Radio-isotope neutron activation analysis for vanadium,manganese and tungsten in alloy steels

An instrumental neutron activation method for V, Mn and W in alloy steels with a ²⁴¹ Am/Be isotopic neutron source is described. The samples were irradiated to induce the nuclear reactions ⁵¹V(n, γ) ⁵²V, ⁵⁵Mn(n, γ)⁵⁶Mn, and ¹⁸⁶W(n, γ)¹⁸⁷W. The activities were measured with a NaI(TI) detector. Interferences on the measured photopeaks were shown to be negligible by measuring the half-lives of ⁶²V, ⁵⁶Mn and ¹⁸⁷W.These thre elementes were determined in the range 1.5–12.9% in special steels; manganese in the range 0.5–1.6% was measured in cast irons. Calibration was done by comparison with results from wet chemistry and x-ray fluorescence spectrometry. The processing times for the vanadium, manganese and tungsten determinations were 11 min, 3 h and 26.3 h, respectively, but these were reduced greatly by intoruding a scheme wherein six samples were simultaneously irradiated and the ⁵⁶Mn and ¹⁸⁷W nuclides were measured sequentially for a series of 66 samples. The average processing time was reduced to 45 min for tungsten with a precision of 4.0% and accuracy of 3.4% and 22.8 min for manganese with a precision of 3.8% and accuracy of 3.1%.     

Impurity correction factor of MnSO4 compound for the determination of neutron emission rate on the manganese bath method

The manganese sulphate bath method is widely used for measurements of neutron source strength. In this study, the analytical chemistry method based on the Inductively Coupled Plasma (ICP) spectrometry was used for examining the impurity contents of MnSO₄·H₂O, to induce⁵⁵Mn(n,γ)⁵⁶Mn reactions. From the analytical results, mainly K, Co, and Zn as well as trace amounts of Cd, Li, etc., have turned out to be the relevant impurities absorbing the neutrons and the fraction of neutron absorbed by the total impurities was determined to be 1.37%.     

Evaluation of an alternative convenient irradiation system for determination of emission rate of radio-isotopic neutron sources

In present work, an alternative irradiation system based on a symmetric cylindrical tank filled with a moderator containing hydrogen, which was equipped with a NaI(Tl) scintillation detector, was proposed for using in determination of neutron flux. This irradiation system was designed by MCNP4C code, with considering a ²⁴¹Am–Be neutron source in several volumes and different materials. When the neutron is captured by hydrogen, a 2.22 MeV prompt gamma-ray is emitted. The gamma pulse-height spectrum shows a photo-peak around 2.22 MeV whose net area is proportional to the total emission rate of neutron. The simulation result showed that a cylindrical tank with 110 cm diameter and height filled with water can be a suitable system for neutron source strength calibration. Furthermore, a proper two-layer shielding must be placed between the source and detector for preventing neutrons and gamma rays to directly enter the detector.     

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.     

Determination of manganese in some pyrolusite ores of Myanmar by neutron activation analysis

Manganese in pyrolusite ores from various regions of Myanmar was determined by thermal neutron activation analysis using an Am(Be) neutron source. The induced activities of⁵⁶Mn were monitored by a -counting technique.     

Neutron beam preparation with Am–Be source for analysis of biological samples with PGNAA method

Material analysis with prompt gamma neutron activation analysis (PGNAA) requires a proper geometrical arrangement for equipments in laboratory. Application of PGNAA in analysis of biological samples, due to small size of sample, needs attention to the dimension of neutron beam. In our work, neutron source has been made of ²⁴¹Am–Be type. Activity of ²⁴¹Am was 20 Ci which lead to neutron source strength of 4.4 × 10⁷ neutrons per second. Water has been considered as the basic shielding material for the neutron source. The effect of various concentration of boric acid in the reduction of intensity of fast and thermal components of the neutron beam and gamma ray has been investigated. Gamma ray is produced by (α, n) reaction in Am–Be source (4.483 MeV), neutron capture by hydrogen (2.224 MeV), and neutron capture by boron (0.483 MeV). Various types of neutron and gamma ray dosimeters have been employed including BF₃ and NE-213 detectors to detect fast and thermal neutrons. BGO scintillation detector has been used for gamma ray spectroscopy. It is shown that the gamma and neutron radiation dose due to direct beam is of the same magnitude as the dose due to radiation scattered in the laboratory ambient. It is concluded that 14 kg boric acid dissolved in 1,000 kg water is the optimum solution to surround the neutron source. The experimental results have been compared with Monte Carlo simulation.     

The comparison of four neutron sources for Prompt Gamma Neutron Activation Analysis (PGNAA) in vivo detections of boron

A Prompt Gamma Ray Neutron Activation Analysis (PGNAA) system, incorporating an isotopic neutron source has been simulated using the MCNPX Monte Carlo code. In order to improve the signal to noise ratio different collimators and a filter were placed between the neutron source and the object. The effect of the positioning of the neutron beam and the detector relative to the object has been studied. In this work the optimisation procedure is demonstrated for boron. Monte Carlo calculations were carried out to compare the performance of the proposed PGNAA system using four different neutron sources (²⁴¹Am/Be, ²⁵²Cf, ²⁴¹Am/B, and DT neutron generator). Among the different systems the ²⁵²Cf neutron based PGNAA system has the best performance.     

Determination of the ratio of the hydrogen and manganese absorption cross sections by the manganese bath technique

The ratio of the hydrogen and manganese neutron absorption cross sections, _H/_Mn, is a most important parameter in the determination of radioactive neutron source strength by the manganese bath technique. The ratio is well measured by observing the change in⁵⁶Mn activity induced in the manganese bath by a fixed neutron source as the manganese concentration of the bath is changed. In the present study, the neutron source was a Maxwellian beam from²⁵²Cf. Concentrations were determined by the two methods: volumetric and gravimetric. The cross section ratio has turned out to be _H/_Mn=0.02506.     

Measurements of nuclear data of minor actinides for transmutation of high-level waste

For nuclear transmutation of minor actinides, delayed neutron emission measurement for²⁴¹Am was carried out in thermal neutron irradiation location. The neutron capture cross sections of²⁴¹Am were also measured radiochemically. The transmutation process of²⁴¹Am in reactor is discussed by calculating the yields of minor actinides with the nuclear data measured in this study and the evaluated values. The accelerator driven transmutation of minor actinides by high-flux neutrons from spallation reactions is also presented.     

Simultaneous determination of 152Eu and 241Am in liquid solution by liquid scintillation counting

¹⁵²Eu and ²⁴¹Am are the most frequently used radiotracers in the separation studies on trivalent minor actinides and lanthanides. In almost all those studies, the determination of ¹⁵²Eu and ²⁴¹Am has been based on measuring their γ radiation by using a NaI(Tl) scintillation detector and/or a germanium detector. In this study, based on measuring the β particles and mono-energy electrons from ¹⁵²Eu and the α particles from ²⁴¹Am, we provide a new option to simultaneously determine the radioactivities of ¹⁵²Eu and ²⁴¹Am by liquid scintillation counting (LSC) with the aid of α/β discrimination. If the count rate ratio of ²⁴¹Am to ¹⁵²Eu is within the range of 100:1–1:100, the radioactivities of ¹⁵²Eu and ²⁴¹Am in mixed samples can be simultaneously determined by LSC with the errors less than 5 %. In addition, the interferences of ²⁴¹Am on Eu are divided into two parts: inside and outside the ²⁴¹Am region of interest. Only if the count rate ratio of ²⁴¹Am to Eu is more than 10:1, should the latter interference be in consideration.     

Study of the <Superscript>241</Superscript>Am(n,2n)<Superscript>240</Superscript>Am reaction cross section in the energy range of <Emphasis Type="Italic">E</Emphasis><Subscript><Emphasis Type="Italic">n</Emphasis></Subscript> = 8.8−11.1 MeV

The measurement of the cross section of the reaction ²⁴¹Am(n,2n)²⁴⁰Am has been performed at neutron energies from 8.8 to 11.1 MeV, implementing the activation technique. The neutron beam was produced at the TANDEM accelerator of NCSR “Demokritos” by the ²H(d,n)³He reaction, using a deuterium gas target. During the 5-day long irradiation, the neutron beam fluctuations were monitored in 100 seconds intervals by a BF₃ counter connected with a multiscaling unit. The radioactive target consisted of a 37 GBq ²⁴¹Am source enclosed in a Pb container. A natural Au foil, a ²⁷Al foil and a ⁹³Nb foil were used as reference materials for the neutron flux determination. After the end of the irradiation the activity induced at the target and the reference foils, was measured off-line by a 56% HPGe detector.     

Spectra and dose with ANN of 252Cf, 241Am–Be, and 239Pu–Be

Artificial neural networks have been applied to unfold the neutron spectra and to calculate the effective dose, the ambient equivalent dose, and the personal dose equivalent for ²⁵²Cf, ²⁴¹Am–Be, and ²³⁹Pu–Be neutron sources. The count rates that these neutron sources produce in a Bonner Sphere Spectrometer with a ⁶LiI(Eu) were utilized as input in both artificial neural networks. Spectra and the ambient dose equivalent were also obtained with BUNKIUT code and the UTA4 response matrix. With both procedures spectra and ambient dose equivalent agrees in less than 10%. The Artificial neural network technology is an alternative procedure to unfold neutron spectra and to perform neutron dosimetry.     

Simultaneous liquid scintillation determination of239Pu and241Am in tissue

A liquid scintillation counting method for the simultaneous determination of Pu and Am, with a two-phase cocktail, has been applied to the analysis of a tissue sample from an accidental exposure incident. The sample contained²³⁹Pu,²⁴¹Pu, and²⁴¹Am. In addition to analysis by two liquid scintillation counting techniques, analysis of the sample was performed by -spectroscopy and ZnS scintillation techniques, and the results were compared. The presence of²⁴¹Pu interfered with the liquid scintillation determination of²⁴¹Am when the two-phase cocktail was used, but the results were in agreement sufficient to be useful in determining what course of treatment, if any, might be necessary for the patient.     

Determination of <Superscript>241</Superscript>Am in urine using sector field inductively coupled plasma mass spectrometry (SF-ICP-MS)

Quantification of ²⁴¹Am in urine at low levels is important for assessment of individuals’ or populations’ accidental, environmental, or terrorism-related internal contamination, but no convenient, precise method has been established to rapidly determine these low levels. Here we report a new analytical method to measure ²⁴¹Am as developed and validated at the Centers for Disease Control and Prevention (CDC) by means of the selective retention of Am from urine directly on DGA resin, followed by SF-ICP-MS detection. The method provides rapid results with a limit of detection (LOD) of 0.22 pg/L (0.028 Bq/L), which is lower than 1/3 of the C/P CDG for ²⁴¹Am at 5 days post-exposure. The results obtained by this method closely agree with CDC values as measured by liquid scintillation counting, and with National Institute of Standards Technology (NIST) Certified Reference Materials (CRM) target values.     

Hot atom chemistry in oxyanion targets,part II1: Recoil behaviour of56Mn in permanganates

Fractional precipitation techniques have been utilized to separate the lower valent and parent forms of⁵⁶Mn in permanganate targets and an attempt is made to study a few aspects of chemical stabilization of recoil⁵⁶Mn in permanganates. Ammonium permanganate, recoil behaviour of which has not been studied previously, is chosen as one of the targets along with the potassium permanganate for initial retention and also for isothermal annealing.⁵⁶Mn initial retentions of about 12% and about 4% are obtained for potassium and ammonium permanganate, respectively, by activation from a Ra–Be neutron source. A usual trend for KMnO₄ and the reduction of recoil fragments by ammonium ions in NH₄MnO₄ are seen through the isothermal annealing study.     

Determination of copper in some Myanmar indigenous medicines by neutron activation analysis

Copper was determined in two Myanmar indigenous medicines by neutron activation analysis using an Am (Be) radionuclide neutron source. The activity of 511 keV peak of the⁶⁴Cu was measured.     

A multi-element serum standard for neutron activation analysis

An annular ²²⁷AcBe isotopic neutron source, containing 6.6-Ci ²²⁷Ac, is described for application in fast and thermal neutron activation analysis, with high accuracy, for major constituents in ores, alloys and industrial concentrates. The characteristics of the neutron output and of the fast, epithermal and thermal flux and flux gradients is described in detail. The determination of manganese in pyrolusite ores and ferro-manganese is compared to results obtained previously with a cylindrical 1-Ci ²²⁶RaBe source. Two new sources of systematic errors have been discovered.     

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.     

An241Am−Be source based thermal neutron activation analysis facility at KFUPM

An instrumental thermal neutron activation analysis facility based, on a 16 Ci²⁴¹ Am–Be source, a high resolution -ray spectrometry setup and a PC-based data acquisition system at KFUPM is described. The thermal neutron flux distribution was determined from the induced activities of high purity indium foils. The absolute thermal neutron flux was calculated from the activities of bare and cadmium-covered gold foils at a position of 3 cm from the soource at which the flux reaches a maximum. The facility tests were carried out with the determination of manganese concentrations in six types of industrially important steel samples. The result of 1.33% manganese in SS-304 steel sample was in excellent agreement with the literature value. The method is nondestructive, economical and ideal for bulk analysis.