Neutron activation analysis without multi-element standards

The paper shows how, from the neutron irradiation of multi-element standards, one can derive neutron flux parameters for the irradiation position and, at the same time, greatly improve knowledge of nuclear data parameters, such as thermal cross sections, resonance integrals and gamma raz emission probabilities, for the nuclides concerned. It is then shown how the resulting neutron flux parameters and nuclear data parameters can be used to carry out neutron activation analysis without further irradiation of multi-element standards. The technique is applied to the analysis of Chinese geochemical reference material     

Determination of uranium element concentration and235U isotope abundance by neutron activation analysis

A new neutron activation technique has been developed for the determination of uranium element concentration and²³⁵U isotope abundance in nuclear safeguards and reference material samples based on the activation of bare and cadmium-covered samples with different thermal to epithermal neutron flux ratios and on the combination of the two corre-sponding delayed-fission neutron measurements. The principle of the new technique can be applied also to improve multi-element neutron activation analysis.     

Counting yield calibration of NaI(Tl) detectors for complex geometry samples by use of californium-252

The counting yield for large volume, complex geometry samples such as solutions in Marinelli beakers as counted on a large NaI(Tl) detector can be calibrated using radionuclides activated by neutrons form a²⁵²Cf source. Calibration may be done by using either a known neutron flux facility or by cross calibration of the activated material as a point source vs. sealed gamma-standard sources. The point source of activated material is dissolved after cross calibration to produce the large volume distributed source.     

BEAMGAA: A chance for high precision analysis of big samples

Summary In activation analysis of traces in small samples, the non-equivalence of the activating radiation doses of sample and calibration material gives rise to sometimes tolerable systematic errors. Conversely, analysis of major components usually demands high trueness and precision. To meet this, beam geometry activation analysis (BEAMGAA) procedures have been developed for instrumental photon (IPAA) and neutron activation analysis (INAA) in which the activating neutron/photon beam exhibits broad, flat-topped characteristics. This results in a very low lateral activating flux gradient compared to known radiation facilities, however, at significantly lower flux density. The axial flux gradient can be accounted for by a monitor-sample-monitor assembly. As a first approach, major components were determined in high purity substances as well as selenium in a cattle fodder additive.     

A feasibility study for a linac-based transuranic waste characterization system

Remote-handled transuranic nuclear waste poses a particular challenge for assaying due to the high neutron and -ray background that emanate from the non-fissile, but highly radioactive material, contained within the waste. The utilization of a Radio Frequency Quadrupole (RFQ) linac with a neutron flux of the order of 10¹⁰ n/s/4 has shown that, in principle, the differential die-away technique can be used to reliably assay this special class of nuclear waste.     

Neutron/gamma discrimination in LiBaF3 scintillator

The detection of neutrons in the presence of significant gamma-radiation is often required in arms control, material accountability, and nuclear smuggling scenarios as well as in basic nuclear research. The new scintillator material LiBaF₃ offers the possibility of measuring neutron count rates and energy spectra simultaneously while measuring gamma-count rates and spectra using a single detector. These capabilities derive from the fact that LiBaF₃ exhibits a very fast core-valence luminescence under gamma-irradiation whereas this component is missing under neutron irradiation. Relatively simple pulse shape analysis techniques can be used to obtain excellent neutron/gamma discrimination. We present our current results illustrating these capabilities.     

Neutron activation analysis without multielement standards

To carry out neutron activation analysis without using multielement standards requires knowledge of (i) the absolute photopeak efficiency as a function of energy of the gamma-ray detector, (ii) nuclear data for each reaction used, and (iii) neutron flux parameters for the irradiation position. The present paper discusses each of these topics and shows an example of the determination of flux parameters and improved nuclear data.     

Evaluation of nuclear effects in the analysis of plastics by neutron activation analysis

Validation of a method for the determination of trace elements in plastic using neutron activation analysis (NAA) required evaluation of nuclear effects. Since plastic is a moderator of neutrons, it is possible that the neutron flux may be affected in such a way as to effectively increase or decrease the measured activity of radionuclides within the plastic. Such effects are likely to depend on the thickness of the sample and so thicknesses of plastic ranging from 1 to 20 mm were used to quantify the effects. A standard containing vanaldium (a thermal neutron absorber) and silver (with epithermal neutron resonances) was used in a sandwich of the plastic. The sandwich was then irradiated and the standard counted. The activity of each radionuclide for each thickness was evaluated. The results showed that a variation of only 10% maximum was observed for thicknesses up to 20 mm. An increase in measured activity was observed for both the thermal and epithermal absorbers.     

Determination of silicon in rocks by fast-neutron activation analysis using internal standards

A technique for using internal standards in the determination of Si in rocks using fast neutron activation is presented. Different weights of barium acetate were irradiated for 30 seconds then cooled for 10 seconds before counting for one minute. A peak area, at 0.662 MeV due to^137mBa, versus weight of barium calibration curve (I) was made. Similarly, barium acetate, which served as the neutron flux monitor, was mixed with known weights of standard rocks, BCR-1 and G-2. Then a peak area (at 1.78 MeV due to²⁸Al) versus weight of silicon (present in the standard rocks) calibration curve (II) was constructed which was corrected for flux variations. Flux corrections were made possible using curve (I). Utilizing curve (I) and curve (II) the percentage Si in granite samples obtained from Llano, Texas, was determined. This technique avoids any external neutron monitor or sample rotation system. The applicability of this approach may be limited to samples in which the internal flux monitor can be dispersed. On assignment from the Philippine Atomic Energy Commission, Manila, NAS-IAEA Fellow.     

Development and implementation of <Emphasis Type="Italic">k</Emphasis><Subscript>0</Subscript>-INAA standardization at PINSTECH

The k ₀ method has been introduced at the 30 kW miniaturized neutron source reactor (MNSR) at the Pakistan Institute of Nuclear Science & Technology (PINSTECH). It involved the full energy peak efficiency calibration of the HPGe detector for different counting geometries and the characterization of the neutron flux at four inner irradiation channels. The latter involved the determination of the thermal to the epithermal flux ratio, epithermal flux shape factor, the modified spectral index, Westcott’s g-factor, the Maxwellian neutron temperature and the fast flux. The method was validated by analyzing IAEA-SL1 (Lake Sediment) and NIST-SRM-1572 (Citrus Leaves) reference materials. All calculations were performed in Excel, including the optimization step. The results revealed that most of the elements were estimated with less than 10% relative deviation from the certified value.     

Precise determination of bromine in PP resin pellet by instrumental neutron activation analysis using internal standardization

Instrumental neutron activation analysis with the internal standardization was applied to the precise determination of Br in polypropylene resin of candidate certified reference material. The known amount of ¹⁹⁷Au was used as an internal standard to compensate for neutron flux inhomogeneity, to improve the γ ray measurement uncertainty and the linearity of the calibration curves. The reliability of the proposed method validated using analytical results of BCR-681. The analytical result of Br in the sample was consistent with that obtained by ID-ICPMS. The relative expanded uncertainty (k = 2) was 1.5 %, and it was equivalent to that of ID-ICPMS.     

Multielement analysis of concrete from nuclear reactors by INAA,ICP-MS and ICP-AES

There is a need to know, before dismantling nuclear plants, how the induced radioactivity is distributed in concrete. Computer codes are used to predict such a distribution, by applying input data like, among others, the chemical composition of the material. Biological shield concretes with particularly high boron or barium concentrations had to be analyzed. Instrumental neutron activation analysis is not able to determine all elements, thus, additional techniques are required, like ICP-MS and ICP-AES. The respective performances of these techniques are discussed. For INAA, the thermal neutron self-shielding was empirically corrected by using comparisons between thermal neutron activation (TNAA), epithermal and fast neutron activation (ENAA), ICP-AES and ICP-MS measurements.     

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.     

Neutron Activation Procedures Used for the Calibration of a Nitrogen-16 Reactor Power Monitor

Personnel from the Pennsylvania State University's Breazeale Nuclear Reactor assisted NRC (Nuclear Research Corporation) personnel in calibrating a new MSL/¹⁶N Nitrogen-16 Monitor. This paper describes the neutron flux calibration procedures utilized and the results obtained for the production of a nitrogen-16 source of known activity for a BGO detector calibration.     

Determination of cadmium in zinc ores by thermal neutron absorption analysis

A method has been developed for routine determination of cadmium in zinc ores by thermal neutron absorption analysis, based on the attenuation of a thermal neutron flux passing through a neutron absorbing material. The thermal neutron flux is related to the⁵²V-activity induced in a vanadium detector, surrounded by pellets pressed from a mixture of powdered material with graphite. Besides cadmium, also the major constitutents zinc, iron and sulfur contribute significantly to the total attenuation of the thermal neutron flux. Calibration lines for these elements are worked out. All irradiations are carried out for 200 s in the partially thermalized neutron flux of a 5 Ci²²⁷Ac—Be isotope neutron source. After a decay of 30 s, the⁵²V-activity of the vanadium detector is measured for 400 s with a NaI(T1) scintillation detector. The analysis sequence, including the computation of the results from the counting data, is automated by means of a LSI—11 microprocessor with 12K×16 bit memory. Zinc ores, containing 0.02 to 1.45% cadmium, have been analyzed with a precision ranging from 12.6% to 0.54% relative. As a test for the reliability of the method, two NBS standard reference materials were analyzed in the same way as the zinc ore samples.     

Application of CR-39™ SSNTD with a boron converter for the characterization of the external neutron beam of a powder diffractometer

In this work, the solid state nuclear track detector (SSNTD) CR-39? has been used in combination with a boron converter screen for the characterization of the neutron distribution in the neutron powder diffractometer of the Es-Salam research reactor. A relationship between the track density in the CR-39? and the neutron flux distribution has been established by a mathematical development. Good agreement was found between the distribution of the thermal neutron flux determined by CR-39? detector and the distribution measured by the activation of Dysprosium foils. On the other hand, the degree of homogeneity of the neutron beam has been determined by the SSNTD and direct neutron radiography techniques. The results obtained by both techniques showed a depression of the flux at the left bottom region of the beam.     

Monte Carlo efficiency calibration of a neutron generator-based total-body irradiator

Many body composition measurement systems are calibrated against a single-sized reference phantom. Prompt-gamma neutron activation (PGNA) provides the only direct measure of total body nitrogen (TBN), an index of the body’s lean tissue mass. In PGNA systems, body size influences neutron flux attenuation, induced gamma signal distribution, and counting efficiency. Thus, calibration based on a single-sized phantom could result in inaccurate TBN values. We used Monte Carlo simulations (MCNP-5; Los Alamos National Laboratory) in order to map a system’s response to the range of body weights (65–160 kg) and body fat distributions (25–60%) in obese humans. Calibration curves were constructed to derive body-size correction factors relative to a standard reference phantom, providing customized adjustments to account for differences in body habitus of obese adults. The use of MCNP-generated calibration curves should allow for a better estimate of the true changes in lean tissue mass that many occur during intervention programs focused only on weight loss.     

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.     

The role of NAA in nuclear chemistry education

One of the missions of our Institute is the promotion of basic nuclear teaching for students as well as professional teaching for workers in nuclear industry and research. For nuclear chemistry education, we present here a one day teaching course on radioactive decay and nuclear reactions, and a two or three days course based on reactor irradiation of uranium oxide, instrumental and radiochemical analysis of fission products. In the first experiment, the neutron capture is presented as an example of nuclear reaction; the neutron activation of a silver coin with a Am-Be neutron source, followed by γ-ray spectrometry, is used to identify three radionuclides of silver and to calculate their half-lives. In the second experiment, our teaching reactor is used as a neutron source with a flux about 10¹⁰ n·cm⁻²·s⁻¹ at a low thermal power (10 kW). This low flux allows us to irradiate a small uranium sample which is usable for spectrometry after a short cooling time of about two hours. The first day is reserved for instrumental analysis of the fission products and a second day for the radiochemical separation of a fission radionuclides. With these experimental results, the students have to calculate the number of fissions in the irradiated sample. On optional third day for postgraduate students is devoted to the presentation of NAA and some applications as uranium determination by the fission product spectrometry.     

Comparison of neutron activation analysis methods

This work shows that nuclear data and neutron flux parameters can be applied to calculate directly the elemental concentrations. Techniques for extracting neutron flux parameters pertaining to the irradiation positions and nuclear data pertaining to the isotopes concerned from measured reaction rates have been previously developed. This method is compared to the comparator and relative methods of activation analysis. The principles, advantages and disadvantages of each method and sources of errors are discussed. It also briefly discusses other factors such as accuracy and precision, sensitivity, detection limits and limit of quantitative determination. The three methods are applied to the analysis of five environmental reference materials. The concentrations of more than 20 elements are determined. The results show a good agreement with the certified and/or literature values.