High conduction neutron absorber to simulate fast reactor environment in an existing test reactor

A new metal matrix composite material has been developed to serve as a thermal neutron absorber for testing fast reactor fuels and materials in an existing pressurized water reactor. The performance of this material was evaluated by placing neutron fluence monitors within shrouded and unshrouded holders and irradiating for up to four cycles. The monitor wires were analyzed by gamma and X-ray spectrometry to determine the activities of the activation products. Adjusted neutron fluences were calculated and grouped into three bins—thermal, epithermal, and fast—to evaluate the spectral shift created by the new material. A comparison of shrouded and unshrouded fluence monitors shows a thermal fluence decrease of ~11 % for the shielded monitors. Radioisotope activity and mass for each of the major activation products is given to provide insight into the evolution of thermal absorption cross-section during irradiation. The thermal neutron absorption capability of the composite material appears to diminish at total neutron fluence levels of ~8 × 10²⁵ n/m². Calculated values for dpa in excess of 2.0 were obtained for two common structural materials (iron and nickel) of interest for future fast flux experiments.     

Development of a fast-pneumatic sample transfer system for DT neutron activation analysis at the University of Michigan

The development of an automated pneumatic transfer system used to quickly acquire data from materials irradiated with a deuterium–tritium (DT) neutron generator is described in this paper. This system was designed to gather data on short-lived activation and fast-fission products, and was used to characterize the generator’s neutron field. The average sample transit time between irradiation and data acquisition is 363.9 ms at an average velocity of 30.92 m/s (101.3 ft/s). The neutron flux profile as a function of depth into the sample capsule is shown to decrease exponentially, having a maximum flux value of 5.662 × 10⁸ ± 0.056 × 10⁸ n/cm² s. The average DT neutron energy in the system’s sample geometry was determined to be 14.250 ± 0.011 MeV using a unique zirconium–niobium “sandwich” technique. A flux surface equation is also presented as a function of accelerator voltage and deuterium beam current. Methods of analysis are discussed with a proof of a linear flux profile assumption for thin foils.     

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 neutron flux parameters in PUSPATI TRIGA Mark II Research Reactor, Malaysia

In standardization NAA, it is necessary to characterize the neutron spectrum parameters such as epithermal neutron flux shape factor (α), thermal to epithermal neutron flux ratio (f), thermal neutron flux (φ _th) and epithermal neutron flux (φ _epi) in the irradiation facility to determine the concentration of an element in the sample using absolute and k ₀ standardization methods. The α and f were determined using Cd-ratio multi monitor method using experimental data obtained in PUSPATI TRIGA Mark II research reactor at four irradiation positions (10, 20, 30 and 40) of the rotary rack. The calculated values of α and f ranged from 0.006 to 0.0281 and 18.56 to 19.12 respectively. The average values of φ _th and φ _epi were found as 2.33 × 10¹² and 1.23 × 10¹¹ n cm^?2 s^?1 respectively. Moreover, a comparison of the neutron flux parameters in the present study shows an acceptable level of consistency with those of previous studies.     

Mapping of neutron flux gradient at NIST reactor to optimize neutron activation analysis of53Mn

Determination of⁵³Mn in meteorites by neutron activation analysis requires a thermal neutron flux high enough to ensure adequate production of⁵⁴Mn from⁵³Mn with a sufficiently low fast neutron component to minimize its production through fast neutron reactions. Thermal and fast neutron fluxes were mapped as a function of sample position within the NIST research reactor in order to determine the optimum position for irradiation of⁵³Mn.     

Improving neutron activation analysis accuracy for the measurement of gold in the characterization of heterogeneous catalysts using a TRIGA reactor

To measure the gold content of a catalyst accurately, neutron activation analysis (NAA) is one of the methods of choice. NAA is preferred for such heterogeneous catalysts because: (1) it requires minimal sample preparation; (2) NAA provides consistent and accurate results; and (3) in most cases results are obtained much quicker than competing methods. NAA is also used as a referee for the other elemental techniques when results do not fall within expected statistical uncertainties. However, at very high gold concentrations, applying NAA to determine the gold in a heterogeneous catalyst is more challenging than a routine NAA procedure. On the one hand, the neutron absorption cross section for gold is very high, resulting in significant self-shielding related errors. On the other hand, gold exhibits low energy resonance neutron absorptions. In this application the self-shielding minimization effort was handled more rigorously than the classic suppression of neutron flux within a specimen. This non-routine approach was used because: (1) for most applications, high accuracy, <3 % relative, is desired, (2) the low energy resonances of gold make its neutron reaction rate complex and (3) the TRIGA reactor flux profile used in this study contains both thermal and significant epithermal neutron fluxes. Accuracy and precision, using this new approach, are expected to improve from 15 % to better than 3 % relative uncertainty. This has been accomplished through a rigorous assessment of the observed effects of low energy resonance on the neutron flux spectral shape within the sample and designing an experiment to minimize the effects.     

Instrumental neutron activation analysis of silicon wafers using the silicon matrix as the comparator

For the instrumental neutron activation analysis of trace impurities in high purity silicon wafer, a modified single comparator method has been applied. The energy distribution of the neutrons at the irradiation position was measured using the two flux monitors, Au and Co, and elemental contents were calculated using the silicon matrix in the wafer as a comparator. This has advantage of reducing the cross contamination from an external monitor during sample preparation and irradiation, the uncertainties from the non-homogeneity of the neutron flux and the error on the weight of comparators. Determination limits for 49 elements were presented under the condition of 72 hours irradiation at a neutron flux of 3.7·10¹³ n·cm^-2·s^-1 and 4000 s measurement. The analytical results obtained by this method and the conventional single comparator method were compared and were found to agree well within 5%.     

The neutron flux of a252Cf neutron activation analysis device using the MCNP code

This report presents results from the application of the Monte Carlo N-Particle (MCNP) computer code to the²⁵²Cf neutron activation analysis (NAA) Device in the Technical Physics Institute of the Heilongjiang Science Academy of the People's Republic of China. The thermal and epithermal neutron flux at the sample positions and the neutron and photon fluxes on the surfaces of the device were calculated. A comparison between the calculated and experimental thermal and epithermal neutron fluxes at sample positions yield relative errors of less than 10% for the thermal neutron flux.     

Fabrication and characterization of an irradiation facility for large-sample geometry

An irradiation facility consisting of a modified beam port shielding plug has been designed, fabricated built and characterized for use in irradiating non-standard sample geometries. The shielding plug features a graphite moderator at the core end with a hole, or “well” drilled of sufficient diameter and depth to accommodate an eight ounce (227 gram) sample bottle. Added shielding behind the graphite consists of castable neutron- and -gamma-ray shielding. The modified shielding plug can be removed relatively quickly from its irradiation position to minimize personnel exposures. It is mounted in close proximity to the Ohio State University Research Reactor reactor core to allow performance of high-sensitivity neutron activation analysis studies. Using the SAND-II unfolding code, the energy-dependent neutron flux has been measured in the sample irradiation position. When operating at 100 % power, the total flux is 3.9 × 10¹² n/cm²/s. Of this, 55 % is thermal (<0.5 eV), 23 % is epithermal (>0.5 eV, <0.5 MeV), and 22 % is “fast” (>0.5 MeV). This makes the facility suitable for neutron activation studies. Recently it has been used for irradiation of filter papers collected in a study of particulate air pollution in the form of atmospheric particulate matter in an urban environment.     

A pilot study to measure levels of selected elements in Thai foods by instrumental neutron activation analysis

A pilot study was carried out to evaluate the scope of instrumental neutron activation analysis (INAA) for measuring the levels of selected elements in a few commonly consumed food items in Thailand. Several varieties of rice, beans, aquatic food items, vegetables and soybean products were bought from major distribution centers in Bangkok, Thailand. Samples were prepared according to the protocols prescribed by the nutritionist for food compositional analysis. Levels of As, Br, Ca, Cd, Cl, Cr, Cu, Fe, K, Mg, Mn, and Zn were measured by INAA using the irradiation and counting facilities available at the Thai Research Reactor with the maximum in-core thermal neutron flux of 3 × 10¹³ cm^?2 s^?1 of the Thailand Institute of Nuclear Technology in Bangkok. Selenium was determined by cyclic INAA using the Dalhousie University SLOWPOKE-2 Reactor facilities in Halifax, Canada at a thermal neutron flux of 2.5 × 10¹¹ cm^?2 s^?1. Both cooked and uncooked foods were analyzed. The elemental composition of food products was found to depend significantly on the raw material as well as the preparation technique.     

Utilization of pneumatic carrier facility of Dhruva reactor for trace element determination by neutron activation analysis

The pneumatic carrier facility (PCF) of Dhruva reactor is being extensively used for neutron activation analysis (NAA) studies pertaining to research work as well as routine sample analysis. It is useful for the determination of trace elements using short and medium half-lives radioisotopes produced in neutron activation with available higher neutron flux (~5 × 10¹³ cm^?1 s^?1). Solid samples placed in high density polypropylene capsule, are irradiated for 1 min duration and radioactive assay is carried out by high resolution gamma ray spectrometry. Design aspects of PCF and various applications to samples of diverse matrices using NAA are presented.     

Rare earth element (REE) in surface mangrove sediment by instrumental neutron activation analysis

A study is carried out on the concentrations of rare earth element (REE) elements present in surface mangrove sediments from 10 locations throughout west coast Malaysia. In carrying out the analysis, the best and most convenient method being the instrumental neutron activation analysis (INAA). Samples were obtained, dried, crushed to powdery form and samples prepared for INAA. All the samples for analysis were weighted approximately 150 mg for short irradiation and 200 mg for long irradiation time. As calibration and quality control procedures, blank samples, standard reference material SL-1 were then irradiated with thermal neutron flux of 4 × 10¹² cm^?2 s^?1 at the MINT TRIGA Mark II research reactor which operated at 750 kW by using a pneumatic transport facility. The REE elements of surface sediment samples in this study are Dy, Sm, Eu,Yb, Lu, Tb, La and Ce. It was found that the level of concentrations of all the REE elements varies in the range (0.35–117.4 mg/kg). The geochemical behavior of REEs in surface sediments and normalized pattern (chondrite and shale) has been studied. The degree of sediments contaminations were computed using an enrichment factor. The results showed that the enrichment factor varied in the range (0.75–6.75).     

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.     

Determination of cobalt in solid matrices by neutron activation analysis

The concentration of cobalt in 2 solid matrices was determined by neutron activation analysis (NAA) using standard solutions which were prepared by dissolving pure cobalt in nitric acid. The matrices assayed were a cobalt-aluminium wire and an iron foil and the respective Co concentrations found were 0.488% and 0.138%. Both solid materials can equally be used as standard references of cobalt in NAA. Subcadmium and epicadmium neutron fluxes in the reactor core were determined using Co?Al and Au?Al alloy wires. Very good agreement was obtained for all irradiation configurations of the target monitors cobalt and gold.     

Second-order interference in the neutron activation analysis of arsenic in a germanium matrix

The second-order interference ⁷⁴Ge(n, γ; β^-, n, γ)⁷⁶ As can occur in the activation analysis of arsenic in a germanium matrix, using thermal neutrons. As the literature data show poor agreement, this interference was determined experimentally. A practical formula was derived, for irradiation times longer than 2 h, which showed that the interference, expressed as an apparent arsenic concentration, is proportional to the neutron flux. Experiments were performed for irradiation times of 10, 15 and 20 h at a neutron flux of 10¹⁴ n/cm²/sec, yielding apparent arsenic concentrations in the germanium matrix of respectively 223, 408 and 597 p.p.b. From these results a value of 0.48 ±0.06 barn could be calculated for the activation cross-section of ⁷⁴Ge for neutron capture.     

Towards routine NAA of materials rich in heavy elements with iterative gamma-ray attenuation and neutron self-shielding calculations

Solids and powders can be analysed directly and with good accuracy by neutron activation analysis without sample preparation because of the excellent penetrating powers of neutrons and gamma rays. However, if the sample contains high concentrations of gamma-absorbing heavy elements or neutron-absorbing elements, the analysis results must be corrected for neutron self-shielding and gamma-ray attenuation. These effects are coupled and depend on the chemical composition of the sample, which is the final result of the analysis. Thus, the correction calculation must be iterative. In this work we performed the first coupled iterative corrections of the two effects. Six test samples were prepared by mixing powders containing compounds of Cd, a neutron absorber, and the rare-earth elements Ce, Pr and Nd with concentrations as high as 47 %. The samples were irradiated in the SLOWPOKE research reactor and counted with a germanium gamma-ray detector. In the samples with the highest heavy element concentrations, the uncorrected Neutron activation analysis results were in error by as much as 55 %. The results were corrected iteratively using the neutron self-shielding model coupled with the gamma-ray attenuation model, and the final corrected results were accurate to 5 % or better.     

Design and construction of a facility for neutron activation analysis using the 14 MeV neutron generator at HMS Sultan

The potential for using a small, sealed tube, DT neutron generator for neutron activation analysis has been well documented but not well demonstrated, except for 14 MeV activation analysis. This paper describes the design, construction and characterization of a neutron irradiation facility incorporating a small sealed tube DT neutron generator producing 14 MeV neutrons with fluence rates of 2·10⁸ s⁻¹ in 4π (steady state) and 10¹¹ s⁻¹ in 4π (pulsed). Monte Carlo modeling using MCNP4c and McBend9 has been used to optimize the design of this facility, including the location of a thermal irradiation facility for conventional neutron activation analysis. A significant factor in designing the facility has been the requirement to conform with Ionising Radiation Regulations and the design has been optimized to keep potential radiation doses to less that 1 μSv/h at the external walls of the facility. Activation of gold foils has been used for flux characterization and the experimental results agree well with the modeling.     

Activation analysis of heavy metals in surface waters using ion exchange filter papers and cyanide complexing

This paper describes a simple, reasonably rapid, and accurate technique developed for the multielement analysis of heavy metals in surface waters using ion exchange filter papers and cyanide complexing in conjunction with neutron activation analysis. The method employs large water volume irradiation for improved sensitivity.¹³⁰Te was used as an internal flux monitor to correct for neutron flux variations in the samples. Proof of the value of this technique was obtained by the analysis of 10 surface water samples previously analyzed using conventional methods by the U. S. Geological Survey of the U. S. Department of Interior. Good agreement resulted from this comparison.     

Determination of 129I in cement-solidified radwastes using neutron activation

The purpose of this study is to develop a neutron activation method to determine trace amounts of ¹²⁹I in cement-solidified radwastes. The radwaste samples were alkaline fused using KOH and then ¹²⁹I and iodine carrier were chemically separated by solvent extraction before and after neutron irradiation. Both stable iodine (¹²⁷I) and ¹²⁹I can be activated by neutrons through ¹²⁷I (n, 2n) ¹²⁶I and ¹²⁹I (n, γ) ¹³⁰I reactions; their activated radionuclides were counted together with a high-purity germanium detector. The chemical recovery yields ranged from 30 to 60 %, and it was found that more than 99.9 % of interfering radionuclides can be removed using solvent extraction after neutron irradiation. The minimum detectable amounts can be lowered to less than 1 mBq g^?1, which is superior to low energy γ-ray spectrometry by a factor of >10², on average. The established technique can be applied to re-evaluation of ¹²⁹I content in radwastes that can be re-classified to lower classes, and the cost for designing a final disposal facility can be significantly reduced.     

Preparation of radio-Sm by neutron activation for accelerator mass spectrometry

Field measurement of isotopic ratios may be used to fingerprint an element’s origin, be it from commercial power, industrial, medical or historical weapons fallout. Samples of samarium radionuclides were prepared by neutron activation for subsequent analysis using accelerator mass spectrometry (AMS). High purity samarium (III) oxide powder was irradiated in the University of Texas at Austin TRIGA reactor to a total neutron fluence of 5 × 10¹⁵ cm^?2. An initial determination of the isotopic ratios was made using activation calculations with a BURN card in an MCNPX-based model of the TRIGA core. Experimental validation of the MCNP results was achieved by analyzing gamma spectra of the irradiated oxide powers after irradiation. Subsequent measurement of ¹⁵¹Sm will be conducted at the CAMS facility at LLNL demonstrating the first measurement of this isotope at this facility.