GEANT4 used for neutron beam design of a neutron imaging facility at TRIGA reactor in Morocco

Neutron imaging has a broad scope of applications and has played a pivotal role in visualizing and quantifying hydrogenous masses in metallic matrices. The field continues to expand into new applications with the installation of new neutron imaging facilities.In this scope, a neutron imaging facility for computed tomography and real-time neutron radiography is currently being developed around 2.0MW TRIGA MARK-II reactor at Maamora Nuclear Research Center in Morocco (Reuscher et al., 1990 [1]; de Menezes et al., 2003 [2]; Deinert et al., 2005 [3]).The neutron imaging facility consists of neutron collimator, real-time neutron imaging system and imaging process systems. In order to reduce the gamma-ray content in the neutron beam, the tangential channel was selected. For power of 250 kW, the corresponding thermal neutron flux measured at the inlet of the tangential channel is around 3×10¹¹ ncm²/s.This facility will be based on a conical neutron collimator with two circular diaphragms with diameters of 4 and 2 cm corresponding to L/D-ratio of 165 and 325, respectively. These diaphragms' sizes allow reaching a compromise between good flux and efficient L/D-ratio. Convergent-divergent collimator geometry has been adopted.The beam line consists of a gamma filter, fast neutrons filter, neutron moderator, neutron and gamma shutters, biological shielding around the collimator and several stages of neutron collimator. Monte Carlo calculations by a fully 3D numerical code GEANT4 were used to design the neutron beam line (http://www.info.cern.ch/asd/geant4/geant4.html[4]).To enhance the neutron thermal beam in terms of quality, several materials, mainly bismuth (Bi) and sapphire (Al₂O₃) were examined as gamma and neutron filters respectively. The GEANT4 simulations showed that the gamma and epithermal and fast neutron could be filtered using the bismuth (Bi) and sapphire (Al₂O₃) filters, respectively.To get a good cadmium ratio, GEANT 4 simulations were used to define the design of the moderator in the inlet of the radiation channel. A graphite block of 22 cm thickness seems to be the optimal neutron moderator.The results showed that the combination of 5 cm of bismuth with 5 cm of sapphire permits the filtration of gamma-rays, epithermal neutrons as well as fast neutrons in a considerable way without affecting the neutron thermal flux.     

The HANARO neutron reflectometer with horizontal sample geometry; engineering designs and performance simulation

A new neutron reflectometer with a horizontal sample geometry was designed and is now under construction at the HANARO, 30 MW research reactor. It was originally built and operated at the H9-A beam port at BNL, and was relocated to HANARO in 2004. We performed simulations of neutron ray-tracing to evaluate the performance of all of the optical components of the instrument with a Monte Carlo technique using McStas code. The feasible wavelength of the incident neutron beam is 2.52 Å. It produces a q-range up to 0.126 Å^− 1 with a supermirror as a deflector. Our studies indicated possibilities to improve the performance of the guide tube. Although the performance is limited (limited q-range and flux due to multiple reflections prior to the deflector), it promises to be the first reflectometer in Korea for the study of free surfaces, which is currently in demand.     

Performance test of Si PIN photodiode line scanner for thermal neutron detection

Thermal neutron imaging using Si PIN photodiode line scanner and Eu-doped LiCaAlF₆ crystal scintillator has been developed. The pixel dimensions of photodiode are 1.18 mm (width)×3.8 mm (length) with 0.4 mm gap and the module has 192 channels in linear array. The emission peaks of Eu-doped LiCaAlF₆ after thermal neutron excitation are placed at 370 and 590 nm, and the corresponding photon sensitivities of photodiode are 0.04 and 0.34 A/W, respectively. Polished scintillator blocks with a size of 1.18 mm (width)×3.8 mm (length)×5.0 mm (thickness) were wrapped by several layers of Teflon tapes as a reflector and optically coupled to the photodiodes by silicone grease. JRR-3 MUSASI beam line emitting 13.5 meV thermal neutrons with the flux of 8×10⁵ n/cm² s was used for the imaging test. As a subject for imaging, a Cd plate was moved at the speed of 50 mm/s perpendicular to the thermal neutron beam. Analog integration time was set to be 416.6 μs, then signals were converted by a delta-sigma A/D converter. After the image processing, we successfully obtained moving Cd plate image under thermal neutron irradiation using PIN photodiode line scanner coupled with Eu-doped LiCaAlF₆ scintillator.     

PKUNIFTY: A neutron imaging facility based on an RFQ accelerator

A neutron imaging facility, PKUNIFTY, based on a radio frequency quadrupole (RFQ) accelerator-driven compact neutron source, presently under construction at the Peking University, is described. It consists of a deuteron linear accelerator, a neutron target-moderator-reflector assembly, and a thermal neutron imaging system. Neutrons are generated via the deuteron-beryllium reaction with an expected fast-neutron yield of 3×10¹² n/s. The thermal neutron flux on the imaging plane is 5×10⁵ n/cm²/s at a nominal L/D ratio of 50, and the L/D ratio can be selectable over a range of 25-200. The corresponding n/γ ratio is close to or higher than 1×10¹⁰ n/cm²/Sv. The field of view is 20 cm×20 cm at 2 m downstream of the collimator entrance aperture where the thermal neutron flux uniformity is better than 7%. The effective Cd ratio can be tuned by using a light chopper and neutron time-of-flight technique without excessive sacrifice of the thermal neutron flux.     

LiF crystals as high spatial resolution neutron imaging detectors

Neutron imaging by color center formation in LiF crystals was applied to a sensitivity indicator (SI) as a standard samples for neutron radiography. The SI was exposed to a 5 mm pinhole-collimated thermal neutron beam with an LiF crystal and a neutron imaging plate (NIP) for 120 min in the JRR-3M thermal neutron radiography facility. The image in the LiF crystal was read out using a laser confocal microscope. All gaps were clearly observed in images for both the LiF crystal and the NIP. The experimental results showed that LiF crystals have excellent characteristics as neutron imaging detectors in areas such as high spatial resolution.     

Neutronic design and simulated performance of Peking University Neutron Imaging Facility (PKUNIFTY)

The Peking University Neutron Imaging Facility (PKUNIFTY) is a Radio Frequency Quadruple (RFQ) accelerator based system. The fast neutrons are produced by 2 MeV deuterons bombarding beryllium target. The moderator, reflector, shielding and collimator have been optimized with Monte-Carlo simulation to improve the neutron beam quality. The neutrons are thermalized in water cylinder of Φ26×26 cm² with a polyethylene disk in front of Be target. The size of deuteron beam spot is optimized considering both the thermal neutron distribution and the demand of target cooling. The shielding is a combination of 8 cm thick lead and 42 cm thick boron doped polyethylene. The thermal neutrons are extracted through a rectangular inner collimator and a divergent outer collimator. The thermal neutron beam axis is perpendicular to the D⁺ beam line in order to reduce the fast neutron and the γ ray components in the imaging beam. When the neutron yield is 3×10¹² n/s and the L/D is 50, the thermal neutron flux is 5×10⁵ n/cm²/s at the imaging plane, the Cd ratio is 1.63 and the n/γ ratio is 1.6×10¹⁰ n/cm²/Sv.     

Coded source neutron imaging with a MURA mask

In coded source neutron imaging the single aperture commonly used in neutron radiography is replaced with a coded mask. Using a coded source can improve the neutron flux at the sample plane when a very high L/D ratio is needed. The coded source imaging is a possible way to reduce the exposure time to get a neutron image with very high L/D ratio. A 17×17 modified uniformly redundant array coded source was tested in this work. There are 144 holes of 0.8 mm diameter on the coded source. The neutron flux from the coded source is as high as from a single 9.6 mm aperture, while its effective L/D is the same as in the case of a 0.8 mm aperture. The Richardson-Lucy maximum likelihood algorithm was used for image reconstruction. Compared to an in-line phase contrast neutron image taken with a 1 mm aperture, it takes much less time for the coded source to get an image of similar quality.     

Thermal neutron imaging with rare-earth-ion-doped LiCaAlF6 scintillators and a sealed Cf source

Thermal neutron imaging with Ce-doped LiCaAlF₆ crystals has been performed. The prototype of the neutron imager using a Ce-doped LiCaAlF₆ scintillating crystal and a position sensitive photomultiplier tube (PSPMT) which had 64 multi-channel anode was developed. The Ce-doped LiCaAlF₆ single crystal was grown by the Czochralski method. A plate with dimensions of a diameter of 50×2 mm² was cut from the grown crystal, polished, and optically coupled to PSPMT by silicone grease. The ²⁵²Cf source (<1 MBq) was sealed with 43 mm of polyethylene for neutron thermalization. Alphabet-shaped Cd pieces with a thickness of 2 mm were used as a mask for the thermal neutrons. After corrections for the pedestals and gain of each pixel, we successfully obtained two-dimensional neutron images using Ce-doped LiCaAlF₆.     

Improving the performance of the cryogenic heat pipe-target system for the COSY-TOF experiment

Very lightweight, thin liquid hydrogen/deuterium heat pipe-target systems are used in the Time Of Flight (TOF) spectrometer at the COSY accelerator facility. The proton beam impinges upon LH₂/LD₂ targets thereby heating the target. The stability of the liquid targets depends on the thermal capacity of the whole system, the energy losses from the proton beam and heat losses from the surrounding of the heat pipe-target system. The radiation heat load has been reduced by a factor of 4.5 by reducing the length of the gas tube from 180 cm (long tube) to 40 cm (short tube). Furthermore, the 40 cm long gas tube was coated with a thin polished gold layer, thereby reducing the heat load by an additional factor 22. The thermal capacity is improved by reducing the mass of the gas tube from 23 g to 5 g. The cool down time of the 7 mm diameter gold coated heat pipe with the gold coated 40 cm gas tube is reduced by 12 min.     

Thermal neutron detection with Ce doped LiCaAlF6 single crystals

Cerium-doped LiCaAlF₆ (Ce:LiCAF) crystals have been studied as scintillators in application to thermal neutron detection. Three crystals: high-doping Ce:LiCAF, low-doping Ce:LiCAF with 50% enrichment of ⁶Li (both 10 mm×10 mm×2 mm, rectangular) and high-doping Ce:LiCAF with 95% enrichment of ⁶Li (Ø50.8 mm×2 mm, discus) coupled to Photonis XP5300B PMT, were tested. The response of these crystals to neutrons emitted from a paraffin moderated ²³⁸PuBe source has been investigated. Thermal neutron peaks have been found at a Gamma Equivalent Energy (GEE) of ∼2.5 MeV for high-doping Ce:LiCAF (50% ⁶Li), ∼2 MeV for low-doping Ce:LiCAF (50% ⁶Li) and ∼1.9 MeV for high-doping Ce:LiCAF (95% ⁶Li). The light output of Ce:LiCAF was also measured (175-250 phe/MeV from sample to sample). Lithium-6 glass GS20 from Saint Gobain was used as a reference scintillator (Ø50 mm×2 mm, circle). Relative neutron efficiency, normalized to that of GS20 lithium glass, as well as gamma-neutron intrinsic efficiency for all tested samples was calculated. Intrinsic efficiency on thermal neutron detection for small Ce:LiCAF samples was estimated at about 32-35% of that of GS20 and for large Ce:LiCAF sample as about 82% of that of GS20.     

Experimental and numerical characterization of the neutron field produced in the n@BTF Frascati photo-neutron source

A photo-neutron irradiation facility is going to be established at the Frascati National Laboratories of INFN on the base of the successful results of the n@BTF experiment. The photo-neutron source is obtained by an electron or positron pulsed beam, tuneable in energy, current and in time structure, impinging on an optimized tungsten target located in a polyethylene-lead shielding assembly. The resulting neutron field, through selectable collimated apertures at different angles, is released into a 100 m² irradiation room. The neutron beam, characterized by an evaporation spectrum peaked at about 1 MeV, can be used in nuclear physics, material science, calibration of neutron detectors, studies of neutron hardness, ageing and study of single event effect. The intensity of the neutron beam obtainable with 510 MeV electrons and its fluence energy distribution at a point of reference in the irradiation room were predicted by Monte Carlo simulations and experimentally determined with a Bonner Sphere Spectrometer (BSS). Due to the large photon contribution and the pulsed time structure of the beam, passive photon-insensitive thermal neutron detectors were used as sensitive elements of the BSS. For this purpose, a set of Dy activation foils was used. This paper presents the numerical simulations and the measurements, and compares their results in terms of both neutron spectrum and total neutron fluence.     

Measurement of neutron flux and beam divergence at the cold neutron guide system of the new Munich research reactor FRM-II

A sophisticated neutron guide system has been installed at the new Munich neutron source FRM-II to transport neutrons from the D₂ cold neutron source to several instruments, which are situated in a separate neutron guide hall. The guide system takes advantage of supermirror coatings and includes a worldwide unique “twisted” guide for a desired phase space transformation of the neutron beam. During the initial reactor commissioning in summer 2004, the integral and differential neutron flux as well as the distribution of beam divergence at the exit of two representative and the twisted neutron guide were measured using time-of-flight spectroscopy and gold-foil activation. The experimental results can be compared to extensive simulation calculations based on MCNP and McStas. The investigated guides fulfill the expectations of providing high neutron fluxes and reveal good quality with respect to the reflective coatings and the installation precision.     

High-fidelity MCNP modeling of a D-T neutron generator for active interrogation of special nuclear material

Fast and robust methods for interrogation of special nuclear material (SNM) are of interest to many agencies and institutions in the United States. It is well known that passive interrogation methods are typically sufficient for plutonium identification because of a relatively high neutron production rate from ²⁴⁰Pu [1]. On the other hand, identification of shielded uranium requires active methods using neutron or photon sources [2]. Deuterium-deuterium (2.45 MeV) and deuterium-tritium (14.1 MeV) neutron-generator sources have been previously tested and proven to be relatively reliable instruments for active interrogation of nuclear materials [3] and [4]. In addition, the newest generators of this type are small enough for applications requiring portable interrogation systems.Active interrogation techniques using high-energy neutrons are being investigated as a method to detect hidden SNM in shielded containers [4] and [5]. Due to the thickness of some containers, penetrating radiation such as high-energy neutrons can provide a potential means of probing shielded SNM. In an effort to develop the capability to assess the signal seen from various forms of shielded nuclear materials, the University of Michigan Neutron Science Laboratory’s D-T neutron generator and its shielding were accurately modeled in MCNP. The generator, while operating at nominal power, produces approximately 1×10¹⁰ neutrons/s, a source intensity which requires a large amount of shielding to minimize the dose rates around the generator. For this reason, the existing shielding completely encompasses the generator and does not include beam ports. Therefore, several MCNP simulations were performed to estimate the yield of uncollided 14.1-MeV neutrons from the generator for active interrogation experiments. Beam port diameters of 5, 10, 15, 20, and 25 cm were modeled to assess the resulting neutron fluxes. The neutron flux outside the beam ports was estimated to be approximately 2×10⁴ n/cm² s.     

Detection of SNM by delayed gamma rays from induced fission

The Pulsed Neutron Interrogation Test Assembly (PUNITA) is an experimental device for research in NDA methods and field applicable instrumentation for nuclear safeguards and security applications. PUNITA incorporates a standard 14-MeV (D-T) pulsed neutron generator inside a large graphite mantle. The generator target is surrounded by a thick tungsten filter with the purpose to increase the neutron output and to tailor the neutron energy spectrum. In this configuration a sample may be exposed to a relatively high average thermal neutron flux of about (2.2±0.1)×10³ s⁻¹ cm⁻² at only 10% of the maximum target neutron emission. The sample cavity is large enough to allow variation of the experimental setup including the fissile sample, neutron and gamma detectors, and shielding materials.The response from SNM samples of different fissile material content was investigated with various field-applicable scintillation gamma detectors such as the 3×2 in. LaBr₃ detector. Shielding in the form of tungsten and cadmium was applied to the detector to improve the signal to background ratio. Gamma and neutron shields surrounding the samples were also tested for the purpose of simulating clandestine conduct. The energy spectra of delayed gamma rays were recorded in the range 100 keV-9 MeV. In addition time spectra of delayed gamma rays in the range 3.3-8 MeV were recorded in the time period of 10 ms-120 s after the 14-MeV neutron burst. The goal of the experiment was to optimize the sample/detector configuration including the energy range and time period for SNM detection. The results show, for example, that a 170 g sample of depleted uranium can be detected with the given setup in less than 3 min of investigation. Samples of higher enrichment or higher mass are detected in much shorter time.     

Hexagonal boron nitride epitaxial layers as neutron detector materials

Micro-strip metal-semiconductor-metal detectors for thermal neutron sensing were fabricated from hexagonal boron nitride (hBN) epilayers synthesized by metal organic chemical vapor deposition. Experimental measurements indicated that the thermal neutron absorption coefficient and length of natural hBN epilayers are about 0.00361 μm⁻¹ and 277 μm, respectively. A continuous irradiation with a thermal neutron beam generated an appreciable current response in hBN detectors, corresponding to an effective conversion efficiency approaching ∼80% for absorbed neutrons. Our results indicate that hBN semiconductors would enable the development of essentially ideal solid-state thermal neutron detectors in which both neutron capture and carrier collection are accomplished in the same hBN semiconductor. These solid-state detectors have the potential to replace ³He gas detectors, which faces the very serious issue of ³He gas shortage.     

A CdTe detector with a Gd converter for thermal neutron detection

A CdTe detector with a Gd converter has been developed and investigated as a neutron detector for neutron imaging. The fabricated Gd/CdTe detector with the 25 μm thick Gd was designed on the basis of simulation results of thermal neutron detection efficiency and spatial resolution. The energy resolution of the Gd/CdTe detector is less than 4 keV, which is enough to discriminate neutron capture gamma rays from background gamma emission. The Gd/CdTe detector shows the detection of neutron capture gamma ray emission in the ¹⁵⁵Gd(n, γ)¹⁵⁶Gd, ¹⁵⁷Gd(n, γ)¹⁵⁸Gd and ¹¹³Cd(n, γ)¹¹⁴Cd reactions and characteristic X-ray emissions due to conversion-electrons generated inside the Gd film. The observed efficient thermal neutron detection with the Gd/CdTe detector shows its promise in neutron radiography application.     

Development of a dedicated ion injector for accelerator-based nanoprobe facilities

The paper discusses possible ways of increasing beam brightness in ion injectors. The argon/helium ion injector comprising a newly designed RF ion source and, a Wien filter has been designed for use in accelerator-based nanoprobe facilities. The phase set degradation due to aberrations in the injector ion-optic system was simulated with allowance for multipole and fringing fields. The RF ion sources with different permanent magnet systems were tested. Experiments were performed with argon and helium. A plasma density of up to 3×10¹¹ cm⁻³ and beam brightness of ∼100 A/(m² rad² eV) were obtained. The ion current density inside an extracting electrode in the source was 10 mA/cm² for an emission hole diameter of 0.6 mm. Measurements of the current value and emittance were performed with ion source testing equipment permitting measurements of the ion beam current, emittance, mass composition, and RF power input into the plasma.     

Basic study of Europium doped LiCaAlF6 scintillator and its capability for thermal neutron imaging application

Eu²⁺ 0.1, 0.5, 1, and 2 mol% doped LiCaAlF₆ single crystalline scintillators were grown by the micro-pulling down (μ-PD) method. Eu²⁺ 2 mol% doped LiCaAlF₆ was also prepared using the Czochralski method. In the transmittance spectra, 4f-5d absorption lines appeared around 200-220 and 290-350 nm. An intense emission at 375 nm due to Eu²⁺ 5d-4f transition was observed under ²⁴¹Am α-ray excitation. When ²⁵²Cf excited pulse height spectra were measured, Eu 2% doped one showed the highest light yield of 29,000 ph/n with 1.15 μs decay time. Using the 2 inchφ Czochralski grown one coupled with the position sensitive photomultiplier tube covered by Cd mask with various size (1, 2, 3, and 5 mm) pin holes, thermal neutron imaging was examined. As a result, the spatial resolution turned out to be better than 1 mm.     

A means to estimate thermal and kinetic parameters of coal dust layer from hot surface ignition tests

A method to estimate thermal and kinetic parameters of Pittsburgh seam coal subject to thermal runaway is presented using the standard ASTM E 2021 hot surface ignition test apparatus. Parameters include thermal conductivity (k), activation energy (E), coupled term (QA) of heat of reaction (Q) and pre-exponential factor (A) which are required, but rarely known input values to determine the thermal runaway propensity of a dust material. Four different dust layer thicknesses: 6.4, 12.7, 19.1 and 25.4 mm, are tested, and among them, a single steady state dust layer temperature profile of 12.7 mm thick dust layer is used to estimate k, E and QA. k is calculated by equating heat flux from the hot surface layer and heat loss rate on the boundary assuming negligible heat generation in the coal dust layer at a low hot surface temperature. E and QA are calculated by optimizing a numerically estimated steady state dust layer temperature distribution to the experimentally obtained temperature profile of a 12.7 mm thick dust layer. Two unknowns, E and QA, are reduced to one from the correlation of E and QA obtained at criticality of thermal runaway. The estimated k is 0.1 W/m K matching the previously reported value. E ranges from 61.7 to 83.1 kJ/mol, and the corresponding QA ranges from 1.7 × 10⁹ to 4.8 × 10¹¹ J/kg s. The mean values of E (72.4 kJ/mol) and QA (2.8 × 10¹⁰ J/kg s) are used to predict the critical hot surface temperatures for other thicknesses, and good agreement is observed between measured and experimental values. Also, the estimated E and QA ranges match the corresponding ranges calculated from the multiple tests method and values reported in previous research.