Neutron spectrum and doses in a 18 MV LINAC

Bonner sphere spectrometer with TLDs pairs has been utilized to measure the neutron spectrum 100 cm from isocenter of a 18 MV LINAC, simultaneously the ambient dose equivalent due to neutrons and photons was measured in the control room area with neutron and gamma-ray area monitors. Measurements were carried out when the LINAC was delivering a dose of 600 MU at the isocentre that was located at 5 cm depth of a head phantom. Undesired neutron field in the treatment room produce activation reactions with nuclei in different materials of LINAC, couch, air, and phantom. To determine the dose due to decay of activation nuclei the ambient dose due to gamma-rays was measured inside the treatment room immediately after dose was delivered. Measured spectrum has two peaks, one between 0.1–1 MeV and other in the thermal region, the ambient dose equivalent in the control room are 3.1 and 0.93 μSv h⁻¹ for photons and neutrons, respectively. In the treatment room the ambient dose equivalent due to photons produced during decay of activation nuclei varies from 6 to 26.1 μSv h⁻¹.     

Characterisation of mixed radiation field produced in medical linear accelerators using foil activation technique

The photon spectrum produced in medical linear accelerators and used for tumour therapy was measured using foil activation techniques in this work. The machine employed is the linear medical accelerator SL-25, Philips, installed at the Walsgrave Hospital Radiotherapy Centre in Coventry, U.K. A number of foil sets, with different energy thresholds were irradiated at different points inside a 400 mm by 400 mm treatment field at a nominal dose rate of 400 MU (4 Gy/min), and photon energy of 25 MV at the machine's isocentre. The induced activity of each foil was measured using a NaI(Tl) detector and a PC-based multichannel analyzer. The spectrum of the photons was unfolded using the computer code LOUHI82. The relative changes in the spectrum across the treatment field, were also measured using foils placed at 2.5°, 5°, 10° and 13° on both sides of the central axis of the treatment field. In order to estimate the extra dose received by the patient due to the neutron component, the neutron flux distribution at different points across the treatment field was measured using gold foils. The results and implications are discussed.     

Shielding design and detection of neutrons from medical and industrial electron accelerators--simple method of design calculation for neutron shielding

The neutron leakage from medical and industrial electron accelerators has become an important problem and its detection and shielding is being performed in their facilities. This study provides a new simple method of design calculation for neutron shielding of those electron accelerator facilities by dividing into the following five categories; neutron dose distribution in the accelerator room, neutron attenuation through the wall and the door in the accelerator room, neutron and secondary photon dose distributions in the maze, neutron and secondary photon attenuation through the door at the end of the maze, neutron leakage outside the facility-skyshine.     

Neutron spectrum correction of k0-factors for k0-based neutron-induced prompt gamma-ray analysis

Neutron spectrum correction has been attempted for the k₀-factors of the non-1/v elements which are affected neutron spectrum difference. Effective g-factors and Westcott g-factors, which are neutron spectrum correction factors obtained from an actual neutron spectrum and the Maxwellian distribution, respectively, for the non-1/v elements were calculated using their neutron cross section data of JENDEL-3.2. The neutron spectrum correction was made for the measured k₀-factors of the non-1/v elements such as Cd, Sm and Gd with the cold and thermal guided neutron beams of JRR-3M using the g-factors. The corrected k₀-factors between the cold and thermal neutron beams using both g-factors for both neutron beams agreed well for Cd. However, 9 to 44% deviations have been found for Sm and Gd, respectively.     

Spectra and dose with ANN of 252Cf, 241Am–Be, and 239Pu–Be

Artificial neural networks have been applied to unfold the neutron spectra and to calculate the effective dose, the ambient equivalent dose, and the personal dose equivalent for ²⁵²Cf, ²⁴¹Am–Be, and ²³⁹Pu–Be neutron sources. The count rates that these neutron sources produce in a Bonner Sphere Spectrometer with a ⁶LiI(Eu) were utilized as input in both artificial neural networks. Spectra and the ambient dose equivalent were also obtained with BUNKIUT code and the UTA4 response matrix. With both procedures spectra and ambient dose equivalent agrees in less than 10%. The Artificial neural network technology is an alternative procedure to unfold neutron spectra and to perform neutron dosimetry.     

The new cold neutron depth profiling instrument at NIST

The Cold Neutron Depth Profiling (CNDP) instrument at the NIST Cold Neutron Research Facility (CNRF) is now operational. The neutron beam originates from a 16 liter D₂O-ice cold source and passes through a filter of 13.5 cm of single crystal sapphire. The neutron energy spectrum may be described by a 65 K Maxwellian distribution. The sample chamber configuration allows for remote controlled scanning of 15 cm×15 cm samples and varying of both sample and detector angle. The improved sensitivity over the current thermal depth profiling instrument has permitted the first nondestructive measurements of¹⁷O profiles. Results of some of the first sample measurements are presented.     

Photoneutron spectra around an 18 MV LINAC

Photoneutron spectra around an 18 MV LINAC were calculated in order to observe the effect produced by media around the accelerator. Calculations were carried out with MCNP 4C code, three different cases were analyzed: Head model, Head and phantom model, and Head, air, phantom and wall model. The spectra were calculated in five detectors located at the irradiation room at different distances from the isocentre. A sixth detector, located near the entrance door was included to analyze how the maze change the neutron spectrum. Neutrons are mainly produced in the LINAC head change the shape of evaporation neutrons from the source term, some of these neutrons leak out the head with lesser energy, another neutrons goes with the treatment beam. At any site near the isocentre neutron spectrum has evaporation and thermal neutrons joined by a set of epithermal neutrons. As the distance from the isocentre increases evaporation neutrons tend to decrease while, epithermal and thermal neutrons tend to remain constant regardless de distance due to room return produced by the walls. The maze contributes to reduce the neutron fluence, reducing the evaporation neutrons; resulting spectrum is mainly the contribution of thermal and epithermal neutrons. Near the door these neutrons can produce activation and prompt gamma rays.     

Determination of boron in materials by cold neutron prompt gamma-ray activation analysis

An instrument for cold neutron prompt gamma-ray activation analysis (PGAA), located at the NIST Center for Neutron Research (NCNR), has proven useful for the measurement of boron in a variety of materials. Neutrons, moderated by passage through liquid hydrogen at 20 K, pass through a (58)Ni coated guide to the PGAA station in the cold neutron guide hall of the NCNR. The thermal equivalent neutron fluence rate at the sample position is 9 x 10(8) cm(-2) s(-1). Prompt gamma rays are measured by a cadmium- and lead-shielded high-purity germanium detector. The instrument has been used to measure boron mass fractions in minerals, in NIST SRM 2175 (Refractory Alloy MP-35-N) for certification of boron, and most recently in semiconductor-grade silicon. The limit of detection for boron in many materials is <10 ng g(-1).     

Measurement of neutron emission from a Ti−D2 system

Neutron emission from the d-d nuclear fusion reaction, D/d,n/³He, in and on titanium metals /titanium sponge and the mixture of titanium powder/ trapped deuterium at about 1 atm has been ascertained by using a high resolution liquid scintillation detector. The neutron emissions from 11 samples which were provided under wide varieties of conditions were measured by temperature change in the range of liquid nitrogen temperature to 350 °C. As a result, it was proved that the neutron emission observed can be divided into two types, such as cooling and heating, by the evolved conditions. Moreover, by estimating the neutron emission efficiencies of samples, it was suggested that the neutron emission reactions are closely related to the deuterium trapped in the surface of titanium metal.     

Neutron Equivalent Dose Rate Measurements of Gypsum-Waste Tire Rubber Layered Structures

The neutron equivalent dose rates (µSv/h) of gypsum, steel-reinforced rubber waste tire, and gypsum-waste tire rubber sandwich composite samples were investigated. Prepared samples were irradiated with ²⁴¹Am-Be neutrons and transmission values were obtained using dose equivalent rates measured with a BF₃ neutron detector. Results were compared to those of concrete, and as a result of neutron shielding, the performance of gypsum, waste tire, and waste tire (steel-reinforced rubber) embedded gypsum samples was higher than that of concrete. This information may be useful for shielding design of nuclear application areas.     

Measurements of fast neutron capture cross sections on <Superscript>63</Superscript>Cu and <Superscript>186</Superscript>W

Neutron capture cross sections on ⁶³Cu and ¹⁸⁶W were measured by fast neutron activation method at neutron energies from 1 to 2 MeV. Monoenergetic fast neutrons were produced by ³H(p,n)³He reaction. Neutron energy spread by target thickness, which was assumed to be the main factor of neutron energy spread, was estimated to be 1.5% at neutron energy of 2.077 MeV. Neutron capture cross sections on ⁶³Cu and ¹⁸⁶W were calculated by reference comparison method on those of ¹⁹⁷Au(n,γ). Not only statistical errors of gamma-counts from samples but also systematic errors in the counting efficiency for HP Ge detector and the uncertainty of areal density of samples were considered in calculating neutron capture cross section. Estimated neutron capture cross sections on ⁶³Cu and ¹⁸⁶W were also compared with ENDF-6 data.     

Measurement of photoneutrons in the output of 15 MV varian clinac 2100C LINAC using bubble detectors

Photoneutron contamination in the output of Varian Clinac 2100C medical linear accelerator (LINAC) operated at 15 MV photon beam energy has been investigated using bubble detectors. Photoneutrons are produced from the photo-disintegration reaction of photons with materials of the head components with threshold energy of approximately 8 MeV. Measurements were conducted in the patient plane at 100 cm source-to-detector distance on beam axis and at stipulated distances outside the irradiated field for 5×5 cm² to 40×40 cm² field sizes for in-air and 5×5 cm² to 20×20 cm² for water phantom measurements. Neutron dose equivalent of 1.57±0.10 mSv·Gy⁻¹ was measured for 10×10 cm² field size for in-air. For in phantom, neutron dose equivalent of 1.42 mSv·Gy⁻¹ was measured for 10×10 cm² field size on the beam axis at a depth of 1 cm but independent of field size at depth >5 cm.     

Study of Doppler broadening of gamma-ray spectra in 14-MeV neutron activation analysis

Neutron reactions producing characteristic photons of isotopes are important for nondestructive analysis of materials. Technique to determine the intensity of neutron induced gamma rays by fitting a spectrum with a Gaussian function using detector resolution curves derived from isotopic sources may fail if the peak is Doppler-broadened. This leads to the miscalculation of the area of the peak and, therefore, to misidentification of the material. This work shows that Doppler broadening occurs in the 14-MeV neutron analysis with photons emitted in inelastic scattering reactions on light nuclei with excited states whose lifetimes are much smaller than the time of flight of a recoiling nucleus in the material. It provides groundwork for analysis of gamma ray spectra utilizing detector response functions measured with a 14-MeV neutron source using actual geometry of an active interrogation system.     

Neutron and gamma ray doses from a 252Cf brachytherapy source in a water phantom

Axial and radial doses of Neutrons and gamma rays from an Isotron ²⁵²Cf Brachytherapy source were calculated in a Water phantom using Maienshein’s prompt fission gamma rays data and Maxwellian neutron energy spectrum. It was observed that neutron dose due to the source casing thickness does not contribute significantly to the total dose. Further the calculated secondary gamma ray dose rate is very small compared to the calculated primary gamma dose rate. Neutron and secondary gamma ray dose calculated in this study agree with the published data. Results of this study will be presented here.     

Optimization of HANARO cold neutron induced prompt gamma activation analysis system by using Monte Carlo code

Prompt gamma activation analysis (PGAA) is a nuclear analytical technique for non-destructive determination of elemental and isotopic compositions. The principle of PGAA technique is based on detection of captured gamma-ray emitted from an analytical sample while being irradiated with neutrons. Use of a cold neutron beam guide greatly reduces the gamma-ray background at the analytical sample while maintaining a neutron capture rate is comparable to that of standard thermal neutron PGAA. A new cold neutron induced prompt gamma activation analysis (CN-PGAA) system has been under construction since April of 2009 at the HANARO Cold Neutron Building (KAERI, Republic of KOREA). In this study, the Compton suppression factor of the CN-PGAA system was estimated to be 5.5 using a ⁶⁰Co radioactive source in conjunction with the MCNPX simulations. Several parameters of the CN-PGAA system were studied to estimate and optimize the performance of the system: scintillation material in the guarded detector of a Compton suppression spectrometer (CSS); the relative positions of the HPGe detector and annular detector; and the distance between the HPGe detector and back catcher BGO detectors of the CSS. In addition, the neutron ray-trace simulation package, McStas, was adopted to predict the neutron flux and wavelength distribution at the end of the cold neutron beam guide. These results served as input for the MCNPX simulation of the CN-PGAA system.     

Computational study of pulsed neutron induced activation analysis of cargo

Pulsed neutron induced activation analysis is a nondestructive technique to detect threats hidden in bulk objects such as cargo pallets, trucks, etc. Isotopic content of cargo can be measured by counting photons emitted with characteristic energies as a result of neutron induced reactions within cargo’s materials. Neutron and gamma radiation transport in active interrogation system consisting of a 14-MeV neutron source, photon detector, and a cargo truck was analyzed with MCNPX code. Gamma ray signatures of cargo with hidden explosive threat were analyzed during the neutron pulse and between neutron pulses for varying system’s geometry and material composition of cargo.     

Production of double-sided targets by electrodeposition: initial evaluation and optimization of performance

The Detector for Advanced Neutron Capture Experiments (DANCE) located at the Los Alamos Neutron Science Center is used to perform neutron capture cross-section measurements on radioactive and non-radioactive isotopes. Thin actinide targets for the DANCE detector are typically prepared by molecular deposition on thin titanium foils. For the preparation of double-sided deposits, a Teflon electrodeposition cell was constructed with two liquid chambers with a foil substrate in between, allowing electrodeposition on both sides of the foil. We have been studying the electrodeposition of uranium from isopropyl alcohol solutions using this cell. Effects of acid composition, uranium concentration, current, and deposition time will be described.     

Developments in accelerator based boron neutron capture therapy

This paper will review the current status of Boron Neutron Capture Therapy (BNCT), from basic physical mechanisms and clinical indications, to neutron beam development and dosimetry. For in-hospital facilities, particle accelerators presently provide the favoured option, and this paper concentrates on this approach to neutron beam production for BNCT. Various accelerator-based approaches will be reviewed, but discussion will concentrate on the Birmingham programme, particularly the design of a suitable neutron beam delivery system and the experimental validation of Monte Carlo simulations on a mock-up neutron beam moderation system. The use of dose modifying factors to evaluate the likely clinical utility of an epithermal neutron beam will also be discussed, with illustrations from the Birmingham programme.     

Neutron spectrum parameters in irradiation channels of the Nigeria Research Reactor-1 (NIRR-1) for the <Emphasis Type="Italic">k</Emphasis><Subscript>0</Subscript>-NAA standardization

Summary The first nuclear research reactor in Nigeria has been commissioned for neutron activation analysis and limited radioisotope production. In order to extend its utilization to include the k₀-standardization method, the following neutron spectrum parameters in inner and outer irradiation channels were determined by the “Cd-ratio for multi-monitor method”: the thermal-to-epithermal flux ratio, f, and the epithermal flux shape factor, α. Neutron spectrum parameters determined in the inner irradiation channel B2, are: α = -0.052±0.002 and f = 19.2±0.5. For the outer irradiation channel B4, the neutron spectrum parameters were found to be α = +0.029±0.003 and f = 48.3±3.3. The results are compared with the neutron spectrum parameters of other reactor facilities with similar core configuration such as the Slowpoke and Miniature Neutron Source Reactor facilities available in the literature.     

Neutron transport simulation (selected topics)

Neutron transport simulation is usually performed for criticality, power distribution, activation, scattering, dosimetry and shielding problems, among others. During the last fifteen years, innovative technological applications have been proposed (Accelerator Driven Systems, Energy Amplifiers, Spallation Neutron Sources, etc.), involving the utilization of intermediate energies (hundreds of MeV) and high-intensity (tens of mA) proton accelerators impinging in targets of high Z elements. Additionally, the use of protons, neutrons and light ions for medical applications (hadrontherapy) impose requirements on neutron dosimetry-related quantities (such as kerma factors) for biologically relevant materials, in the energy range starting at several tens of MeV. Shielding and activation related problems associated to the operation of high-energy proton accelerators, emerging space-related applications and aircrew dosimetry-related topics are also fields of intense activity requiring as accurate as possible medium- and high-energy neutron (and other hadrons) transport simulation. These applications impose specific requirements on cross-section data for structural materials, targets, actinides and biologically relevant materials.Emerging nuclear energy systems and next generation nuclear reactors also impose requirements on accurate neutron transport calculations and on cross-section data needs for structural materials, coolants and nuclear fuel materials, aiming at improved safety and detailed thermal-hydraulics and radiation damage studies.In this review paper, the state-of-the-art in the computational tools and methodologies available to perform neutron transport simulation is presented. Proton- and neutron-induced cross-section data needs and requirements are discussed. Hot topics are pinpointed, prospective views are provided and future trends identified.