Beryllium Neutron Research Articles

Total neutron emission from deuteron fusion and plasma pinch compression in a medium-sized plasma focus operating with D2 and D2 + Ne gas mixtures—Experimental results

Newly obtained results on hot and dense deuterium and deuterium-neon plasma compression in a z-pinch electrical discharge configuration are presented. The investigated plasma was generated and compressed using 269 high-current discharges in a medium-sized (dense) plasma focus device. The experimental chamber of the device was filled with deuterium and deuterium-neon gas mixtures under constant total mass/density conditions. Magnetic and electric probes, beryllium neutron activation counter, and high-speed four-frame vacuum ultraviolet/soft x-ray pinhole camera were used to study the plasma dynamics and radiation emission. The results obtained experimentally for the first time confirmed clearly a decrease in the minimum radius of plasma columns with an increase in initial neon fraction. Simultaneously, a decrease in the total neutron emission from deuteron fusion was found. The observed plasma/discharge evolution revealed that the classical description of plasma-focus discharges can be approximately correct up to the moment of maximum compression. Including, existence of quasi-equilibrium plasma compression is probable. It is also possible that the homogeneity of plasma columns during the slow compression phase and maximum compression moment increases with the increase in initial neon fraction. The effect of higher stabilization (repeatability) of discharges was confirmed, for higher initial neon fractions. The dependency of the total neutron emission yield on the parameters describing the full discharge dynamics and the maximum discharge voltage was confirmed. The existence of this type of dependency, for a minimum pinch radius is also possible. In contrast, there was little dependency to the total discharge current parameters measured in the collector area.

Systematic evaluation of fast neutron sensing with Cesium Hafnium Chloride

Cesium Hafnium Chloride (CHC) is a promising new scintillator for dual mode sensing of both neutrons and gamma-rays. The high chlorine content allows for both 35Cl(n,p)35S and 35Cl(n, α)32P reaction channels leading to reliable detection of fast neutrons. By utilizing pulse shape discrimination (PSD), these neutron interactions may be reliably separated from gamma-ray signals with a high figure-of-merit (FOM) after optimization of the PSD algorithm. In this study, the PSD algorithm settings for CHC were systematically investigated using a bare 252Cf and a lead-shielded plutonium beryllium neutron source. It was found that the PSD algorithm settings affects both the FOM and the neutron detection efficiency, but a FOM as high as 4.5 for alpha particles was observed in one data processing scenario. Further, an intrinsic 5 parts per million alpha-emitting contamination was observed in the sample, which we attribute to natural uranium.

Design and construction of a 9 MeV [formula omitted]-ray source based on capture of moderated plutonium–beryllium neutrons in nickel

We have designed and constructed a high-energy γ-ray source for detector characterisation and calibration. The source is a composite type based on a plutonium–beryllium neutron emitter embedded in a paraffin moderator, which is homogeneously mixed with nickel powder. The 9 MeV γ-ray source produces approximately 450 photons per second in 4π when 2.2×105 neutrons per second are emitted, corresponding to a surface flux of 9 MeV γ-rays of approximately 2.5×10−6 cm−2 per emitted neutron. Here we discuss the properties and design of this source, including the characterisation of homogeneity and high-energy γ-ray emission spectra.

Study on the Design, Preparation, and Performance Evaluation of Heat-Resistant Interlayer-Polyimide-Resin-Based Neutron-Shielding Materials.

Polymers have an excellent effect in terms of moderating fast neutrons with rich hydrogen and carbon, which plays an indispensable role in shielding devices. As the shielding of neutrons is typically accompanied by the generation of γ-rays, shielding materials are developed from monomers to multi-component composites, multi-layer structures, and even complex structures. In this paper, based on the typical multilayer structure, the integrated design of the shield component structure and the preparation and performance evaluation of the materials is carried out based on the design sample of the heat-resistant lightweight polymer-based interlayer. Through calculation, the component structure of the polymer-based materials and the three-layer thickness of the shield are obtained. The mass fraction of boron carbide accounts for 11% of the polymer-based material. Since the polymer-based material is the weak link of heat resistance of the multilayer shield, in terms of material selection and modification, the B4C/TiO2/polyimide molded plate was prepared by the hot-pressing method, and characterization analysis was conducted for its structure and properties. The results show that the ball milling method can mix the materials well and realize the uniform dispersion of B4C and TiO2 in the polyimide matrices. Boron carbide particles are evenly distributed in the material. Except for Ti, the other elemental content of the selected areas for mapping is in good agreement with the theoretical values of the elemental content of the system. The prepared B4C/TiO2/polyimide molded plate presents excellent thermal properties, and its glass transition temperature and initial thermal decomposition temperature are as high as 363.6 °C and 572.8 °C, respectively. In addition, the molded plate has good toughness performs well in compression resistance, shock resistance, and thermal aging resistance, which allows it to be used for a long time under 300 °C. Finally, the prepared materials are tested experimentally on an americium beryllium neutron source. The experimental results match the simulation results well.

Computational analysis of a neutron trap in the center of the WWR-K reactor core for irradiation tests of large-sized objects

Nowadays, the WWR-K research reactor is operated with low-enriched uranium fuel and a beryllium neutron reflector. Irradiation positions located in the core make it possible to irradiate objects no larger than 50 mm in size. In the reactor tank, behind the beryllium reflector, there are irradiation positions with a diameter of 200 mm, but in them, in comparison with irradiation positions in the core, the neutron intensity is lower. This motivated us to carry out computational studies on the generation of a large diameter neutron trap in the center of the WWR-K reactor core. The paper presents the results of neutron-physical calculations on the generation of a neutron trap in the center of the WWR-K reactor core. The generation of a large diameter neutron trap in the core of the WWR-K reactor will make it possible to carry out radiation tests of large objects. For this, it is necessary to change the layout of the core. The influence of the transformation of the core on the experimental characteristics of the WWR-K reactor is shown.

Status of HCCR TBM program for DEMO blanket

Development of the Helium Cooled Ceramic Reflector (HCCR) breeding blanket is underway according to the development strategy of core technology for K-DEMO under the fusion energy development roadmap in Korea. The main goals are to validate integrated design tools and to develop core technologies in the field of materials, manufacturing and joining technologies, helium cooling and tritium technologies, system integration, and safety technologies. A unique graphite reflector concept was adopted to save cost by reducing the amount of beryllium neutron multiplier. The main functions and design concepts of the HCCR breeding blanket are to be verified in ITER through the HCCR Test Blanket Module (TBM) program. Not only breeder modules are to be tested in ITER but also the technologies of cooling, coolant purification, tritium extraction, etc. are to be tested or proved in ITER before employed in DEMO. This paper introduces the current updates of the design and R&D activities of the HCCR TBM systems in the preliminary design stage.

Simulations for practical measurement methods of spatial neutron distribution inside blanket mock-up irradiated with DT neutrons

Measuring methods of spatial neutron distribution have been required for the evaluation of TBR inside the blanket. In this paper, neutron transport inside a blanket mock-up, mainly assembled by beryllium neutron multiplying layers and lithium breeding layers, was simulated for measurements of the neutron energy spectrum and the spatial distribution with activation analyses using activation foils and an imaging plate. Transport of deuterium-tritium (DT) fusion reaction neutrons were simulated by MCNP6 and activation calculations were performed by DCHAIN. These results of neutron transport indicate that activation foils were reacted to neutrons with specific energies; dysprosium for thermal, manganese with cadmium for epi-thermal and niobium or cobalt for fast neutrons. The results of activation calculations indicate each activity of activation foils inside the blanket, which proposes the exposure conditions for measurements using imaging plate.

Feasibility study on tritium recoil barrier for neutron reflectors of research and test reactors

Tritium release into the primary coolant of the JMTR and the JRR-3 M had been studied, and it is found that tritium recoil release from the chain reaction of beryllium neutron reflectors is dominant. To prevent the tritium recoil release, Al, Ti, V, Ni and Zr are selected as the candidate tritium recoil barrier materials in this feasibility study. It is clear that 20∼40 μm thickness is required depending on the material to reduce by 3 orders, and that an impact on the effective multiplication factor is about 0.2 % at most. Total evaluation including the activities, fabrication and usage experiences, suggests the selection of Al as the first candidate may have the least development risk as the tritium recoil barrier.

Neutron shielding properties of TPX/B4C composites based on Monte Carlo method

As the key and difficult point of shielding protection, many domestic and foreign scholars have carried out a lot of experimental research on neutron shielding materials. In this paper, the research progress, advantages and disadvantages of common neutron shielding materials are discussed, and the neutron shielding process of TPX/B4C composites is simulated by Monte Carlo method, and the results are compared with the Boron polyethylene, lead boron polyethylene, aluminum based boron carbide mentioned in the literature. Based on the simulation results, it is concluded that the thermal neutron absorption rate of TPX/B4C composites with the thickness of 1cm is 98.3%, the fast neutron shielding rate of 1MeV is 1.457, and the neutron shielding reduction factor and neutron dose reduction factor of americium beryllium neutron source are obviously better than the other three materials. The simulation results show that TPX/B4C Composite can be used as a new kind of high temperature neutron shielding material.

Characterization of a plutonium–beryllium neutron source

Here we present an investigation of a plutonium–beryllium neutron source available at the Horia Hulubei National Institute of Physics and Nuclear Engineering, Romania, to be used for detector characterization during the implementation of the Extreme Light Infrastructure – Nuclear Physics project. Using several different techniques and instruments, we have measured the isotopic composition for plutonium to be 75% 239Pu and 24% 240Pu, with a minor contribution from other isotopes. Furthermore, we have measured the source activity as of November 20th 2019 to be 2.220(5)×105 neutrons per second with a mean energy of 3.25(17) MeV. We have also measured both the γ-tagged and full neutron energy spectra, and discuss the origin of the observed structure in the neutron energies based on these. All these parameters are of importance both for traceability of nuclear material, radioprotection, and accurate detector characterization.

A Fizikai Intézet Ra-Be neutronforrásának felhasználása és eltemetése

The period of service of our laboratory Radium-226 – Beryllium neutron source, used in our education for 44 years, is expired so its disposal became necessary. The disassembling of the source, its preparation for disposal was a unique, professional challenge. In this paper, we show the performance of this work, with regard to the referring regulations. On the basis of our measurements we delivered an uninjured closed isotope to the radioactive waste disposal facility.

Feasibility study of Tritium recoil barrier forneutron reflectors

Tritium release into the primary coolant of the research and test reactors during operation had been studied, and it is found that the beryllium components used as a neutron reflector in the core strongly affect the tritium release into the primary coolant, and it is also found the recoil release from chain reaction of 9Be is dominant. To reduce tritium concentration of the primary coolant, feasibility study of the tritium recoil barrier for the beryllium neutron reflectors was carried out, and the tritium recoils of various materials such as Al, Ti, V, Ni, Zr, etc., were calculated by PHITS.  From these calculation results, it is clear that the thickness of tritium recoil barrier depends on the material and 20-40 µm is required for three orders reduction. Additionally, the denser materials have shorter recoil length in general, however, it is clear that the recoil lengths of Zr and Pb are similar with Ti.

Neutron spectrometry of a lightly encapsulated 241Americium–beryllium neutron source using two different Bonner Sphere Spectrometers

Abstract 241Americium–beryllium neutron sources are used in a variety of applications, spanning from the civil engineering to the oil industry, and to the radiation dosimetry. The ISO 8529-1 Standard reports a reference distribution for the energy of the emitted neutrons, which extends from 0.1 to 11 MeV. Nevertheless every source exhibits a slightly different spectrum, depending on the material and dimension of the capsule, and on the amount of active material. This source-to-source variability is expected to mainly affect the energy region below 0.1 MeV. As these differences can be important in the fields of neutron dosimetry and metrology, some relevant source formats have been investigated with spectrometric methods. The neutron calibration laboratory of the Politecnico di Milano recently acquired a CZ/1003/S — 96 type 241 Am–Be source (nominal emission rate 2.2E+6 s−1). As this type of source is in use in a number of European laboratories, its spectrum was determined through a neutron spectrometry experiment, involving two Bonner sphere spectrometers with well-known responses. The experimental spectrum is compared to that published for a similar source and to the ISO-recommended one.

Beam dynamics and commissioning of CW RFQ for a compact deuteron–beryllium neutron source

The Institute of Modern Physics (IMP) of the Chinese Academy of Sciences (CAS) is building a compact neutron source using a superconducting linac. The injector RFQ will operate in continuous wave (CW) mode, accelerating a 10-mA deuteron beam from 20 keV/u to 1.5 MeV/u. To minimize the possibility of beam loss in the downstream superconducting linac and maintain high acceleration efficiency, we take the 99.9% longitudinal beam emittance as the key optimization parameter for the beam dynamics design of the RFQ, and include an internal buncher to reduce the longitudinal beam emittance. This paper describes the design procedures, beam dynamics simulations and preliminary beam commissioning results for this RFQ. Simulation results show that 99.9% longitudinal beam emittance at the RFQ exit is optimized to 3.5 pi mm mrad, which is 0.13 times the longitudinal acceptance of the downstream superconducting linac. Beam commissioning results demonstrate that the RFQ can accelerate 7.8 mA H2+ to 3.11 MeV. The transmission efficiency from a Faraday cup before the RFQ to the beam dump after the RFQ is 97.6%. The measured transmission efficiency agrees well with simulation results at different values for the inter-vane voltage.

Study on the PGNAA measurement of heavy metals in aqueous solution by the Monte Carlo–Library Least-Squares (MCLLS) approach

In the present work, a prompt gamma neutron activation analysis (PGNAA) setup, which consists of a 300mCi 241Americium–Beryllium (Am–Be) neutron source and a 4 × 4-in. Bismuth germanium oxide (BGO) detector, was developed for heavy metal detection in aqueous solutions. A series of standard samples with analytical purity were prepared by dissolving heavy metals in deionized water. Quantitative spectrum analysis was performed by the Monte Carlo–Least-Squares (MCLLS) approach to measure the standard samples. The detector response functions of 4 × 4-in. BGO detector were generated by using the CEARDRF code. The element libraries were simulated in silico by the CEARCPG code, which was developed by Dr. Gardner. The simulation results presented were in very good agreement with the experimental results. The correlation coefficients were very close to 1 when the fitted spectrum was compared with the experimental spectrum. By applying the MCLLS approach, the relative deviation of the measurement accuracy was less than 2.27% for Ni, Mn, and Cu and up to 69.33% for Pb.

Electrical Properties of MWCNT/HDPE Composite-Based MSM Structure Under Neutron Irradiation

Because of their low cost, low energy consumption, high performance, and exceptional electrical properties, nanocomposites containing carbon nanotubes are suitable for use in many applications such as sensing systems. In this research work, a metal–semiconductor–metal (MSM) structure based on a multiwall carbon nanotube/high-density polyethylene (MWCNT/HDPE) nanocomposite is introduced as a neutron sensor. Scanning electron microscopy, Fourier-transform infrared, and infrared spectroscopy techniques were used to characterize the morphology and structure of the fabricated device. Current–voltage (I–V) characteristic modeling showed that the device can be assumed to be a reversed-biased Schottky diode, if the voltage is high enough. To estimate the depletion layer length of the Schottky contact, impedance spectroscopy was employed. Therefore, the real and imaginary parts of the impedance of the MSM system were used to obtain electrical parameters such as the carrier mobility and dielectric constant. Experimental observations of the MSM structure under irradiation from an americium–beryllium (Am–Be) neutron source showed that the current level in the device decreased significantly. Subsequently, current pulses appeared in situ I–V and current–time (I–t) curve measurements when increasing voltage was applied to the MSM system. The experimentally determined depletion region length as well as the space-charge-limited current mechanism for carrier transport were compared with the range for protons calculated using Monte Carlo n-particle extended (MCNPX) code, yielding the maximum energy of recoiled protons detectable by the device.

Chemical investigation of three plutonium–beryllium neutron sources

Thorough physical and chemical characterization of plutonium–beryllium (PuBe) neutron sources is an important capability with applications ranging from material accountancy to nuclear forensics. Characterization of PuBe sources is not trivial owing to range of existing source designs and the need for adequate infrastructure to deal with radiation and protect the analyst. This study demonstrates a method for characterization of three PuBe sources that includes physical inspection and imaging followed by controlled disassembly and destructive analysis.

Developments for neutron-induced fission at IGISOL-4

At the IGISOL-4 facility, neutron-rich, medium mass nuclei have usually been produced via charged particle-induced fission of natural uranium and thorium. Neutron-induced fission is expected to have a higher production cross section of the most neutron-rich species. Development of a neutron source along with a new ion guide continues to be one of the major goals since the commissioning of IGISOL-4. Neutron intensities at different angles from a beryllium neutron source have been measured in an on-line experiment with a 30MeV proton beam. Recently, the new ion guide coupled to the neutron source has been tested as well. Details of the neutron source and ion guide design together with preliminary results from the first neutron-induced fission experiment at IGISOL-4 are presented in this report.

Neutron producing reactions in PuBe neutron sources

There are a plenty of out-of-use plutonium–beryllium neutron sources in Eastern Europe presenting both nuclear safeguards and security issues. Typically, their actual Pu content is not known. In the last couple of years different non-destructive methods were developed for their characterization. For such methods detailed knowledge of the nuclear reactions taking place within the source is necessary. In this paper we investigate the role of the neutron producing reactions, their contribution to the neutron yield and their dependence on the properties of the source.

Characterization of neutron calibration fields at the TINT's 50 Ci americium-241/beryllium neutron irradiator

Reliable measurement of neutron radiation is important for monitoring and protection in workplace where neutrons are present. Although Thailand has been familiar with applications of neutron sources and neutron beams for many decades, there is no calibration facility dedicated to neutron measuring devices available in the country. Recently, Thailand Institute of Nuclear Technology (TINT) has set up a multi-purpose irradiation facility equipped with a 50 Ci americium-241/beryllium neutron irradiator. The facility is planned to be used for research, nuclear analytical techniques and, among other applications, calibration of neutron measuring devices. In this work, the neutron calibration fields were investigated in terms of neutron energy spectra and dose equivalent rates using Monte Carlo simulations, an in-house developed neutron spectrometer and commercial survey meters. The characterized neutron fields can generate neutron dose equivalent rates ranging from 156 μSv/h to 3.5 mSv/h with nearly 100% of dose contributed by neutrons of energies larger than 0.01 MeV. The gamma contamination was less than 4.2-7.5% depending on the irradiation configuration. It is possible to use the described neutron fields for calibration test and routine quality assurance of neutron dose rate meters and passive dosemeters commonly used in radiation protection dosimetry.