand of facilities taking up small spaces and Compact Proton Therapy Centers (CPTC), with one or two treatment rooms, supposing the technical response of manufacturers to this request. A large amount of stray radiation is yielded in the interaction of proton beam used in therapy, neutrons mainly, hence, optimal design of shielding and verifications must be carried out in commissioning phases. Currently, almost 50 proton centers are under construction and start up in several countries, including ten in Spain. In the present work the effectiveness of shielding in two CPTC was verified with the Monte Carlo code MCNP6 by calculating the ambient dose equivalent, H*(10) due to secondary neutrons, outside the enclosures and walls of the center. The facilities modelled were the two centers currently operating in Spain, the first, since December 2019, with a superconductor synchrocyclotron, and the second, since March 2020, with a compact synchrotron. The geometry and materials are based on dimensions proposed a priori by the vendors, therefore, the paper is focused on check the suitability of the materials and thickness of the walls of the centers. Several models of the radiation sources were simulated, starting from a conservative assumptions, followed by more realistic scenarios. In all cases, the results reached for the ambient dose equivalent, H*(10), were below 1 mSv/year, which is the legal limit considered for the public in international references. Finally, considering that the recent ICRU Report 95 proposes changes in the operational quantities, the dose outside shieldingt has been evaluated in terms of the new next area surveillance quantity, H*, known as ambient dose, in the process of implementation.

Application of an algebraic methodology for the combination of Berthold LB6411 and WENDI-II for neutron area monitoring in D-T neutron generators and fusion facilities.

Neutron area monitors do not often have a good adjustment of their dose response functions to the ICRP74 neutron fluence-to-H*(10) conversion function between 10 and 20 MeV. The objective of this work is to establish a methodology to combine the dose response functions of Berthold LB6411 and WENDI-II, adjusting this combined function to the ICRP74 conversion function: this combination shows an almost perfect adjustment between 0.5 and 20 MeV. Thus, this article presents an easy and cheap alternative to the recalibration in D-T generators.

A novel conceptualization in the analysis and design of passive neutron area monitors based on gold foil activation.

In the detection and measurement of neutron fields, with energies between 10 and 20 MeV, current passive neutron area monitors based on gold foil sensors usually do not have a perfect fitting of their dose response functions with the neutron fluence-to-ambient dose equivalent conversion function from ICRP74. However, apart from the radiative capture in 197Au, the common channel considered in these monitors, other nuclear reactions in 197Au can be considered to improve the fit between both functions. Therefore, this work aimed to develop a mathematical combination of response functions in passive monitors with gold foils, considering the (n, γ) and (n, 2n) channels in 197Au, to extend their response up to 20 MeV, improving their performance under neutron fields with high energies. The proposed methodology avoids introducing modifications in the original device, such as the insertion of sheets with high-Z materials, and simplifies the design and manufacturing of passive monitors, while reducing costs.

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.

Determination of the uncertainties associated to the use of different nuclear data libraries in the analysis of extended-range rem-meters.

In proton therapy centers stray neutron radiation of up to 230-250 MeV is yielded by (p, Xn) nuclear reactions. To monitor ambient dose in such facilities, extended-range rem-meters are needed. The aim of this project was to characterize the response of two extended-range rem-meters, WENDI-II and LUPIN-II, by Monte Carlo methods, for energies ranging from 10-9 MeV up to 230 MeV. Different nuclear data libraries (ENDF/B-VII.1, TENDL2017, TENDL2019, JEFF-3.3) have been used, determining the uncertainties associated with the application of the libraries in the calculation of the response functions of both monitors. The differences found are very significant at energies around 150-200 MeV. This is an issue for predicting by Monte Carlo methods the response of these instruments at high energies. The results point to the necessity of testing experimentally the response of rem-meters at 150 MeV-200 MeV neutron beams.

Neutron dosimetry and shielding verification in commissioning of Compact Proton Therapy Centers (CPTC) using MCNP6.2 Monte Carlo code.

Proton therapy (PT) is an external radiotherapy using proton beams with energies between 70 and 230 MeV to treat some type of tumours with outstanding benefits, due to its energy transfer plot. There is a growing demand of facilities taking up small spaces and Compact Proton Therapy Centers (CPTC), with one or two treatment rooms, supposing the technical response of manufacturers to this request. A large amount of stray radiation is produced in the interaction of protons used in therapy, neutrons mainly, hence, optimal design of shielding and verifications must be carried out in commissioning stages. Currently, almost 50 CPTC are under construction and start up in many countries, including several in Spain. In the present work, the effectiveness of shielding in a CPTC was verified with the Monte Carlo code MCNP6 by calculating the ambient dose equivalent, H*(10) due to secondary neutrons, outside the enclosures and walls of the center. The facility modelled was similar to one planned to start operating in 2019 in Spain, a CPTC, made up of a superconducting synchrocyclotron and one treatment room, with a configuration standard, shielding and width of barriers based on dimensions proposed a priori by the vendor. Therefore, the paper is focused in check the suitability of the materials and thickness of the walls of the center and develop the assessment of enclosures. Several models of the radiation sources and type of concrete in walls were simulated, starting from a conservative assumptions, followed by more realistic models. In all cases, the results were below 1 mSv/year, which is the international legal limit considered for the general public. This work is part of the project Contributions to Shielding and Dosimetry of Neutrons in Compact Proton Therapy Centers (CPTC).

Experimental characterization of FANT, a new thermal neutron source.

FANT is the acronym of Enhanced Thermal Neutron Source (Fuente Ampliada de Neutrones Térmicos, in Spanish). This is a parallelepiped box of high-density polyethylene moderator and an isotopic neutron source. The moderator has a cylindrical irradiation chamber where a rather uniform thermal neutron flux is obtained. The FANT design was previously optimized and the neutron spectra were estimated by Monte Carlo calculations with the MCNP6.1 code. To check the characteristics of the FANT thermal neutron field, measurements have been performed at the reference point inside the irradiation chamber with a Bonner sphere spectrometer holding a small 6LiI(Eu) thermal neutron detector. To unfold the neutron spectrum BUNKIUT with UTA4 response matrix and NSDann Ver 4.0 codes were used. Some issues have been found and recommendations are made about the use of large BSS inside narrow spaces, and about the capacity of NSDann code to unfold these kind of spectra. However, the results confirm that the moderation process in FANT is very effective and allows obtaining useful thermal neutron fluence rates.

Analysis by Monte Carlo methods of the response of an extended-range Bonner Sphere Spectrometer.

High-energy neutrons up to 230 MeV are generated as a consequence of (p,n) nuclear reactions in proton therapy facilities. The aim of this work is to evaluate the potential extension of the UPM Bonner Sphere Spectrometer (BSS) through the use of spallation materials, for its future application to determine neutron fluence spectra in such facilities. Monte Carlo methods have been used to model the response of the actual and modified spheres with the introduction of a spallation material layer, lead or copper, analyzing their response functions. An analysis is also made of the neutron fluence spectra over the 6LiI(Eu) scintillator volume and the uncertainty that can be associated to the response functions at high energies as a consequence of the different physics models that can be applied for their analysis.

Monte Carlo characterization and benchmarking of extended range REM meters for its application in shielding and radiation area monitoring in Compact Proton Therapy Centers (CPTC).

Compact Proton Therapy Centers, CPTC, have a single treatment room, and are technologically more affordable, smaller, advanced and easier to use. From a radiological protection point of view, the leading concern in CPTC are interactions of protons with components of the facility and patients that yield a broad emission of secondary particles, mainly high-energy neutrons, up to 230 MeV, and photons. Optimal design of shielding involves theoretical assumptions in the design phase and, consequently, experimental measurements with extended range neutron detectors must be carried out in the facility during the commissioning period to verify the design, assumptions and building of the enclosures. There are almost 50 CPTC under construction and planning around the world, hence the improvement of methodologies to verify the shielding and to evaluate the dose to workers and general public in CPTC is a trending issue. The aim of this work was to evaluate and compare the response of two commercial extended range REM meters, WENDI-II and LUPIN-II, for their application in shielding verification and radiation area monitoring in CPTC facilities, by estimating the ambient dose equivalent, H*(10), through the Monte Carlo code MCNP6. The results have been compared with previous works. Likewise, the performance evaluation of these devices in continuous energy neutron field have been carried out, using the AmBe/241 neutron source of the Neutronics Hall (NH) of the Neutron Measurements Laboratory of the Energy Engineering Department of Universidad Politecnica de Madrid (LMN-UPM), through Monte Carlo simulation with the MCNP6 code and experimental measurements. The work is framed into the project Contributions to Shielding and Dosimetry of Neutrons in CPTC.

Design by Monte Carlo method of a thermal neutron device using a 241Am/9Be source and high-density polyethylene moderator.

A thermal neutron system intended to be used in neutron activation analysis has been designed by Monte Carlo methods. The device is based on a241Am/9Be neutron source of 111 GBq, placed inside a cylindrical cavity open inside a parallelepiped of moderator material. Three different moderator materials, water, graphite and high-density polyethylene (HDPE), were simulated to check what is the most suitable for the detection system, concluding that HDPE reach the better performance. The device achieves an increased thermal neutron flux by taking advantage of neutron moderation in the polyethylene and the neutron scattering in the irradiation chamber walls. The thermal fluence rates obtained were 904 cm-2  s-1, i.e. 8.144 cm-2 s-1 GBq-1, with a fraction of thermal neutrons at the best point of 83% of pristine fast neutrons emitted by the source. The device has been designed by Monte Carlo techniques using the MCNP6 code, and the main tasks developed were to select the moderator material and to maximize the thermal neutrons flux in the irradiation chamber.

Analysis by Monte Carlo of thermal neutron flux from a 241Am/9Be source for a system of trace analysis in materials.

Neutron techniques to characterize materials have a wide range of applications, one of the major developments being the identification of terrorist threats with chemical, biological, radiological, nuclear and explosives (CBRNE) materials. In this work, a thermal neutron irradiation system, using a241Am/9Be source of 111 GBq inside polyethylene cylindrical moderators, has been designed, built and tested. The geometry of moderator and the neutron source position were fixed trying to maximize the thermal neutrons flux emitted from the system. Therefore, the system is in fact a thermalized neutron source taking advantage of the backscattered neutrons, achieving thermal fluence rates of up to 5.3x102 cm-2 s-1, with dominantly thermal spectra. Samples can be placed there for several hours and thereafter be measured to identify their component elements by NAA (Neutron Activation Analysis). Through Monte Carlo techniques employing the MCNP6 code (Pelowitz et al., 2014), four different configurations with polyethylene cylinders were simulated to choose the most adequate geometry. The theoretical model was then replicated in the neutronics hall of the Neutron Measurements Laboratory of the Energy Engineering Department of Universidad Politécnica de Madrid (LMN-UPM), carrying out experimental measurements using a BF3 neutron detector. A high agreement between MCNP6 results and the experimental values measured was observed. Consequently, the system developed could be employed in future laboratory experiments, both for the identification of trace substances by NAA and for the calibration of neutron detection equipment.

Study by Monte Carlo methods of an explosives detection system made up with a D-D neutron generator and NaI(Tl) gamma detectors.

Detection of hidden explosives is of utmost importance for homeland security. Several configurations of an Explosives Detection System (EDS) to intercept hidden threats, made up with a Deuterium-Deuterium (D-D) compact neutron generator and NaI (Tl) scintillation detectors, have been evaluated using MCNP6 code. The system's response to various samples of explosives, such as RDX and Ammonium Nitrate, is analysed. The D-D generator is able to produce fast neutrons with 2.5 MeV energy in a maximum yield of 1010 n/s. It is surrounded by high-density polyethylene to thermalize the fast neutrons and to optimize interactions with the sample inspected, whose emission of gamma rays gives a characteristic spectrum of the elements that constitute it. This procedure allows to determine its chemical composition and to identify the type of substance. The necessary shielding is evaluated to estimate its thicknesses depending on the admissible dose of operation, using lead and polyethylene. The results show that its functionality is promising in the field of national security for explosives inspection.