Commercial grade aluminium: A versatile reactor flux monitor

With the advent of Ge(Li) spectrometry, a high standard of purity for neutron flux monitors no longer remains an imperative “must” and becomes rather superfluous. From this standpoint, commercial grade Al was investigated for its suitability as a reactor flux monitor and was found to have a much greater practical utility than most of the monitors reported. Three Al foils and one wire were randomly selected from four different commercial sources, and analysed for their Fe, Ga, Mn and Na contents by neutron activation and high-resolution gamma-spectrometry. While Na was found to have a very heterogeneous distribution, Fe, Mn and Ga concentrations in different splits of each type of Al were consistently uniform within ±2–3%. Eight possible monitor reactions on Al, Mn, Fe and Ga have been recommended as neutron flux integrators for all the 3 components of a reactor spectrum, viz. thermal, epithermal and fast, covering a wide range of flux levels from 10⁷ to 10¹⁴n·cm⁻²·sec⁻¹. The advantages and versatility of commercial grade Al as a pile neutron dosimeter are discussed.     

Characterization of the neutron flux gradients in typical irradiation channels of a TRIGA Mark II reactor

The neutron distribution in a defined volume (gradient) for different matrices (air, water, cellulose, biological material and silicon dioxide) in two typical irradiation channels (pneumatic tube (PT) and IC40-channel in the carousel facility) in the TRIGA Mark II reactor at the Joef Stefan Institute (IJS) was studied. Our experiment was based on inserting Fe wires (flux monitors) into the chosen matrices. The wires were cut into small pieces after irradiation and the induced activities of ⁵⁹Fe measured. The results showed that for the studied geometry the average spatial thermal neutron flux inhomogeneities (for five studied matrices) are about 2.3% in the PT-channel and about 2.9% in the IC40-channel.     

A new reflector structure for facility thermalizing D–T neutrons

A facility for thermalization of fast neutrons (14.2 MeV) emitted by compact deuterium–tritium (D–T) neutron generators (NGs) for thermal neutron activation analysis is proposed. Its final design is based on Monte Carlo calculations (MCNP5). To maximize the ratio between the thermal neutron flux and the total neutron flux and simultaneously to ensure the highest possible value of the thermal neutron flux at the output surface, the facility should consist of a two-layer reflector [tungsten (W)—the inner part, molybdenum—the outer part], a two-layer multiplier (W followed by lead), a moderator (polyethylene followed by magnesium fluoride) and a collimator (molybdenum and nickel near the output surface). For the D–T NG producing the maximum available neutron yield 10¹⁵ n s^?1, the facility provides the thermal neutron flux 2.0 × 10¹¹ n cm^?2 s ^?1 and a slightly higher fast neutron flux 2.3 × 10¹¹ n cm^?2 s^?1. To improve the ratio of the thermal neutron flux to the fast neutron flux (above 2.7) an addition of a silicon layer to the moderator and especially a proper adjustment and a threefold increase of the multiplier thickness is necessary.     

Fast co-pyrolysis of coal and biomass in a fluidized-bed reactor

Co-pyrolysis is one of the most promising options for the utilization of coal and biomass. Coal/biomass blends were prepared using Yilan subbituminous (YL) and corncob and the mass ratios of coal in mixtures varied between 0 and 100 %. Co-pyrolysis characteristics were investigated in a thermogravimetric analyzer from 303 to 973 K under the nitrogen flow of 100 mL min^?1. The co-pyrolysis residues were less than the sum simply added of the solid yields of individuals. With heating rate increased from 10 to 40 K min^?1, the residues decreased more severely compared to the expected under various blending ratios. For fast pyrolysis in fluidized-bed reactor, gas volumes and char yields of co-pyrolysis showed a significant linearity. But pyrolysis-oil yields were higher than the expected from the additive model when the YL blending ratios were less than 60 %. The co-pyrolysis evolved more H₂, CH₄, C₂ + C₃, and less CO than an additive pyrolysis process of individual fuel. The GC/MS results indicated that co-pyrolysis-oil contained more alcohols, ketones, aldehydes, or acids than that of individual fuel. All of that suggested the H/OH in volatiles produced from rapid pyrolysis of biomass transferred to the radicals of coal pyrolysis. The possible reaction mechanism also was provided in the paper.     

BCR certified reference materials for reactor neutron dosimetry

A series of reference materials intended for use as activation or fission monitors for neutron fluence rate measurements has been prepared by the Joint Research Centre of the European Commission. Certification has been carried out by expert European laboratories and distribution of the certified reference materials (CRMs) is through the BCR programme of the Commission. The list (18 CRMs) includes materials to cover the complete energy spectrum, and suitable for different irradiation times. Fission monitors are ²³⁸UO₂ or ²³⁷NpO₂ in the form of microspheres. Activation monitors are high purity metals (Ni, Cu, Al, Fe, Nb, Rh, or Ti), certified for interfering trace impurities, or dilute aluminium-based alloys, where aluminium is chosen as a suitable matrix for reducing the neutron self-shielding effect. Newly certified materials are IRMM-530R Al-0.1%Au, replacing the exhausted IRMM-530 material, used as comparator for k ₀-standardization, and three new Al-Co alloys (0.01-1%Co). Two others, in the process of certification are Al-0.1%Ag and Al-2%Sc for thermal and epithermal fluence rate measurements. Other candidate reference materials currently being certified are two uranium-doped glass intended for dosimetry by the fission-track technique.     

Determination of aluminium contents in selected food samples by instrumental neutron activation analysis

Food and food products are the main sources of Aluminium entering the human body. In order to know aluminium contents in food and food products, selected 26 samples from local market were analyzed by instrumental neutron activation analysis (INAA) using reactor neutrons and high resolution gamma-ray spectrometry. INAA using 1,779 keV γ-ray of ²⁸Al (2.24 min) was used for aluminium concentrations in the range of 33–529 mg kg^?1. Two NIST standard reference materials (SRMs) and two IAEA reference materials (RMs) were analyzed by INAA for quantification of aluminium as a part of method validation.     

Development and implementation of Høgdahl convention and Westcott formalism for k<Subscript>0</Subscript>-INAA application at Malaysian Nuclear Agency reactor

The present work shows the development of k₀-instrumental neutron activation analysis (k₀-INAA) method at the Malaysian Nuclear Agency research reactor. To use the k₀-INAA method, two formalisms were regulated according to 1/ν and non-1/ν (n, γ) reaction nuclides. The reactor neutron spectrum parameters, the thermal to epithermal neutron flux ratio (f), the epithermal neutron flux shape factor (α) were measured using the bare bi-isotopic monitor and bare triple monitor methods, respectively, based on the Høgdahl convention. In addition, the modified spectral index \( r(\alpha )\sqrt {T_{n} /T_{0} } \), the Westcott \( g_{Lu} (T_{n} ) \) factor and the absolute neutron temperature T _n parameters were determined using the Westcott formalism. ¹⁷⁶Lu was used as non- 1/ν monitor while ¹⁹⁷Au, ⁹⁶Zr and ⁹⁴Zr were used as 1/ν monitors. The average values of \( r(\alpha )\sqrt {T_{n} /T_{0} } ,\,g_{Lu} (T_{n} ) \) and T _n were determined to be 0.1795 ± 0.0044, 1.9729 ± 0.0234 and 50.12 ± 3.21°C, respectively. The accuracy of the method was evaluated by analysing IAEA-Soil 7, IAEA-SL 1, NBS SRM 1633A-1 and IAEA-Soil 375 as reference materials. The results show an acceptable level of consistency.     

Utilization of a mixed-bed column for the removal of iodine from radioactive process waste solutions

A boron carbide capsule was previously designed and tested by Pacific Northwest National Laboratory (PNNL) and Washington State University (WSU) for spectral-tailoring in mixed spectrum reactors. The presented work used this B₄C capsule to create a fission product sample from the irradiation of highly enriched uranium (HEU) with a fast fission neutron spectrum. An HEU foil was irradiated inside of the capsule in WSU’s 1 MW TRIGA reactor at full power for 200 min to produce 5.8 × 10¹³ fissions. After 3 days of cooling, the sample was shipped to PNNL for radiochemical separations and analysis by gamma and beta spectroscopy. Fission yields for products were calculated from the radiometric measurements and compared to measurements from thermal neutron induced fission (analyzed in parallel with the non-thermal sample at PNNL) and published evaluated fast-pooled and thermal nuclear data. Reactor dosimetry measurements were also completed to fully characterize the neutron spectrum and total fluence of the irradiation.     

Preparation,thermal property,and thermal stability of microencapsulated n-octadecane with poly(stearyl methacrylate) as shell

This study focused on preparation and thermal properties of poly(stearyl methacrylate) shell (PSMA) microcapsules containing n-octadecane as a phase change material (PCM). Pentaerythritol triacrylate (PETA) and divinylbenzene (DVB) were employed as crosslinking agents. The surface morphologies, particle sizes, and distributions of the microencapsulated phase change material (microPCM) were studied by scanning electron microscopy. The thermal properties, thermal reliabilities, and thermal stabilities of the microPCMs were investigated by differential scanning calorimetry and thermal gravimetric analysis. The microPCM with DVB exhibits higher phase change enthalpies of melting (87.9 J g^?1) and crystallization (94.8 J g^?1) and a greater thermal stability in comparison with the microPCM with PETA. The phase change temperatures and enthalpies of the microPCMs varied little after thermal cycles. Thermal images showed that the gypsum board with PSMA/n-octadecane microPCM possessed temperature-regulated property. Therefore, microencapsulated n-octadecane with PSMA as shell has good thermal energy storage and thermal regulation potential.     

Rapid analysis of trinitite with nuclear forensic applications for post-detonation material analyses

Analysis of post-nuclear detonation materials provides information on the type of device and its origin. Compositional analysis of trinitite glass, fused silicate material produced from the above ground plasma during the detonation of the Trinity nuclear bomb, reveals gross scale chemical and isotopic heterogeneities indicative of limited convective re-homogenization during accumulation into a melt pool at ground zero. Regions rich in weapons grade Pu have also been identified on the surface of the trinitite sample. The absolute and relative abundances of the lanthanoids in the glass are comparable to that of average upper crust composition, whereas the isotopic abundances of key lanthanoids are distinctly non-normal. The trinitite glass has a non-normal Nd isotope composition, with deviations of ?1.75 ± 0.60 ε (differences in parts in 10⁴) in ¹⁴²Nd/¹⁴⁴Nd, +2.24 ± 0.75 ε in ¹⁴⁵Nd/¹⁴⁴Nd, and +1.01 ± 0.38 ε in ¹⁴⁸Nd/¹⁴⁴Nd (all errors cited at 2σ) relative to reference materials: BHVO-2 and Nd-Ames metal. Greater isotopic deviations are found in Gd, with enrichments of +4 ± 1 ε in ¹⁵⁵Gd/¹⁶⁰Gd, +4.19 ± 0.75 ε in ¹⁵⁶Gd/¹⁶⁰Gd, and +3.48 ± 0.52 ε in ¹⁵⁸Gd/¹⁶⁰Gd compared to BHVO-2. The isotopic deviations are consistent with a ²³⁹Pu based fission device with additional ²³⁵U fission contribution and a thermal neutron fluence between 1.4 and 0.97 × 10¹⁵ neutrons/cm².     

Evaluation of natural gamma radiation and absorbed gamma dose in soil and rocks of Perambalur district (Tamil Nadu,India)

The activity concentrations and absorbed gamma dose of primordial radionuclides ²³⁸U, ²³²Th and ⁴⁰K were determined employing γ-ray spectrometry in 31 soil samples from the land area earmarked for house construction in Perambalur district and 14 rock samples from quarries that supply stones for the entire district. The soil samples registered relatively a higher mean value of 13.2 Bq kg^?1 for ²³⁸U, 66 Bq kg^?1 for ²³²Th and 340.3 Bq kg^?1 for ⁴⁰K as compared to mean values for rock samples (²³⁸U—8.0 Bq kg^?1; ²³²Th—65.1 Bq kg^?1; ⁴⁰K—199.1 Bq kg^?1). The mean absorbed gamma dose rate for soil (61.4 nGy h^?1) marginally exceeded the prescribed limit of 55 nGy h^?1 while, rocks registered the mean absorbed gamma dose rate of 10.4 nGy h^?1. The mean radium equivalent activity was distinctly higher in soil (130.6 Bq kg^?1) than in rock (20.0 Bq kg^?1). However, these values were lower than the limit (370 Bq kg^?1) set by OECD for building materials. It is evident from the data that the soil and rocks do not pose any radiological risk for house constructions in Perambalur district.     

Average cross-sections for (n,p) reactions on calcium in a fission-type reactor spectrum

The average cross-section in a fission-type reactor spectrum \(\bar \sigma _F\) was experimentally determined for the reactions⁴²Ca(n, p)⁴²K,⁴³Ca(n, p)⁴³K and⁴⁴Ca(n, p)⁴⁴K. Calcium carbonate samples and fast neutron flux monitors were irradiated with and without cadmium shielding in the Thetis reactor (Institute for Nuclear Sciences, Rijksuniversiteit, Gent). The potassium activities induced in the calcium carbonate samples were separated and purified by tetraphenylborate precipitation, after which they were measured with a Ge(Li)-detector of calibrated detection efficiency. On the basis of \(\bar \sigma _F = 0.64\) mb for the reaction²⁷Al(n, α)²⁴Na, the average cross-sections were as follows:⁴²Ca(n, p)⁴²K: 2.82±0.07mb;⁴³Ca(n, p)⁴³K: 1.89±0.05 mb;⁴⁴Ca(n, p)⁴⁴Ca(n, p)⁴⁴K: 0.065±0.003 mb.     

Ion beam irradiation of ZnS dispersed in PMMA and its photocatalytic application

ZnS nanoparticles implanted with 45 keV O⁵⁺ ion beam exhibited 83.6 % degradation of methyl blue in 2 h. This idea was utilized to fabricate nanocomposite system of ZnS and PMMA where ZnS nanoparticles were immobilized in PMMA film and irradiated with 45 keV O⁵⁺ ion beam at particle fluence of 2.5 × 10¹⁵, 1 × 10¹⁶ and 4 × 10¹⁶ particles/cm². These irradiated batches of ZnS nanoparticle immobilized in PMMA batches revealed formation of porous structure characterized by scanning electron microscopy and these batches exhibited 54 % photocatalytic degradation of methyl blue in 80 min which was higher as compared to the pristine ZnS nanoparticles.     

Titanium dioxide nanoparticle based solid phase extraction of trace Alizarin Violet,followed by its specrophotometric determination

Nanometer-sized titanium dioxide (nano-TiO₂) is shown to be a viable material for the preconcentration of Alizarin Violet (AV, a common dye and biological stain). In the preconcentration step, a 5-ring cyclic ester is formed between the ortho-dihydroxy groups of AV and two hydroxy groups of the titanic acid on the surface of the nano-TiO₂. Under optimized conditions, the adsorption capacity of nano-TiO₂ is?~?20 μg?·?mg^?1, the adsorption efficiency is 98 %. The adsorbed AV can be eluted with 5 mL of 5 mol?·?L^?1 5-sulfosalicylic acid with an elution efficiency of more than 91.8 %. The preconcentration factor is 50 in case of 250 mL samples. Spectrophotometric determination of AV in the eluate gives a linear calibration plot in the range between 18.8 μg?·?L^?1 and 10 mg?·?L^?1 and a detection limit (3 s; for n?=?11) of 18.8 μg?·?L^?1. The method is simple and fast. It was successfully applied to the analysis of AV in spiked natural waters, and recoveries were found to range between 94.2 and 97.3 %.

Nanometer-sized titanium dioxide is a viable material for the preconcentration of Alizarin Violet (AV), before its spectrophotometrical determination. The method is simple and fast. It was successfully applied to the analysis of AV in spiked natural waters, and recoveries were found to range between 94.2 and 97.3 %.     

The effect of the addition of polypropylene-grafted SiO2 nanoparticle on the thermal conductivity of isotactic polypropylene

The thermal conductivity of Isotactic polypropylene (iPP)/silica particle (SiO₂, 26 nm) nanocomposite has been investigated. The untreated SiO₂ and iPP grafted onto SiO₂ were dispersed in the iPP (M _w = 2.5 × 10⁵) matrix. The molecular mass of the iPP-grafted chain, M _n, was precisely controlled to be 5.8 × 10³, 1.2 × 10⁴, and 4.6 × 10⁴. It was found that the thermal conductivities of graft-treated nanocomposites were higher than that of untreated SiO₂ composites. This implied that it is possible to achieve even higher thermal conductivity using the graft treatment. A thermal conductivity analysis conducted using a three-phase model, with considerations for thermal conductivity at interfacial layers, showed that the thermal conductivity of the interfacial layer increased significantly when a graft chain was incorporated. Moreover, the thermal conductivity per graft chain was proportional to the 1/2 power of the molecular mass ( \( M_{\text{n}}^{0.5} \) ). The results strongly suggest that the thermal conductivity pathway of interfacial layer was the main chain direction of iPP-grafted molecular chains.     

Utilization of low-cost sorbent for removal and separation of 134Cs, 60Co and 152+154Eu radionuclides from aqueous solution

Eggshell material was used as low-cost and eco-friendly biosorbent for removal of ¹³⁴Cs, ⁶⁰Co and ^152+154Eu radionuclides from aqueous solution. The eggshell material was calcined at 500 and 800 °C, and then characterized. Comparative studies on the natural and calcined eggshell for sorption of the three radionuclides were carried out. It was found that, the uptake is in the order: Eu(III) > Co(II) > Cs(I). Further, column chromatography was used in separation of ¹³⁴Cs, ⁶⁰Co and ^152+154Eu using 0.15, 0.2 and 0.5 mol/l nitric acid, respectively. Eggshell material can be considered as a promising material for separation of radionuclides from radioactive waste solution.     

Kinetic behavior of immobilized aldolase in a continuously stirred tank reactor

Aldolase was bound covalently to CNBr-activated Sepharose. The immobilized enzyme was investigated in a continuously stirred tank reactor (CSTR). In spite of the complexity of the system, the stationary performance of the CSTR can be described by a very simple equation. There are essentially two control parameters,V _m andK _eq— that govern the stationary performance of the aldolase reactor.     

Analytical possibilities of the BOR 60 reactor

The analytical potential of a fast neutron reactor has been studied. The maximum density of the neutron flux is 2.5·10¹⁵ cm^–2·s^–1 at a maximum energy of 450 keV. The determining reaction for activation analysis at this reactor is an (n, n) reaction. The possibility of selectively determining Pb, Hg, Cd, Au, Ag and other elements at a level of 10^–5–10^–8% is demonstrated. This activation technique allows rapid determination of these elements.     

Radiolanthanide-labeled HA particles in the treatment of rheumatoid arthritis: ready-to-use cold kits for rapid formulation in hospital radiopharmacy

Hydroxyapatite (HA) [Ca₁₀(PO₄)₆(OH)₂] particles radiolabeled with a variety of β^? emitting lanthanide radionuclides and also pseudolanthanide ⁹⁰Y have been proposed for the treatment of arthritis. A ready-to-use cold kit of HA particles (1–10 µm size) was developed for fast and convenient formulation of radiolanthanide-labeled HA particles at hospital radiopharmacy. Six radionuclides namely, ¹⁶⁹Er, ¹⁷⁷Lu, ¹⁵³Sm, ¹⁶⁶Ho, ¹⁴²Pr and ⁹⁰Y, having β^? emissions of a wide range of energy [E_β(max) = 0.34–2.28 MeV] were identified and produced by thermal neutron activation. Clinical doses of HA particles labeled with these radionuclides were prepared in high yield (>97 %) and radiochemical purity (>99 %) using the cold kits. Pre-clinical studies of ¹⁷⁷Lu–HA carried out in Wistar rats bearing arthritis in knee joints revealed no leakage of the activity from the joints. In preliminary clinical investigation using 333 ± 46 MBq doses of the same preparation, significant improvement in the disease conditions was reported in patients with chronic rheumatoid arthritis of knee joints.     

Moderator for neutron activation with the photoneutrons produced by a LINAC

Using Monte Carlo methods a polyethylene moderator has been designed to induce activation using the photoneutrons field of a 15 MV linear accelerator for radiotherapy. In the calculations the photoneutron spectrum at 1 m from the isocenter was used as a source term and the neutron spectra were calculated in the center of different size cylindrical moderators. The best size was selected defining the thermal-to-fast-neutron ratios as a figure of merit. The moderator was built and its performance was evaluated by inducing the activation of Mn dissolved in water, silver coins and souvenir coin. The thermal neutron fluence rate was determined with the Mn samples being 9.96 × 10⁵ cm^?2 Gy _x ^?1 .