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.     

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 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.     

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.     

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.     

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.     

Gamma ray spectra and sensitivities for 14 MeV neutron activation analysis

The purpose of this study was to define experimentally the sensitivity of determination for 63 different elements by 14 MeV neutron activation, with a 150 kV Cockroft-Walton accelerator at a neutron flux of 2·10⁸ n·cm⁻²·sec⁻¹ on the sample. The obtained gamma ray spectra are given, and the origin of the photopeaks observed are explained. A maximum irradiation time of five minutes was used as a convenient experimental limit to obtain the maximum sensitivity, considering, however, that the tritium target life is limited, and that the time to perform an analysis has to be reasonable. The practical use of 14 MeV neutron activation analysis is demonstrated by the detection limits obtained.     

Fast-neutron activation analysis with short-lived nuclides

At the GKSS Research Center Geesthacht, a new 14 MeV activation facility—a 5·10¹² n/s neutron generator combined with a fast rabbit system (KORONA)—is being installed. Homogeneous neutron flux at a level of 5·10¹⁰ n·cm⁻²·s⁻¹ and sample transfer times of 140 ms to a 16m distant detector station are characteristic features of the facility described in the paper. With special consideration of short-lived nuclides and including cyclic activation, the analytical prospects with the intense neutron source are discussed, and sensitivities for 78 elements are presented.     

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.     

Neutron bombardment of single wall carbon nanohorn (SWCNH): DSC determination of the stored Wigner-Szilard energy

Single wall carbon nanohorn (SWCNH) were neutron-bombarded to a dose of 3.28 × 10¹⁶ n/cm². The Wigner or stored energy was determined by a differential scanning calorimeter and was found 5.49 J/g, 50 times higher than the Wigner energy measured on graphite flakes treated at the same neutron dose. The activation energy for the thermal annealing of the accumulated radiation damage in SWCNH was determined in the range 6.3–6.6 eV against a typical activation energy for the annealing of the radiation-damaged graphite which is in the range of 1.4–1.5 eV. Furthermore the stored energy in neutron-damaged SWCNH is released at 400–430 °C while the main peak in the neutron-damaged graphite occurs at 200–220 °C. The radiation damaged SWCNH were examined with FT-IR spectroscopy showing the formation of acetylenic and aliphatic moieties suggesting the aromatic C=C breakdown caused by the neutron bombardment.     

Applications of k 0 NAA at FRM II with high f values

The research reactor FRM II offers different irradiation facilities with highly thermalized neutron flux. 3 facilities for the k ₀ neutron activation analysis (k ₀ NAA) will be introduced shortly. The influence of flux parameter α on the concentration calculation of samples irradiated in a neutron field with very high ratio of thermal to epi-thermal neutron flux f > 1,000 are here investigated. Even for the most k ₀ isotopes with big Q ₀ values, the uncertainty of a concentration calculation without α correction is <3 %, when the f value larger than 3,000. The uncertainty is about 5 % for the isotope ⁹⁶Zr in this case. The k ₀ library of the computer program MULTINAA is updated. A standard reference material IAEA/soil-7 was analyzed to verify the k ₀ NAA at FRM II.     

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.     

The effects induced by proton irradiation on structural characteristics of nuclear graphite

A grade TSX graphite was irradiated by a 2.5 MeV proton and a dose of 1.47 × 10¹⁸ ion cm⁻² at 330 K. The displacement per atom under this irradiation condition was about 0.02. The lattice parameter, crystallite size and the vacancies density in the graphite was measured before and after irradiation. It was found that the proton irradiation led to an increase in the volume of the sample. The volume change in the irradiated sample was confirmed by atomic force and scanning electron microscopes observations as increased roughness and pore size. Also, FTIR results showed that graphite is slightly oxidized by irradiation.

     

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.     

A pre-concentration neutron activation analysis method for the determination of 232Th utilizing a dry-tube irradiation facility

A revized method for determining ²³²Th using a pre-concentration neutron activation analysis (PCNAA) procedure was developed to accommodate irradiation in a dry irradiation tube environment. ²³²Th extracted by ion-exchange from a sample was electrodeposited onto 5/8″ diameter vanadium planchets, which are arranged in a stack and irradiated in the dry tube central irradiation facility (CIF) of the reactor. The higher neutron fluence of this facility improved sensitivity by approximately 37%, however, the higher temperatures required modifications to the irradiation procedure. Because the heat in the CIF would melt the plastic spacers used in the original method, a tube of high-purity quartz was used to contain samples, and high purity quartz spacers were used to separate the vanadium planchets during the irradiation. Test irradiations have determined that no significant transfer occurred from the disks to the silica disks and no significant variation in the neutron flux was observed. Finally, a thin film barrier was tested for its ability to reduce recoil contamination from ²²⁹Th onto the detector during alpha spectroscopy. The film was shown to reduce contamination to levels indistinguishable from normal background.     

Sulfur determination in coal by 14 MeV neutron activation analysis

A procedure involving the irradiation of coal samples with 14 MeV neutrons and subsequent gamma-ray spectrometry of the irradiated sample for the estimation of solfur in coal, has been outlined. The samples were irradiated with 14MeV neutrons from a Cockroft-Walton type generator for one minute and then subjected to gamma-ray spectrometry for another minute using an automated transfer cyclic system. Ten such cycles were repeated for accumulating events under the 2130 keV gamma ray photopeak belonging to³⁴P (T=12.4 s) produced by the³⁴S(n, p)³⁴P reaction for assessing the lower level of detection, LLD, of Sulfur. Interferences due to the presence of other elements in coal were also determined. Sulfur can be determined at LLD of 0.25% in coal provided a 5 g sample of the coal is irradiated with a neutron flux of 5·10⁹ n·cm⁻²·sec⁻¹ assayed with a gamma ray spectrometer having a large hollow core Ge(Li) detector and an anti-Compton shield.     

Measurements of activation cross-sections for the 101Ru(n,p)101Tc reaction for neutrons with energies between 13 and 15 MeV

In this study, activation cross-sections were measured for the ¹⁰¹Ru(n,p)¹⁰¹Tc reaction at three different neutron energies from 13.5 to 14.8 MeV. The fast neutrons were produced via the ³H(d,n)⁴He reaction on K-400 neutron generator. Induced gamma activities were measured by a high-resolution gamma-ray spectrometer with high-purity germanium detector. Measurements were corrected for gamma-ray attenuations, random coincidence (pile-up), dead time and fluctuation of neutron flux. The data for ¹⁰¹Ru(n,p)¹⁰¹Tc reaction cross-sections are reported to be 15.7 ± 2.0, 18.4 ± 2.7 and 22.0 ± 2.4 mb at 13.5 ± 0.2, 14.1 ± 0.2, and 14.8 ± 0.2 MeV incident neutron energies, respectively. Results were compared with the previous works.     

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%.