A method for neutron-induced gamma spectra decomposition analysis based on Geant4 simulation

In various monitoring and detection tools that use pulsed neutron generators as radiation sources, the gamma rays induced by the interaction with various nuclei at different stages of neutron transport can reflect information about the medium. These gamma rays are generated in two major interactions: inelastic scattering of fast neutrons and radiative capture of thermal neutrons, corresponding to the inelastic and capture gamma rays,respectively. However, the two types of gamma rays that reflect...     

Re-estimation of absolute gamma ray intensities of ^56Mn using k0-standardization

The thermal neutron capture gamma ray facility at Pakistan Research Reactor （PARR-1） is being used for the re-estimation of various properties like capture cross-sections, resonance integral, absolute gamma intensities, etc. of different isotopes. The data for gamma ray transitions from the capture of thermal neutrons by ^55Mn are not in good agreement specifically below 2 MeV. So there is a need to re-estimate its intensities with better accuracy. Analytical grade MnCl2 powder and high purity Mn metal pieces were used in this study. Standard ^152Eu and ^60Co radioactive sources as well as thermal neutron capture γ-rays in chlorine were chosen for efficiency calibration. The k0 standardization technique was applied for these measurements to eliminate systematic errors in efficiencies. Chlorine also acted as a comparator in k0-factor calculations. The results have been tabulated for the main gamma rays from ^56Mn in the low as well as in the medium energy regions. The absolute intensities are in good agreement with most of the reported values.     

Design of Neutron Absorber for Gamma Total Absorption Facility

正Gamma Total Absorption Facility (GTAF)will be used to accurately measure the cross section of neutron capture by prompt gamma meth-od.As shown in Fig.1,cascade gamma rays emitted by the compound nucleus of capture reac-     

Development of Prototype Neutron Flux Monitor for ITER

The prototype neutron flux monitor consists of a high purity 235U fission chamber detector and a “blank” detector, which is a fissile material free detector with the same dimension as the fission chamber detector to identify noise issues such as noise coming from gamma rays. The main parameters of the fission chamber assembly that have been measured in the laboratory are confirmed to approach the technological level of the International Thermonuclear Experimental Reactor (ITER) in the near future. This prototype neutron flux monitor will be further developed to become a neutron flux monitor suitable for the operation phase of D-D fusion on the ITER.     

Analyzing the transient effects of ~(60)Co gamma rays in a CIS by Monte Carlo method

The objective of this work is to analyze the transient effects of ~(60)Co gamma rays in the CMOS image sensor(CIS) using the Monte Carlo method, based on Geant4. The track, energy spectrum, and angle of produced electrons when gamma rays traversed a silicon or silicon dioxide cube were calculated. A simplified model of a 500 × 500 CIS array was established, and the transient effects of gamma rays in the CIS were simulated. The raw images were captured when the CIS was irradiated by gamma rays. The experimental results were compared with the simulation results. The characteristics of the typical events induced by transient effects were analyzed.     

Compact lithium-glass neutron beam monitor for SANS at CSNS

A small-angle scattering neutron spectrometer for material research is under construction at the China Spallation Neutron Source. An intervening neutron beam monitor behind the sample is needed to measure the beam intensity in order to reduce the measurement uncertainty caused by beam fluctuation. Considering the mobility requirement and limited space, we proposed a compact monitor using a type of lithium-glass scintillator provided by China Building Materials Academy. Its performance was studied experimentally using ~(252)Cf and ~(60)Co sources.The neutron light yield of the selected scintillator was measured to be 5:3 × 10~3 photons/neutron. The feasibility of n-gamma discrimination using the charge comparison method was verified. By using the Geant4 toolkit, themonitor was modeled with precise physical processes including neutron tracking, scintillation, and optical photon transmission. The gamma sensitivity and detection efficiency were investigated in the simulation. It was concluded that a 0.5-mm-thick lithium-glass scintillator with natural lithium is an appropriate choice to satisfy both the neutron detection efficiency and gamma elimination requirements.     

A study of PFBR auxiliary neutron source strength activation and its variability with respect to the neutron spectrum and ~(123)Sb capture cross section

In fast reactors, the inherent neutron source strength is often insufficient for monitoring the reactor start-up operation with ex-core detectors. To increase the subcritical neutron flux, an auxiliary neutron source subassembly(SSA) is generally used to overcome this problem. In this study, the estimated neutron source strength and detector count rate of an antimony-beryllium-based SSA are obtained using the deterministic transport code DORT and Monte Carlo calculations. Because the antimony activation rate is a critical parameter, its sensitivity to the capture cross section and neutron flux spectrum is studied. The reaction cross section sensitivity is studied by considering data from different evaluated nuclear data files.It is observed that, because of the variation in the cross sections from different evaluated nuclear data files, the values of the saturation gamma( 1.67 MeV) activity and neutron strength predicted by ORIGEN2 lie within ±2%.The obtained antimony activation rate and sensitivity to the neutron flux are partially validated by irradiating samples of antimony in the KAMINI reactor. The average onegroup capture cross sections of bare and cadmium-covered ~(123)Sb samples obtained by the ratio method are 4.0 and 1.78 b, respectively. The results of the calculation predicting the activated neutron source strength as a function of operating time and sensitivity to the neutron spectrum in the irradiation region are also presented.     

Measurements of fast—neutron capture cross sections for ^159Tb and ^169Tm

The neutron capture cross sections for ^159Tb and ^169Tm relative to the ^197Au (n,γ)^198Au reaction are measured at neutron energies of 0.57,1.10 and 1.60 MeV by using the activation method.The activities of the products are measured with a high resolution HPGe detector gamma-ray spectrometer.The errors of the present work are 5-6% for Tb,6-7% for Tm.The recommended data in energy region of 0.4-3.0MeV are given as compared with other data published previously.     

Coal analysis using the pulsed neutron generator

A prototype of elemental analyzer for coal has been developed by using a PFTNA (pulse fast thermal neutron analysis) system. The PFTNA technology is based on the reactions such as (n, γ), (n, n'γ), (n, pγ), etc. by examining the characteristic gamma rays emitted. In our prototype a pulsed neutron generator provides 14 MeV pulse neutrons, which contribute to the separation of spectrum II (the sum of capture and activation spectrum) from spectrum I (the sum of inelastic, capture and activation spectrum), and thus to the measurement of C and O contents in coal. Data management is completed by computer program using the least-square regression method. The experiment in Changshan Power Plant for 3 months showed that the precision of calorific value, whole water, volatile content and ash content is 0.5 kJ/kg, 1.0 wt%, 2.0 wt% and 1.5 wt%, respectively.     

MEASUREMENT OF PROMPT NEUTRON SPECTRA OF ^238U FISSION INDUCED BY 10.17 AND 12.12 MeV NEUTRONS

Experimental method to measure the prompt neutron spectra of 238U fission induced by fast neutrons has been developed at HI-13 Tandem Van de Graaff Accelerator Laboratory of CIAE.These techniques employ a multi-segment fission chamber and two liquid scintillator neutron detectors.TOF(time of flight)techniques are used for primary neutrons to select the fission events induced by monoenergetic neutron from 2H(d,n) reactions instead of breakup neutrons from 2H(d,np) reactions.The fission neutron TOF spectra are measured in coincidence with the fission fragments to distinguish fission neutrons from other secondary neutrons.The method permits measurements to a fairly good accuracy under large neutron and gamma ray background.The techniques are described and experimental spectra are presented.     

Design of a shielding collimator device for a small-angle monoenergetic neutron source

To obtain a small-angle monoenergetic neutron source,a shielding collimator device is designed for the neutron source generated by a neutron tube.The device is divided into the collimator and the capture cave.The collimator is made of three layers of stainless steel and borated polyethylene and is used to constrain neutrons in a small angle.The capture cave is used to increase the number of times neutron inelastic scattering occurs in the opposite direction of the radiation field,thereby reducing the proportion of scattered neutrons in the radiation field.Material thickness,aperture size,and the optimum structure of the capture cave were simulated using MCNP.The design features a neutron emission angle within a range of 3° and neutron fluxes in the radiation field,which are higher by two orders of magnitude than those outside the radiation field.This research has practical value for the generation of monoenergetic small-angle neutron sources and neutron applications.     

Neutron Radioactive Capture Cross Section Calculations of Natural Zr in the Incident Energy Region from 0.01 to 20 MeV

The neutron radiate capture cross sections of natural Zr are calculated in the neutron incident energy region from 0.01 MeV to 20 MeV. In the calculations, photon transmission coefficients of the compound statistical processes are calculated by solving the cascade gamma deexcitation processes of the compound nucleus. The nonstatistical effects of the neutron radiate captures, the radiate captures in the shape elastic scattering channels and the compound elastic scattering channels as well as the direct-semidirect captures are considered carefully in the calculated energy region. The calculated results are in better agreement with the experimental values.     

Measurement of the prompt neutron spectrum from thermalneutron-induced fission in U-235 using the recoil proton method

A measurement of the ~(235)U prompt fission neutron spectrum(PFNS) by the recoil proton method was performed at the Institute of Nuclear Physics and Chemistry, China. Details of the method, which include the calculation and validation of the response matrix, are presented. The PFNS for ~(235)U in the energy range 1–12 MeV,induced by thermal neutrons, was obtained. The measured spectrum in the low-energy region was in good agreement with previous work and the ENDF/B-VII library, except for minor differences. In the high-energy region, however, the relative height of the measured spectrum was greater, and an analysis of the experiment indicated uncertainties of 13% at 10 MeV and 24% at 12 MeV. Experimental results showed that the recoil proton method could be used to measure prompt fission neutron spectra. Some directions for future work are included.     

Optimization of moderator assembly for neutron flux measurement: experimental and theoretical approaches

A moderator of paraffin wax assembly has been demonstrated where its thickness can be optimized to thermalize fast neutrons. The assembly is used for measuring fast neutron flux of a neutron probe at different neutron energies, using BF03(U10and 200) and3He(U0.500)neutron detectors. The paraffin wax thickness was optimized at 6 cm for the neutron probe which contains an Am–Be neutron source. The experimental data are compared with Monte Carlo simulation results using MCNP5 version 1.4. Neutron flux comparison and neutron activation techniques are used for measuring neutron flux of the neutron probe to validate the optimum paraffin moderator thickness in the assembly. The neutron fluxes are measured at(1.17 ± 0.09) 9 105 and(1.19 ± 0.1) 9 105n/s, being in agreement with the simulated values. The moderator assembly can easily be utilized for essential requirements of neutron flux measurements.     

Experimental Studies on the Self-Shielding Effect in Fissile Fuel Breeding Measurement in Thorium Oxide Pellets Irradiated with 14 MeV Neutrons

The 14 MeV neutrons produced in the D-T fusion reactions have the potential of breeding Uranium-233 fissile fuel from fertile material Thorium-232. In order to estimate the amount of U-233 produced, experiments are carried out by irradiating thorium dioxide pellets with neutrons produced from a 14 MeV neutron generator. The objective of the present work is to measure the reaction rates of 232 Th + 1 n → 233 Th → 233 Pa → 233 U in different pellet thicknesses to study the self-shielding effects and adopt a procedure for correction. An appropriate assembly consisting of high-density polyethylene is designed and fabricated to slow down the high-energy neutrons, in which Thorium pellets are irradiated. The amount of fissile fuel ( 233 U) produced is estimated by measuring the 312 keV gammas emitted by Protactinium-233 (half-life of 27 days). A calibrated High Purity Germanium (HPGe) detector is used to measure the gamma ray spectrum. The amount of 233 U produced by Th 232 (n, γ) is calculated using MCNP code. The self-shielding effect is evaluated by calculating the reaction rates for different foil thickness. MCNP calculation results are compared with the experimental values and appropriate correction factors are estimated for self-shielding of neutrons and absorption of gamma rays.     

Diagnostic Technique of Neutron Doppler Broadening for Indirectly-Driven Implosions

The neutron Doppler broadening in inertial confinement fusion has been acquired from the time of flight for the neutron, from which the fuel ion temperature can be derived. An ultrafast-quenched plastic scintillation detector was used to measure the time of flight for the neutron at a low-imploded DT neutron yield (5×107-1×108) in the experiment performed on the Shenguang Ⅱ laser facility. The typical temperatures of ablating targets for indirect drive were around 2.8 keV and the uncertainties were ±30 % - ±40%. The detection efficiency of the detector for DT neutrons was calibrated at a K-400 accelerator. The time response function of the detection system was calibrated by imploded neutrons from a DT-filled capsule, which can be regarded as a S function pulsed neutron source due to its much narrower pulse width than that of the measured neutron time-of-flight spectrum.     

Yield ratio of neutrons to protons in ~(12)C(d,n)~(13)N and ~(12)C(d,p)~(13)C from 0.6 to 3 MeV

The neutron yield in the~(12)C(d,n)~(13)N reaction and the proton yield in the~(12)C(d,p)~(13)C reaction have been measured using deuteron beams of energies 0.6-3 MeV.The deuteron beam is delivered from a 4-MeV electrostatic accelerator and bombarded on a thick carbon target.The neutrons are detected at 0°,24°,and 48°and the protons at135°in the laboratory frame.Further,the ratio of the neutron yield to the proton yield was calculated.This can be used to effectively recognize the resonances.The resonances are found at 1.4 MeV,1.7 MeV,and 2.5 MeV in the~(12)C(d,p)~(13)C reaction,and at 1.6 MeV and 2.7 MeV in the~(12)C(d,n)~(13)N reaction.The proposed method provides a way to reduce systematic uncertainty and helps confirm more resonances in compound nuclei.     

Gamma ray multiplicity of a 240Pu solid sphere simulated by JMCT

Determining the mass of plutonium metal is an important research objective in the field of nuclear material accounting and control. Based on the 3D neutron and photon transport code JMCT（Jointed Monte Carlo Transport）, the gamma ray multiplicity of²⁴⁰Pu was simulated in this study, and the average number of gamma rays leaking from²⁴⁰Pu solid spheres with different masses was also obtained. The simulation results show that there is a oneto-one correspondence between the aver...     

Gamma irradiation-induced effects on the properties of TiO_2 on fluorine-doped tin oxide prepared by atomic layer deposition

The effect of gamma irradiation with different doses(25–75 kGy) on TiO_2 thin films deposited by atomic layer deposition has been studied and characterized by X-ray diffraction(XRD),photoluminescence measurements,ultraviolet–visible(UV–Vis) spectroscopy,and impedance measurements.The XRD results for the TiO_2 films indicate an enhancement of crystallization after irradiation,which can be clearly observed from the increase in the peak intensities upon increasing the gamma irradiation doses.The UV–Vis spectra demonstrate a decrease in transmittance,and the band gap of the TiO_2 thin films decreases with an increase in the gamma irradiation doses.The Nyquist plots reveal that the overall charge-transfer resistance increases upon increasing the gamma irradiation doses.The equivalent circuit,series resistance,contact resistance,and interface capacitance are measured by simulation using Z-view software.The present work demonstrates that gamma irradiation-induced defects play a major role in the modification of thestructural,electrical,and optical properties of the TiO_2 thin films.     

Tokamak fusion neutron spectrometer based on PXI bus

In order to realize on-line real-time measurement of dynamic and time-sharing neutron spectrum of HL-2A,a tokamak fusion neutron spectrometer based on PXI bus was developed.It consists of electronics system and eight thermal neutron detectors,namely SP9 3He proportional counter,embedded in eight polyethylene spheres in different diameters.Response function of the eight polyethylene spheres was the key to calculate the neutron spectrum accurately.In this paper,response function of the eight polyethylene spheres is simulated by adopting Geant4 code,and neutron counts from an 241Am-Be neutron source are measured by the eight detectors.The calculated spectrum of the Am-Be neutron is accurate in 0-2 MeV region,and is similar to the theoretical spectrum.The tokamak fusion neutron spectrometer was used in HL-2A device to monitor the dynamic neutron spectrum of HL-2A on-line and real-time.