Neutron energy spectra of 252Cf, Am-Be source and of the D(d,n)3He reaction

The neutron energy spectrum of the following sources were measured using a fast neutron spectrometer with the NE-213 liquid scintillator: ²⁵²Cf, Am-Be and D(d,n)³He reaction from a 3 MeV Pelletron accelerator in Tokyo Institute of Technology. The measured proton recoil pulse height data of ²⁵²Cf, Am-Be and D(d,n)³He were unfolded using the mathematical program to obtain the neutron energy spectrum. The ²⁵²Cf and Am-Be neutron energy spectra were measured and the results obtained showed a good agreement with the spectra usually published in the literature. The neutron energy spectrum from D(d,n)³He was measured and the results obtained also showed a good agreement with the calculation by time of flight (TOF) methods.     

A method for experimental determination of effective resonance energies related to (n, γ) reactions

When making use of some single comparator or absolute standardization methods in reactor neutron and in epicadmium neutron activation analysis, the knowledge of the effective resonance energy ( ) is essential to correct for the effect of the nonideal epithermal flux distribution on the analysis result. can be calculated from neutron resonance data, but when these are incomplete, not accurate or even not known at all, experimental determination should be considered. Such a method, providing both and the resonance integral to 2200 ms^–1 cross-section ratio (Q_O), is described in this paper. Results are given for 11 isotopes.     

Production of monoenergetic MeV-range neutrons by <Superscript>3</Superscript>H(p,n)<Superscript>3</Superscript>He reaction

Monochromatic MeV-energy neutron source for secondary reaction was developed utilizing tritium embedded titanium (Ti-³H) thin film via ³H(p,n)³He reaction. We have measured the neutron energies and the energy spread by resonance reactions of ¹²C(n,tot) and ²⁸Si(n,tot). The available energy was within the range from 0.6 to 2.6 MeV. Energy spread was 1.6% at energy of 2.077 MeV. The flux in the beam direction was determined to be 3.76·10⁷ n/s/sr by irradiating ¹⁹⁷Au by about 2 MeV neutrons. This source was shown to be useful for measurements of nuclear data by measuring the total cross sections of neutrons on Fe and comparing these data to the data of ENDF-6.     

The single-comparator method in thermal neutron activation analysis extended for some (n,p) reactions

From experimental studies of the rate of the reactions⁴⁷Ti(n, p)⁴⁷Sc,⁴⁸Ti(n, p)⁴⁸Sc and⁵⁸Ni(n, p)⁵⁸Co in four nuclear reactors, it is concluded that for the irradiation positions of the light water moderated reactors BR2 (Mol, Belgium) and HFR (Petten, Netherlands) a simple empirical relation exists between the fast neutron flux on the one hand and the thermal and epithermal neutron flux on the other. The graphite moderated reactor BR1 (Mol, Belgium) and the heavy water reactor FRJ2 (Jülich, Federal Republic of Germany), however, have relatively much lower fast fluxes and their irradiation facilities do not obey the empirical relation determined.     

ko-measurements and related nuclear data compilation for (n, γ) reactor neutron activation analysis

k_o-factors of 35 isotopes used in reactor neutron activation analysis were measured with a high degree of accuracy (1–2%). To minimize systematic errors, measurements were carried out using different reactor types, irradiation conditions (18 < Φ_s/Φ_e), Ge(Li) detectors, sample detector geometry, etc. Analyst-oriented tabulations including all necessary nuclear data, “best values”, as well as recommended k_o-values are given to facilitate analytical work with the new method. Some practical aspects as well as limitations of the k_o-method are also outlined together with the applied neutron flux and cross-section conventions. Research associate of the National Fonds voor Wetenschappelijk Onderzoek, Belgium     

Determination of lithium by reactor activation analysis using the reaction chain:6Li(n,t)4He;32S(t,n)34mCl

The occurrence, in a nuclear reactor, of the reaction chain:⁶Li(n,t)⁴He;³²S(t,n)^34mCl has been experimentally established. Experimentation for its application to the activation analysis determination of lithium has been carried out, and a radiochemical method for separation of^34mCl is presented. The sensitivity is 0.4 g for the following conditions; 15 min irradiation (thermal flux: 1.5·10¹³ n·cm^–2·s^–1); 30 min decay; 2,000 s measurement (semiconductor detector).     

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.     

Determination of the 209Bi(n,γ) capture cross section at a cold neutron beam

Summary The total capture cross section of ²⁰⁹Bi was determined at the cold neutron beam PGAA-NIPS facilities at the Budapest Neutron Centre. The measurements were performed using a coaxial HPGe detector with Compton suppression. The total and partial gamma-ray production cross sections were deduced relative to the ¹⁴N(n,γ) partial gamma-ray production cross section. By using a bismuth nitrate stoichiometric compound as the sample, we excluded various systematic uncertainties. The total capture cross section is in very good agreement with the compilation of Mughaghab, but is slightly lower than the most recent value determined at the high flux reactor of the ILL in Grenoble, France. We also performed measurements using a 0.5 mm thin Bi metal disc. The relative intensities determined from the Bi disc and the compound samples are in good agreement.     

Using synthetic multi-element standards (SMELS) for calibration and quality control of the irradiation facilities in the BR1 reactor

The thermal to epithermal neutron flux ratio (f) and the deviation of the epithermal neutron spectrum from the 1/E shape (α) are essential parameters for the correct application of k ₀-standardized neutron activation analysis. Several methods are applied for the determination of f and α. They are based on Cd-covered multi-monitor or on bare-irradiations methods. The recently developed and characterized synthetic multi-element standards (SMELS) were designed as a validation tool for the proper implementation of the k ₀-NAA method in a laboratory. In particular, SMELS Type III contains Au and Zr, thus allowing the direct determination of f and α. It could, therefore, replace the traditional flux monitors. Furthermore, it could be used as a quality control material to monitor the stability of the irradiation facility and the detector. This paper presents the accuracy of the f and α determination and the feasibility of quality control using SMELS for irradiation channel Y4 of the BR1 reactor.     

14.8 MeV neutron activation cross-sections for (n,p) and (n, α) reactions of some rare earth nuclides

The activation cross-sections for sixteen (n, p) and (n, α) reactions on rare earth nuclides were measured at 14.8 MeV neutron energy. A Ge(Li) detector was used for the radioactivity measurements. The measured cross-sections are compared with the semi-empirical predictions of LEVKOVSKII³ and BAYHURST and PRESTWOOD.¹⁶     

The Characterization of a Neutron Radiography Triga Reactor for NAA of Chlorine in an Iron Oxide Matrix

An irradiation position in the 250 kW Triga reactor was characterized for instrumental neutron activation analysis of chlorine in an iron oxide matrix. Factors that affect the accuracy of the determination include variations in the reactor neutron spectrum and flux as a function of spatial position and the presence of chlorine impurities. Gold wire and foils were used to determine the neutron flux and cadmium ratio as a function of height in an air-filled irradiation tube.     

Calibration of the irradiation channel DBVK at BER-II reactor and feasibility of applying the k0-method at this device

This work is about k ₀-INAA using unstable neutron flux for sample irradiation. Due to slow transport, each irradiation in the channel DBVK consists of three phases: stable irradiation at the final position, and two additional irradiations during travelling by exposure to an increasing neutron flux in the delivery course and to a decreasing neutron flux in the fetch course. In this work, the neutron flux distribution along this channel was calibrated and the neutron flux variation with irradiation time was calculated, making it possible to evaluate activity growth during a complete irradiation period. The feasibility of the k ₀-method was checked by analyses of four SRM-materials and three multi-element standards at three DBVK-positions. An accuracy of better than ±10% was found for nearly all determined elements in each determination.     

Activation cross-sections for some (n, 2n), (n,p) and (n, γ)-reactions

The average fission neutron cross-sections of the reactions²³³U(n, 2n)²³²U,⁶⁰Ni(n, p)⁶⁰Co and²⁷Al(n, p)²⁷Mg and the resonance integrals of the (n, γ)-reactions of the nuclides¹⁸¹Ta,¹⁷⁶Lu,¹⁷⁵Lu,⁶⁴Ni,⁵⁹Co and²⁶Mg have been determined by the activation method following the well-known conventions. The results verify some of the existing values and present data for hitherto unknown or poorly known reactions.     

Possibilites et limites de l’analyse d’echantillons biologiques par activation neutronique derriere ecran de cadmium

The analytical possibilities offered by long-term irradiation of biological samples in a fission neutron flux filtered by a cadmium screen are examined and the preparation of samples for irradiation is described. This specific activation technique is applied to the determination of molybdenum and nickel in the following three reference substances: human serum, NBS bovine liver and BOWEN’s kale.     

Non-destructive determination of zinc in biological certified reference materials

Non-destructive neutron activation analysis was employed to determine zinc in ten biological standard reference materials from the National Bureau of Standards and the National Research Council of Canada. The use of a 4 h. irradiation at a medium neutron flux allows excellent accuracies, precision and sensitivities to be attained for all the samples analyzed. It is suggested that neutron activation analysis be one of the benchmark methods for the determination of zinc in biological reference materials for concentrations as low as the one part per million level.     

Instrumental neutron activation determination of Pt,Ru, Ir,Sc, Cr,Fe, Co,La, Ce,Nd, Sm,Yb, Lu,Hf and Ta employing CIRUS reactor as a thermal neutron source

An accurate and simple method has been developed for the determination of Pt, Ru, Ir and other elements employing instrumental neutron activation analysis (INAA). Nondestructive analysis has been carried out for the determination of these elements in different rock samples such as Dolerite, Dyke Dolerite, Country Basalt, Hyaloclastite, Trachyte, Ijolite, Spillite, Diorite and Lamprophyre from various locations like Daman, Tapti, Murud, Talasari, Ranala and Bassein in Maharashtra state. High flux provided by the CIRUS reactor (1×10¹³ n cm^?2s^?1) has been used for thermal neutron bombardment followed by radioassaying of the (n,γ) products on a HPGe detector coupled to a PC-based MCA unit.     

The determination of ko factors and their application to the neutron activation analysis of pure zirconium metal by the single-comparator method

The procedures followed for the determination of standardization factors and their use in the single-comparator method for multi-element reactor neutron activation analysis are described. The method is applied to the determination of Cr, Mn, Fe, Ni and Mo in pure zirconium metal. The matrix could serve as the single comparator and as a bi-isotopic thermal-to-epithermal neutron flux ratio monitor.     

Investigation of gamma-spectra of biological materials irradiated in the epithermal range of the reactor neutron spectrum

Irradiation of various biological materials in the nuclear reactor neutron epithermal flux allows the increase of the selectivity of the pure instrumental method of activation analysis for the definite set of tracers either not determined instrumentally during sample irradiation by the thermal neutron flux or determined with low sensitivity. The given paper describes a method of increasing the sensitivity of the instrumental neutron activation determination of As, Br, Mo, Cd, Sb, W and Au in biological materials at the level of the whole blood, tissue and subcellular components.     

Determination of manganese in natural waters by atomic absorption spectrometry with a carbon tube atomizer.

A method is described for the accurate and precise determination of alumina and silica in bauxites with the aid of a ²²⁷Ac-Be isotopic neutron source with a total neutron output of 10⁸ n s^-1. Three ores can be analysed in triplicate within 4 h, including the determination of the natural radioactivity of the ores. Samples are pellets pressed from a mixture of 4.5 g of powdered bauxite and 0.9 g of a wax as pelleting agent. The special flux distributions of the source allow irradiations at very different fast-to-thermal flux ratios without cadmium neutron absorbers. The drying and water re-uptake of bauxite and the natural radioactivity in these ores are discussed. The method was tested with several certified standards. A relative precision of 0.7% for the alumina determination was obtained for triplicate analyses. Vanadium is the only interfering element, the concentration of which should be determined separately or estimated; a correction procedure is given.     

Monte Carlo calculations and neutron spectrometry in quantitative prompt gamma neutron activation analysis (PGNAA) of bulk samples using an isotopic neutron source

An activation analysis facility based on an isotopic neutron source (185 GBq ²⁴¹Am/Be) which can perform both prompt and cyclic activation analysis on bulk samples, has been used for more than 20 years in many applications including 'in vivo' activation analysis and the determination of the composition of bio-environmental samples, such as, landfill waste and coal. Although the comparator method is often employed, because of the variety in shape, size and elemental composition of these bulk samples, it is often difficult and time consuming to construct appropriate comparator samples for reference. One of the obvious problems is the distribution and energy of the neutron flux in these bulk and comparator samples. In recent years, we have attempted to adopt the absolute method based on a monostandard and to make calculations using a Monte Carlo code (MCNP4C2) to explore this further. In particular, a model of the irradiation facility has been made using the MCNP4C2 code in order to investigate the factors contributing to the quantitative determination of the elemental concentrations through prompt gamma neutron activation analysis (PGNAA) and most importantly, to estimate how the neutron energy spectrum and neutron dose vary with penetration depth into the sample. This simulation is compared against the scattered and transmitted neutron energy spectra that are experimentally and empirically determined using a portable neutron spectrometry system. This revised version was published online in August 2006 with corrections to the Cover Date.