Incident Thermal Neutrons Research Articles

The response of an Al-10B4C coupled neutron detector based on PIPS technology to Cf-252

After decades of discussions, it has been firmly established that detectors made of silicon-based semiconductor materials can be effectively used for neutron detection by simply coating them with suitable substances. The incident thermal neutrons interact with the coating neutron-sensitive materials such as 10B and 6LiF, resulting in the production of secondary charged particles which can be effectively detected in the sequencial silicon substrate. In this article, the detector system was designed with a coupled neutron detector structure which combined a silicon detector with a 10B4C film in various forms. The 10B4C layer was deposited on the substract with electron beam evaporation method. Two kinds of structrue were discussed: (1) one was the direct contact neutron detector by depositing 10B4C directly onto the front surface of silicon-based detectors; (2) the other was the coupled neutron detectors by depositing 10B4C onto substrates made from different materials such as Al and glass which then coupled with silicon-based detectors. The responses of these neutron detectors to neutrons (Cf-252) were measured individually. It's showen that the detection capability of direct contact neutron detectors was lower than the coupled neutron detectors. For the coupled detectors, the detector by depositing 10B4C on the aluminum substrate was found to be superior than that by depositing 10B4C on the glass substrate.

Simultaneous detection of fast and thermal neutrons with a stilbene-6Li glass composite scintillator

For simultaneous detection of fast and thermal neutrons, a composite scintillator consisting of stilbene crystal and 6Li glass was designed. Energy calibration was finished with a 152Eu γ source through Compton coincidence measurement. A 241Am-Be neutron source with polyethylene moderator was used to test it. The detection efficiency was examined through Monte Carlo simulation. The measured results show that the composite scintillator has excellent triple pulse shape discrimination performance. With the threshold of 100 keVee, figure of merits for γ-fast neutron and fast neutron-thermal neutron are 1.571 ± 0.009 and 2.376 ± 0.010 respectively. Two thermal neutron peaks emerge at 0.245 MeVee and 0.291 MeVee in the energy spectra. 79.24% of the incident thermal neutrons can be detected by 6Li glass. 39.56% of 1 MeV neutrons can be detected by stilbene.

Response of TLD-600/TLD-700 and CR-39 to neutrons for medical dosimetry.

In this study, thermoluminescence detectors (TLD)-600 and TLD-700 were used under different conditions to study neutron dosimetry for its application in medical dosimetry as albedo dosimeters. 6 LiF has a high cross-section for neutrons and is more sensitive to neutrons than to gamma radiation. Conversely, TLD-700 showed a better response to gamma radiation. Therefore, to obtain the response for neutrons, the responses for TLD-600 and TLD-700 were subtracted. A cadmium sheet was used to absorb incident thermal neutrons so that detector measured only backscattered neutrons from the albedo dosimeter. A Perspex sheet was used as a moderator to thermalize the fast neutron source Am-Be used in the experiment. CR-39 was used to detect fast neutrons and act as an albedo dosimeter when covered with a 10 B disc. Both TLD and CR-39 dosimeters were successful in detecting scattered neutrons in radiotherapy rooms.

Multilayered Solid-State Neutron Sensor

Recent advances in semiconductor processing technology have enabled development of a solid-state neutron sensor with exceptional efficiency for its unique reduced size. The boron-10 ( n, alpha) capture reaction is used to detect incident thermal neutrons. This reaction offers significant advantages for a small detector because the reaction products have high energies which can generate a large number of charge carriers in semiconductor materials by impact ionization. The enabling and differentiating technology of the sensor design is a stack consisting of multiple bilayers alternating between converter material (boron) and collector semiconductor material (silicon). As the number of these bilayers increases, the probability of a neutron interacting with boron-10 in at least one of the bilayers-so it is captured in the detector before it has a chance to pass through-approaches unity. The sensor can be made very thin while also remaining highly efficient, offering a unique possibility for low-profile in situ dosimetry. Calculations have shown that a multilayer sensor with an efficiency greater than 70% is feasible in a sensor stack 2 mm thick.

Thermal neutron detector based on COTS CMOS imagers and a conversion layer containing Gadolinium

In this work we will introduce a novel low cost position sensitive thermal neutron detection technique, based on a Commercial Off The Shelf CMOS image sensor covered with a Gadolinium containing conversion layer. The feasibility of the neutron detection technique implemented in this work has been experimentally demonstrated. A thermal neutron detection efficiency of 11.3% has been experimentally obtained with a conversion layer of 11.6 μm. It was experimentally verified that the thermal neutron detection efficiency of this technique is independent on the intensity of the incident thermal neutron flux, which was confirmed for conversion layers of different thicknesses. Based on the experimental results, a spatial resolution better than 25 μm is expected. This spatial resolution makes the proposed technique specially useful for neutron beam characterization, neutron beam dosimetry, high resolution neutron imaging, and several neutron scattering techniques.

Effect of particle size and percentages of Boron carbide on the thermal neutron radiation shielding properties of HDPE/B4C composite: Experimental and simulation studies

In this paper the effects of particle size and weight percentage of the reinforcement phase on the absorption ability of thermal neutron by HDPE/B4C composites were investigated by means of Monte-Carlo simulation method using MCNP code and experimental studies. The composite samples were prepared using the HDPE filled with different weight percentages of Boron carbide powder in the form of micro and nano particles. Micro and nano composite were prepared under the similar mixing and moulding processes. The samples were subjected to thermal neutron radiation. Neutron shielding efficiency in terms of the neutron transmission fractions of the composite samples were investigated and compared with simulation results.According to the simulation results, the particle size of the radiation shielding material has an important role on the shielding efficiency. By decreasing the particle size of shielding material in each weight percentages of the reinforcement phase, better radiation shielding properties were obtained.It seems that, decreasing the particle size and homogeneous distribution of nano forms of B4C particles, cause to increase the collision probability between the incident thermal neutron and the shielding material which consequently improve the radiation shielding properties. So, this result, propose the feasibility of nano composite as shielding material to have a high performance shielding characteristic, low weight and low thick shielding along with economical benefit.

A simulation study on angular and micro pattern effects in GEM detectors

A useful approach for the enhancement of thermal neutrons detection has been reported here. This technique, based on the angular and micro pattern effects, has been developed and applied to the boron-coated (10B) Gas Electron multiplier (GEM) detector. In the angular effect case, as a general rule, the detector device is turned at an angle which improves the device response per unit area of the detector. While for the latter case, a regular pattern in the form of micrometer deep grooves is fabricated onto the converter coating, consequently it enhances the capture probability of the detector. For the current study, both of these techniques using a 10B-coated GEM detector have been simulated for low energy neutrons. For the evaluation of detector response thermal neutrons in the energy ranges from 25meV to 100meV were transported onto the detector surface. For this work, FLUKA MC code has been utilized. The output in both cases has been estimated as a function of incident thermal neutron energies. By employing both techniques, the angle and the micro pattern dependent efficiencies for 10B-coated GEM detectors are presented, which indicate an improved efficiency response of the device. We anticipate that by using these modifications can lead a further forward step in the development and improvement of thermal neutron detection technology.

Thermal neutron response of a boron-coated GEM detector via GEANT4 Monte Carlo code

In this work, we report the design configuration and the performance of the hybrid Gas Electron Multiplier (GEM) detector. In order to make the detector sensitive to thermal neutrons, the forward electrode of the GEM has been coated with the enriched boron-10 material, which works as a neutron converter. A total of 5×5cm2 configuration of GEM has been used for thermal neutron studies. The response of the detector has been estimated via using GEANT4 MC code with two different physics lists. Using the QGSP_BIC_HP physics list, the neutron detection efficiency was determined to be about 3%, while with QGSP_BERT_HP physics list the efficiency was around 2.5%, at the incident thermal neutron energies of 25meV. The higher response of the detector proves that GEM-coated with boron converter improves the efficiency for thermal neutrons detection.

Characterization of a mid-sized Li foil multi-wire proportional counter neutron detector

A 550cm2 thermal neutron detector was constructed with five parallel sheets of 75µm thick 6Li foil (95% enrichment) spaced 1.63cm apart. Anode wire banks containing a plurality of anode wires were strung on both sides of each foil, six banks in total. The chamber was backfilled with P-10 proportional gas and over-pressured to 1.1, 1.5, 2.0, and 2.8atm (111, 151, 202, and 284kPa). The design was tailored to allow the products from the 6Li(n,t)4He reaction to escape both sides of the Li foil simultaneously, thereby, allowing for concurrent measurement in the proportional gas. The measured intrinsic thermal neutron detection efficiency of the detector with normal incident thermal neutrons to the foil sheets was 53.8±0.20%. When the detector was angled (55° from normal) such that a 0.5cm diameter thermal neutron beam intersected all of the foil layers, the intrinsic thermal neutron detection efficiency increased to 58.6±0.21%. A 252Cf neutron source positioned at a distance of 2.0m yielded an absolute neutron detection efficiency of 0.73cpsng−1. The gamma-ray rejection ratio (GRR) was 7.67×10–9 as measured from a 60Co source for an exposure rate of 40mRhr−1. Theoretical pulse-height spectra obtained with MCNP6 agreed well with experimental data and allowed pulse-height spectra and discriminator settings to be energy-calibrated. These results demonstrate the potential for the Li foil multi-wire proportional counter (MWPC) as a viable 3He neutron detector replacement.

A method for bulk hydrogen analysis based on transmission and back scattering of fast neutrons

A method was proposed for bulk hydrogen analysis. It is based on simultaneous detection of transmitted fast neutrons and back scattered thermal neutrons from the investigated samples by 3He detectors. The fast neutron beams were obtained from 252Cf and Pu–Be neutron sources. The experimental set-up as well as samples preparation were described. Incident thermal neutrons beams obtained from either 252Cf or Pu–Be sources, were used to investigate the samples by neutron backscattering. The results obtained from transmission and backscattering of fast neutrons were compared and discussed. The advantage and capabilities of the proposed method were presented. The results obtained using fast neutron beams are more sensitive than those obtained using thermal neutron beams. Validation procedures were proposed to credit the results.

A numerical model of a coated capillary-plate thermal neutron collimator

A novel thermal neutron collimator was successfully fabricated by coating the inner surface of the capillary plate (CP) with gadolinium oxide using atomic layer deposition (ALD) technology. This CP-based collimator is efficient and compact. A numerical model is presented in the paper to estimate the main performance characteristics of the collimator and to optimize the design for specific applications. According to the results of the calculation based on currently available CPs, the FWHM of the collimator's rocking curve can be smaller than 0.15° while suppressing more than 99.9% of the incident thermal neutrons on the double wings of the curve. Such a coated CP is as thin as 1.25 mm or even thinner, providing high angular resolution with good transmission in a very limited space.

Estimation of the response function of a PADC based neutron dosimeter in terms of fluence and Hp(10)

The UAB PADC-based neutron dosimeter consists of a 500 μm thick Poly Allyl Diglicol Carbonate layer, covered with polyethylene, to originate protons from fast neutrons through elastic (n,p) scattering, with Makrofol, to flatten the energy response of the dosimeter, especially at high neutron energies, and with Nylon, where protons of 626 keV are produced through the 14N(n,p)14C reaction from incident thermal neutrons. Extensive work was performed in the past to characterise the Hp(10) response of this neutron dosimeter, with exposures to reference monoenergetic fields as well as to ISO radionuclide based sources. In the present work, all known information about the dosimeter response R is gathered in order to build the response curves as a function of incident neutron energy in terms of fluence, RΦ(E), and in terms of personal dose equivalent, RHp(10)(E), for normally incident neutrons. Once RΦ(E) is known, the calculated dosimeter readings when exposed to known neutron fields (SIGMA source (IRSN, Cadarache) and several ISO sources) are evaluated and compared to experimental measurements. Calculated and experimental values agree within uncertainties for the SIGMA source and the 241Am–Be and 252Cf sources, while differences about 20% appear for the D2O moderated sources, which have an important contribution of intermediate energy neutrons. An average value of 350 cm−2 mSv−1 for the coefficient K relating dosimeter reading with personal dose equivalent allows determining neutron personal equivalent doses in the range ∼200 keV–12 MeV with an overall standard uncertainty below 50%, as required for routine operational neutron personal dosimetry.

Simulation study using GEANT4 Monte Carlo code for a Gd-coated resistive plate chamber as a thermal neutron detector

The Resistive Plate Chamber (RPC) has been developed in many application areas ever since its introduction, from high energy physics experiments to positron emission tomography. Such detectors can be coated with a Gd layer that enables them to detect thermal neutrons. Consequently these RPCs can be utilized for industrial and medical purposes. Here, we present the configuration of a resistive plate chamber which is utilized to detect thermal neutrons by employing GEANT4 Monte Carlo code. The response of the RPC was evaluated as a function of neutron energy in the GEANT4 Monte Carlo code. The simulation results are taken for incident neutron energy in the energy range from 25 meV to 100 meV. The detection efficiency was found to be between 10% and 20%, depending on the detector configuration, for incident thermal neutrons of 25 meV energy.

Gadolinium loaded plastic scintillators for high efficiency neutron detection

Gadolinium has the highest thermal neutron absorption cross section of any naturally occurring element, and emits conversion electrons as well as atomic X-rays in over 50% of its neutron captures, which makes it a useful dopant in scintillators for detecting thermal neutrons. Gadolinium isopropoxide was studied as a possible dopant for styrene-based plastic scintillators as a convenient and inexpensive method to produce high-efficiency thermal neutron detectors. Plastic scintillators with gadolinium weight concentrations of up to 3% were transparent, uniform and defect-free and were characterized with spectral measurements performed under x-ray and neutron irradiation. The new material has the same characteristic emission of styrene with a maximum at approximately 425 nm, and a light output of 76% relative to the undoped plastic. A 13 mm thick sample containing 0.5% gadolinium by weight detected 46% of incident thermal neutrons, which makes this an attractive material for a variety of applications.

Neutron detection using large area silicon detectors

We present large-area silicon detectors for neutron detection. To detect neutrons, the silicon is coated with a thin reactive layer, such as 10B or 6Li, which has a high probability of interacting with incident thermal neutrons. Each neutron interaction in the reactive layer results in an alpha particle to be detected by the semiconductor, plus a recoil nucleus. A large-area thermal neutron detector of this type was recently constructed at the Naval Research Laboratory by coating a 9.5 cm×9.5 cm, 0.5 mm thick silicon strip detector with a thin (1.5 μm) layer of 10B. This device was used to detect the 1.47 and 1.78 MeV alpha particles that result from the interaction of incident thermal neutrons with 10B. The device also detected tritons when coated with 6LiF. This has prompted the development of a custom large-area thermal neutron detector. This design is fabricated using Silicon-on-Insulator (SOI) wafers, potentially enabling low-cost and high-volume manufacture of the detector arrays. Furthermore, the SOI technology allows for thinner (7 μm) silicon layers that significantly reduce the instrument's sensitivity to gamma ray backgrounds.

Performance of the 36 m small-angle neutron scattering spectrometer at BATAN, Serpong, Indonesia

A 36 m small-angle neutron scattering (SANS) spectrometer (SMARTer) has been installed at the end of a 49 m long neutron guide tube and located in the neutron guide hall at the Neutron Scattering Laboratory (NSL), Serpong, Indonesia. At present, this is the largest SANS spectrometer in the Asia–Pacific region and consists of an 18 m long tube collimation system and another 18 m long tube to accommodate a 128 × 128 3He two-dimensional position-sensitive detector. The detector can be moved continuously from 1.5 to 18 m and can be shifted laterally by 0.1 m to cover a large range of Q, where Q is the magnitude of the scattering vector given by Q = (4\pi /\lambda) \sin (\theta /2), θ is the scattering angle and λ is the wavelength. By selecting the rotational speed of the velocity selector, the incident thermal neutron beam will have a wavelength λ in the range 3–6 A and a Q range of 0.002–0.6 A−1. The maximum neutron flux at the sample position is 4 × 106 neutrons cm−2 s−1. Measurements of some standard samples using SMARTer are reported for inter-laboratory comparisons that show, for the first time, how SMARTer's capabilities compare with those of other prominent SANS instruments.

Neutron beams at the spin revolution

Spherical neutron polarimetry (SNP), as implemented at ILL with the second-generation zero-field polarimeter CRYOPAD II, enables to orient precisely the spin polarisation vector in the incident thermal neutron beam and to measure how it is changed (amplitude and direction) by interaction with the sample. For neutrons scattered by excitations in low-dimension magnetic systems, the components which are transverse to the initial polarisation direction are large and could be recently and systematically observed by Regnault et al. (Physica B 267–268 (1999) 227.). Maleyev has shown that such transverse components can arise uniquely from the existence of mixed nuclear-magnetic Van–Hove pair correlation functions. Therefore, neutron scattering is telling us that the excitations are strongly hybridized. We discuss what are the most important technical features of this ultimate polarised neutron technique in view of this fascinating application of inelastic neutron scattering for the years to come.

The feasibility of boron containing phosphors in thermal neutron image plates, in particular the systems M 2B 5O 9X : Eu 2+ (M=Ca, Sr, Ba; X=Cl, Br) Part II: experimental results

The storage phosphor properties of the compounds M 2 B 5 O 9 X : Eu 2+ ( M= Ca, Sr, Ba; X= Cl, Br ) have been studied. The best compounds, Sr 2 B 5 O 9 Br : Eu 2+ and Ca 2 B 5 O 9 Br : Eu 2+ , show a photostimulated luminescence yield of, respectively, 21 and 14 photons per incident thermal neutron. They contain natural boron. With these yields, image plates with detective quantum efficiency values of more than 40% are envisaged when using enriched 10B, but the stimulation energy at 633 nm is too high.

Semi-insulating bulk GaAs thermal neutron imaging arrays

Prototype thermal neutron imaging arrays have been fabricated from semi-insulating (SI) bulk GaAs. The arrays are 1 mm square Schottky diodes arranged in a 5/spl times/5 matrix. GaAs Schottky barrier radiation detectors are relatively radiation hard and can withstand higher neutron and gamma ray exposure fields than MOS-based Si diode imaging arrays. The devices use /sup 10/B to convert incident thermal neutrons to energetic Li ions and alpha particles. The truncated field effect observed with SI bulk GaAs detectors produces high and low field regions in the device. Electron-hole pairs produced in the active (or high field) region of the device contribute to the observed induced charge, whereas electron-hole pairs produced in the low field region contribute very little to the induced charge. The effect is manipulated to reduce the background gamma ray interaction rate in the devices. Preliminary results show no indication of device degradation after exposure to a total thermal neutron fluence of 1.73/spl times/10/sup 13/ n/cm/sup 2/. Images have been formed of 1, 1.5, and 2 mm holes and crosses from 2 mm thick Cd templates.

Application of BGO Scintillators to Absolute Measurement of Neutron Capture Cross Sections between 0.01 eV and 10 eV.

Applying a total energy absorption γ-ray detector composed of 12 bricks (5 × 5 cm2, 7.5 cm thick) of BGO scintillators, the absolute measurement of capture cross sections for Au and Sb has been made in an energy region between 0.01 and 10eV using the linac time-of-flight method. Incident thermal neutron flux was absolutely determined by using the BGO detection system with a Sm sample. To extend the neutron flux measurement from the thermal neutron region to higher neutron energies, the 10B(n, αγ) reaction was applied. Absolute capture yield for the relevant capture sample was obtained by the saturated capture yield at a large resonance of the sample. Gold was selected to investigate the application of the BGO detection system to the absolute measurement of the capture cross sections, since the 197Au(n, γ)198 Au reaction cross section is a well known standard one. The result of the 197Au(n, γ)198 Au reaction cross section showed good agreement with the evaluated data in JENDL Dosimetry File and ENDF/B-VI. ...