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.

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.

Estimating the neutron component of radiation properties of the IVG.1M research reactor irradiated low-enriched fuel.

Safe irradiated nuclear fuel (INF) storage of research reactors is ensured by solving issues of protection against γ-irradiation while the neutron component is usually without consideration due to significantly lower intensity. Regarding the low-enriched composite uranium fuel of the IVG.1M reactor that is characterized by a set of elements with low and mean atomic weight where reaction is possible (α, n), evaluation of the neutron component is an indispensable procedure for ensuring radiation safety INF storage. This research suggests a method for neutron component calculation of radiation properties of fresh and irradiated fuel of the IVG.1M reactor, the α-n-component was evaluated. The research results will be useful when choosing a technology for INF storage and for analysis of feasibility to use neutron irradiation with a purpose to control fuel burnout.

Fluorescent organic particle doped polymer-based gel dosimeter for neutron detection.

The purpose of this work is to develop a material capable of detecting neutrons produced by photodisintegration in a linear accelerator for its medical use. In this study, we have developed a gel-like material doped with fluorescent organic particles. PPO at 1 wt% is used as primary dopant and POPOP as secondary one at 0.03 wt%. A set of four samples is produced, with boric acid concentrations of 0, 400, 800 and 1200 ppm. The viscoelastic properties of the material are characterized with rheological measurements, finding a gel-like behavior, i.e., a material that can keep its original shape if no stresses are applied, but can also be deformed by applying a moderate shear rate. Furthermore, the material was irradiated with gamma, electron, and neutron emission sources from 137Cs, 22Na, 60Co, 210Po, 90Sr and 241AmBe, and its response was measured in two different experimental settings, in two different institutions, for comparative purposes. From these measurements, one can clearly establish that the new material detects neutrons, electrons, and gammas within the MeV regions and below. Thus, our findings show that the developed material and its properties make it a promising technology for its use in a neutron detector.

Characterizing the coaxial HPGe detector using Monte Carlo simulations and evolutionary algorithms.

A standard procedure for characterizing the high-purity germanium detector (HPGe), manufactured by Canberra Industries Inc., is performed directly by the company using patented methods. However, the procedure is usually expensive and must be repeated because the characteristics of the HPGe crystal changes over time. In this work, the principles of a technique for use in obtaining and optimizing the detector characteristics based on a cost-effective procedure in a standard research laboratory were developed. The technique required the geometrical parameters of the detector to be determined as precisely as possible by the Monte Carlo method in parallel with the optimization process based on evolutionary algorithms. The development of this approach facilitated the modeling of the HPGe detector as a standardized procedure. The results would be also beneficial in the development of gamma spectrometers and/or their calibrations before routine measurements.

A multi-moderator neutron spectrometer for use in BNCT studies of the Tehran research reactor.

The Tehran Research Reactor is the only appropriate and available neutron source in Iran for clinical boron neutron capture therapy (BNCT). One of the requirements for BNCT is to carefully evaluate and measure the therapeutic neutron beam (epithermal neutrons) as well as the fast and thermal neutron components for successful treatment. In this research, a multi-moderator neutron spectrometer (MMNS) with LiI(Eu) scintillator as neutron counter was proposed for these measurements in the range of 10-11 eV to 15 MeV. The results confirmed promising precision of the designed MMNS for the epithermal spectrum; however, the angular dependency of the therapeutic beam due to any probable change in the beam-shaping assembly should be considered.

Neutron beam preparation for soil moisture measurements: Nedis-PHITS and artificial neural networks study.

The angular distribution of thermal neutrons scattered from the surface of a soil sample was used to determine soil moisture content. All simulations were performed assuming optimal dimensions of the designed collimator and AmO2-Be neutron source. Moreover, an assembly consisting of five Amersham X.l4 type capsular neutron sources was studied to obtain faster, more accurate data. The results showed promising agreement with previously published data and provided insights into potential applications for the construction of biological shielding of nuclear reactors.

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).

Angular distribution of scattered neutrons as a tool for soil moisture measurement: A feasibility study.

The conceptual design of a soil-moisture measurement instrument using a rectangular soil sample and an almost collimated 241Am-9Be source was proposed. Unlike previous studies and in a different simulation approach, the soil moisture was determined using the angular distribution of thermal neutrons using MCNPX2.6 Monte Carlo code, where a cylindrical BF3 proportional counter located at different polar angles was responsible for thermal neutron detection. Both Monte Carlo library least-squares method (MCLLS) and artificial neural networks (ANN) were used to calculate the soil moisture based on BF3 count rates with small relative error, about 2% and 10% maximum relative errors, respectively.

Nedis-Serpent simulation of a neutron source assembly with complex internal heterogeneous structure.

The hybrid use of Nedis-2m and Serpent 2.1.30 codes to predict the radiation characteristics (i.e., neutron yield and energy spectrum) of an Am-Be source with a fine-grained mixture of americium dioxide (AmO2) and beryllium (Be) core was studied with a focus on the grain size influence on the simulation results. The study showed that the fine-grained structure of the source core would decrease the number of alpha particles participating in the nuclear reactions with 17,18O and 9Be nuclei, which softened the neutron energy spectrum and reduced the neutron yield. The simulations also confirmed that the source core made of the stable crystals of AmBe13 intermetallic alloy would improve the neutron yield to maximum 50% compared to the core made of AmO2. Moreover, a source with a variable neutron yield was proposed with a heterogeneous core of AmO2 rods embedded in Be. The neutron energy spectrum of heterogeneous source resembled the energy spectrum of Deuterium-Tritium (D-T) neutrons which were generated in a long magnetic trap with high-temperature plasma. The subcritical irradiation facility assembled from the nth number of heterogeneous Am-Be source can be used to study the properties of materials and the equipment operating in the epithermal and fast neutron spectra. The use of a heterogeneous Am-Be assembly, as a basic element of an irradiation installation, simplifies the handling and operation procedures because it is easily disabled by removing the Be layer, or by inserting a sheet of the appropriate size and material between the Be and Am rod.

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.