Search

In homeland security, neutron detection is used to prevent the smuggling of Special Nuclear Materials in the fight against nuclear terrorism. Thermal neutrons are normally detected with 3 He proportional counters surrounded by a polyethylene box. However, due to the 3 He shortage in the radiation portal monitors reported in 2009, new procedures are being studied. In this work, Monte Carlo methods (using the MCNP6 code) have been used to study the neutron detection features of a 10 B + ZnS(Ag) under real conditions inside a radiation portal monitor. The performance of neutron detection was carried out for 252 Cf, HEU 235 U (enriched at 70% and 94%) and 239 Pu (enriched at 93%) under two different conditions, bare and shielded with lead and polyethylene. To mimic an actual situation occurring at border areas, when a vehicle was passing through the detection zone, a sample of Special Nuclear Materials, SNM, sited inside a cargo vehicle mixed with scrap was simulated, and the radiation portal monitors contained 10 B + ZnS(Ag) neutron detectors. The responses were calculated for four different neutron detectors, two real detectors (N-15 and N-48) and two prototype detectors (N-15 plus and N-48 plus), which are based on the N-15 and N-48 geometries with an improvement in the amount of 10 B and a higher polymethyl methacrylate, PMMA, thickness. Measurements were reported for the actuals detectors, and the response was greater than 200 cm from a neutron source of 252 Cf in c/s-ng of 252 Cf. An array of three N-15 or one array of N-48 are close to detect 2.5 c/s-ng of 252 Cf as required the Pacific Northwest National Laboratory, PNNL. The PNNL is the laboratory that tests almost every alternative detector. It requires that each detector must have an absolute efficiency greater than 2.5 c/s-ng of 252 Cf for such a 252 Cf source, located 200 cm perpendicular to the geometric midpoint of the neutron sensor, complementing the ANSI recommendations that the detector must be able to detect the pass of a neutron source of 252 Cf of 20,000 n/s at a distance of 200 cm, to be considered an attractive alternative for the 3 He neutron detectors. The N-15 plus system reached more than 2.5 c/s-ng of 252 Cf. Therefore, the 10 B + ZnS(Ag) detectors are an innovative and viable replacement for the 3 He detectors in the radiation portal monitor. [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]

A passive neutron area monitor has been designed using Monte Carlo methods; the monitor is a polyethylene cylinder with pairs of thermoluminescent dosimeters (TLD600 and TLD700) as thermal neutron detector. The monitor was calibrated with a bare and a thermalzed 241 AmBe neutron sources and its performance was evaluated measuring the ambient dose equivalent due to photoneutrons produced by a 15 MV linear accelerator for radiotherapy and the neutrons in the output of a TRIGA Mark III radial beam port. [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]

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]

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]