Simulation of the Traweling Wave Burning Regime on Epithermal Neutrons

New results of two computer experiments on modeling of superthermal neutron-nuclear combustion of natural uranium for two different flux densities of external neutron source and duration of half a year each are presented. The simulation results demonstrate the dependence of the autowave combustion modes on the parameters of the external source.

Fabrication and performance evaluation of GaN thermal neutron detectors with bm^6LiF conversion

A GaN-based pin neutron detector with a 6LiF conversion layer was fabricated, and can be used to detect thermal neutrons. Measurement of the electrical characteristic of the GaN-based pin neutron detector showed that the reverse leakage current of the neutron detector was reduced significantly after deposition of a 6LiF conversion layer on the detector surface. The thermal neutrons used in this experiment were obtained from an 241Am-Be fast neutron source after being moderated by 100-mm-thick high-density polyethylene. The experimental results show that the detector with 16.9-μm thick 6LiF achieved a maximum neutron detection efficiency of 1.9% at a reverse bias of 0 V, which is less than the theoretical detection efficiency of 4.1% calculated for our GaN neutron detectors.

Determining absolute value of thermal neutron flux density based on monocrystalline silicon in nuclear reactors

A new type of neutron detector based on monocrystalline Si is developed to measure the fluence and flux density of thermal and fast neutrons.The principle of this detector is based on the relationship between changes in electrical conductivity and neutron fluence during irradiation.Therefore,the absolute values of thermal neutron fluence and flux density are measured in a facile manner with high reliability.Compared with activation methods,our method not only possesses a similar accuracy,but also demonstrates superior application potential for the investigation of neutron fields in nuclear reactors owing to its suitable half-life.

Efficiency-determined method for thermal neutron detection with inorganic scintillator

Because of 3He shortage,sintillator is a promising alternative choice for neutron detection in the field of thermal neutron scattering and imaging.Also,the neutron detection efficiency is difficult to be determined.In this paper,the efficiency for thermal neutron detection is presented by inorganic scintillator using probability principles,supposed that the material of scintillator is uniform in element distribution,and that attenuation length of scintillation light is longer than that of its thickness in the scintillator.The efficiencies for two pieces of lithium glass are determined by this method,indicating the method is useful for determining efficiency of thermal neutron detections.

B4C/NRL flexible films for thermal neutron shielding

Boron carbide/natural rubber latex(B_4 C/NRL)flexible films were prepared via dip-molding with B_4 C content in the range of 5–55 wt% for thermal neutron(0.0253 e V) shielding. B_4 C was well dispersed in NRL according to microscopic observation. Both the inside and outside surfaces of the film were smooth. For B_4 C/NRL flexible films, the minimum elongation at break was greater than 600%, the minimum tensile strength was greater than 12 MPa, and the hardness was in the range of 35–55 HA,which were suitable for preparing flexible wearable products. The attenuation efficiencies of the B_4 C/NRL flexible films for thermal neutrons were also calculated. The B_4 C/NRL flexible films exhibit good attenuation effect for thermal neutrons.

Calculation of thermal neutron albedo for mono-material and bi-material reflectors

Thermal neutron albedo has been investigated for different thicknesses of mono-material and bi-material reflectors. An equation has been obtained for a bi-material reflector by considering the neutron diffusion equation. The bi-material reflector consists of binary combinations of water, graphite, lead, and polyethylene. An experimental measurement of thermal neutron albedo has also been conducted for mono-material and bi-material reflectors by using a^(241) Am–Be(5.2 Ci) neutron source and a BF3 detector. The maximum value of thermal neutron albedo was obtained for a polyethylene–water combination(0.95 ± 0.02).

Analysis of Nuclear Track Parameters of CN-85 Detector Irradiated to Thermal Neutrons by Using MATLAB Program

CN-85 detector which covered with boric acid H3Bo3 pellete has been irradiated by thermal neutrons from (241Am-9Be) source with activity 12 Ci and neutron flux 105 n. cm-2. s-1. The irradiation times-TD for detector were 4 h, 8 h, 16 h and 24 h. The track detector has been etched with sodium hydroxide. After chemical etching of the irradiated CN-85 detector, the images have been taken from a digital camera connected to the optical microscope. Image processing for the output images has been performed using MATALB program, and these images were analyzed and we had found the following relations: a) The relation between summation of opened track or surface density for tracks (intensity-IT) varies with radius of opening (track radius-RT). b) The relation between the tracks number-NT varies with the tracks diameter-DT (in micrometer) and tracks area-AT. That analysis of image processing was obtained, and the track intensity-IT was decreased with increase of track radius-RT at all of the irradiation time-TD. And the track intensity-IT was increased with increasing irradiation time-TD (h) for different track radius-RT (0.4225, 0.845, 1.2675 and 1.69 μm). The study indicates the possibility of using the analysis of image processing to CN-85 detector for classification of α-particle emitters through limitation of radius of track-RT, in addition to the contribution of these techniques in preparation of nano-filters and nono-membrane in nanotechnology fields.

The latest progress of thermal neutron tomography at SPRR-300 reactor

Thermal neutron radiography is a useful complementary tool of the other non-destructive testing methods for the hydrogenous materials and heavy metal subassembly.By the use of MC method program and the thermal neutron digital radiography facility at SPRR-300 reactor,the simulation and experimental study of the thermal neutron tomography has been developed.Its method and result has been introduced and analyzed.

Determination the Effect of Gamma Radiation and Thermal Neutron on PM-355 Detector by Using FTIR Spectroscopy

The effect of gamma on nuclear track detector type PM-355 (at the dose range 200 to 1600 kGy) and thermal neutron (flux 105 n·cm-2·s-1) was calculated by using of two irradiation methods. First method (G + N) was an irradiation PM-355 detector by gamma radiation and then irradiation by thermal neutrons, and another method (N + G) was irradiated by thermal neutrons and then gamma radiation. FTIR-spectroscopy was used to measure the change in deferent of transmission percent ΔT% at the wavenumber 1260 cm-1 with wavenumber 2962 cm-1 [ΔT%]1260-2962 and wavenumber 1138 cm-1 [ΔT%]1260-1138. The values of [ΔT%]1260-2962 and [ΔT%]1260-1138 were increasing with the increase of gamma irradiation with maximum response at 820 kGy and then drop after this dose until to 1600 kGy. This study determined the linear equations relation between the effect of gamma radiation on PM-355 detector and the change of [ΔT%]1260-2962 and [ΔT%]1260-1138, and this change appeared in (N + G) irradiation method better than in (G + N) irradiation method. The appearance of wavenumber 2964 cm-1 in (G + N) irradiation method referred to alkyl asymmetry C-H bond stretched out of skelated plane after changes in chemical structure of PM-355 detector by gamma or neutrons radiation.

Investigation the Effects of Neutron Source Number and Arrangement in Landmines Detection by Thermal Neutron Capture Gamma Ray Analysis

Thermal neutron capture can be used as a successful technique for detection of non-metallic landmines via the detection of their constituent like nitrogen. Recently, it has been shown that the detection of 10.829 MeV photons from the 14N(n,γ)15N reaction can be used for finding the landmines. In this method a high-energy neutron source like 241Am-Be inside water as a moderator is used to have thermal neutron. In this paper we have investigated the effects of the number of neutron sources and their orientation on the gamma ray spectrum by using MCNP4C code. The best case for number of sources and their positions and orientations have been achieved corresponding to maximum flux of 10.829 MeV photons.

High efficiency event-counting thermal neutron imaging using a Gd-doped micro-channel plate

An event-counting thermal neutron imaging detector based on 3 mol % nattGd2O3-doped micro-channel plate （MCP） has been developed and tested. A thermal neutron imaging experiment was carried out with a low flux neutron beam. Detection efficiency of 33% was achieved with only one doped MCP. The spatial resolution of 72μ m RMS is currently limited by the readout anode. A detector with larger area and improved readout method is now being developed.

A Monte Carlo simulation and setup optimization of output ef f iciency to PGNAA thermal neutron using ^(252)Cf neutrons

We present the design and optimization of a prompt T-ray neutron activation analysis （PGNAA） thermal neutron output setup based on Monte Carlo simulations using MCNP5 computer code. In these simulations, the moderator materials, reflective materials, and structure of the PCNAA 2526f neutrons of thermal neutron output setup are optimized. The simulation results reveal that the thin layer paraffin and the thick layer of heavy water moderating effect work best for the 252Cf neutron spectrum. Our new design shows a significantly improved per- formance of the thermal neutron flux and flux rate, that are increased by 3.02 times and 3.27 times, respectively, compared with the conventional neutron source design.

Study on the gamma rays and neutrons energy response optimization of a scintillating fiber detector for EAST with Geant4

A new scintillating fiber detector inside magnetic shielding tube was designed and assembled for use in the next round of fusion experiments in the experimental advanced superconducting tokamak to provide D–T neutron yield with time resolution.In this study,Geant4 simulations were used to obtain the pulse height spectra for ideal signals produced when detecting neutrons and gamma rays of multiple energies.One of the main sources of interference was found to be low-energy neutrons below 10–5 MeV,which can generate numerous secondary particles in the detector components,such as the magnetic shielding tube,leading to high-amplitude output signals.To address this issue,a compact thermal neutron shield containing a 1-mm Cd layer outside the magnetic shielding tube and a 5-mm inner Pb layer was specifically designed.Adverse effects on the measurement of fast neutrons and the shielding effect on gamma rays were considered.This can suppress the height of the signals caused by thermal neutrons to a level below the height corresponding to neutrons above 4 MeV because the yield of the latter is used for detector calibration.In addition,the detector has relatively flat sensitivity curves in the fast neutron region,with the intrinsic detection efficiencies(IDEs)of approximately 40%.For gamma rays with energies that are not too high(<8 MeV),the IDEs of the detector are only approximately 20%,whereas for gamma rays below 1 MeV,the response curve cuts off earlier in the low-energy region,which is beneficial for avoiding counting saturation and signal accumulation.

Experimental research on a THGEM-based thermal neutron detector

A new thermal neutron detector with a domestically produced THGEM （Thick Gas Electron Multiplier） was developed as an alternative to 3He to meet the needs of the next generation of neutron facilities. One type of Au-coated THGEM was designed specifically for neutron detection. A detector prototype has been developed and the preliminary experimental tests are presented, including the performance of the Au-coated THGEM working in At/CO2 gas mixtures and the neutron imaging test with 252 Cf source, which can provide the reference for experimental data for research in the future.

Comparison of thermal neutron detection ef f iciency of 6Li scintillation glass and 3He gas proportional tube

We report on a comparison study of the 3He gas proportional tube and the 6Li incorporated scintillation glasses on thermal neutron detection efficiency. Both 3He and 6Li are used commonly for thermal neutron detection because of their high neutron capture absorption coefficient. By using a neutron source 252Cf and a paraffin moderator in an alignment system, we can get a small beam of thermal neutrons. A flash ADC is used to measure the thermal neutron spectrum of each detector, and the detected number of events is determined from the spectrum, then we can calculate the detection efficiency of different detectors. Meanwhile, the experiment has been modeled with GEANT4 to validate the results against the Monte Carlo simulation.

Study of the RPC-Gd as thermal neutron detector

The BESⅢ ·RPC with Gd coating as thermal neutron detector was designed and constructed. Three prototypes were built with different techniques of producing the gadolinium converter. The performance of the cosmic ray test, the signal and the radiation spectrum were discussed in this paper. Lastly, the efficiency of one prototype with the best performance for detecting the thermal neutron was tested as 8.7%.

Visualization and quantification of water movement in porous cement-based materials by real time thermal neutron radiography:Theoretical analysis and experimental study

Water movement in porous cement-based materials is of great importance when studying their deterioration processes and durability.Many traditional methods based on mass changes,electricity or nuclear magnetic resonances are available for studying water transport in cement-based materials.In this research,an advanced technique i.e.thermal neutron radiography was utilized to achieve visualization and quantification of time dependent water movement including water penetration and moisture vapor in porous cement-based materials through theoretical analysis and experimental study.Because thermal neutrons ex-perience a strong attenuation by hydrogen,neutron radiography exhibits high sensitivity to small amounts of water.A neutron transmission analysis for quantitative evaluation of raw radiographic measurements was developed and optimized based on point scattered functions(PScF).The determinations of the real time and space dependent water penetration into uncracked and cracked mortar samples,as well as the drying process have been presented in this paper.It is illustrated that thermal neutron radiography can be a useful research tool for visualization and quantification of water movement in porous building materials.The obtained results will help us to better understand deteriorating processes of cement-based materials and to find ways to improve their durability.

Electrospun and in situ self-polymerization of polyacrylonitrile containing gadolinium nanofibers for thermal neutron protection

In this work, the polyacrylonitrile containing gadolinium nanofibers for thermal neutron protection were successfully fabricated by electrospunning and followed by in situ self-polymerization. Scanning electron microscopy(SEM) and energy-dispersive spectroscopy(EDS) results show that there are no beads on the smooth surface of the nanofibers and gadolinium elements are uniformly dispersed in the matrix. The thermal analysis and FTIR results prove that gadolinium methacrylate is induced in situ selfpolymerization during the heat treatment. The leaching rate of Gd^(3+) decreases from 79.97% to 10.74% tested by lowfield nuclear magnetic resonance(LF-NMR) method after the self-polymerization of gadolinium methacrylate in the matrix when the nanofibers were immersed in water for7 days. The thermal neutron shielding analysis calculated by MCNP program shows that above 99% thermal neutrons are absorbed when traveling through the 2-mm-thick polyacrylonitrile containing gadolinium nanofibers.

Parameters measurement for the thermal neutron beam in the thermal column hole of Xi'an pulse reactor

The distribution of the neutron spectra in the thermal column hole of Xi'an pulse reactor was measured with the time-of-flight method.Compared with the thermal Maxwellian theory neutron spectra,the thermal neutron spectra measured is a little softer,and the average neutron energy of the experimental spectra is about 0.042±0.01 eV.The thermal neutron fluence rate at the front end of thermal column hole,measured with gold foil activation techniques,is about 1.18×105 cm-2 s-1.The standard uncertainty of the measured thermal neutron fluence is about 3%.The spectra-averaged cross section of 197Au(n,γ) determined by the experimental thermal neutron spectra is(92.8±0.93) ×10-24 cm2.

Analysis of Neutronics and Thermal-Hydraulic Behavior in a Fuel Pin of Pressurized Water Reactor (PWR)

This paper presents a comparative analysis of different parameters such as enthalpy, moderator temperature, moderator density, flow velocity, pressure, and fuel temperature profile at the fuel pin cell level of PWR. Moreover, in this paper pitches to fuel pin radius ratio are varied from 2.3 to 4. The methods and implementation strategy are such that the coupled neutronic and thermal-hydraulic analysis is executed in a fully one dimensional (1D) manner. The thermal hydraulic is based on moderator/coolant mass and enthalpy equation together with one group diffusion equation for fuel pin. Modelling of fuel pin cell and subchannel is executed in two steps. First, the governing equations are derived assuming that all the parameters appearing in the equations are temperature independent. Fuel pin centerline temperature and radially averaged temperature equations are derived from Fourier laws of thermal conductivity. Finally, diffusion coefficient, fission cross-section and absorbing cross-section are evaluated with respect to the fuel pin temperature. The outcome will be helpful for further neutronics and thermal analysis of PWR. Thermal hydraulics parameter varies the maximum 30 percentage from the lowermost value.