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.     

Fine structure of mass and charge distribution in low energy fission

Studies of finer details in mass and charge distribution fission leads to a better understanding of the fission process. Experimental determination of independent and cumulative yields using radiochemical techniques as well as mass spectrometers and fission product recoil separators form the basis of such studies. It has been established that closed shells as well as an even number of nucleons influence both mass and charge distributions. The magnitudes of these effects may be estimated from existing experimental yield data and various fission models. Using our measurements of several fission yields and those existing in the literature we have calculated even-odd proton and neutron effects for various low energy fissioning systems. Where enough data existed, direct calculations were made, whereas for other cases the Z_p-model of WAHL has been used. It is found that the even-odd proton effect is well established and pronounced in thermal neutron fission of²³⁵U and²³³U. Lesser effects were found for reactor neutron induced fission of²³²Th, thermal neutron fission of²³⁹Pu and spontaneous fission of²⁴⁵Cm and²⁴⁹Cf. No effect seems to exist in the thermal neutron fission of²⁴¹Pu and the spontaneous fission of²⁵²Cf. The even-odd neutron effect is found to be much lower than the corresponding proton effect in²³⁵U and²³³U fissions and is nonexistent in the rest of the fissioning systems.     

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.     

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.     

The analysis and attribution of the time-dependent neutron background resultant from sample irradiation in a SLOWPOKE-2 reactor

The Royal Military College of Canada (RMCC) has commissioned a Delayed Neutron Counting (DNC) system for the analysis of special nuclear materials. A significant, time-dependent neutron background with an initial maximum count rate, more than 50 times that of the time-independent background, was characterised during the validation of this system. This time-dependent background was found to be dependent on the presence of the polyethylene (PE) vials used to transport the fissile samples, yet was not an activation product of vial impurities. The magnitude of the time-dependent background was found to be irradiation site specific and independent of the mass of PE. The capability of RMCC’s DNC system to analyze the neutron count rates in time intervals <1 s facilitated a more detailed data analysis than that obtained in previous DNC systems recording cumulative neutron counts. An analysis of the time-dependent background behaviour suggested that an equivalent of 120 ng of ²³⁵U contamination was present on each irradiated vial. However, Inductively Coupled Plasma—Mass Spectroscopy measurements of material leached from the outer vial surfaces after their irradiations found only trace amounts of uranium, 0.118 ± 0.048 ng of ²³⁵U derived from natural uranium. These quantities are insufficient to account for the time-independent background, and in fact could not be discriminated from the noise associated with time-independent background. It is suggested that delayed neutron emitters are deposited in the vial surface following fission recoil, leaving the main body of uranium within the irradiation site. This hypothesis is supported by the physical cleaning of the site with materials soaked in distilled water and HNO₃, which lowered the background from a nominal ²³⁵U mass equivalent of 120 to 50 ng per vial.     

Moderator for neutron activation with the photoneutrons produced by a LINAC

Using Monte Carlo methods a polyethylene moderator has been designed to induce activation using the photoneutrons field of a 15 MV linear accelerator for radiotherapy. In the calculations the photoneutron spectrum at 1 m from the isocenter was used as a source term and the neutron spectra were calculated in the center of different size cylindrical moderators. The best size was selected defining the thermal-to-fast-neutron ratios as a figure of merit. The moderator was built and its performance was evaluated by inducing the activation of Mn dissolved in water, silver coins and souvenir coin. The thermal neutron fluence rate was determined with the Mn samples being 9.96 × 10⁵ cm^?2 Gy _x ^?1 .     

Performance evaluation of k 0-instrumental neutron activation analysis and flame atomic absorption spectrophotometry in the characterization of various types of alloys

Certified alloys of Ni–Cu based, Fe based and Cu–Sn based were analysed by semi-absolute, standardless k ₀-instrumental neutron activation analysis (k ₀-INAA) and flame atomic absorption spectrophotometry (FAAS) aiming at evaluating their comparative performances. In k ₀-INAA measurements, the irradiations were performed at miniaturized neutron source reactor having thermal neutron flux of about 1 × 10¹² cm^?2 s^?1. The experimentally optimized parameters for INAA suggested a maximum of three irradiations for the quantification of 21 elements within 5 days. The same experiments also produced quantitative results of 13 elements not reported in the certificates of the reference materials. AAS was, however, unable to determine any of those elements. Accuracy of the two techniques was assessed by comparing their average root mean squared errors. The data analysis concluded that k ₀-INAA had better sensitivity and accuracy than FAAS.     

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.     

Study of the gamma spectrum of 16N with a BGO detector, for the purpose of calibration and of determining the fluorine grade of mineral samples

Over the past several years, the Pacific Northwest National Laboratory (PNNL) has developed an ultra-low-background proportional counter (ULBPC) technology. The resulting detector is the product of an effort to produce a low-background, physically robust gas proportional counter for applications like radon emanation measurements, groundwater tritium, and ³⁷Ar. In order to fully take advantage of the inherent low-background properties designed into the ULBPC, a comparably low-background dedicated counting system is required. An ultra-low-background counting system (ULBCS) was recently built in the new shallow underground laboratory at PNNL. With a design depth of 30 m water-equivalent, the shallow underground laboratory provides approximately 100× fewer fast neutrons and 6× fewer muons than a surface location. The ULBCS itself provides additional shielding in the form of active anti-cosmic veto (via 2-in-thick plastic scintillator paddles) and passive borated poly (1 in.), lead (6 in.), and copper (~3 in.) shielding. This work will provide details on PNNL’s new shallow underground laboratory, examine the motivation for the design of the counting system, and provide results from the characterization of the ULBCS, including initial detector background.     

A compilation of the fission spectrum average neutron cross-sections for (n, α) reactions

The (n, α) cross-sections averaged over the²³⁵U thermal fission neutron spectrum are compiled. The original values and whenever possible renormalized values are presented in parallel. The methods used for measurements and calculations are briefly discussed.     

The new cold neutron depth profiling instrument at NIST

The Cold Neutron Depth Profiling (CNDP) instrument at the NIST Cold Neutron Research Facility (CNRF) is now operational. The neutron beam originates from a 16 liter D₂O-ice cold source and passes through a filter of 13.5 cm of single crystal sapphire. The neutron energy spectrum may be described by a 65 K Maxwellian distribution. The sample chamber configuration allows for remote controlled scanning of 15 cm×15 cm samples and varying of both sample and detector angle. The improved sensitivity over the current thermal depth profiling instrument has permitted the first nondestructive measurements of¹⁷O profiles. Results of some of the first sample measurements are presented.     

A new reflector structure for facility thermalizing D–T neutrons

A facility for thermalization of fast neutrons (14.2 MeV) emitted by compact deuterium–tritium (D–T) neutron generators (NGs) for thermal neutron activation analysis is proposed. Its final design is based on Monte Carlo calculations (MCNP5). To maximize the ratio between the thermal neutron flux and the total neutron flux and simultaneously to ensure the highest possible value of the thermal neutron flux at the output surface, the facility should consist of a two-layer reflector [tungsten (W)—the inner part, molybdenum—the outer part], a two-layer multiplier (W followed by lead), a moderator (polyethylene followed by magnesium fluoride) and a collimator (molybdenum and nickel near the output surface). For the D–T NG producing the maximum available neutron yield 10¹⁵ n s^?1, the facility provides the thermal neutron flux 2.0 × 10¹¹ n cm^?2 s ^?1 and a slightly higher fast neutron flux 2.3 × 10¹¹ n cm^?2 s^?1. To improve the ratio of the thermal neutron flux to the fast neutron flux (above 2.7) an addition of a silicon layer to the moderator and especially a proper adjustment and a threefold increase of the multiplier thickness is necessary.     

Syntheses and thermal analyses of curium trichloride

Curium trichloride was synthesized by the solid state reaction of curium nitride with cadmium chloride heated up to 748 K in a dynamic vacuum. The product was hexagonal ²⁴⁴CmCl₃, of which lattice parameters were determined to be a = 0.7385 ± 0.0005 and c = 0.4201 ± 0.0005 nm. The melting temperature of the ²⁴⁴CmCl₃ sample was determined to be 970 ± 3 K by differential thermal analyses using a gold crucible. These values are close to those reported in literature. The results show that mg-scale CmCl₃ samples for thermochemical measurements were prepared from the purified oxide sample without the use of corrosive reagents.     

Selenium content of Argentinean infant formulae and baby foods by pseudo-cyclic instrumental neutron activation analysis coupled to Compton suppression

The selenium levels of Argentinean infant formulae and baby food were measured using the 162-keV gamma-ray of ^77mSe (t _½ = 17.4 s) by a pseudo-cyclic instrumental neutron activation analysis (PC-INAA) method in conjunction with Compton suppression spectrometry (CSS). For comparison purposes, 5 selected infant formulae were also analyzed for selenium by a radiochemical neutron activation analysis (RNAA) method. The selenium levels for three samples agreed between ±2.8 and 6.5 % while the other two differed by 12 and 17 % which could perhaps be attributed to sample inhomogeneity. The selenium content of cow milk-based infant formulae varied from 42–146 μg kg^?1 compared to 52–63 μg kg^?1 for soy-based milk formulae. In the case of baby foods, the selenium levels varied from 34 to 74 μg kg^?1. The detection limits for selenium by PC-INAA–CSS for all the samples analyzed in this work were between 8.5 and 65 μg kg^?1 depending on the major elements present in the samples, while it was 20 μg kg^?1 for the RNAA method. The expanded uncertainty (κ = 2) of the PC-INAA–CSS method was 7.0 % at the end of cycle #4 for a sample containing 73.7 μg kg^?1 selenium compared to the RNAA value of 24.2 % for a sample of 67.0 μg kg^?1 selenium content.     

Design calculation of a horizontal thermal neutronic beam for neutron radiography at the Syrian MNSR

The computer code MCNP4C and the ENDF/B-V cross-section library were used to design calculation of a horizontal thermal beam for neutron radiography (NR) at Syrian MNSR and to evaluate the safety of the reactor after installation of the NR facility (NRF). Thermal, epithermal and fast neutron energy ranges were selected as <0.30 eV, 0.30 eV–10.0 keV and >10.0 keV, respectively. To produce a good neutron beam in terms of intensity and quality, bismuth (Bi) and silicon (Si) were used as photon and neutron filters, respectively. The ratio of L/D of the NRF ranges between 90 and 125. The thermal neutron flux at the beam exit plane can be varied from 1.836 × 10⁵ to 3.057 × 10⁵ n/cm² s. If such thermal neutron beam would be built into the Syrian MNSR, many scientific applications of the NR would be available.     

Design,construction and characterization of a prompt gamma activation analysis facility at the Oregon State University TRIGA<Superscript>®</Superscript> reactor

A prompt gamma neutron activation analysis facility has been designed, built, and characterized at the Oregon State University TRIGA^® reactor. This facility was designed for versatile multi-elemental analyses. The facility utilizes the leakage neutrons originating from beam port #4 of the Oregon State University TRIGA^® reactor. The neutrons are collimated through a series of lead and Boral^® collimators, and filtered through both a bismuth filter and single-crystal sapphire. Samples are irradiated in a sample chamber outside the biological shielding of the reactor, and the resulting gamma radiation produced from neutron interactions within the sample is monitored using a high-purity germanium detector (HPGe). The thermal and epithermal neutron fluxes were measured using gold-foil irradiations and found to be 2.81 × 10⁷ and 1.70 × 10⁴ cm^?2 s^?1, respectively. The resulting cadmium ratio was 106. Measured detection limits for boron, chlorine, and potassium in a NIST SRM 1571 orchard leaf were 5.6 × 10^?4 mg/g, 8.2 × 10^?2 mg/g, and 1.0 mg/g, respectively. Detection limits for additional elements and samples are presented.     

Evaluation of INAA–anticoincidence counting for measuring nanogram levels of vanadium in relatively high-salt nutritional reference materials

An instrumental neutron activation analysis method in conjunction with anticoincidence counting (INAA–AC) gamma-ray spectrometry was developed for the determination of ppb levels of V in biological, mostly nutritional, reference materials containing varying amounts of salt. The method involved irradiation in the Dalhousie University SLOWPOKE-2 reactor facility at a fission neutron flux of 5 × 10¹¹ cm^?2 s^?1 for 1 min, decay for 1 min, and counting for 10 min. In order to fully investigate the extent of improvement that can possibly be obtained for V determination by INAA–AC, a theoretical term called the analytical figure of merit was developed and applied to 16 National Institute of Standards and Technology and International Atomic Energy Agency reference materials. The overall background around the 1,434.1-keV photopeak of ⁵²V was reduced by a factor of 5–10 for several materials in the anticoincidence counting mode. The detection limits were lowered by factors of 3–5 in INAA–AC (0.61–9.4 μg kg^?1) compared to conventional INAA (1.9–79 μg kg^?1) in samples with varying ratios of Na/V (0.24–1,000), Cl/V (0.12–1,827), Al/V (7.45–115) and Mn/V (1.84–66.9) making rapid and reliable V measurements possible at sub-ppb levels without any chemical separation.     

BCR certified reference materials for reactor neutron dosimetry

A series of reference materials intended for use as activation or fission monitors for neutron fluence rate measurements has been prepared by the Joint Research Centre of the European Commission. Certification has been carried out by expert European laboratories and distribution of the certified reference materials (CRMs) is through the BCR programme of the Commission. The list (18 CRMs) includes materials to cover the complete energy spectrum, and suitable for different irradiation times. Fission monitors are ²³⁸UO₂ or ²³⁷NpO₂ in the form of microspheres. Activation monitors are high purity metals (Ni, Cu, Al, Fe, Nb, Rh, or Ti), certified for interfering trace impurities, or dilute aluminium-based alloys, where aluminium is chosen as a suitable matrix for reducing the neutron self-shielding effect. Newly certified materials are IRMM-530R Al-0.1%Au, replacing the exhausted IRMM-530 material, used as comparator for k ₀-standardization, and three new Al-Co alloys (0.01-1%Co). Two others, in the process of certification are Al-0.1%Ag and Al-2%Sc for thermal and epithermal fluence rate measurements. Other candidate reference materials currently being certified are two uranium-doped glass intended for dosimetry by the fission-track technique.     

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.     

Radioanalytical prediction of radiative capture in 99Mo production via transmutation adiabatic resonance crossing by cyclotron

In this study, the transmutation adiabatic resonance crossing (TARC) concept was estimated in ⁹⁹Mo radioisotope production via radiative capture reaction in two designs. The TARC method was composed of moderating neutrons in lead or a composition of lead and water. Additionally, the target was surrounded by a moderator assembly and a graphite reflector district. Produced neutrons were investigated by (p,xn) interactions with 30 MeV and 300 μA proton beam on tungsten, beryllium, and tantalum targets. The ⁹⁹Mo production yield was related to the moderator property, cross section, and sample positioning inside the distinct region of neutron storage as must be proper to achieve gains. Gathered thermal flux of neutrons can contribute to molybdenum isotope production. Moreover, the sample positioning to gain higher production yield was dependent on a greater flux in the length of thermal neutrons and region materials inside the moderator or reflector. When the sample radial distance from Be was 38 cm inside the graphite region using a lead moderator design, the production yield had the greatest value of activity, compared with the other regions, equal to 608.72 MBq/g. Comparison of the two designs using a Be target revealed that the maximum yield occurred inside the graphite region for the first design at 38 cm and inside the lead region for the second design at 10 cm. The results and modeling of the new neutron activator were very encouraging and seem to confirm that the TARC concept can be used for ⁹⁹Mo production in nuclear medicine.