Rare earth element (REE) in surface mangrove sediment by instrumental neutron activation analysis

A study is carried out on the concentrations of rare earth element (REE) elements present in surface mangrove sediments from 10 locations throughout west coast Malaysia. In carrying out the analysis, the best and most convenient method being the instrumental neutron activation analysis (INAA). Samples were obtained, dried, crushed to powdery form and samples prepared for INAA. All the samples for analysis were weighted approximately 150 mg for short irradiation and 200 mg for long irradiation time. As calibration and quality control procedures, blank samples, standard reference material SL-1 were then irradiated with thermal neutron flux of 4 × 10¹² cm^?2 s^?1 at the MINT TRIGA Mark II research reactor which operated at 750 kW by using a pneumatic transport facility. The REE elements of surface sediment samples in this study are Dy, Sm, Eu,Yb, Lu, Tb, La and Ce. It was found that the level of concentrations of all the REE elements varies in the range (0.35–117.4 mg/kg). The geochemical behavior of REEs in surface sediments and normalized pattern (chondrite and shale) has been studied. The degree of sediments contaminations were computed using an enrichment factor. The results showed that the enrichment factor varied in the range (0.75–6.75).     

Radioargon production through the irradiation of calcium oxide

The production of radioargon through the irradiation of CaO was performed in an in-core facility within the The University of Texas at Austin MARK II TRIGA reactor. The major radioargon isotope produced was ³⁷Ar via the ⁴⁰Ca(n,α)³⁷Ar reaction pathway. The CaO powder was vacuumed sealed in a quartz ampoule. The sample was irradiated in a Cd-lined Al alloy canister for 2 h at 500 kW. After the irradiation, the sample was counted using an HPGe detector five times with increasing count times. ⁴¹Ar, ⁴²K, ⁴³K, and ⁴⁷Ca were detected in the spectra. The induced activities of ³⁷Ar, ³⁹Ar, ⁴¹Ar, ⁴²K, ⁴³K, and ⁴⁷Ca were calculated using a 63-group energy-dependent neutron flux determined utilizing a neutron energy flux profile calculated from a MCNPX model of the TRIGA reactor core. The production ratios generated from the model allowed for the estimation of induced ³⁷Ar and ³⁹Ar activities through the measured induced activities of ⁴¹Ar, ⁴²K, ⁴³K, and ⁴⁷Ca.     

Research reactors as sources of atmospheric radioxenon

Radioxenon emissions of the TRIGA Mark II research reactor in Vienna were investigated with respect to a possible impact on the verification of the Comprehensive Nuclear Test-Ban-Treaty. Using the Swedish Automatic Unit for Noble Gas Acquisition (SAUNA II), five radioxenon isotopes ¹²⁵Xe, ^131mXe, ^133mXe, ¹³³Xe and ¹³⁵Xe were detected, of which ¹²⁵Xe is solely produced by neutron capture in stable atmospheric ¹²⁴Xe and hence acts as an indicator for neutron activation processes. The other nuclides are produced in both fission and neutron capture reactions. The detected activity concentrations ranged from 0.0010 to 190 Bq/m³. The source of the radioxenon is not yet fully clarified, but it could be micro-cracks in the fuel cladding, fission of ²³⁵U contaminations on the outside of the fuel elements or neutron activation of atmospheric Xe. Neutron deficient ¹²⁵Xe with its highly complex decay scheme was seen for the first time in a SAUNA system. In many experiments the activity ratios of the radioxenon nuclides carry the signature of nuclear explosions, if ^131mXe is omitted. Only if ^131mXe is included into the calculations of the isotopic activity ratios, the majority of the measurements revealed a “civil” signature (typical for a NPP). A significant contribution of the TRIGA Vienna to the global or European radioxenon inventory can be excluded. Due to the very low activities, the emissions are far below any concern for human health.     

Design,construction and characterization of a prompt gamma activation analysis facility at the Oregon State University TRIGA<Superscript>®</Superscript> reactor

A prompt gamma neutron activation analysis facility has been designed, built, and characterized at the Oregon State University TRIGA^® reactor. This facility was designed for versatile multi-elemental analyses. The facility utilizes the leakage neutrons originating from beam port #4 of the Oregon State University TRIGA^® reactor. The neutrons are collimated through a series of lead and Boral^® collimators, and filtered through both a bismuth filter and single-crystal sapphire. Samples are irradiated in a sample chamber outside the biological shielding of the reactor, and the resulting gamma radiation produced from neutron interactions within the sample is monitored using a high-purity germanium detector (HPGe). The thermal and epithermal neutron fluxes were measured using gold-foil irradiations and found to be 2.81 × 10⁷ and 1.70 × 10⁴ cm^?2 s^?1, respectively. The resulting cadmium ratio was 106. Measured detection limits for boron, chlorine, and potassium in a NIST SRM 1571 orchard leaf were 5.6 × 10^?4 mg/g, 8.2 × 10^?2 mg/g, and 1.0 mg/g, respectively. Detection limits for additional elements and samples are presented.     

Utilization of pneumatic carrier facility of Dhruva reactor for trace element determination by neutron activation analysis

The pneumatic carrier facility (PCF) of Dhruva reactor is being extensively used for neutron activation analysis (NAA) studies pertaining to research work as well as routine sample analysis. It is useful for the determination of trace elements using short and medium half-lives radioisotopes produced in neutron activation with available higher neutron flux (~5 × 10¹³ cm^?1 s^?1). Solid samples placed in high density polypropylene capsule, are irradiated for 1 min duration and radioactive assay is carried out by high resolution gamma ray spectrometry. Design aspects of PCF and various applications to samples of diverse matrices using NAA are presented.     

A low cost neutron capture prompt gamma-ray analysis facility at a research reactor

A low cost neutron capture prompt gamma activation analysis facility has been constructed at The University of Michigan's Pheonix Memorial Laboratory. Although the neutron beam used has a fairly large epithermal component (Cd ratio 7.1), background levels are low enough to result in satisfactory measurement of over 16 different elements. For the elements of greatest sensitivity (samarium, boron, gadolinium, and cadmium) minimum detectable levels of 3.6·10⁻⁵ to 1.4·10⁻⁵ gram for a one hour measurement are possible. The fast neutrons incident to the detector were found to be minimal. Estimates of up to 3 years of continuous operation before measurable damage is expected.     

Determination of neutron flux parameters in PUSPATI TRIGA Mark II Research Reactor, Malaysia

In standardization NAA, it is necessary to characterize the neutron spectrum parameters such as epithermal neutron flux shape factor (α), thermal to epithermal neutron flux ratio (f), thermal neutron flux (φ _th) and epithermal neutron flux (φ _epi) in the irradiation facility to determine the concentration of an element in the sample using absolute and k ₀ standardization methods. The α and f were determined using Cd-ratio multi monitor method using experimental data obtained in PUSPATI TRIGA Mark II research reactor at four irradiation positions (10, 20, 30 and 40) of the rotary rack. The calculated values of α and f ranged from 0.006 to 0.0281 and 18.56 to 19.12 respectively. The average values of φ _th and φ _epi were found as 2.33 × 10¹² and 1.23 × 10¹¹ n cm^?2 s^?1 respectively. Moreover, a comparison of the neutron flux parameters in the present study shows an acceptable level of consistency with those of previous studies.     

Elemental analysis of limestone by instrumental neutron activation analysis

Limestone samples from Assuit Governrate in Upper Egypt were subjected to elemental analysis by instrumental neutron activation analysis and X-ray fluorescence techniques. The samples were properly prepared together with their standards and simultaneously irradiated in a neutron flux of the order 7 × 1011 n/cm² s using TRIGA research reactor at Mainz. After activation the samples were subjected to γ-ray spectrometry using a high purity germanium detection system and computerized multichannel analyzer. Nineteen elements: Na, Ca, Mn, Fe, Sc, Cr, Co, Zn, Sn, La, Ce, Nd, Eu, Sm, Yb, Lu, Hf, Th and U were analyzed. X-ray fluorescence spectrometry have been also used. The presence of any elements in higher or lower levels in certain limestone samples is contingent on the occurrence of its bearing minerals, nature of parent sediments and depositional environments of these sediments. The major elements in the samples were also observed to be among the elements that had high enrichment factors in the study of suspended dust particulate within and around cement industries. This confirms cement as the major contributor to the airborne particulate matter in the environs.     

Application of INAA for phyto-accumulation study of selenium by chickpea plant

The phyto-accumulation efficacy of selenium (Se) from soil by chickpea plant is reported. Chickpea plants were grown in soil having different concentrations (1–4 mg kg^?1) of Se. Samples of soil and different parts of chickpea plants in Se rich soil were analyzed for determination of Se concentrations by instrumental neutron activation analysis (INAA). Samples were irradiated in self-serve facility of CIRUS reactor, BARC, Mumbai at a neutron flux of the order of 10¹³ cm^?2 s^?1. The gamma activity at 264.7 keV of ⁷⁵Se (119.8 d) was measured using a 45% relative efficiency HPGe detector coupled to MCA. Dependence of Se distribution in soil and plants on its spiking concentration was evaluated in this work. The Se concentrations determined in plant parts grown in control soil and in soil spiked with Se (4 mg kg^?1) are in the range of 0.6–0.8 and 65–68 mg kg^?1 respectively.     

Experimental and Monte Carlo simulation on new manganese bath facility for absolute neutron source emission rate measurement at ENEA-INMRI

This work deals with the absolute measurement of the neutron emission rate from a ²⁴¹Am–Be source by means of the manganese sulphate bath technique, which is the principal method for the absolute determination of the neutron emission rate from radionuclide neutron sources. The facility consists of a spherical container filled with an aqueous solution of manganese sulphate with a ²⁴¹Am–Be neutron source placed at the center. As well known, neutrons from the source, after having been thermalized by the aqueous solution, undergo neutron capture by hydrogen, manganese, sulphur, and oxygen nuclei, thus inducing a certain activity to the solution. Subsequent gamma spectrometry measurements of ⁵⁶Mn activity generated by ⁵⁵Mn neutron activation allows to determine the neutron emission rate of the source, The experimental activity has involved a variety of measurement techniques and calculation procedures, ranging from neutron reactor activation to liquid scintillation counting and Monte Carlo calculations. Neutron activations of ⁵⁵Mn samples has been carried out with the TRIGA reactor of the ENEA-Casaccia Research Centre, and ⁵⁶Mn activated samples were subsequently characterized by liquid scintillation counting, in order to obtain reference standards for the calibration of the NaI(Tl) scintillation detectors utilized to record gamma-ray emission from ⁵⁶Mn. Monte Carlo calculations, carried out by the MCNPX code, were required to calculate neutron transport within the sulphate manganese bath, in particular to determine ⁵⁵Mn neutron capture probability, and (n, α) and (n, p) concurrent reactions, as well as the neutron leakage. Such a procedure has allowed to maintaining the neutron emission rate uncertainty well below 1 %. All the measurements have been carried out at the ENEA-Casaccia Research Centre by the Italian National Institute of Ionizing Radiation Metrology.     

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.     

Determination of neutron fluence using the plot of isotope variation of cadmium and gadolinium

The method described to determine the neutron fluence is based on the plot of the isotopic variation of Cd and Gd subjected to neutron irradiation in a research reactor. The isotopic ratios are measured by thermal ionization mass spectrometry. The results indicate that the fluence values obtained, using the variation in the ratios¹¹⁴Cd/¹¹³Cd,¹⁵⁶Gd/¹⁵⁵Gd and¹⁵⁸Gd/¹⁵⁷Gd show standard deviations varying from 0.3 to 6.6%. These values agree with the extrapolated values calculated using the short time Au activation method. The method appears to be useful for determining paleo neutron flux in natural samples and irradiated fuels.     

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.     

Large sample neutron activation analysis of dross for gold and silver

Large sample neutron activation analysis of dross from India Government Mint, Mumbai was carried out for quantification of gold (Au) and silver (Ag) using graphite reflector position of Advanced Heavy Water Reactor critical facility at Bhabha Atomic Research Centre, Mumbai. The k ₀-based internal monostandard NAA was used to calculate concentration ratios of Au and Ag with respect to sodium (Na), which was used as an internal monostandard. The concentration ratio values of Au to Na of varying mass of dross showed that mass ≥2 g was the representative sample size for analysis. Concentrations of gold and silver were found to be in the range of 200–400 and 1200–1700 mg kg^?1, respectively in three different samples.     

Development of a fast-pneumatic sample transfer system for DT neutron activation analysis at the University of Michigan

The development of an automated pneumatic transfer system used to quickly acquire data from materials irradiated with a deuterium–tritium (DT) neutron generator is described in this paper. This system was designed to gather data on short-lived activation and fast-fission products, and was used to characterize the generator’s neutron field. The average sample transit time between irradiation and data acquisition is 363.9 ms at an average velocity of 30.92 m/s (101.3 ft/s). The neutron flux profile as a function of depth into the sample capsule is shown to decrease exponentially, having a maximum flux value of 5.662 × 10⁸ ± 0.056 × 10⁸ n/cm² s. The average DT neutron energy in the system’s sample geometry was determined to be 14.250 ± 0.011 MeV using a unique zirconium–niobium “sandwich” technique. A flux surface equation is also presented as a function of accelerator voltage and deuterium beam current. Methods of analysis are discussed with a proof of a linear flux profile assumption for thin foils.     

A feasibility study to measure low levels of boron in selected Canadian and Japanese foods by prompt gamma activation analysis using the JAEA JRR-3 facility

A feasibility study was carried out to evaluate the potential of the thermal neutron capture prompt-gamma activation analysis (PGAA) for the measurement of low levels of boron in selected Canadian and Japanese foods using the PGA facility at the JRR-3 reactor of the Japan Atomic Energy Agency (JAEA) in Tokai. A method was optimized for this purpose. It is rapid and can be used without any chemical separation. The precision and accuracy of the method are good. The detection limit is around 0.5 mg kg^?1.     

Determination of concentrations of U and Th in their mixed oxides by INAA methods

Uranium and thorium mixed oxides are being prepared using natural U and Th for studies on fuels for Advanced Heavy Water Reactors, wherein composition of U and Th is specific and requires strict control in terms their contents and homogeneity. Chemical quality control necessitates accurate and precise compositional characterization of the fuel material by a suitable analytical method. Among various analytical methods for U and Th, instrumental neutron activation analysis (INAA) is one of the best methods for their simultaneous determination without chemical dissolution and separation. INAA methods using reactor neutrons namely thermal NAA and epithermal NAA were standardized for the determination of U and Th in their mixed oxides. Standards, synthetic samples and U–Th mixed oxide samples, prepared in cellulose matrix, were irradiated at pneumatic carrier facility of Dhruva reactor as well as at self serve facility of CIRUS reactor under cadmium cover (0.5 mm). Radioactive assay was carried out using a 40% relative efficiency HPGe detector. Both activation and daughter products of ²³⁸U (²³⁹U and ²³⁹Np) and ²³²Th (²³³Th and ²³³Pa) were used for their concentration determination. The method was validated by analyzing synthetic samples of 6–48%U–Th mixed oxides. The standardized method was used for the concentration determination of U and Th in 4–30%U–Th mixed oxide samples. Results of U and Th concentrations including associated uncertainties obtained from the INAA methods are presented in this paper.     

Radioxenon production through neutron irradiation of stable xenon gas

The Spectral Deconvolution Analysis Tool (SDAT) software was developed to improve counting statistics and detection limits for nuclear explosion radionuclide measurements. SDAT utilizes spectral deconvolution spectroscopy techniques and can analyze both β–γ coincidence spectra for radioxenon isotopes and high-resolution HPGe spectra from aerosol monitors. The deconvolution algorithm of the SDAT requires a library of β–γ coincidence spectra of individual radioxenon isotopes to determine isotopic ratios in a sample. In order to get experimentally produced spectra of the individual isotopes, we have irradiated enriched samples of ¹³⁰Xe, ¹³²Xe, and ¹³⁴Xe gas with a neutron beam from the TRIGA reactor at The University of Texas. The samples were counted in an Automated Radioxenon Sampler/Analyzer (ARSA) style β–γ coincidence detector. The spectra produced show that this method of radioxenon production yields samples with very high purity of the individual isotopes for ^131mXe and ¹³⁵Xe and a sample with a substantial ^133mXe to ¹³³Xe ratio.     

Production of a <Superscript>173</Superscript>Lu target for neutron capture cross section measurements

Lutetium has been used as a radiochemistry detector to measure neutron fluence in NTS tests. A measure of the neutron capture cross sections on ¹⁷³Lu is needed to improve the interpretation value of the Lu radiochemistry isotopic ratios. A natural hafnium target was irradiated with protons to produce neutron poor lutetium radioisotopes. The short lived species were allowed to decay prior to chemical processing resulting in predominantly ¹⁷³Lu with a small amount of ¹⁷⁴Lu. This material was deposited on a titanium foil for use in the neutron capture cross section measurement.     

Fabrication and characterization of an irradiation facility for large-sample geometry

An irradiation facility consisting of a modified beam port shielding plug has been designed, fabricated built and characterized for use in irradiating non-standard sample geometries. The shielding plug features a graphite moderator at the core end with a hole, or “well” drilled of sufficient diameter and depth to accommodate an eight ounce (227 gram) sample bottle. Added shielding behind the graphite consists of castable neutron- and -gamma-ray shielding. The modified shielding plug can be removed relatively quickly from its irradiation position to minimize personnel exposures. It is mounted in close proximity to the Ohio State University Research Reactor reactor core to allow performance of high-sensitivity neutron activation analysis studies. Using the SAND-II unfolding code, the energy-dependent neutron flux has been measured in the sample irradiation position. When operating at 100 % power, the total flux is 3.9 × 10¹² n/cm²/s. Of this, 55 % is thermal (<0.5 eV), 23 % is epithermal (>0.5 eV, <0.5 MeV), and 22 % is “fast” (>0.5 MeV). This makes the facility suitable for neutron activation studies. Recently it has been used for irradiation of filter papers collected in a study of particulate air pollution in the form of atmospheric particulate matter in an urban environment.