Search

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. • Compact Proton Therapy Centers (CPTC) are a new trend in the development of proton facilities. • Benchmarking of ambient dose equivalent, H*(10), and ambient dose, H*, in CPTC, was carried out through MCNP6.2 code. • CPTC benchmarked were those with compact synchrocyclotrons and with compact synchrotrons operating in Spain in 2020. • Different scenarios and materials in barriers were considered in the assessment of shielding of CPTC. [ABSTRACT FROM AUTHOR]

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 cheap and easy alternative to have new ambient dose equivalent measurements. • Combination of the responses of WENDI-II and Berthold LB6411. • Monitoring of neutron fields with energies between 10 MeV and 20 MeV. • An alternative solution to the calibration of monitors in D-T neutron generators. [ABSTRACT FROM AUTHOR]

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). • Current proton therapy trends involve the building of Compact Proton Therapy Centers (CPTC). • The effectiveness of the shielding of a CPTC was carried out with the code MCNP6.2. • Even with the most conservative assumptions H*(10) is below 500 uSv/year. • Several scenarios were simulated as radiation sources, physical models or types of concrete. • The main sources of uncertainty were concrete density and composition and nuclear models. [ABSTRACT FROM AUTHOR]

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. • MCNP benchmarking with experimental H*(10) measurements: γ-rays and neutrons. • H*(10) around thermal neutron irradiation facilities: total, γ-rays and neutrons. • Berthold LB6411 and LB1236 area monitors. • Radiological protection assurance around thermal neutron irradiation facilities. [ABSTRACT FROM AUTHOR]

Abstract 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. Highlights • The performance of a system for explosives detection was determined. • Detection is based on prompt gamma-rays induced by neutrons produced by a D-D neutron generator. • The system was tested with 1.5″ × 1.0″ and 3″ × 3″ NaI(Tl) scintillators. • Two explosives, RDX and Ammonium nitrate, were used. • The performance was evaluated with the MCNP6 code. [ABSTRACT FROM AUTHOR]

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 A m 95 241 / B e 4 9 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 C f 98 252 neutron source is used and were compared with the results obtained with the A m 95 241 / B e 4 9 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 C f 98 252 produces a larger amount of neutrons. • 241Am/9Be neutron source. • Neutrons sources 241Am/9Be and 252Cf are moderated and scattered in the system. • 252Cf neutron source. • MCNP6.2 [ABSTRACT FROM AUTHOR]

The response of a scintillation neutron detector of ZnS(Ag) with 10 B was calculated, using the MCNPX Monte Carlo Code. The detector consists of four panels of polymethyl methacrylate (PMMA) and five thin layers of ~0.017 cm thick 10 B+ZnS(Ag) in contact with the PMMA. The response was calculated for the bare detector and with different thicknesses of High Density Polyethylene, HDPE, moderator for 29 monoenergetic sources as well as 241 AmBe and 252 Cf neutrons sources. In these calculations the reaction rate 10 B(n, α) 7 Li and the neutron fluence in the sensitive area of the detector 10 B+ZnS(Ag) was estimated. Measurements were made at the Neutron Measurements Laboratory, Universidad Politécnica de Madrid, LMN-UPM, to quantify the detections in counts per second in response to a 252 Cf neutron source separated 200 cm. The MCNPX computations were compared with measurements to estimate the efficiency of ZnS(Ag) for detecting the α that is created in the 10 B(n, α) 7 Li reaction. After validating new models with different geometries it will be possible to improve the detector response trying to achieve a sensitivity of 2.5 cps-ng 252 Cf comparable with the response requirements for 3 He detectors installed in the Radiation Portal Monitors, RPMs. This type of detector can be considered an alternative to the 3 He detectors for detection of Special Nuclear Material, SNM. [ABSTRACT FROM AUTHOR]

The use of different nuclear data libraries and physics models can be a source of discrepancies in neutron transport simulation. Different Monte Carlo simulation toolkits can be used to characterize neutron monitors, these codes usually employ by default different nuclear data libraries and physics models. This work presents, for the first time, a comparison of MCNP and PHITS for the characterization of a LUPIN-II neutron rem-meter. The most significant discrepancies between the codes have been found around 100 MeV. • Comparison of Monte Carlo simulation toolkits: MCNP and PHITS. • Characterization by MCNP and PHITS of a LUPIN-II neutron area monitor. • Nuclear data libraries and physics models: sources of discrepancies. • Discrepancies in the simulation of neutron rem-meters in high-energy fields. [ABSTRACT FROM AUTHOR]

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. • Employment of spallation nuclear reactions in gold foils for neutron detection. • Combination of nuclear reactions, (n,γ) and (n,2n), in gold foil sensors to improve the response of monitors up to 20 MeV. • Comparison of different lateral irradiation models, point and plane, for a cylindrical detector. • Anisotropy of cylindrical rem-meters, response functions and elastic scattering reactions. [ABSTRACT FROM AUTHOR]

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. • Employment of extended-range neutron rem-meters in proton therapy facilities. • Analysis by MCNP6 of the response functions of WENDI-II and LUPIN-II. • Comparison of nuclear data libraries: ENDF/B – JEFF -TENDL. • Nuclear data libraries: Discrepancies at high-energies in Monte Carlo simulation. • Correlations in neutron monitors: fluence, energy deposition and response function. [ABSTRACT FROM AUTHOR]

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. • CPTC involve advanced technologies and smaller footprint and represents the next step in the evolution of proton therapy. • Response of WENDI-II and LUPIN-II was assessed for its application in verification of shielding against neutrons in CPTC. • The benchmark in monoenergetic neutron fields was carried out calculating the ambient dose equivalent H*(10) with MCNP6. • The performance in continuous neutron fields have been carried out with experimental measurements using an AmBe/241 source. [ABSTRACT FROM AUTHOR]

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 BF 3 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. • Design of an enhanced thermal neutron source for use in trace detection in substances or materials, through NAA method. • The thermal neutron irradiation system is using a241Am/9Be source. • Different configurations of polyethylene cylinders were simulated to choose the one giving the highest thermal neutron flux. • The performance was evaluated with the MCNP6 code and validated with laboratory measurements with a miniature BF 3 detector. [ABSTRACT FROM AUTHOR]

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. • A device has been designed to increase thermal neutron flux from a 241Am/9Be source, through the neutron moderation. • The system can be used to identify trace materials by Neutron Activation Analysis. • Three different alternative moderator materials, water, graphite and high-density polyethylene (HDPE), were simulated. • The system performance was evaluated with the MCNP6 code. [ABSTRACT FROM AUTHOR]

When linacs operate above 8 MV an undesirable neutron field is produced whose spectrum has three main components: the direct spectrum due to those neutrons leaking out from the linac head, the scattered spectrum due to neutrons produced in the head that collides with the nuclei in the head losing energy and the third spectrum due to room-return effect. The third category of spectrum has mainly epithermal and thermal neutrons being constant at any location in the treatment hall. These neutrons induce activation in the linac components, the concrete walls and in the patient body. Here the induced radioisotopes have been identified in concrete samples located in the hall and in one of the wedges. The identification has been carried out using a gamma-ray spectrometer. [ABSTRACT FROM AUTHOR]

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 A 95 241 m / B 4 9 e 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,β) 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. • A 95 241 m / B 4 9 e Thermal neutron source. • Neutrons from a A 95 241 m / B 4 9 e source are moderated and scattered in the system. • ENDF/B-VII.1, JEFF 3-3 and TENDL2017. • MCNP6.1 [ABSTRACT FROM AUTHOR]

FANT is the acronym of Enhanced Thermal Neutron Source (F uente A mpliada de N eutrones T érmicos , 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. • FANT is a thermal neutron source based on 241Am/9Be source inside HDPE cavity. • Neutrons from a241Am/9Be source are moderated and scattered in the system. • Designed by Monte Carlo to increase the thermal neutron flux. • Bonner Spheres Spectrometer (BSS) used to determine the neutron spectrum. • BUNKIUT (UTA4) and NSDann Ver 4.0 spectral unfolding codes used. [ABSTRACT FROM AUTHOR]

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. • MCNP6 analysis of Bonner spheres to characterize high-energy neutron fields. • New spheres design with spallation material layer to extend their energy range. • Absolute response functions of different extended-range Bonner spheres. • Analysis of the neutron spectra in the 6LiI(Eu) crystal scintillator. • Analysis of uncertainty associated to different physics models applied. [ABSTRACT FROM AUTHOR]