Finding a suitable shield for mixed neutron and photon fields based on an Am–Be source

This study aimed to investigate a shielding design against neutron and photon rays from neutron irradiators based on Am–Be sources, using the Monte Carlo simulation. Different shielding materials were studied, including borated polyethylene, DaGa concrete, and epoxy resin with colemanite. The Monte Carlo N-particle code (MCNP) was used to design shielding. A new type of neutron and photon shielding material based on 40 % galena, 55 % polyethylene, and 5 % boric acid is proposed. The results show that the total effective dose of radiation is significantly reduced by the optimum radius of this shielding system.     

X-Ray,Gamma, and Neutron Radiation Tests on Epoxy-Ferrochromium Slag Composites by Experiments and Monte Carlo Simulations

Radiation shielding effects of ferrochromium slag loading hardened epoxy resin samples were investigated. Five different samples including different percentages of epoxy resin and ferrochromium slag were produced. X-ray, gamma ray, and neutron particle transmission experiments were performed for epoxy-ferrochromium slag composites. Also, FLUKA Monte Carlo simulations were made to obtain absorbed doses. As a result, radiation shielding performance increases with increasing ferrochromium slag additive in epoxy.     

Investigation of fast neutron shielding characteristics depending on boron percentages of MgB<Subscript>2</Subscript>, NaBH<Subscript>4</Subscript> and KBH<Subscript>4</Subscript>

The macroscopic cross-section Σ and average neutron fluence in matter Φ are usable factors to comment neutron shielding property of samples. In this paper, we have used MgB₂, NaBH₄ and KBH₄ samples including different percentages of boron. Neutron macroscopic cross-section measurements of them have been done by using a source of mono-energetic neutrons (E _eff = 4.5 MeV ²⁴¹Am–Be). Average neutron fluence values and double differential fast neutron flux distributions of each samples calculated by using FLUKA Monte Carlo code. Also half value layers (HVLs) of samples are compared to paraffin which is one of the most neutron moderators. As a result, growing boron concentration can raise neutron shielding property of materials.     

Characterization of some boron minerals against neutron shielding and 12 year performance of neutron permeability

In this study, the nuclear radiation permeability properties of various boron minerals are evaluated because of their high neutron absorption and lowest transmission properties. Because of these properties boron minerals can be used at the area of neutron shielding. X-ray diffraction (XRD) analyses are done for the identification of the minerals, and then their B₂O₃ contents are determined experimentally. In addition, X-ray florescence (XRF) analyses are made for quantitative determination of calcium, iron, zinc and arsenic contents. The methods of Differential Thermal Analysis, Thermal Gravimetry (TG/DTA) and Differential Scanning Calorimeter (DSC) are used for obtaining the enthalpy and weight changes with temperature. Additionally, neutron permeability experiments are conducted. From the experimental results, the highest boron oxide content was found in clay containing colemanite. Iron, zinc and arsenic contents were not affecting the neutron shielding. The lowest permeability is provided by the kurnakovite mineral. Also it is observed that all of the minerals show an increase in their permeability in 12 years. It can be stated that boron minerals, specifically kurnakovite, is determined to yield the lowest neutron permeability value and therefore, the use of these materials for neutron shielding would be suitable.     

Potential use of NR and wood/NR composites as thermal neutron shielding materials

This work investigated thermal neutron shielding, cure characteristics and mechanical properties of natural rubber (NR) and wood/NR composites with addition of either boron oxide (B₂O₃) or boric acid (H₃BO₃) for potential use as flexible shielding materials. The results showed that increase in the B₂O₃ or H₃BO₃ content from 0 to 80 phr and 0–50 phr in 10-phr increments, respectively, could improve thermal neutron shielding properties but reduced overall tensile properties, while the addition of 20-phr wood particles in wood/NR composites improved surface hardness and dimensional stability. Furthermore, the values of the Half Value Layer (HVL), which represent the required thickness of material to attenuate half of the incoming neutrons, were evaluated at a content of 80-phr B₂O₃ by varying thickness of both NR and wood/NR composites from 2.5 mm to 20.0 mm in 2.5-mm increments. The results indicated that the HVL values were approximately the same at 3.5 mm. Hence, the overall properties investigated in this work suggested great potential of these composites to be used as effective thermal neutron shielding materials.     

Neutron self-shielding correction for prompt gamma neutron activation analysis of large samples

Summary Facilities and methods for INAA of large samples (up to 30 kg) have been developed and successfully tested at IRI, Delft. The methods encompass corrections for neutron self shielding in an isotropic neutron field and gamma self-absorption. The sample’s neutron absorption and scattering characteristics are determined by monitoring the neutron fluence rate around the sample and comparing the neutron densities measured with unperturbed fluence rates. We report on the possibility of developing similar methods for PGNAA. Relative self-shielding factors were measured as well as obtained from Monte Carlo computations. The agreement is good except for the most extreme case, with respect to absorption, attempted (CCl₄).     

The soil moisture and its effect on the detection of buried hydrogenous material by neutron backscattering technique

Among the available nuclear techniques, the neutron backscattering technique, based on the detection of the produced thermal neutrons, is thought to be the most promising for landmine detections.The results obtained from Monte Carlo simulation were used for selection of BF₃ detector and Am–Be neutron source shielding. In addition, soil moisture was discussed as a limitation of the neutron backscattering technique. It was experimentally found that this technique is useful for soil whose water content is lower than 14%.     

Neutron Equivalent Dose Rate Measurements of Gypsum-Waste Tire Rubber Layered Structures

The neutron equivalent dose rates (µSv/h) of gypsum, steel-reinforced rubber waste tire, and gypsum-waste tire rubber sandwich composite samples were investigated. Prepared samples were irradiated with ²⁴¹Am-Be neutrons and transmission values were obtained using dose equivalent rates measured with a BF₃ neutron detector. Results were compared to those of concrete, and as a result of neutron shielding, the performance of gypsum, waste tire, and waste tire (steel-reinforced rubber) embedded gypsum samples was higher than that of concrete. This information may be useful for shielding design of nuclear application areas.     

Nanocomposite based on epoxy and MWCNTs modified with NiFe2O4 nanoparticles as efficient microwave absorbing material

In order to make a microwave absorbent material with good dielectric and magnetic properties, well dispersed microwave absorbing hybrid epoxy polymer composites containing nickel doped Fe₃O₄ nanocrystals coated on carbon nanotubes (NiFe₂O₄‐MWCNTs/epoxy) were synthesized by the combined precipitation‐hydrothermal method in 1‐30 wt.% of nanoparticles. Nickel possess well interaction with microwave radiation and represents fine electromagnetic interference (EMI) shielding and by dopping it into ferrite spinel structures, does not show any tendency to oxidation. Well‐dispersed NiFe₂O₄–MWCNTs/epoxy nanocomposite prepared by new in‐situ polymerization method. First, NiFe₂O₄–MWCNT nanoparticles ultrasonicated in acetone and after mixing with epoxy resin ultrasonicated again. Finally, hardner added to the composite and tuned temperature for evaporating solvent. X‐ray diffraction (XRD) and energy dispersive spectroscopy (EDS) confirmed the synthesizing NiFe₂O₄ nanoparticles. Saturation magnetization value of NiFe₂O₄‐MWCNTs is about 29 emu/g with very low remanence and coercivity content, which revealed that the NiFe₂O₄‐MWCNTs is ferromagnetic nanocrystal. Transmission electron microscopy (TEM) used to characterize the distribution of NiFe₂O₄ nanocrystals on the surface of MWCNTs. The TEM images show that NiFe₂O₄ nanocrystals have a mean size of 12 nm, and completely coated on the exterior surface of MWCNTs. The obtained results of reflection loss revealed that the maximum values of reflection loss of the NiFe₂O₄‐MWCNTs/epoxy increase by enhancing the content of nanoparticles until 10 wt.% and decreases in 30 wt.%.     

Gamma ray attenuation coefficient measurement for neutron-absorbent materials

The compounds Na₂B₄O₇, H₃BO₃, CdCl₂ and NaCl and their solutions attenuate gamma rays in addition to neutron absorption. These compounds are widely used in the shielding of neutron sources, reactor control and neutron converters. Mass attenuation coefficients of gamma related to the four compounds aforementioned, in energies 662, 778.9, 867.38, 964.1, 1085.9, 1173, 1212.9, 1299.1,1332 and 1408 keV, have been determined by the γ rays transmission method in a good geometry setup; also, these coefficients were calculated by MCNP code. A comparison between experiments, simulations and Xcom code has shown that the study has potential application for determining the attenuation coefficient of various compound materials. Experiment and computation show that H₃BO₃ with the lowest average Z has the highest gamma ray attenuation coefficient among the aforementioned compounds.     

Conductivity and electromagnetic shielding effectiveness of flaky Ni/Ni-Cu-coated glass fiber/epoxy resin composite coating

Flaky Ni/Ni-Cu-coated glass fiber/epoxy resin composite coatings were prepared using the glass fibers and flaky nickel powders as fillers and epoxy resin as binder. The conductivity and electromagnetic shielding effectiveness of the coatings are as follows: (1) the appropriate content of Ni-Cu-coated glass fibers is 6 wt % in the composite filler and the optimum ratio of the filler to epoxy resin is 4: 1; (2) electrical conductivity of the coating with a thickness of 300 μm has a minimum value of 0.72 Ω cm; (3) shielding effectiveness of the coatings is up to 50.21–55.43 dB in the frequency range of 0.3–1000 MHz. This offers a new idea to enhance the added value of the glass fibers and raise the level of electromagnetic radiation protection.     

Estimation of background interference in prompt-gamma neutron activation using MCNP

Prompt-gamma neutron activation (PGNA) is used to measure total-bodynitrogen and hydrogen in humans. Background interference in the gamma spectraarises from both subject and shielding. A Monte Carlo simulation program (MCNP4B2)was used to examine the neutron and gamma signals in the PGNA system ( ²⁴¹AmBe source). N and H peak regions were assessed in the presenceand absence of calibration phantoms. The simulations suggested extracorporealH peak contributions of up to 30%, depending on subject body habitus. MostN background could be attributed to detector pileup events. The MCNP resultsallowed us to improve shielding design and develop background correction algorithmsto improve measurement precision.     

Study of γ‐Fast Neutron Attenuation and Mechanical Characteristics of Modified Concretes for Shielding and Sheltering Purposes

The study of γ‐neutrons attenuation and mechanical characteristics of modified concrete are vital and crucial parameters for the construction of civilian radiological, nuclear shielding, and/or shelters. In this work, fifteen samples of ordinary concretes with five different additives; steel fibers, polypropylene, silica fume, and fly ash, with variation of cement percentages, were prepared and used for performing the mechanical and radiation attenuation investigations. The compressive strength, tensile strength, slump test, bulk density, and water permeability were also carried out for the prepared concrete mixes. Collimated coherent beams from ⁶⁰Co and Pu‐Be fast neutron sources were used to check the radiation penetrability through the syntheized mixed concrete‐additives. Very sensitive and well calibrated gamma‐neutron pulse shape discriminating spectrometer with its electronic componenets and stilbene organic detector and 3′′ × 3′′ NaI scintillation crystal was used to measure the radiation before and after attenuation and transmission. The integrated fast neutron removal macroscopic cross section (Σ_r) and linear attenuation coefficient of total gamma rays (μ) were calculated for all the analysed concrete mixes. The results of measurements, tests, analyses and calculations are given and explained. The investigated modified concrete mixes show good workability and properties from the view point of mechanical loads and γ‐fast neutrons penetrability and resistance. These results can be used for shielding and sheltering design.     

The Argonne low level14C counting system

A low level¹⁴CO₂ counting system is described. This system was used to process several thousand CO₂ samples derived from atmospheric collections at various altitudes. Special features include counter construction utilizing electrolytic copper and shielding with neutron moderating and absorbing paraffin containing sodium metaborate. The effect of steel shielding thickness is shown, and the anticoincidence counters are described. Purification of the CO₂ for proportional counting is discussed.Work performed under the auspices of the U.S. Department of Energy, under Contract No. W-31-109-ENG-38.     

Shielding design and detection of neutrons from medical and industrial electron accelerators--simple method of design calculation for neutron shielding

The neutron leakage from medical and industrial electron accelerators has become an important problem and its detection and shielding is being performed in their facilities. This study provides a new simple method of design calculation for neutron shielding of those electron accelerator facilities by dividing into the following five categories; neutron dose distribution in the accelerator room, neutron attenuation through the wall and the door in the accelerator room, neutron and secondary photon dose distributions in the maze, neutron and secondary photon attenuation through the door at the end of the maze, neutron leakage outside the facility-skyshine.     

Fluorination effects of MWCNT additives for EMI shielding efficiency by developed conductive network in epoxy complex

To improve the efficiency in shielding electromagnetic interference in electronic devices, multi-walled carbon nanotubes were used, due to their excellent electric and magnetic properties at high aspect ratios, and were added to an epoxy matrix. Fluorination was carried out to achieve excellent dispersion and adhesion of the additives in the epoxy matrix. The improved dispersion was confirmed by UV spectra. The permittivity and permeability were also significantly improved based on the effects of the additives and the fluorination treatment. The efficiency of shielding electromagnetic interference increased up to 28 dB. This improved efficiency of shielding electromagnetic interference may be caused by a well-organized conductive network of additives in epoxy.     

Interfacial structure and properties in polystyrene/epoxy bilayer films

The interfacial structure and properties of immiscible deuterated polystyrene (dPS)/epoxy bilayer films were investigated with neutron reflectivity as functions of the composition of the epoxy layer, the thickness of the dPS layer, and the annealing time. We have found that the interfacial width and its growth rate depend strongly on the compositions of the epoxy layer but only weakly on the thickness of the dPS layer. The effect of the resin/crosslinker composition on the interfacial width and its growth rate is likely due to the different near‐surface structures that result for different epoxy stoichiometries. For an ultra‐thin dPS film (thickness = 2R_g), the data suggest a slight suppression of the growth of the interfacial width that could be due to confinement effects for the long‐chain molecules such as have been previously reported for a thickness of less than approximately 4R_g, where R_g is the radius of gyration of polymer molecules. © 2002 Wiley Periodicals, Inc. J Polym Sci Part B: Polym Phys 40: 2653–2660, 2002     

Preparation of activated carbon filled epoxy nanocomposites

Activated carbon derived from oil palm empty fruit bunch (AC-EFB), bamboo stem (AC-BS), and coconut shells (AC-CNS) were obtained by pyrolysis of agricultural wastes using two chemical reagents (H₃PO₄ or KOH). The AC-EFB, AC-BS and AC-CNS were used as filler in preparation of epoxy nanocomposites. Epoxy nanocomposites prepared at 1, 5 and 10 % activated carbons filler loading using KOH and H₃PO₄ chemical agents. Transmission electron microscopy confirms better dispersion of the nano-activated carbons in the epoxy matrix at 5 % activated carbon. The presence of 5 % AC-CNS in the epoxy matrix using H₃PO₄ chemical reagent resulted in an improvement of the thermal stability of epoxy matrix. KOH treated AC filled epoxy nanocomposites were slightly better in thermal stability as compared to H₃PO₄ treated AC filled epoxy nanocomposites, may be due to better interaction of filler with epoxy matrix. Thermal analysis results showed that thermal stability of the activated carbon filled epoxy nanocomposites improved as compared to the neat epoxy matrix. The degree of crystallinity of epoxy matrix was improved by adding the activated carbon due to interfacial interaction between AC and epoxy matrix rather than loading of AC alone. Developed nanocomposites from biomass (agricultural wastes) materials will help to reduce the overall cost of the materials for its demanding applications as insulating material.     

Design considerations for a neutron generator-based total-body irradiator

Summary The prompt- and delayed-gamma neutron activation techniques have been used for the non-invasive measurement of human body composition. In recent years, neutron irradiators have used only transuranic isotopic sources (^{238PuBe, 241}AmBe, ²⁵²Cf). However, in today’s security-minded environment, the use of alternate neutron sources may provide some advantages. We have examined several designs for an irradiator that would use a high-output, miniature D-T neutron generator (MF Physics). The use of this type of neutron source will lessen the storage, security, and transport issues associated with continuous-output isotopic neutron sources. To determine the scientific impact of this decision, we have performed Monte Carlo simulations (MCNP-4B2; Los Alamos National Laboratory) to aid in the design of the irradiator system, evaluating shielding materials, collimation, and source-to-subject distance, for the measurement of total body nitrogen (TBN). Based on internal flux distributions within the simulated body region of a subject, several design options were identified. The final design will be selected based on the optimization of precision, dose, and exposure time.     

Decreasing the detection limits to ppq levels for high purity silicon analysis by instrumental neutron activation analysis with kilogram sampling weight

The amount of silicon sample used in instrumental neutron activation analysis (INAA) is generally less than twenty grams and the detection limits are in the range of ppb to ppt. The detection limits can be decreased further in several ways. Increasing neutron flux density, extending irradiation period and/or using more effective detector can improve the detection limits to some extent. Increasing sample weight, however, is a more feasible way to decrease the detection limits by a factor of hundreds with no new investment in existing irradiation/counting systems. In this work, two 8-inch high-purity silicon samples were analyzed by INAA to evaluate the validity and limitations of this concept in respect to neutron flux inhomogeneity, neutron shielding and -ray counting efficiency.