Development of a new electronic neutron imaging system

A moveable neutron imaging facility using D-T neutron source based on a compact accelerator

Thermal neutron imaging with CsBr storage phosphors

Assessment of fast and thermal neutron ambient dose equivalents around the KFUPM neutron source storage area using nuclear track detectors

The amount of activation by thermal and epithermal neutrons of silver in a photographic emulsion was calculated in terms of thermal neutron flux and cadmium ratio as determined by gold foils 1.27 cm in diameter by 0.0297 cm thick. The dose rate distributions around point sources of Ag108 and Ag110 were also calculated and used to determine the dose to a photographic emulsion resulting from activation of its own silver by thermal and epithermal neutrons. Experimental support for the calculations was obtained using Dupont Type 555 film. It was found that the dose to an emulsion due to thermal neutrons only could be expressed as D(th) = 0.52 × 10−10 (nvt) tdg rads, where (nvt) is the average thermal neutron flux in neutrons/cm2 and tdg is the silver content of the emulsion in mg/cm2. The ratio of the total dose to an emulsion due to both thermal and epithermal neutrons to the dose due to thermal neutrons only was obtained as a function of cadmium ratio.

Boron Neutron Capture Therapy (BNCT) is a relatively safer technology for killing cancer cells, one of which is the Glioblastoma multiforme. One of the main components of the BNCT equipment is the collimator which functions as an exit point for epithermal neutron particles that hit cancer cells. In addition to the experimental method, BNCT research can be carried out by modeling, including using the MCNPX software based on the Monte Carlo Method. This research aimed to determine the flux distribution of fast and epithermal neutrons and the dose rate of fast neutrons and gamma that hit the target cancer cells in the phantom head of ORNL MIRD. Modeling using the MCNPX software has three main parts: cell cards, surface cards, and data cards. A tally is used on the data card to calculate the neutron flux. Based on the calculation of the modeling results, the flux of epithermal neutron is 2.87 x 109 n/cm2.s. The dose ratio of the epithermal to the fast neutron flux is 2.29 x 10-14 Gy.cm2/n. Then, the balance of the dose rate of the epithermal to the gamma is 1.64 x 10-14 Gy.cm2/n, and the ratio of epithermal to thermal neutron flux is 0.004. In this study, the epithermal neutron flux hitting the target cancer cells in cell target was moderated at 4 cm so that at a depth of 8 cm, the energy was converted into thermal neutrons. Based on the analysis of the results, it can be concluded that the neutron flux that will interact with cancer tissue is thermal neutrons, not epithermal neutron flux.

Thermal neutron dosimetry for fission track dating

Improved efficiency of high resolution thermal and cold neutron imaging

Preclinical studies for boron neutron capture therapy (BNCT) using epithermal neutrons are ongoing at several laboratories. The absorbed dose in tumor cells is a function of the thermal neutron flux at depth, the microscopic boron concentration, and the size of the cell. Dosimetry is therefore complicated by the admixture of thermal, epithermal, and fast neutrons, plus gamma rays, and the array of secondary high-linear-energy-transfer particles produced within the patient from neutron interactions. Microdosimetry can be a viable technique for determining absorbed dose and radiation quality. A 2.5-cm-diameter tissue-equivalent gas proportional counter has been built with 50 parts per million (ppm) 10B incorporated into the walls and counting gas to simulate the boron uptake anticipated in tumors. Measurements of lineal energy (y) spectra for BNCT in simulated volumes of 1-10 microns diameter show a dose enhancement factor of 4.3 for 30 ppm boron, and a "y" of 250 keV/microns for the boron capture process. Chamber design plus details of experimental and calculated linear energy spectra will be presented.

Fast and thermal neutron fluence rates from a 15 MV X-ray beams of a Siemens Primus Linac were measured using bare and moderated BF3 proportional counter inside the treatment room at different locations. Fluence rate values were converted to dose equivalent rate (DER) utilizing conversion factors of American Association of Physicist in Medicine's (AAPM) report number 19. For thermal neutrons, maximum and minimum DERs were 3.46 × 10-6 (3 m from isocenter in +Y direction, 0 × 0 field size) and 8.36 × 10-8 Sv/min (in maze, 40 × 40 field size), respectively. For fast neutrons, maximum DERs using 9” and 3” moderators were 1.6 × 10-5 and 1.74 × 10-5 Sv/min (2 m from isocenter in +Y direction, 0 × 0 field size), respectively. By changing the field size, the variation in thermal neutron DER was more than the fast neutron DER and the changes in fast neutron DER were not significant in the bunker except inside the radiation field. This study showed that at all points and distances, by decreasing field size of the beam, thermal and fast neutron DER increases and the number of thermal neutrons is more than fast neutrons.

A variety of opportunities for characterization of fresh nuclear fuels using thermal (~25meV) and epithermal (~10eV) neutrons have been documented at Los Alamos National Laboratory. They include spatially resolved non-destructive characterization of features, isotopic enrichment, chemical heterogeneity and stoichiometry. The LANSCE spallation neutron source is well suited in neutron fluence and temporal characteristics for studies of fuels. However, recent advances in high power short pulse lasers suggest that compact neutron sources might, over the next decade, become viable at a price point that would permit their consideration for poolside characterization on site at irradiation facilities. In a laser-driven neutron source the laser is used to accelerate deuterium ions into a beryllium target where neutrons are produced. At this time, the technology is new and their total neutron production is approximately four orders of magnitude less than a facility like LANSCE. However, recent measurements on a sub-optimized system demonstrated >1010 neutrons in sub-nanosecond pulses in predominantly forward direction. The compactness of the target system compared to a spallation target may allow exchanging the target during a measurement to e.g. characterize a highly radioactive sample with thermal, epithermal, and fast neutrons as well as hard X-rays, thus avoiding sample handling. At this time several groups are working on laser-driven neutron production and are advancing concepts for lasers, laser targets, and optimized neutron target/moderator systems. Advances in performance sufficient to enable poolside fuels characterization with LANSCE-like fluence on sample within a decade may be possible. This report describes the underlying physics and state-of-the-art of the laser-driven neutron production process from the perspective of the DOE/NE mission. It also discusses the development and understanding that will be necessary to provide customized capability for characterization of irradiated fuels. Potential operational advantages compared to a spallation neutron source include reduced shielding complexity, reduced energy requirements, and a production target free of fission products. Contributors to this report include experts in laser-driven neutron production (Roth, Fernandez), laser design (Haefner, Siders, Leemans), laser target design (Glenzer), spallation target/moderator design (Mocko), neutron instrumentation and characterization applications (Vogel, Bourke).

This report presents an investigation of beam holes to be provided in a medical reactor for Boron Neutron Capture Therapy. The principal requirement for the beam holes is to deliver the therapeutic doses of thermal and epithermal neutrons in a modest time (30 to 60min) with minimal fast neutron and γ-contaminants. Characteristics of the beam holes have been evaluated by 2-dim. n-γ coupling S N transport calculations. Reexaminations and revisions of the beam hole design have brought improvements of the characteristics, especially an increase of the thermal neutron flux at the horizontal thermal neutron beam port and a decrease of the fast neutron flux at the vertical epithermal neutron beam port. The design objectives for the beam holes set up in this study may be achievable even if the thermal power of the reactor is reduced from 2 to 1MW.

A series of experiments has been conducted at the Frank Laboratory of Neutron Physics (FLNP) of the Joint Institute for Nuclear Research (JINR) in order to study the possibility of determining the moisture content of coke using a standard neutron source. The proposed method is based on a measurement of the spectrum of prompt γ rays emitted when samples are irradiated by fast and/or thermal neutrons. The moisture content is determined from the area of the peaks of characteristic γ rays produced in the radiative capture of thermal neutrons by the proton (Eγ = 2.223 MeV) and inelastic scattering of fast neutrons by 16O (Eγ = 6.109 MeV). The 239Pu–Be neutron source (〈E n 〉 ~ 4.5 MeV) with an intensity of ~5 × 106 n/s was used to irradiate the samples under study. A scintillation detector based on a BGO crystal was used to register the characteristic γ radiation from the inelastic fast neutron scattering and slow (thermal) neutron capture. This paper presents the results of humidity measurement in the range of 2–50% [1, 2].

We have developed a thermal-neutron coded-aperture imager that reveals the locations of hydrogenous materials from which thermal neutrons are being emitted. This imaging detector can be combined with an accelerator to form an active interrogation system in which fast neutrons are produced in a heavy metal target by means of xcitation by high energy photons. The photo-induced neutrons can be either prompt or delayed, depending on whether neutronemitting fission products are generated. Provided that there are hydrogenous materials close to the target, some of the photo-induced neutrons slow down and emerge from the surface at thermal energies. These neutrons can be used to create images that show the location and shape of the thermalizing materials. Analysis of the temporal response of the neutron flux provides information about delayed neutrons from induced fission if there are fissionable materials in the target. The combination of imaging and time-of-flight discrimination helps to improve the signal-to-background ratio. It is also possible to interrogate the target with neutrons, for example using a D-T generator. In this case, an image can be obtained from hydrogenous material in a target without the presence of heavy metal. In addition, if fissionable material is present in the target, probing with fast neutrons can stimulate delayed neutrons from fission, and the imager can detect and locate the object of interest, using appropriate time gating. Operation of this sensitive detection equipment in the vicinity of an accelerator presents a number of challenges, because the accelerator emits electromagnetic interference as well as stray ionizing radiation, which can mask the signals of interest.

A new application of the Sohrabi albedo neutron dosimeters is reported for the first time for determination of very low-level neutron ambient dose equivalents on and around a 3.5 kJ plasma focus device (PFD). The Sohrabi dosimeters basically use a polycarbonate track detector as bare and/or in contact with B convertor(s) under special cadmium cover arrangements. Its sensitivity was improved by using enriched B under new cadmium arrangements in order to detect epithermal neutrons in addition to fast and thermal neutrons. Results of 12 dosimeters installed externally around the PFD at different azimuthal (φ) and polar (θ) angles showed that azimuthal (φ) fast, epithermal, thermal, and total neutron ambient dose equivalents were symmetric and isotropic, respectively, with values 55.15 ± 8.36, 1.36 ± 02, 0.53 ± 03, and 57.04 ± 8.62 μSv/shot at ~25 cm from anode top. Polar (θ) neutron ambient dose equivalent values on z-axis relative to 90 angle were relatively higher. Results of 38 dosimeters placed on PFD facility walls for workplace monitoring and on a BOMAB phantom at operator's location for personal dose equivalent determination showed values below minimum detection limits after exposure to 130 PFD shots. However, an operator's personal dose equivalents at ~1.0 and ~3.0 m from the anode top were estimated to be, respectively, ~13.7 and ~1.52 mSv y using azimuthal angle (φ) values if the PFD operates, for example, up to 20 shots per day for 200 d y. Even under such an extreme assumption, annual personal dose equivalent is still much below 20 mSv, the annual ICRP dose limit for workers.

Neutron counters using He3 gas at absolute pressures up to 10 atm have been constructed and tested. The relative detection efficiency for thermal, epicadmium, and fast neutrons was measured as a function of pressure from 1 to 10 atm, and compared with similar data from a boron trifluoride counter of the same size filled to a pressure of 65 cm of B10F3 gas. The ratio He3/BF3 increases approximately linearly with pressure for epicadmium and fast neutrons, reaching values of 26.6 and 21.6, respectively, at 10 atm He3 pressure. This ratio shows a saturation effect for thermal neutrons, becoming 4.45 at 10 atm He3 pressure. The gamma-ray response has been found to be low with proper pulse shaping. High voltage plateau curves for 3- and 7-atm counters are given and indicate operating regions at least 100 V wide at a counter high voltage of about 1200 V.