Fast pulse sampling module for real-time neutron–gamma discrimination

An adaptable and compact fast pulse sampling module was developed for the neutron–gamma discrimination. The developed module is well suited for low-cost and low-power consumption applications. It is based on the Domino Ring Sampler 4(DRS4) chip, which offers fast sampling speeds up to 5.12 giga samples per second(GSPS) to digitize pulses from front-end detectors. The high-resolution GSPS data is useful for obtaining precise real-time neutron–gamma discrimination results directly in this module. In this study, we have implemented real-time data analysis in a field programmable gate array. Real-time data analysis involves two aspects: digital waveform integral and digital pulse shape discrimination(PSD). It can significantly reduce the system dead time and data rate processed offline. Plastic scintillators(EJ-299-33), which have proven capable of PSD, were adopted as neutron detectors in the experiments. A photomultiplier tube(PMT)(model #XP2020) was coupled to one end of a detector to collect the output light from it. The pulse output from the anode of the PMT was directly passed onto the fast sampling module. The fast pulse sampling module was operated at 1 GSPS and 2 GSPS in these experiments, and the AmBe-241 source was used to examine the neutron–gamma discrimination quality. The PSD results with different sampling rates and energy thresholds were evaluated. The figure of merit(FOM) was used to describe the neutron–gamma discrimination quality. The best FOM value of 0.91 was obtained at 2 GSPS and 1 GSPS sampling rates with an energy threshold of 1.5 MeV_(ee)(electron equivalent).     

Geant4 analysis and optimization of a double crystal phoswich detector for beta–gamma coincidence detection

In this study, a novel phoswich detector for beta–gamma coincidence detection is designed. Unlike the triple crystal phoswich detector designed by researchers at the University of Missouri, Columbia, this phoswich detector is of the semi-well type, so it has a higher detection efficiency. The detector consists of BC-400 and NaI:Tl with decay time constants of 2.4 and 230 ns, respectively.The BC-400 scintillator detects beta particles, and the Na I:Tl cell is used for gamma detection. Geant4 simulations of this phoswich detector find that a 2-mm-thick BC-400 scintillator can absorb nearly all of the beta particles whose energies are below 700 keV. Further, for a 2.00-cmthick NaI:Tl crystal, the gamma source peak efficiency for photons ranges from a maximum of nearly 90% at 30 keV to 10% at 1 MeV. The self-absorption effect is also discussed in this paper in order to determine the carrier gas' s influence.     

Characterization of a new shielding rubber for use in neutron–gamma mixed fields

A variety of formulations was investigated for the fabrication of an appropriate shielding rubber to be used in neutron–gamma mixed fields. Having considered the required mechanical properties together with tungsten as the gamma-ray absorbing element, calculations with MCNPX 2.6 code confirmed that the incorporation of 5 weight percentage(wt%) of boron carbide exhibited the best performance as a thermal neutron absorber. A series of both experimental and simulation results are provided for comparison.     

A test device for isotopic γ-ray imaging with CdZnTe detector

A test device for isotopic γ-ray imaging, which consists of an isotope γ-ray source, a CdZnTe γ-ray spectrometer and other auxiliary equipment, is studied here. Compared with the conventional X-ray, the isotope γ-ray, which is utilized in this project, has its own advantages in imaging. Furthermore, with a room-temperature high-energy-resolution CdZnTe detector and a modem imaging processing technique, this device is capable of effectively suppressing the background and gaining more information, thus it can obtain a better image than conventional X-ray devices. In the experiment of PCB imaging, all soldered points and chip components are sharply demonstrated.     

The extraction and smoothing algorithms for γ-ray spectrum of a CdZnTe detector system

The extraction algorithms for pulse amplitude and smoothing of energy spectrum have a great influence on energy spectrum of γ-rays during the digital detection and analysis procedure. For a CdZnTe digital γ detector system, different extraction algorithms for pulse amplitude and smoothing of energy spectrum are discussed in this paper. The results show that extraction of pulse amplitude using the first-order derivative method and smoothing of energy spectrum using the wavelet transformation method may obtain energy spectrum with good performance.     

Research of CdZnTe detector based portable energy dispersive spectrometer

A kind of excellent CdZnTe crystal has been grown in Yinnel Tech, Inc. in recent years. Based on these CdZnTe crystals and some new techniques, a portable energy-dispersive spectrometer has been constructed which has yielded good results. CdZnTe detector has a 3% relative resolution in high-energy field and can detect gamma rays at room temperature. An integrated circuit based on preamplifier and shaping amplifier chips is connected to the detector. Voltage pulses are transformed into digital signals in MCA （multichannel analyzer） and are then transmitted to computer via USB bus. Data process algorithms are improved in this spectrometer. Fast Fourier transform （FFT） and numerical differentiation （ND） are used in energy peak＇s searching program. Sampling-based correction technique is used in X-ray energy calibration. Modified Gaussian-Newton algorithm is a classical method to solve nonlinear curve fitting problems, and it is used to compute absolute intensity of each detected characteristic line.     

Investigations on landmine detection by BF3 detector

Experiments were carried out to investigate the possible use of neutron backscattering for the detection of polyethylene (PE) sample buried in the soil. In detection of landmine by neutrons, the neutron detector and its shield play an important role. In this paper, the effects of graphite, heavy water, polyethylene and boric acid moderators on the flux of back scattered neutrons were investigated. We have also experimentally verified the effect of BF3 detector shield and obtained good agreement with theory.     

Theoretical estimates of cross sections for neutron–nucleus collisions

We construct an analytical model derived from nuclear reaction theory and having a simple functional form to demonstrate the quantitative agreement with the measured cross sections for neutron induced reactions. The neutron–nucleus total, reaction and scattering cross sections, for energies ranging from 5 to 700 MeV and for several nuclei spanning a wide mass range are estimated. Systematics of neutron scattering cross sections on various materials for neutron energies up to several hundred MeV are important for ADSS applications. The reaction cross sections of neutrons are useful for determining the neutron induced fission yields in actinides and pre-actinides. The present model based on nuclear reaction theory provides good estimates of the total cross section for neutron induced reaction.     

Experimental test of a new neutron threshold detector and its application

The Possibility of using 209Bi as a new threshold detector to measure high-energy neutrons was investigated for the first time.At the same time the experiment measured successfully the emitted neutron fluence rate,energy spectrum and dose equivalent rate distributions in the heavy ion target area using a detector complex including 209Bi,^115In,^27Al,^19Fand ^12C samples.     

Experiments on mine—detection with fast neutron activation

Fast neutron activation of nitrogen and oxygen contained in the explosives used for simulated mine samples has been preliminarily carried out in our laboratory.By spectroscopic analysis of characteristic γ-rays emitted from activated nitrogen and oxygen,mine can be idenfitied almost instantly.This technique integrated with robotics would be a method for mine scavenging.     

Progress of neutron induced prompt gamma analysis techniquein 1988-2003

This paper describes new development of the neutron induced prompt gamma-ray analysis (NIPGA) technology in 1988-2003. The pulse fast-thermal neutron activation analysis method, which utilized the inelastic reaction and capture reaction jointly, was employed to measure the elemental contents more efficiently. Lifetime of the neutron generator was more than 10000h and the performance of detector and MCA reached a high level. At the same time, Monte Carlo library least-square method was used to solve the nonlinearity problem in the NIPGA.     

Progress of neutron induced prompt gamma analysis technique in 1988～2003

This paper describes new development of the neutron induced prompt gamma-ray analysis (NIPGA) technology in 1988～2003. The pulse fast-thermal neutron activation analysis method, which utilized the inelastic re action and capture reaction jointly, was employed to measure the elemental contents more efficiently. Lifetime of the neutron generator was more than 10000h and the performance of detector and MCA reached a high level. At the same time, Monte Carlo library least-square method was used to solve the nonlinearity problem in the NIPGA.     

Optimization of the steady neutron source technique for absorption cross section measurement by using an 124Sb–Be neutron source

An improved experimental approach has been developed to determine thermal neutron absorption cross sections. It uses an ¹²⁴Sb–Be neutron source which has an average neutron energy of only about 12 keV. It can be moderated in either a water tank or a paraffin filled box and can be used for aqueous or powder samples. This new design is first optimized by MCNP simulation and then benchmarked and calibrated with experiments to verify the simulations and realize the predicted improved measurement sensitivity and reproducibility. The ¹²⁴Sb–Be source device is from 1.35 to 1.71 times more sensitive than the previous method based on the use of a ²⁵²Cf source.     

Elemental PGNAA analysis using gamma–gamma coincidence counting with the library least-squares approach

An accurate method for determining elemental analysis using gamma–gamma coincidence counting is presented. To demonstrate the feasibility of this method for PGNAA, a system of three radioisotopes (Na-24, Co-60 and Cs-134) that emit coincident gamma rays was used. Two HPGe detectors were connected to a system that allowed both singles and coincidences to be collected simultaneously. A known mixture of the three radioisotopes was used and data was deliberately collected at relatively high counting rates to determine the effect of pulse pile-up distortion. The results obtained, with the library least-squares analysis, of both the normal and coincidence counting are presented and compared to the known amounts. The coincidence results are shown to give much better accuracy. It appears that in addition to the expected advantage of reduced background, the coincidence approach is considerably more resistant to pulse pile-up distortion.     

Response functions of a Si(Li) detector for 1.3–4.0 keV monochromatic photons

The response functions of a Si(Li) detector have been measured for monochromatic photons in the energy range 1.3–4.0 keV using synchrotron radiation from a 2.5 GeV electron storage ring. The contribution of the low-energy tail, just below the Gaussian full-energy peak, changed drastically at the boundary energy of the Si-K. absorption edge. The structure of the low-energy tail and the tail-to-peak ratio can be described by introducing a new simple geometrical model. The results of a fit based on the model were excellent for all data obtained over a wide energy range.     

Absolute efficiency calibration function for the energy range 63–3054 keV for a coaxial Ge(Li) detector

A new absolute efficiency calibration function for the energy range 63–3054 keV is reported. The proposed function works very well for different geometries (Petri, Marinelli, 750 cm³ and 100 cm³ beakers). We have used our own data as well as those from several other authors.     

Geant4 simulation of multi-sphere spectrometer response function and the detection of ~(241)Am–Be neutron spectrum

This paper is aimed at detecting the neutron spectrum of~(241)Am–Be, a widely used neutron source, with the SP9 ~3He proportional counter, which is a multi-sphere spectrometer system of eight thermal neutron detectors embedded in eight polyethylene(PE) spheres of varying diameters. The transport processes of a neutron in the multi-sphere spectrometer are simulated using the Geant4 code. Two sets of response functions of the PE spheres are obtained for calculating the~(241)Am–Be neutron spectrum.Response Function 1 utilizes the thermal neutron scattering model G4 Neutron HPThermal Scattering for neutron energies of ≤4 eV, and Response Function 2 has no thermal treatment. Neutron spectra of an~(241)Am–Be neutron source are measured and compared to those calculated by using the response functions. The results show that response function with thermal treatment is more accurate and closer to the real spectrum.     

The integral form of the neutron transport equation for backward–forward scattering in bare spheres

An integral form of the transport equation for bare spheres is developed which incorporates fission plus an arbitrary proportion of isotropic, backward and forward scattering. The integral equation is solved numerically and the critical radius of the sphere is obtained. The results are compared with those from the P_N approximation as reported by Yildiz and Alcan (Yildiz, C., Alcan, E., 1995. The effect of strong anisotropic scattering on the critical sphere problem in neutron transport theory using a synthetic kernel. Annals of Nuclear Energy, 22, 671). and some interesting anomalies are found and discussed.     

Two-dimensional “transXend” detector for third-generation energy-resolved computed tomography

The “transXend” detector measures X-rays as electric currents and provides the energy distribution of the measured X-rays after analysis. Capabilities of material distinction, effective atomic number measurement, and low-dose exposure computed tomography (CT) with high K-edge contrast agent from the use of the transXend detector in energy-resolved CT have been demonstrated via the first-generation CT measurements. For application of the principle of the transXend detector to the third-generation CT for human subjects in future work, a method for fabrication of a two-dimensional transXend detector is proposed and demonstrated using a commercial two-dimensional detector and two kinds of strip absorbers. The energy-resolved CT is performed by placing a proposed absorber system in front of a two-dimensional detector, which is used for conventional current measurement CT.