A method for bulk hydrogen analysis based on transmission and back scattering of fast neutrons

A method was proposed for bulk hydrogen analysis. It is based on simultaneous detection of transmitted fast neutrons and back scattered thermal neutrons from the investigated samples by ³He detectors. The fast neutron beams were obtained from ²⁵²Cf and Pu–Be neutron sources. The experimental set-up as well as samples preparation were described. Incident thermal neutrons beams obtained from either ²⁵²Cf or Pu–Be sources, were used to investigate the samples by neutron backscattering. The results obtained from transmission and backscattering of fast neutrons were compared and discussed. The advantage and capabilities of the proposed method were presented. The results obtained using fast neutron beams are more sensitive than those obtained using thermal neutron beams. Validation procedures were proposed to credit the results.     

Measurements of fast neutron capture cross sections on <Superscript>63</Superscript>Cu and <Superscript>186</Superscript>W

Neutron capture cross sections on ⁶³Cu and ¹⁸⁶W were measured by fast neutron activation method at neutron energies from 1 to 2 MeV. Monoenergetic fast neutrons were produced by ³H(p,n)³He reaction. Neutron energy spread by target thickness, which was assumed to be the main factor of neutron energy spread, was estimated to be 1.5% at neutron energy of 2.077 MeV. Neutron capture cross sections on ⁶³Cu and ¹⁸⁶W were calculated by reference comparison method on those of ¹⁹⁷Au(n,γ). Not only statistical errors of gamma-counts from samples but also systematic errors in the counting efficiency for HP Ge detector and the uncertainty of areal density of samples were considered in calculating neutron capture cross section. Estimated neutron capture cross sections on ⁶³Cu and ¹⁸⁶W were also compared with ENDF-6 data.     

Determination of manganese in ores by activation analysis using a californium-252 neutron source

A²⁵²Cf neutron source has been used to analyse manganese in ores such as pyrolusite, rodonite (manganese silicate) and blends used in dry-batteries. Samples with about 150 mg and standards of manganese dioxide were irradiated for about 20 min and counted using a well-type NaI(Tl) scintillation counter and scaler, with or without pulse-height discriminator between the detector and the scaler. The interferences of nuclear reactions⁵⁶Fe(n,p)⁵⁶Mn and⁵⁹Co(n,α)⁵⁶Mn were studied, as well as problems in connection with neutron shadowing during irradiation, gamma-rays attenuation during counting and influence of granulometry of samples. Some of the samples were also analysed by wet-chemical method (sodium bismuthate) in order to compare results.     

Design and simulation of neutron radiography system based on 241Am–Be source

Neutron imaging is extended rapidly as a means of non-destructive testing (NDT) of materials. Various effective parameters on the image quality are needed to be studied for neutron radiography system with good resolution. In the present study a portable system of neutron radiography has been designed using ²⁴¹Am–Be neutron source. The effective collimator parameters were calculated to obtain relatively pure, collimated and uniform neutron beam. All simulations were carried out in two stages using MCNPX Monte Carlo code. In the first stage, different collimator configurations were investigated and the appropriate design was selected based on maximum intensity and uniformity of neutron flux at the image plane in the outlet of collimator. Then, the overall system including source, collimator and sample was simulated for achieving radiographic images of standard samples. Normalized thermal neutron fluence of 2.61×10^?5 cm^?2 per source particle with n/γ ratio of 1.92×10⁵ cm^?2 μSv^?1 could be obtained at beam port of the designed collimator. Quality of images was assessed for two standard samples, using radiographic imaging capability in MCNPX. The collimated neutron beam in the designed system could be useful in a transportable exposure module for neutron radiography application.     

Neutron damage of hexagonal boron nitride: h-BN

Hexagonal boron nitride (h-BN) was neutron damaged at an integral flux of 2.40 × 10¹² n cm⁻² s⁻¹ for 1, 2, 3 and 4 h. The h-BN samples undergo a transition from sp² to sp³ hybridization as a consequence of the neutron induced damage with the formation of cubic boron nitride (c-BN) spots, as suggested both by FT–IR and Raman spectroscopy. In addition to c-BN, also a certain degree of amorphization is achieved by h-BN already at the lowest neutron fluence of 8.64 × 10¹⁵ n cm⁻² as clearly evidenced by Raman spectroscopy. The Wigner or stored energy to the radiation-damaged h-BN samples was studied by DSC and also in this case there was a clear evidence that the neutron damage was partly irreversible and insensitive to the thermal annealing up to 630 °C. Electron spin resonance (ESR) was employed to further study the structural defects induced by the neutron bombardment of h-BN. Two kinds of paramagnetic defective structures centered on ¹¹B atoms were identified.

     

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.     

Determination of Si,Al, Ti,Fe and Zr in glass sand samples by 14 MeV neutron activation analysis

Fast neutron activation analysis technique was applied for the determination of Si, Al, Ti, Fe and Zr in glass sand rock samples. The samples and standards were irradiated with a mono-energetic neutron flux of 10⁸n · cm^– · s^–1. Pneumatic facility was used. The gamma activities from samples and standards were counted using a 30 cm³ Ge(Li) detector, with FWHM of 2.9 keV at 1.332 MeV, coupled to an on-line computer facility.     

Non-destructive neutron activation determination of silver and antimony in bismuth byγ.γ-coincidence spectrometry

A non-destructive neutron activation method was developed for the determination of silver and antimony in high-purity bismuth ; γ,γ-coincidence counting of ^110mAg or ¹²⁴Sb was applied. The activity of the matrix or other impurities did not interfere. When 1-g samples were irradiated at a neutron flux of 4·10¹¹ n cm^-2 sec^-1 for 25 days, the sensitivity of the method was in the p.p.b. region.     

Determination of total mercury and methylmercury in hair samples from residents of Kuala Lumpur, Malaysia by neutron activation analysis

to estimate the level of total mercury and methylmercury in Kuala Lumpur residents, 400 hair samples were analysed by neutron activation analysis. Separation of methylmercury from hair samples were carried out prior to neutron activation. The average level of total mercury and methylmercury in hair samples were 3.38 mg^.kg^-1 (in range of 0.59-18.73 mg^.kg^-1) and 1.13 mg^.kg^-1 (in range of 0-4.65 mg^.kg^-1), respectively. The average percentage ratio of methylmercury to total mercury was 31.15% (in range of 0 to 75.81%). This revised version was published online in August 2006 with corrections to the Cover Date.     

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.     

The determination of rhenium in flue dust by neutron activation analysis

Rhenium was determined by radiochemical neutron activation analysis in flue dust samples of complex chemical composition. Two separation techniques were used; distillation from perchloric-hydrochloric acid solution and hydroxide scavenging followed by extraction with tetraphenylarsonium chloride in chloroform. ¹⁸⁶Re and ¹⁸⁸Re were measured with a Ge(Li) low-energy photon detector. Chemical yields were determined by reactivation. For a neutron flux of 5·10¹¹ n cm^-2 s^-1, an irradiation time of 6 h and samples of ca. l g, down to 50 p.p.b. of rhenium could be determined.     

Simplified evaluation of214Bi/238U activity ratios in fossil bones using non-destructive activation analysis

A simple and reliable method has been developed for the evaluation of radioactive disequilibrium state in fossil bones. The fossil bone samples were irradiated with an extremely low neutron fluence, together with a standard pitchblende prevailing the secular equilibrium among the uranium series. The²³⁹Np activity induced from²³⁸U in both samples were adjusted to be gamma-ray spectrometrically nearly equivalent to the naturally occurring radioactivities by controlling the neutron flux and cooling time. Using single gamma-ray spectrometry of the irradiated samples, the determination of²¹⁴Bi/²³⁸U in a fossil bone was carried out by comparing the photopeak ratios of²¹⁴Bi /609 keV/ and²³⁹Np /278 keV/ instead of²³⁸U with the same ratios from the pitchblende standard sample.     

Determination of metals in alloys by neutron capture gamma-ray spectrometry

A collimated neutron beam capable of providing a thermal neutron flux of 4.75·10⁷ n·cm⁻²·sec⁻¹ has been used to analyze alloy samples of 1–5 g during relatively short irradiation times of 30 min by the use of neutron capture gamma-ray spectrometry. The analyses were performed by using a mathematical treatment that relates the count ratio of every constituent present in the matrix with the concentration and thus it requires no standards. The technique was applied to the analysis of steel and gold alloy samples. Errors ranged from 0.8%–10%.     

Performance comparison of 2.8 MeV and <Superscript>241</Superscript>Am-Be neutrons based moisture measurement setups

Performance of a ²⁴¹Am-Be neutron source-based and 2.8 MeV neutrons-based moisture measurement setups have been compared using Monte Carlo simulation. In the setup fast neutrons transmitted through the sample were detected by a fast neutron detector, which was placed behind a massive long double truncated collimator. The setup geometry was optimized to detect maximum effect of 1–7 wt.% moisture on the neutron intensity transmitted through the sample. The yield of neutrons transmitted through concrete, coal, wood and soil samples containing 1–7 wt.% moisture was calculated for 2.8 MeV neutrons and neutrons from an ²⁴¹Am-Be source. The slopes of the fast neutron intensities transmitted through the samples vs. their moisture contents are very sensitive to the neutron energy and the sample composition. Higher slopes have been observed for the samples with larger bulk density. The slopes of fast neutron yield show dependence on the incident neutron energy. Larger slopes have been observed for neutrons with samller energy. Due to the overall large slopes of the transmitted intensity data of the samples for 2.8 MeV neutrons, it is expected to achieve better sensitivity in moisture measurements for a 2.8 MeV neutrons based moisture setup.     

Estimation of fast neutron fluence released by the Tokai-mura criticality accident from 54Mn in soils collected from the JCO grounds

To estimate fast neutron fluence released by the JCO criticality accident in Tokai-mura, ⁵⁴Mn, which is produced by the ⁵⁴Fe(n,p)⁵⁴Mn reaction, was determined in soil samples by ultra low background g-ray spectrometry after radiochemical separation. Activities of ⁵⁴Mn, using 30-200 g soil samples were detected in soil samples from 13 points in all directions within a 20 m zone from the precipitation vessel we have used. The levels of ⁵⁴Mn ranged from 0.015 to 2.12 mBq/g soil. By using these data and MCNP, the fast neutron fluences were estimated.     

Neutron activation analysis of manganese in Japanese iron reference standard materials, with large amounts of manganese

Effects of pile-up and neutron self-shielding were experimentally corrected in INAA of manganese and vanadium in iron and ferro- and silicon manganese samples supplied by the Japan Iron and Steel Federation. A pile-up correction curve was drawn as a function of total gamma count rate for each Ge detector by a ⁶⁰Co source set at a fixed position and ²⁴Na and ¹³⁷Cs sources moved to vary the gamma-ray intensities. The neutron self-shielding effect was examined by varying the weight of irradiated samples. The effect was negligible for iron samples containing a few percent manganese when samples of less than 0.2 g were irradiated. Good results were obtained for silicon and ferro-manganese by repeated analyses of samples less than 0.02 g in order to avoid the influence of the neutron self-shielding and sample inhomogeneity. A good result can be also obtained from the maximum point of a quadratic equation derived from the relationship between apparent manganese concentrations and sample weight in the range clearly affected by self-shielding. The water dilution after acid dissolution was also effective for samples of high manganese content if special caution was given to regulate the concentration of chloride having a large neutron absorption cross section.     

Serum aluminum determination by instrumental neutron activation analysis in long-term haemodialysis patients

Serum aluminum levels were determined by instrumental neutron activation analysis in 31 patients undergoing long-term haemodialysis. Aluminum-28 1.778 MeV (T _1/2=2.24 min) γ-rays produced by the thermal neutron reaction²⁷Al(n,γ)²⁸Al were detected. Successive irradiation of the samples at epithermal neutron fluence was performed to correct for the interference from the fast neutron reaction³¹P(n,α)²⁸Al. Serum aluminum level in this group of subjects was adequately represented by a lognormal distribution with a mean and variance of 16.5 μg/l and 16.8 μg/l, respectively. The results obtained were found to be in agreement with serum aluminum determination performed by electrothermal atomic absorption spectrophotometry (r ²=0.97). Instrumental neutron activation can provide a rapid technique to routinely monitor long-term haemodialysis patients in order to identify individuals at greater risk to develop aluminum toxicity.     

Apparent Neutron Emissions from Polyethylene Capsules during Neutron Activation and Delayed Neutron Counting

At Imperial College uranium is determined at very low levels in environmental samples by delayed neutron counting. High density polyethylene capsules are used for transporting the samples in a pneumatic transfer system, from the reactor, to the neutron detection rig. The detector has a background of 3.6 counts per minute, however the empty transport capsules produce 11.0 counts above the background. We have determined that neither uranium in the polyethylene or from contamination of the capsule nor ¹⁷N from the ¹⁷O(n,p)¹⁷N and ¹⁸O(n,d)¹⁷N reactions are the cause. This paper reports the results of the investigation into the source of the extra counts.     

The leachability of some natural and man-made radionuclides from soil and sediments on acid attack

A radiotracer study was made of the leachability of some natural and man-made radionuclides from soils and sediments subjected to attack by various acid mixtures. Particular attention was paid to the nuclides²³⁸U,²³²Th, Pa (as²³³Pa) and Np (as²³⁹Np), since for these neutron activation can be used to study recoveries and/or to induce in situ radionuclides in samples. Thus conventional NAA allowed determination of total²³⁸U and²³²Th instrumentally, and also enabled analysis of leachates and residues by radiochemical or instrumental NAA. Activation of these nuclides produced samples endogenously labelled with²³³Pa and²³⁹Np whose behaviour on acid dissolution/leaching could be examined. Furthermore, comparison of neutron irradiated and non-irradiated samples allowed us to investigate the possibility of increased leachability induced by nuclear recoil reactions; however, this effect was negligible. We also investigated the acid leaching of americium adsorbed on the surface of sediment. In general, unsatisfactory leaching recoveries were found for²³³Pa,²³⁹Np and²³²Th for most materials, indicating the need for total dissolution procedures.     

Determination of thorium concentrations in soil and sand samples using instrumental neutron activation analysis

Thorium along with its daughter products present in the soil is one of the major contributors to the external gamma dose in the environment. To establish the dose levels, quantification of thorium contents in soil samples is very important. As a part of pre-operational environmental radiological surveillance, a total of 23 soil and six sand samples were collected from different locations around the proposed nuclear power plant site of Jaitapur, Maharashtra. Thorium concentrations in these samples were determined by instrumental neutron activation analysis (INAA). Samples were irradiated with neutrons in Apsara reactor at a neutron flux of?~5?×?10¹¹ cm^?2 s^?1 and radioactive assay was carried out using high resolution gamma ray spectrometry. Relative method of INAA was used for quantification of thorium utilizing 311.9?keV gamma ray of ²³³Pa, the daughter product of ²³³Th. The concentrations of thorium in the soil and sand samples were in the ranges of 4.0?C18.8 and 1.2?C6.2?mg?kg^?1 respectively.