252Cf Neutron Source Research Articles

A new neutron detector based on ceramic THGEM and boron-coated meshes

Boron-coated micro-pattern gaseous detectors, such as gaseous electron multiplier (GEM), are widely studied as the neutron detectors. The existing solution to improve the neutron detection efficiency (NDE) is to coat the boron on the surfaces of the cascade GEM films. So the GEM films themselves will be affected by the complicated coating process and the boron layers, and the $$^{10}$$ B material will be wasted once the GEM films are damaged. Since the neutron conversion and the electron multiplication regions are not separated, the final gains of different layer ionized electrons are very difficult to control and keep consistent. A new neutron detector based on the ceramic thick GEM (THGEM) was proposed. The boron-coated cathode and multilayer meshes were used as the neutron conversion region and separated from the electron multiplication region consisting of only one THGEM film. The sawtooth surface treatment and the double-sided readout method were applied to improve neutron conversion efficiency (NCE). The neutron conversion and the electron collection efficiencies were simulated and studied based on the new structure. The prototype detector was test by using a $$^{252}$$ Cf neutron source. The simulation results show that the cascade boron-coated meshes have good neutron conversion and electron collection efficiencies, and the double-sided readout and sawtooth structure can greatly increase the NCE. The neutron test results show that the detection efficiency of the double-sided readout is 1.6 times higher than that of the common single-sided readout. These indicate that the new detector has high neutron conversion and detection efficiencies and is promising for practical applications.

Calibration experiment and the neutronics analyses on the LHD neutron flux monitors for the deuterium plasma experiment

Prior to the deuterium experiments of the Large Helical Device (LHD), an in situ neutron calibration experiment was carried out by using an 800 MBq 252Cf neutron source in November 2016, where a neutron flux monitor, a neutron activation system and a vertical neutron camera are calibrated. This paper concentrates on the in situ calibration for the neutron flux monitor. The neutron flux monitor based on 235U fission chambers and 10B and 3He proportional counters are positioned at three locations outside the LHD cryostat. The counts are measured during continuous rotation of the 252Cf neutron source on the typical magnetic axis of the plasma by a toy train, where the neutron source can be regarded as a ring neutron source. Detection efficiencies for the 252Cf ring source are derived by (total counts)/(total emitted neutron number) in the continuous rotation of the neutron source. Detection efficiencies for the real plasma neutron source are obtained from the measured detection efficiencies with the correction factor taking account of the deference between the 252Cf ring source and the real plasma neutron source evaluated by MCNP calculations with the three-dimensional model of the LHD. Finally, the uncertainty of the total neutron emission rate measurement is discussed.

Nonnegative Tensor Factorization Approach Applied to Fission Chamber’s Output Signals Blind Source Separation

Inside nuclear reactors, gamma-rays emitted from nuclei together with the neutrons introduce unwanted backgrounds in neutron spectra. For this reason, powerful extraction methods are needed to extract useful neutron signal from recorded mixture and thus to obtain clearer neutron flux spectrum. Actually, several techniques have been developed to discriminate between neutrons and gamma-rays in a mixed radiation field. Most of these techniques, tackle using analogue discrimination methods. Others propose to use some organic scintillators to achieve the discrimination task. Recently, systems based on digital signal processors are commercially available to replace the analog systems. As alternative to these systems, we aim in this work to verify the feasibility of using a Nonnegative Tensor Factorization (NTF) to blind extract neutron component from mixture signals recorded at the output of fission chamber (WL-7657). This last have been simulated through the Geant4 linked to Garfield++ using a 252Cf neutron source. To achieve our objective of obtaining the best possible neutron-gamma discrimination, we have applied the two different NTF algorithms, which have been found to be the best methods that allow us to analyse this kind of nuclear data.

Evaluation of a new neutron energy spectrum unfolding code based on an Adaptive Neuro-Fuzzy Inference System (ANFIS).

The purpose of the present study was to reconstruct the energy spectrum of a poly-energetic neutron source using an algorithm developed based on an Adaptive Neuro-Fuzzy Inference System (ANFIS). ANFIS is a kind of artificial neural network based on the Takagi–Sugeno fuzzy inference system. The ANFIS algorithm uses the advantages of both fuzzy inference systems and artificial neural networks to improve the effectiveness of algorithms in various applications such as modeling, control and classification. The neutron pulse height distributions used as input data in the training procedure for the ANFIS algorithm were obtained from the simulations performed by MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). Taking into account the normalization condition of each energy spectrum, 4300 neutron energy spectra were generated randomly. (The value in each bin was generated randomly, and finally a normalization of each generated energy spectrum was performed). The randomly generated neutron energy spectra were considered as output data of the developed ANFIS computational code in the training step. To calculate the neutron energy spectrum using conventional methods, an inverse problem with an approximately singular response matrix (with the determinant of the matrix close to zero) should be solved. The solution of the inverse problem using the conventional methods unfold neutron energy spectrum with low accuracy. Application of the iterative algorithms in the solution of such a problem, or utilizing the intelligent algorithms (in which there is no need to solve the problem), is usually preferred for unfolding of the energy spectrum. Therefore, the main reason for development of intelligent algorithms like ANFIS for unfolding of neutron energy spectra is to avoid solving the inverse problem. In the present study, the unfolded neutron energy spectra of 252Cf and 241Am-9Be neutron sources using the developed computational code were found to have excellent agreement with the reference data. Also, the unfolded energy spectra of the neutron sources as obtained using ANFIS were more accurate than the results reported from calculations performed using artificial neural networks in previously published papers.

Validation and evaluation of the ADVANTG hybrid code on the ICSBEP labyrinth benchmark experiment

Validation and evaluation of the recently released ADVANTG code is presented in the paper. ADVANTG is a hybrid transport code, which combines a deterministic neutron transport code Denovo with MCNP, a well-known Monte Carlo (MC) transport code. ADVANTG was tested on an ICSBEP labyrinth benchmark where neutron fluxes were measured and calculated in a three-section concrete labyrinth for a 252Cf neutron source. The speed-up factors or the increases in relative efficiency of up to 1400 compared to analog MCNP simulations were achieved using the ADVANTG-generated variance reduction parameters. ADVANTG accelerated the calculations in terms of the CPU time and the calculated values were in good agreement with the experimental values. Additionally, it was shown that no bias was introduced by the ADVANTG code, as the mean values obtained by the ADVANTG accelerated calculation are within statistical uncertainties of the analog calculation. The presented work has shown that ADVANTG is a useful tool for use in solving neutron transport problems in geometries featuring labyrinth.

Development of high-pressure multi-wire proportional chamber neutron detector for the China Spallation Neutron Source multipurpose reflectometer

As a multipurpose reflectometer device, the two-dimensional (2D) position resolution neutron detector with a 200 mm×200 mm effective area is developed for China Spallation Neutron Source (CSNS) in Dongguang, China. Due to the requirements for the specific parameters of the multipurpose reflectometer, it should be designed to have a more than 50% (@2 Å) detection efficiency, better than 2 mm position resolution and 3 times n/γ resolution ability during the whole operation period of 10 years. The high pressure multi-wire proportional chamber (MWPC) neutron detector filling 3He gas is used as a key detector. Some simulation results and the experimental results show that the optimized thickness of the neutron entrance window should be 9 mm with using the 7075 aluminum alloy, the high pressure chamber should be sealed by the aluminous ring and a gas mixture should be filled with 6 bar 3He+2.5 bar C3H8. The assembled detector can achieve a more than 54% (@2 Å) detection efficiency in the normal operation.With the 100 μm wide collimator slit, the position resolution for X-rays is about 0.235 mm. Therefore, the position resolution for neutron is about 1.4 mm when 2.5 bar propane is used as the stopping gas for proton and triton. In the chamber, the water vapor, the oxygen and the organic impurity gases will reduce the gas gain, cause the detector electrodes to break down and the detector to speed up aging. To solve the outgassing effect of the detector components and keep the stable operation, the recycled device is designed to have the purification function for the working gases. It could purify the working gas at a flow rate of 2 L/min to remove the oxygen, the water vapor and the organic impurity gases. The detector gain increases about 27% with the purification function. Finally, the n/γ resolution and 2D imaging ability of the detector are tested with the 252Cf neutron source in Institute of High Energy Physics, Chinese Academy of Sciences, the peak ratio of neutron to gamma is obtained to be above 5 from the energy spectra and the detecor has good 2D imaging ability. The performance of the high-pressure MWPC neutron detector could meet the requirements for the multipurpose reflectometer, and the detector will be mounted in the CSNS in this year.

Assembly Neutron Moderation System for BNCT Based on a 252Cf Neutron Source

Assembly Neutron Moderation System for BNCT Based on a ²⁵²Cf Neutron Source

Development of a bump cathode element for two-dimensional neutron detection

A neutron-detection element using circular cathode bumps was developed for two-dimensional neutron measurement, and an irradiation experiment was performed using a Cf-252 neutron source. The element has triangularly arranged small-sized circular cathode bumps, and the bumps are linked together in the x- and y-directions for the detection of incident neutrons. Preliminary irradiation experiments for the developed element were performed using a neutron-detection system consisting of a pressure vessel, amplifier-shaper-discriminator boards, optical signal transmission devices, position encoders, and a data acquisition device. The element was arranged in the pressure vessel with a fill gas of composition He/(15%)CF4 at 0.7 MPa. The detector system exhibited a two-dimensional uniformity with a response of 10.1%. The average intrinsic spatial resolution was 1.89 mm full width at half maximum in the sensitive region, calculated by taking into account the track lengths of the secondary particles.

In situ calibration of neutron activation system on the large helical device.

In situ calibration of the neutron activation system on the Large Helical Device (LHD) was performed by using an intense 252Cf neutron source. To simulate a ring-shaped neutron source, we installed a railway inside the LHD vacuum vessel and made a train loaded with the 252Cf source run along a typical magnetic axis position. Three activation capsules loaded with thirty pieces of indium foils stacked with total mass of approximately 18 g were prepared. Each capsule was irradiated over 15 h while the train was circulating. The activation response coefficient (9.4 ± 1.2) × 10-8 of 115In(n, n')115mIn reaction obtained from the experiment is in good agreement with results from three-dimensional neutron transport calculations using the Monte Carlo neutron transport simulation code 6. The activation response coefficients of 2.45 MeV birth neutron and secondary 14.1 MeV neutron from deuterium plasma were evaluated from the activation response coefficient obtained in this calibration experiment with results from three-dimensional neutron calculations using the Monte Carlo neutron transport simulation code 6.

Monte Carlo based study of radiation field in a deep geological repository for high-level nuclear waste with different host rock types

Disposal in deep geological formations with arrangements for monitoring and retrieval has been considered as one of the most promising management concepts of heat-generating, high-level nuclear waste. Typical design of a deep geological repository requires a shielding cask containing the waste, which is in turn located underground in an emplacement drift of a host rock formation. Various rock types, such as rock salt, clay stone and granite, are possible candidates as host rock for a deep geological repository. The massive host rock formation around the emplacement drift is the most important barrier against radionuclide release into the biosphere. However, the host rock layers have also influence on the radiation field in the emplacement drift, which should be taken into account when assessing occupational exposure in the repository. In the current study, impact of two different materials, i.e. rock salt and concrete (building material of the supporting liner for a drift in clay stone or granite), on radiation field around a nuclear waste package disposed in an emplacement drift was investigated with Monte Carlo method. The high-level nuclear waste was simulated with a 252Cf neutron source. Both neutron and gamma dose rates and spectra around the waste package were calculated. It was found out that neutrons dominate the radiation field in the drift and the dose rates in the drift are enhanced due to backscattered radiation by the surrounding material layers. Furthermore, due to different material compositions of rock salt and concrete, the resulted neutron spectra have also different characteristics. In general, concrete moderates neutrons better than rock salt, which leads to a lower dose rates in the emplacement drift.

Design of a novel nondestructive portable mobile neutron activation system

A new non-destructive portable mobile system was designed to be used for neutron activation assay according to the Safeguards purposes. The sample is irradiated in a neutron field and its composition is determined by identifying the characteristic induced gamma radiation emitted by the fission and activation products. For the design of a neutron activation system, many aspects must be considered such as choosing the best materials to assure the protection against radiation and estimating the thickness required to shield the neutron source based on its flux or strength. The polyethylene and paraffin wax were selected as neutron shielding material, lead being chosen as gamma shielding material. Several calculations were performed using the Monte Carlo code MCNP5 to estimate the required shield thickness and distribution of thermal neutron flux at different irradiation position inside the shield. The available 252Cf neutron source with 12µCi of activity was used. The novel neutron activation analysis system was proposed, designed and fabricated to meet the maximum flexibility for neutron activation analysis, at minimum cost and with adequate optimized shielding and portability.

Thin CVD diamond film detector for slow neutrons with buried graphitic electrode

We fabricated neutron detectors based on thin epitaxial CVD diamond films grown on single crystal diamond substrates, with a buried graphitic electrode produced by ion implantation and annealing. The spectra of 5.5 MeV alpha particles were measured with ≈3.5% energy resolution, and, based on the features of those spectra, a precise assessment of the film thickness was able achieved. A 10 μm thick detector was tested for 252Cf neutron and γ-radiation source, using 10B and 6Li isotopes as converters. Due to the small film thickness the contribution from the γ-quanta background to the spectra was restricted by energies below 200 keV, and could be easily discriminated from the neutron related signals. The performance of the detectors was tested for 5.5 MeV α-particles at fluences up to 109cm−2, and no detrimental polarization effect was observed for the best sample.

Feasibility study for wax deposition imaging in oil pipelines by PGNAA technique

Wax deposition in pipelines is a crucial problem in the oil industry. A method based on the prompt gamma-ray neutron activation analysis technique was applied to reconstruct the image of wax deposition in oil pipelines. The 2.223MeV hydrogen capture gamma rays were used to reconstruct the wax deposition image. To validate the method, both MCNP simulation and experiments were performed for wax deposited with a maximum thickness of 20cm. The performance of the method was simulated using the MCNP code. The experiment was conducted with a 252Cf neutron source and a LaBr3: Ce detector. A good correspondence between the simulations and the experiments was observed. The results obtained indicate that the present approach is efficient for wax deposition imaging in oil pipelines.

Pulse shape discrimination characteristics of stilbene crystal, pure and 6Li loaded plastic scintillators for a high resolution coded-aperture neutron imager

Pulse shape discrimination performances of single stilbene crystal, pure plastic and 6Li loaded plastic scintillators have been compared. Three pulse shape discrimination algorithms have been tested for each scintillator sample, assessing their quality of neutron/gamma separation. Additionally, the digital implementation feasibility of each algorithm in a real-time embedded system was evaluated. Considering the pixelated architecture of the coded-aperture imaging system, a reliable method of simultaneous multi-channel neutron/gamma discrimination was sought, accounting for the short data analysis window available for each individual channel. In this study, each scintillator sample was irradiated with a 252Cf neutron source and a bespoke digitiser system was used to collect the data allowing detailed offline examination of the sampled pulses. The figure-of-merit was utilised to compare the discrimination quality of the collected events with respect to various discrimination algorithms. Single stilbene crystal presents superior neutron/gamma separation performance when compared to the plastic scintillator samples.

The design and fabricate a pulse shape discriminator apply signal processing method using EJ-301 detector

The high quality measurements of neutron energy spectra are required in various fields of research and applications. However, in many cases the contribution of gamma background causes the inaccuracy of neutron spectrum. Therefore, the discrimination of gamma-ray events in neutron spectrum is necessary. In this article, an algorithm for digital implementation of the chargecomparison method for n/γ discrimination based on Digital Signal Processing technique is described. Furthermore, the APX-500 board was used as a hardware for the development of a Pulse Shape Disciminator, and is equipped with ADC ADM-414 14 bit-100 MSPS. The fully system has been tested with EJ-301 detector, using 252Cf neutron source.

First Demonstration of a Scintillating Xenon Bubble Chamber for Detecting Dark Matter and Coherent Elastic Neutrino-Nucleus Scattering.

A 30-g xenon bubble chamber, operated at Northwestern University in June and November 2016, has for the first time observed simultaneous bubble nucleation and scintillation by nuclear recoils in a superheated liquid. This chamber is instrumented with a CCD camera for near-IR bubble imaging, a solar-blind photomultiplier tube to detect 175-nm xenon scintillation light, and a piezoelectric acoustic transducer to detect the ultrasonic emission from a growing bubble. The time of nucleation determined from the acoustic signal is used to correlate specific scintillation pulses with bubble-nucleating events. We report on data from this chamber for thermodynamic "Seitz" thresholds from 4.2 to 15.0keV. The observed single- and multiple-bubble rates when exposed to a ^{252}Cf neutron source indicate that, for an 8.3-keV thermodynamic threshold, the minimum nuclear recoil energy required to nucleate a bubble is 19±6 keV (1σ uncertainty). This is consistent with the observed scintillation spectrum for bubble-nucleating events. We see no evidence for bubble nucleation by gamma rays at any of the thresholds studied, setting a 90%C.L. upper limit of 6.3×10^{-7} bubbles per gamma interaction at a 4.2-keV thermodynamic threshold. This indicates stronger gamma discrimination than in CF_{3}I bubble chambers, supporting the hypothesis that scintillation production suppresses bubble nucleation by electron recoils, while nuclear recoils nucleate bubbles as usual. These measurements establish the noble-liquid bubble chamber as a promising new technology for the detection of weakly interacting massive particle dark matter and coherent elastic neutrino-nucleus scattering.

Process effects on leakage current of Si‐PIN neutron detectors with porous microstructure

Using the technique of Microfabrication, such as deep silicon dry etching, lithography, etc. Si‐PIN neutron detectors with porous microstructure have been successfully fabricated. In order to lower the leakage current, the key fabrication processes, including the Al windows opening, deep silicon etching and the porous side wall smoothing, have been optimized. The cross‐section morphology and current–voltage characteristics have been measured to evaluate the microfabrication processes. With the optimized conditions presented by the measurements, a neutron detector with a leakage current density of 2.67 μA cm−2 at a bias of −20 V is obtained. A preliminary neutron irradiation test with 252Cf neutron source has also been carried out. The neutron irradiation test shows that the neutron detection efficiency of the microstructured neutron detectors is almost 3.6 times higher than that of the planar ones.

Neutron detection using a water Cherenkov detector with pure water and a single PMT

We present the performance of a novel neutron detector based on a water Cherenkov detector (WCD) employing pure water and a single photomultiplier tube (PMT). The experiments presented in this work were performed using 241AmBe and 252Cf neutron sources in different neutron moderator and shielding configurations. We show that fast neutrons from the 241AmBe and 241Cf sources, as well as thermal neutrons from a neutron moderator, despite having different spectral characteristics, produce essentially the same pulse histogram shape. This characteristic pulse-height histogram shapes are recorded as a clear signature of neutrons with energies lower than ≃11MeV. This is verified in different experimental conditions. Our estimation of the neutron detection efficiency is at the level of (15±5)%, for fast neutrons. Since water is the material employed as active volume, the results of this study are of interest for the construction of low cost and large active volume neutron detectors for various applications. Of special importance are those related with space weather phenomena monitoring as well as those for the detection of fissile special nuclear material, including uranium or plutonium.

Stabilization of Particle Discrimination Efficiencies for Neutron Spectrum Unfolding With Organic Scintillators

The literature discussing pulse-shape discrimination (PSD) in organic scintillators dates back several decades. However, little has been written about PSD techniques that are optimized for neutron spectrum unfolding. Variation in $n$ - $\gamma $ misclassification rates and in $\gamma /n$ ratio of incident fields can distort the neutron pulse-height response of scintillators and these distortions can in turn cause large errors in unfolded spectra. New applications in arms-control verification call for detection of lower-energy neutrons, for which PSD is particularly problematic. In this article, we propose techniques for removing distortions on pulse-height response that result from the merging of PSD distributions in the low-pulse-height region. These techniques take advantage of the repeatable shapes of PSD distributions that are governed by the counting statistics of scintillation-photon populations. We validate the proposed techniques using accelerator-based time-of-flight measurements and then demonstrate them by unfolding the Watt spectrum from measurement with a 252Cf neutron source.

Energy spectra unfolding of fast neutron sources using the group method of data handling and decision tree algorithms

Accurate unfolding of the energy spectrum of a neutron source gives important information about unknown neutron sources. The obtained information is useful in many areas like nuclear safeguards, nuclear nonproliferation, and homeland security. In the present study, the energy spectrum of a poly-energetic fast neutron source is reconstructed using the developed computational codes based on the Group Method of Data Handling (GMDH) and Decision Tree (DT) algorithms. The neutron pulse height distribution (neutron response function) in the considered NE-213 liquid organic scintillator has been simulated using the developed MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). The developed computational codes based on the GMDH and DT algorithms use some data for training, testing and validation steps. In order to prepare the required data, 4000 randomly generated energy spectra distributed over 52 bins are used. The randomly generated energy spectra and the simulated neutron pulse height distributions by MCNPX-ESUT for each energy spectrum are used as the output and input data. Since there is no need to solve the inverse problem with an ill-conditioned response matrix, the unfolded energy spectrum has the highest accuracy. The 241Am-9Be and 252Cf neutron sources are used in the validation step of the calculation. The unfolded energy spectra for the used fast neutron sources have an excellent agreement with the reference ones. Also, the accuracy of the unfolded energy spectra obtained using the GMDH is slightly better than those obtained from the DT. The results obtained in the present study have good accuracy in comparison with the previously published paper based on the logsig and tansig transfer functions.