Pu-Be Neutron Source Research Articles

Thermal and Fast Neutron Measurement in the STU Mini Labyrinth Experiment

The Mini Labyrinth experiment is a neutron and gamma shielding experiment constructed at the Slovak University of Technology in Bratislava (STU). The STU Mini Labyrinth consists of NEUTRONSTOP shielding, blocks of moderators, various neutron sources, a graphite prism, and the detector handling robot. It was designed for research and education purposes, while several experiments are also available online or a hybrid form. There have been several versions of the Mini Labyrinth developed, while currently the V3 version is available. This paper presents the methodology to effectively perform thermal and fast neutron measurements using a PuBe neutron source. In the experiment presented in this paper also active and passive detectors were used, and moderator materials were investigated to slightly shift the neutron spectrum. As for active detectors, the SNM-11 boron coated corona detector was used. Among passive detectors the CR-39 track detectors were investigated. The measurements were carried out in two configurations and the results were evaluated by simulations using the SCALE6 system

Perovskite-based detector for reactor dosimetry monitoring

Hybrid halide perovskites have demonstrated significant efficiency in detecting a broad spectrum of high-energy radiation, including X-rays, gamma rays (γ-rays), and neutrons. Given the common occurrence of mixed radiation fields, we investigated the performance of a perovskite-based detector in a neutron-gamma mixed field. A large methylammonium lead tribromide (MAPbBr3) single crystal (SC) was synthesized via the oriented crystal-crystal intergrowth method. This SC was used to fabricate a gamma detector with carbon electrodes, which was tested in the CROCUS zero-power reactor cavity. The detector’s photocurrent response exhibited a strong correlation with known gamma dose rates, as measured by an ambient Berthold LB 112 gamma probe, facilitating the accurate conversion of photocurrent to dose rate. Notably, the device did not exhibit degradation under neutron radiation exposure. To further assess the impact of neutrons, X-ray diffraction and electron paramagnetic resonance analyses were performed on small MAPbBr3 SCs grown by inverse temperature crystallization. These SCs were irradiated within the CROCUS reactor core and by a Pu-Be neutron source at liquid nitrogen temperature. Our findings indicate that the perovskite material can withstand the nominal in-core operation conditions of the CROCUS reactor. Additionally, it endures irradiation at liquid nitrogen temperature, corresponding to a fast neutron fluence of approximately 1010 cm-2 and a gamma radiation dose of about 50 Gy, confirming only the temporary creation of defects. No signs of long-term deterioration were observed, suggesting a potential self-healing mechanism. This resilience positions perovskite SCs as viable candidates for in-core radiation detection, supporting the further development of miniaturized MAPbBr3 SC devices for such applications.

The acceptance testing of PADC detectors for fast neutron dosimetry using PuBe neutron source

Solid-state nuclear track detectors (SSNTD) represent a passive detection method suitable for fast neutron dosimetry. The paper describes a series of acceptance tests for TASTRAK PADC detectors irradiated with a PuBe neutron source. Individual analyses aimed to improve the understanding of PADC performance under specific conditions and to identify possible problematic factors, such as influence of handling, influence of usage of detectors from different sheets etc. The experiment was conducted with two sets of detectors in cooperation with the Slovak Institute of Metrology using a PuBe neutron source. The equivalent ambient dose delivered to the detectors ranged from 0.14 to 14.99 mSv. The tests focused on detector detection limits, sensitivity, and the ability to determine the measured value. The performance of TASTRAK detectors in a fast neutron field was in agreement with the expectations, problematic areas were identified, and the focus of future research in the field of SSNTD dosimetry at Slovak Technical University in Bratislava (STU) was determined.

On modeling a plutonium-beryllium neutron source

In the present work, the neutron yield of a Pu-Be neutron source has been calculated by using two methods, calculation with the SOURCES4C code developed at LANL and via an equation, respectively. The present work aims to determine and compare the two methods in obtaining the yields and spectra of an (α,n) type neutron source. The source used for this work is being held at IFIN-HH, and was well experimentally characterized in previous work. The results of the present work compare well with the experimental data.

Simultaneous inter-calibration of 160 MiMi neutron detectors

The 160 MiMi neutron detectors of SAFFRON, a 3D full-core mapping system developed at EPFL for the CROCUS zero-power reactor, have been simultaneously inter-calibrated in preparation for their in-core installation. An experimental setup was built to distribute up to 180 MiMi detectors radially at 15 cm around the Pu-Be neutron source of the CARROUSEL facility. An efficient inter-calibration methodology is presented: first, the azimuthal shape of the relative source strength is characterized by rotating 18 MiMi detectors distributed every 20º around the PuBe, followed by the simultaneous determination of the relative sensitivity of 160 MiMi detectors, distributed all around the Pu-Be source.

Rod-shaped pulse shape discrimination plastic scintillation detectors applied for neutron source direction survey

A fast neutron detector system that can figure out neutron source direction is developed. The system consists of two rod-shaped fast neutron detectors and one polyethylene moderator. Pulse shape discrimination plastic scintillators are applied for reduced gamma-ray background and improved system sensitivity for fast neutron detection. The neutron detectors are placed on the back and front sides of the moderator facing a neutron source. Experiments are conducted for a single rod-shaped detector and the system changing the PuBe neutron source direction. Simulation-based studies are also performed using the Monte Carlo simulation code, PHITS (Particle and Heavy ion Transport code System). Taking the ratio of counts of the front and back system detectors, linear change with an increased detector angle is observed, except for near 0° and 90°. This enables us to deduce neutron source direction from several measurements. The method is probably useful for a quick and efficient survey of neutron sources or special nuclear materials.

MEASUREMENT AND SIMULATION OF THE RADIATION DOSES AROUND THE 'MINI LABYRINTH' EXPERIMENTAL WORKPLACE AT STU.

One of the most effective ways of gaining user experience and minimizing user effects on modelling and simulation is benchmarking. Currently (2022) a new experimental workplace, the so called 'Mini Labyrinth' is being developed at Slovak University of Technology in Bratislava (STU). It is a simple neutron and gamma shielding benchmark, inspired by the ALARM-CF-AIR-LAB-001 ICSBEP experiment. The latest V2 setup consists of a PuBe neutron source, several NEUTRONSTOP C5 shielding blocks, H2O filled polylactic acid (PLA) tank, plastic source holder and detectors placed inside and outside the Mini Labyrinth. This paper is focused on the monitoring of the radiation doses around the workplace using the NuDET neutron detector and on the comparison of the measured quantities with ones simulated by SCALE 6 MONACO. The influence of the size of the model and the definition of the detector is studied in this analysis. The achieved results are presented and some discussions on further needed development are also included.

The Mini Labyrinth Benchmark for Radiation Protection and Shielding Analysis

The Mini Labyrinth experiment is a neutron and gamma shielding experiment currently being developed at the Slovak University of Technology, Bratislava (STU). The STU Mini Labyrinth consists of NEUTRONSTOP shielding blocks, water tank, PuBe neutron source, and graphite prism. This article describes the second construction version of the Mini Labyrinth experiment and presents the newest results of the neutron and gamma fields’ simulation and measurement. The PuBe neutron source with the emission rate of 1.06E7 <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\text{n}\cdot \text{s}^{-1}$ </tex-math></inline-formula> was used in the experiment. The measurement of gamma ambient dose equivalent H*(10) and neutron count rates is performed by the Thermo Scientific RadEye portable survey meter and the SNM-11 BF3-filled corona detector. The simulation part was carried out using the state-of-the-art MCNP6 and SCALE6 MONACO stochastic calculation tools, considering the detailed geometry of the Mini Labyrinth and a combined neutron and gamma source of particles. The comparisons were performed between codes and the experiment. The propagation of cross-sectional uncertainties was investigated through the shielding analysis. Almost perfect agreements between simulation codes were achieved. The comparison with measurement suggests the further needs of room effect assessment.

Comparison of detectors with pulse shape discrimination capability for simultaneous detection of gamma-rays, slow and fast neutrons

In this paper, five detectors (EJ-270 — single plastic scintillator, EJ-309B5 — single liquid scintillator, and 3 phoswich detectors: EJ-276+GS20, GS20+EJ-276, EJ-309+GS20) were studied in terms of simultaneous detection of gamma-rays and slow and fast neutrons. The measurements were carried out in a mixed field of neutron and gamma radiation from an intense (∼106 neutrons/s/4π) PuBe source. The PuBe neutron source was placed in an 18 cm diameter paraffin ball in order to increase the number of slow neutron detections. This paper covers the measurements of the neutron/gamma discrimination, photoelectron yield and analysis of the light pulse shapes originating from events related to gamma-rays, slow neutrons and fast neutrons. Sensitivity to slow neutrons is also briefly discussed. A comparison of n/γ discrimination for all of the tested detectors shows excellent performance of all phoswich detectors. Phoswich configurations tested within the scope of this study allow unambiguously resolve all three types of radiation and show superior performance in comparison with single loaded plastic or liquid scintillators.

A new method for measurement of moisture transport in porous media based on forward and backward scattering of epithermal neutrons

We present a new method for determining the spatial distribution and transport of water in porous media. It is based on the detection of both forward and backward scattered neutrons from the wet regions of the samples under investigation. The experimental set-up is based on a Pu-Be neutron source and He-3 neutron detector assemblies. The results obtained showed that back scattered neutrons are more sensitive than the forward scattered neutrons to determine water content. Moreover, both forward and back scattered neutrons are more sensitive than either back or forward neutrons for determining water content. The method was used to measure moisture transport in sand columns and brick samples. Forward and backward scattered neutrons from different wet regions along the water flow path (x) are recorded as the sample absorbs water. Water saturates the regions of the samples tested near the inlet of water faster than the others. The water front positions were found to follow the square root behavior of the absorption time, and capillary penetration coefficients were determined for the samples investigated. The developed method can be used to investigate water absorption at various flow rates in porous samples of various sizes.

The Mini Labyrinth – a Simple Benchmark for Radiation Protection and Shielding Analysis

The Mini Labyrinth experiment is a simple neutron and gamma shielding experiment developed at STU, inspired by the ALARM-CF-AIR-LAB-001 ICSBEP benchmark experiment. The STU Mini Labyrinth is approximately ten times smaller and consists of NEUTRONSTOP shielding blocks. This paper describes the second version of the Mini Labyrinth experiment and presents the results of the neutron and gamma fields simulation and measurement. The PuBe neutron source with the emission rate of 1.0E7 n.s-1 was utilized in the experiment. The measurement of gamma ambient dose equivalent H*(10) and neutron count rates is performed by the Thermo Scientific RadEye portable survey meter. The simulation part was carried out using the state-of-the-art MCNP6 and SCALE6 MONACO stochastic calculation tools taking into account the detailed geometry of the labyrinth and a combined neutron and gamma source of particles. The comparisons were performed between codes and an experiment, based on the dose rate in the unique detection positions and using a 2D map of neutron and photon fluxes. The propagation of cross-section uncertainties was investigated through the shielding analysis. Partial agreement between codes and measurement was achieved, however serious discrepancies near the PuBe source were identified.

EVALUATION OF THE DOSE LEVELS IN THE IRRADIATION CHANNELS OF PU-BE NEUTRON SOURCES STORAGE DEVICE BY USING NEUTRON ACTIVATION ANALYSIS AND MCNP6 MODELLING

PuBeneutrons sources of Educational Neutron Laboratory in the Center for Physical Sciences and Technology are used in several activities including neutron activation for educational purposes and laboratory irradiation experiments. In this work we address the dose rate estimation in the irradiation channels of reconstructed PuBe neutron sources storage device in the particular positions dedicated for neutron irradiation experiments using neutron activation analysis (NAA) and MCNP6 modelling.The results on neutron activation of V2O5, MnO2, Al and Na2CO3were analyzed and compared with MCNP6 model prediction results. The conclusions on the actual neutron flux energy distribution and model corrections are drawn. The absolute neutron intensity at dedicated irradiation points as well as neutron and gamma dose rates are obtained.

Identification of neutrons and gamma rays using a combination of three algorithms for separating signals of the scintillation detector

Scintillation detectors with organic scintillators are widely used for fast neutrons detection in high gamma ray background. The peculiarity of this type of detector is that the pulse shape depends on the type of the detected particle. Traditionally, the Pulse Shape Discrimination (PSD) histogram is used to determine the number of detected neutrons. The PSD parameter is calculated from the shape of the detector pulse and assigned to each pulse. A typical PSD histogram contains two peaks corresponding to neutrons and gamma rays that overlap in the region between the peaks. With this approach, it is impossible to identify each individual signal in the area between the peaks. Therefore, it is not possible to calculate the overall signal identification coefficient. We have proposed a new method for the identification of neutrons and gamma quanta, which includes a combination of three signal separation algorithms: the traditional histogram PSD, the dependence of the area of signals on their amplitude, Tau histogram (tau means the fall constant of the detector pulses). This combination of three algorithms makes it possible to calculate the value of the signal identification coefficient. To test a new method for identifying neutrons and gamma quanta, we used a Pu-Be neutron source, a scintillation detector with a p-terphenyl crystal and a CAEN DT5730 Digitizer (14 bit, 500 MHz). When a scintillation detector registered neutron from a Pu-Be source, the signal identification coefficient was 91.6%. A new method for identifying signals from a scintillation detector is used to register neutrons at the light ion accelerator.

Separation of signals from neutrons and gamma quanta by the method of normalized signals

The solution of the problem how to register fast neutrons in the presence of intense gamma radiation is required when solving such fundamental and applied problems as registration of the neutron and gamma background in underground low-background experiments (the low background detectors of the neutrino and dark matter); beam diagnostic at particle accelerators; radiation monitoring at nuclear facilities, nuclear medicine; environmental monitoring. To separate signals from neutrons and gamma quanta, scintillation detectors with organic scintillators are used. The best scintillators are organic crystals of stilbene and p-terphenyl. The efficiency of separating signals from neutrons and gamma quanta can be increased using various methods of digital signal processing of the pulse shapes of the registered signals. A parameter traditionally called the Figure of Merit (FOM) is used to compare these methods. The experimental setup consisted of a Pu-Be neutron source, a scintillation detector with organic crystal p-terphenyl, a Hamamatsu R6094 photomultiplier, a CAEN DT5730 Digitizer (500 MHz, 14bit), which store the shape of each pulse for the following digital processing. A new “method of normalized signals” was developed. Three variants of the new method of normalized signals are described, which give the following FOM values: 1.6, 1.7, and 2.1. The traditional method of signals separation on the same array of experimental data showed the efficiency FOM = 1.6. The new method of signal separation is used to register fast neutrons in the installation dedicated for the development of a compact neutron generator, which is necessary for the calibration of low-background detectors of neutrinos and dark matter particles.

Insights into the effect of the mineralogical composition of serpentine aggregates on the radiation attenuation properties of their concretes

Antigorite, lizardite, and chrysotile are the most important polymorphsof serpentine. In this work, we investigate the physical and radiation shielding properties of three serpentine concrete mixes containing the three serpentine polymorphs. The physical properties were evaluated in terms of density, water absorption and porosity. The shielding characteristics against fast neutrons and γ-rays were determined. Attenuation measurements were conducted via the transmission method using a PuBe neutron source, which was utilized as a mixed-field source for fast neutrons and γ-rays over the energy regions of 0.8–11 and 0.4–8.3 MeV, respectively. The experimental data were compared with the NXcom program results, and an excellent agreement was observed. This study showed that the antigorite-based concrete has a relatively higher attenuation efficiency against γ-rays and fast neutrons compared with the lizardite- and chrysotile-based concretes. This finding is discussed in terms of the effects of minerals associated with serpentine, namely dolomite (CaMg(CO3)2) and magnetite (Fe3O4). Further, the contributions from dolomite should be considered for the precise determination of loss on ignition (LOI). These results suggest that associated minerals with serpentine aggregates can lead to a significant variation in the efficiency of γ-rays and fast neutrons shielding of serpentine concretes.

Modernization of the pulse shape discrimination method for neutron and gamma quanta in scintillation detector

In this paper, we investigated the efficiency of several known and new methods of digital pulse shape discrimination for neutrons and gamma quanta. Experimental data were obtained on a setup consists of a Pu-Be neutron source, organic p-terphenyl scintillation detector and 14 bits, 500 MHz sampling rate flash-ADC with capability to store and upload to the host computer long waveforms for further analysis. A comparison is made in between the results of using traditional and new methods for calculating the signal separation efficiency of Figure of Merit (FOM). The best known from the literature value of the efficiency of neutron and gamma quanta discrimination for the Pu-Be source is FOM = 1.5. We obtained the separation efficiency FOM = 1.77 in the scintillation detector with the p-terphenyl crystal, by a new method. Note also that for the known liquid scintillator BC-501A FOM≈1. A new method of scintillation detector pulse shape discrimination from neutrons and gamma quanta is used to detect the neutron yield from compact neutron generator that is created on the basis of carbon nanotubes.

Qualitative measurement of spatial shielding isotopics via Compton imaging neutron-induced gamma rays using 3-D CdZnTe detectors

Abstract Characterization of shielding around special nuclear materials (SNMs) in unknown objects is important for homeland security, international safeguards and potential nuclear disarmament verification. Shielding modulates source photons and neutrons, complicating radiation-based SNM characterization, and indicates object purpose and guides response. Previous gamma-ray-based shielding detection algorithms only estimate effective shielding atomic number and areal thickness, complicating the characterization of compound shields comprised of multiple elements. Neutron-induced gamma rays, from inelastic scatter or capture, provide a complementary, isotope-specific shielding signal which can be localized using Compton imaging. Qualitative, angularly-resolved isotopic distributions in a compound shield were measured using polyethylene and polyvinyl chloride targets, excited via a single PuBe neutron source, and a 3-D, position-sensitive CdZnTe imaging spectrometer. Hydrogen capture gamma rays, which predominately originate from the polyethylene target, were well separated in angular space from chlorine capture gamma rays, originating in the polyvinyl chloride target, using Compton imaging. The described qualitative probe of spatial shielding isotopics can be immediately applied on any commercial gamma-ray imaging spectrometer without modification.

New composite material based on heavy concrete reinforced by basalt-boron fiber for radioactive waste management

A new composite material with neutron radiation shielding properties is presented. This fiber reinforced concrete material incorporates basalt-boron fiber, with different concentrations of boron oxide in fiber, and is applicable to nuclear energy and nuclear waste management. The methodology for production of boron oxide (B2O3) infused basalt fiber has been developed. First experimental samples of basalt boron fiber containing 6% of B2O3 and 12% B2O3 have been produced in laboratory conditions. The concrete samples reinforced by two types of basalt-boron fiber with different dosages have been prepared for neutron experiment. The neutron experimental investigations on radiation shielding properties of concrete reinforced by basalt-boron fiber have been performed by means of Pu-Be neutron source. The prepared samples have been tested in the course of several series of tests. It is shown that basalt-boron fibers in concrete improve neutron radiation shielding properties for neutrons with different energies, but it appears to be most effective when it comes to thermal neutrons.

Fast neutron detectors with silicon photomultiplier readouts

Abstract This work summarizes a fast neutron detection performance of two different silicon photomultipliers from two manufacturers. The first SiPM (MAPD-3NK) from Zecotek Photonics consists of deeply buried cells with the active area 3 . 7 × 3 . 7 mm2. The second one (MPPC-S12572-010P) from Hamamatsu, however, consists of surface cell structure which the active area is 3 × 3 mm2. Both SiPMs have the same pixel density of 10 000 mm−2. Both SiPMs coupled to Stilbene ( 5 ∗ 5 ∗ 5 mm3) and p-terphenyl ( 5 ∗ 5 ∗ 5 mm3 ) plastic scintillators were evaluated for detection ability of fast neutrons using a PuBe neutron source . Charge comparison and zero crossing neutron/gamma discrimination techniques were performed for these detectors and the results were compared. The obtained results prove a good fast neutron detection performance of the SiPMs which makes it possible to use these types of neutron detectors in fast neutron detection applications.

P197] Measurement of neutron-induced radioactivity in animal tissue

Purpose The basis of neutron radiotherapy is to induce the decay of 10B to 7Li, an alpha particle and gamma ray. Emitted particles deposit dose in surrounding cancer cells what lead to their death. Though neutron beam can also lead to production other radioactive isotopes with lives longer than 11B∗. The main goal of out project is to measure the induced radioactivity in human tissues after the irradiation and identify the produced radioisotopes and estimate the dose from those nuclides. The secondary aim of this work is the validation of simulation calculations in comparison with experimental results. Methods In order to estimate which radioisotopes are produced during the irradiation the GATE/Geant4 Monte Carlo code was used. For tissue materials, the pig liver and beef bone were chosen. The irradiation will be performed using PuBe neutron source and neutron beam which is available in BNCT treatment room at National Center for Nuclear Research, Otwock. The samples will be irradiated with high dose in order to create a large number of short-lived isotopes or with fractionated dose in order to express the radiation of long-lived isotopes. For gamma spectroscopy measurements the HPGe or LaBr3 detectors will be used. Results The Monte Carlo calculations show the most intense, for gamma spectroscopy, radioisotopes produced within the liver sample by neutron irradiation are 24Na, 28Al and 31Si. In case of bone sample neutron irradiation mostly produced gamma radioisotopes are 28Al, 31Si and 42K. As a result of high occupancy of neutron sources the experimental data will be obtained in April. Conclusions Based on theoretical calculations, one can conclude that a lot of radioisotopes are created in human tissues. For example, the 24Na can wash out from the irradiated volume and circulate through patients body with blood. That is why the dose from induced radioactivity cannot be omitted in calculations of received total dose. In addition, some of the mentioned above nuclides (high cross-section for production) can be useful for off-line measurements of fast neutron fluence.