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.     

Design of a permanent Cd-shielded epithermal neutron irradiation site in the Syrian Miniature Neutron Source Reactor

A Cd-shield (cylindrical shell 1 mm in thickness, 34 mm in diameter and 180 mm in length) was used to design a permanent epithermal neutron irradiation site for epithermal neutron activation analysis (ENAA) in the Syrian Miniature Neutron Source Reactor (MNSR). This site was achieved by shielding the surface of the aluminum tube of one of the outer irradiation sites. The calculated depression ratio of thermal neutron flux was 1/10. Homogeneity of the neutron flux in the first outer irradiation site has been found numerically using the WIMSD4 and CITATION codes and experimentally by irradiating five short copper wires using the outer irradiation capsule. Good agreement was obtained between the calculated and the measured results of the neutron flux distributions.     

Destructive and nondestructive analysis of some elements in tissue and environmental samples by thermal neutron activation analysis technique

A rapid and selective method has been developed and applied for determining elements present in tissue and environmental samples by both destructive and nondestructive activation analysis. Nondestructive activation analysis involves the irradiation of the sample and standard of the elements such as Mn, Na, and K with thermal neutrons from²⁵²Cf neutron source followed by radioassay of the (n,r) products on a HPGe detector coupled to a PC-based MCA. Elements are determined by irradiating the samples in the thermal neutron flux of the CIRUS reactor and radiochemically separating the isotopes of interest using substoichiometric extraction and precipitation technique. The statistical evaluation of the method with respect to accuracy and precision of the method and its sensitivity are discussed.     

Flux calculation in LSNAA using an <Superscript>241</Superscript>Am-Be source

The CITATION code based on neutron diffusion theory is used for flux calculation inside voluminous sample in prompt gamma activation analysis with an isotopic neutron source (²⁴¹Am-Be). The code used the specific parameters related to energy spectrum source, irradiation system materials (shielding, reflector, etc.), geometry and elemental composition of the sample. The flux distribution (thermal and fast) was calculated on three-dimensional geometry for the system: source, air, and polyethylene and water cylindrical sample of 125 liters. The thermal flux was calculated in series of points inside the sample, and agreed with the results obtained by measurements with good statistical uncertainty. The maximum thermal flux was measured at distance of 4.1 cm and calculated at 4.3 cm by the CITATION code. Beyond a depth of 7.2 cm, the ratio of thermal flux to fast flux increases up to twice and allows us the optimization of the detection system in the scope of in-situ PGNAA.     

Experimental and MCNP calculations of neutron flux parameters in irradiation channel at Es-Salam reactor

The Algerian research reactor (Es-Salam) is a 15 MW heavy water reactor type, operating since 1992. It became essential to characterize the neutron field in the most useful irradiation positions, in order to guarantee the accuracy in the application of k ₀-neutron activation analysis (k ₀-NAA). Experimental value of the thermal to epithermal neutron flux ratio (f) and of the deviation of the epithermal neutron spectrum from 1/E shape (α) were determined using different methods. This work focuses the verification of Monte Carlo neutron flux calculation in typical irradiation channel. Comparison of the results for parameter f obtained experimentally and by Monte Carlo simulations shows good agreement in the irradiation channel studied. The difference between both results is about 2.08%.     

Photoneutrons from a beryllium reflector: a potential source of problems with Zr–Au flux monitors in <Emphasis Type="Italic">k</Emphasis><Subscript>0</Subscript> standardization based neutron activation analysis

The assumption that the shape of the epithermal neutron spectrum can be described, in any research reactor, by the 1/E ^1+α function is a fundamental starting point of the k ₀ standardization. This assumption may be questioned from a reactor physics viewpoint. The type of moderator, the existence of neutron reflectors, the additional production of (γ, n) neutrons and resonance capture by construction materials may be different for each reactor, with consequences for the shape of the neutron spectrum. This dependency may explain that various practitioners reported contradicting experiences with the use of Zr–Au flux monitors for the determination of the α-parameter. An objective view on the influence of the design of the reactor and irradiation facility on the shape of the neutron spectrum can be obtained by modeling. This has been applied in the Reactor Institute Delft for reactor configurations in which the irradiation facilities face the fuel elements with the presence of beryllium reflector elements. The Monte Carlo calculations indicate a distortion of the 1/E ^1+α relationship at the higher energy edge of the epithermal neutron spectrum. This distortion is attributed to the formation and thermalisation of both photoneutrons and (n, 2n) produced fast neutrons in the beryllium, and has a direct impact on the resonance activation of ⁹⁵Zr, other than represented by the 1/E ^1+α function. The obtained relationship between neutron flux and neutron energy was also used for estimating the f-value and compared with the value obtained by the Delft Cr–Mo–Au flux monitor.     

Use of the semi-absolute method in neutron activation analysis

A new standardization method has been developed for neutron activation analysis. In this method, experimental activation constants, are determined for a given reactor power level and irradiation and counting position. The unusual feature of this technique is the fact that no flux monitor or standards are needed due to the exceptional stability of the reactor used. The semi-absolute method was tested over a three month period and its reliability was demonstrated for 6 elements of different neutrons cross-section characteristics.     

Non-destructive neutron-activation analysis for determining the chlorine content of paper-pulp

Non-destructive neutron-activation analysis is used for determining chlorine in paper-pulp. Numerical data have been obtained for bleached and unbleached paper-pulps of different types and origins. The sensitivity of this method is 100 ppm for an irradiation time of 30 min and a neutron flux of 6 x 10(10) neutrons.cm(-2).sec(-1) and 10 ppm for an irradiation time of 1 min and a neutron flux of 2 x 10(12) neutrons.cm(-2).sec(-1). In both cases the amount of chlorine that can be determined depends on the presence of the interfering elements manganese and sodium in the paper-pulp. The time required for a complete analysis, after irradiation, is 5 min.     

Fast neutron activation analysis of silicon in aluminum alloys

The silicon content in an aluminum-silicon alloy was measured by nondestructive fast neutron activation analysis with fission spectrum neutrons. A boron nitride irradiation container reduced the flux of thermal and epithermal neutrons at the sample position, enhancing the²⁹Si (n, p)²⁹Al reaction. A detection limit of 0.4% silicon in a 0.5 g alloy sample was obtained.     

Instrumental neutron activation analysis of geological and pedological samples. Further investigation of epithermal neutron activation analysis using monostandard method

A comparative study is presented on neutron activation analysis of rock and soil samples using whole reactor neutron spectrum and epithermal neutrons with both relative and monostandard procedures. The latter procedure used with epithermal neutron activation analysis of soil samples necessitated the use of the “effective resonance integrals” which were determined experimentally. The incorporation of the β factor, representing deviation of reactor epithermal neutron flux from 1/E law, is developed in the present work. The main criteria for the choice of one or more of the procedures studied for a given purpose are also indicated. Analysis of 15 trace elements, Ca and Fe in the standard Japanese granite JC-1 using monostandard epithermal neutron activation gave results in good agreement with the average literature values. This paper is dedicated to the 80th birthday of Professor Dr. Robert Klement, University of Munich.     

Problems and solutions in the standardization of reactor neutron activation analysis

A discussion is held on some selected topics of importance in the standardization of reactor neutron activation analysis. The topics are related to the accuracy and the consistency of the nuclear data, the occurrence of flux gradients in the irradiation container, the extensions and modifications of the k₀-method, and the determination of uranium via the counting of²³⁹Np and²³⁹U with a large Ge-detector or a LEPD.     

Monitoring of neutron flux changes in short-lived neutron activation analysis

It is well known variations in neutron fluxes can adversely affect the final result in neutron activation analysis. The monitoring of neutron flux changes are usually described for medium and long-lived NAA using foils of cobalt, gold, zirconium, etc. However, for short-lived neutron activation analysis there appears to be no systematic study of the variations of the neutron flux. With our new automatic pneumatic system, where irradiation timing, decay and counting and position are very reproducible, we have performed a series of experiments using thermal and epithermal neutrons using aluminum wire as a monitor to monitor the neutron fluxes. Our experiments confirm that neutron flux fluctuations in the worst case can be up to ±12 % with a SD of 2–3 %. This effect can be seen regardless of the irradiation time and must be taken into consideration to achieve the best result.     

Measurement of neutron spectrum parameters for NAA irradiation holes in the Jordan research and training reactor

Three irradiation holes coupled to a pneumatic transfer system were installed for neutron activation analysis in the Jordan Research and Training Reactor (JRTR), which is the first research reactor in Jordan. To perform instrumental neutron activation analysis, neutron spectrum parameters, such as thermal neutron flux, α and f for the irradiation holes, should be measured. The Cd-ratio method was applied for the determination of the aforementioned parameters. For this purpose, 0.1% Au–Al wires and Zr foils were irradiated with and without Cd-cover, and the Cd ratios were determined for Au-198, Zr-95, and Zr-97/Nb-97m nuclides. Then, the parameters were calculated and determined at three irradiation holes.

     

An oxygen standard for the determination of oxygen in steel by 14-MeV neutron activation analysis

A relative method for the determination of oxygen in steel via the ¹⁶O(n,p)¹⁶N reaction by means of 14-MeV neutrons is described. A standard is irradiated immediately behind the sample and the induced activities are counted simultaneously with two separate but identical detector systems. The standard mixture (ca. 5 g of graphite plus iron oxide containing 80 mg of oxygen per g) is compressed into a steel capsule of the same external dimensions as the samples (26 mm diameter, 9 mm thick). Dimensional tolerances, choice and purity control of the oxygen compound and preparation of the standard mixture are discussed. Fast neutron shielding, absorption of fast neutrons, self-absorption of the ¹⁶N /gg-rays and the average neutron flux in sample and standards are considered and a total correction factor is established. Flux inhomogeneities and differences in counting geometry and discriminator setting can be determined by irradiation and counting of two identical standards. The accuracy of this method was checked by comparison of the results with those of the reducing fusion method; satisfactory agreement was observed, although the activation results tended to be slightly higher. The mean long-term standard deviation for analysis of a given sample over a period of 6 months was found to be ±3%.     

Evaluation of high purity graphite sample containers for use with rapid instrumental epithermal neutron activation analysis

The suitability of graphite of spectral quality as a sample container in short irradiations for neutron activation analysis has been investigated. The material is found to be a viable alternative to polyethylene containers especially in high flux irradiation positions where the properties of the latter deteriorate as a result of high neutron and gamma dose rates. Examples of application are demonstrated in epithermal neutron irradiations of biological materials and the accuracy and precision of the determinations were assessed.     

A review of intercomparison exercises run by NPL from 1989–1997

The flux of cold neutrons that is obtainable from various high energy netron sources is studied for a particular model of a cold neutron source when the cold moderation region of the apparatus is at 20, 70, and 298K. The maximum flux obtained with a californium-252 source was 2.7·10^?3 cold neutron per (cm²·second (source neutron)). This flux was obtained when the cold moderation region of the apparatus was at 20K and when the thermal moderator is either polyethylene or trimethylbenzene and the cold moderator is polyethylene. This flux should allow sensitive prompt and delayed neutron activation analysis measurements.     

Rapid analysis of biological reference materials by instrumental neutron activation

Summary Instrumental neutron activation analysis (INAA) was applied to the rapid determination of cadmium and other elements in the IAEA biological reference material horse-kidney (H-8). Nuclear reactor neutrons and epithermal neutrons were used as neutron sources. Cadmium, bromide, iodine and phosphorus were determined by epithermal neutron activation analysis. Aluminum was determined by reactor neutron activation analysis taking into account the contribution of phosphorus to the ²⁸Al activity.     

Neutron spectrum parameters in irradiation channels of the Nigeria Research Reactor-1 (NIRR-1) for the <Emphasis Type="Italic">k</Emphasis><Subscript>0</Subscript>-NAA standardization

Summary The first nuclear research reactor in Nigeria has been commissioned for neutron activation analysis and limited radioisotope production. In order to extend its utilization to include the k₀-standardization method, the following neutron spectrum parameters in inner and outer irradiation channels were determined by the “Cd-ratio for multi-monitor method”: the thermal-to-epithermal flux ratio, f, and the epithermal flux shape factor, α. Neutron spectrum parameters determined in the inner irradiation channel B2, are: α = -0.052±0.002 and f = 19.2±0.5. For the outer irradiation channel B4, the neutron spectrum parameters were found to be α = +0.029±0.003 and f = 48.3±3.3. The results are compared with the neutron spectrum parameters of other reactor facilities with similar core configuration such as the Slowpoke and Miniature Neutron Source Reactor facilities available in the literature.     

Multiparameter neutron activation analysis

An advanced neutron activation technique has been developed for the accurate analysis of elemental and isotopic fissile material required in nuclear safeguards, nuclear material standardization and other applications. It is based on reactor neutron flux spectrum differentiation by cadmium screening and multistandard calibration, including the solution of a second order equation system or of computerized calibration curve fitting, taking into account the thermal neutron flux depression. Some discrepancies at high enrichments have still to be eliminated in order to achieve the required measurement accuracy.     

Activation analysis with fast neutrons using a cyclotron

A systematic study of activation analysis with cyclotron-produced neutrons for (n, 2n), (n, p) and (n, α) reactions is presented. The limits of detection for elements of atomic number from 6 to 80 are given. The possibilities of optimization of irradiation conditions by the choice of the most suitable neutron spectrum are discussed. The potential of this fast neutron activation analysis method is compared with that of 14 MeV neutron activation analysis.