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1.
An instrumental thermal neutron activation analysis facility based, on a 16 Ci 241 Am–Be source, a high resolution -ray spectrometry setup and a PC-based data acquisition system at KFUPM is described. The thermal neutron flux distribution was determined from the induced activities of high purity indium foils. The absolute thermal neutron flux was calculated from the activities of bare and cadmium-covered gold foils at a position of 3 cm from the soource at which the flux reaches a maximum. The facility tests were carried out with the determination of manganese concentrations in six types of industrially important steel samples. The result of 1.33% manganese in SS-304 steel sample was in excellent agreement with the literature value. The method is nondestructive, economical and ideal for bulk analysis. 相似文献
2.
Thermal neutron analysis (TNA) technology has been used for the non-destructive detection of explosives. The system uses a relatively weak 252Cf neutron source (1.03·10 7 n/s) and two 3"×3" NaI(Tl) detectors. The presence of explosives is confirmed via detection of the 10.83 MeV prompt gamma-ray associated with nitrogen decay. The MCNP4A code was used to simulate the neutron and gamma transport through the system. The thermal neutron flux in the activation position was measured using gold and indium foils. The measured thermal neutron flux was lower, by not more than 9.5%, than that of simulation. In this report the results of the preliminary tests on the system are described. 相似文献
3.
The spatial distribution of neutrons was measured at the muon science laboratory of KEK by the activation detector method
using an imaging plate for the radioactivity measurement. It was confirmed that this method is highly sensitive to detect
the average neutron dose of 10 μSv/h. The distribution of thermal and epithermal neutrons was also measured in the experimental
room. The cadmium ratio inside the experimental room is one except for the neutron leakage point. The spatial distribution
of neutrons inside the concrete shield of KENS was measured by the same method. Aluminum and gold foils were used for the
measurement of fast and thermal neutrons, respectively. Two dimensional change of the reaction rate of the 27Al(n,α) 24Na reaction shows a good agreement with the results calculated by the Monte Carlo simulation using MARS14 code. Thermal and
epithermal neutron flux ratio on the beam axis was measured by the cadmium ratio method. The flux ratios were about 30 and
almost constant for every slot except for the surface of the shield, because the cadmium ratio is 2. This method was very
useful to measure the activity of many pieces of detector simultaneously without any efficiency and decay correction. Wide
dynamic range and high sensitivity are also the merit of this method. 相似文献
4.
An alternative method of approach has been developed for the measurement of thermal neutron flux. The method depends only
on the activity of the bare foil if the cadmium ratio at the irradiation position is known. The method has been tested on
the GHARR-1 facility at the Ghana Atomic Energy Commission using gold and indium foils for the measurement of the thermal
neutron flux in the flux range of 10 10–10 12 n·cm −2·s −1 and the results compare very well with those obtained using the conventional method (cadmium separation method). 相似文献
5.
With the advent of Ge(Li) spectrometry, a high standard of purity for neutron flux monitors no longer remains an imperative
“must” and becomes rather superfluous. From this standpoint, commercial grade Al was investigated for its suitability as a
reactor flux monitor and was found to have a much greater practical utility than most of the monitors reported. Three Al foils
and one wire were randomly selected from four different commercial sources, and analysed for their Fe, Ga, Mn and Na contents
by neutron activation and high-resolution gamma-spectrometry. While Na was found to have a very heterogeneous distribution,
Fe, Mn and Ga concentrations in different splits of each type of Al were consistently uniform within ±2–3%. Eight possible
monitor reactions on Al, Mn, Fe and Ga have been recommended as neutron flux integrators for all the 3 components of a reactor
spectrum, viz. thermal, epithermal and fast, covering a wide range of flux levels from 10 7 to 10 14n·cm −2·sec −1. The advantages and versatility of commercial grade Al as a pile neutron dosimeter are discussed. 相似文献
6.
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. 相似文献
7.
Measurement of neutron flux micro distribution within absorbing samples have been made using Dy and Au foils activated in
cadmium filter for both thermal and epithermal regions, respectively. Self-shielding effect for highly absorbing material
has been investigated for the purpose of activation analysis in the Syrian Miniature Neutron Source Reactor (MNSR). Self-shielding
factors for both thermal and resonance regions have been evaluated. Measured and theoretical values of self-shielding factors
have been compared. Correction expression of the neutron activation analysis measurement for self-shielding has been proposed.
This revised version was published online in July 2006 with corrections to the Cover Date. 相似文献
8.
A subcritical nuclear reactor, Model 9000, Nuclear Chicago, is installed and operating at the Aristotle University of Thessaloniki,
in the Atomic and Nuclear Physics Laboratory, at Thessaloniki, Northern Greece. The fuel is about 5500 lbs (2495 kgs) natural
uranium metal (U 3U 8), the moderator about 3600 lbs (1633 kgs) light water, H 2O and the reflector is also light, water, H 2O. The lattice core is hexagonal, 42 inches (1.07 m) high and of 35 inches (88.90 cm) maximum diameter. The neutron source
at the core is Am-Be 5 curies (185 GBq), 1.1·10 7n·s −1. The reactor is used for the activation of various materials by neutrons such as indium, the determination of the thermal
neutrons flux, the horizontal and the vertical distribution of the neutron flux, material buckling, B, and geometric buckling, B, the parameters of the reactor, and the albedo of water for thermal neutrons with foils of indium.
This revised version was published online in July 2006 with corrections to the Cover Date. 相似文献
9.
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 2·10 8 s −1 in 4π (steady state) and 10 11 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. 相似文献
10.
Characteristics of a source of thermal neutrons based on an evacuated NG-400 neutron generator with the maximum flux (Φ f) 2 × 10 11 neutron/s for 14 MeV neutrons and 2 × 10 9 neutrons/s for 3 meV neutrons have been investigated. The possibilities of its application for neutron activation analysis
have been estimated. The distribution, composition, and density (φ T) values of the thermal neutron flux have been measured in the inner cavity of the moderator using activation detectors. φ T was 2 × 10 8 and 2 × 10 6 neutrons/cm 2 s for thermalized neutrons with energies of 14 and 3 MeV, respectively. The possibilities of the apparatus have been estimated
theoretically and experimentally for the cases of thermalized neutrons of 14 MeV and 3 MeV. 相似文献
11.
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. 相似文献
12.
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 × 10 8 ± 0.056 × 10 8 n/cm 2 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. 相似文献
13.
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. 相似文献
14.
An annular 227AcBe isotopic neutron source, containing 6.6-Ci 227Ac, is described for application in fast and thermal neutron activation analysis, with high accuracy, for major constituents in ores, alloys and industrial concentrates. The characteristics of the neutron output and of the fast, epithermal and thermal flux and flux gradients is described in detail. The determination of manganese in pyrolusite ores and ferro-manganese is compared to results obtained previously with a cylindrical 1-Ci 226RaBe source. Two new sources of systematic errors have been discovered. 相似文献
15.
It has been difficult to characterize the thermal to epithermal neutron flux ratio ( f) and the measure of the nonideal epithermal neutron flux distribution ( α) for the RT-2 pneumatic rabbit facility at the NIST National Bureau of Standards Reactor (NBSR). In a previous paper, only cadmium-covered irradiations yielded physically reasonable parameters. New measurements were performed using chromium, manganese, cobalt, zinc, zirconium, molybdenum, antimony, gadolinium, lutetium, and gold. The neutron temperature ( T n ) in RT-2 measured using bare lutetium and gold foils gave unphysical values. The bare foil methods for measuring f and α gave inconsistent results. The underlying reasons are demonstrated via MCNP simulation results for cumulative reaction rates of selected isotopes. To determine expected intervals for f, α and T, parametric methods were explored. Measured reaction rate probability per target atom ( R p ) values for the listed elements were fitted to a modified Westcott curve using an iterative least-squares method to verify consistency of measurements and nuclear data. An advanced parametric approach using a detailed MCNP model of the NBSR was used to calculate neutron flux characterization parameters. 相似文献
16.
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. 相似文献
17.
In this study, the total thermal neutron macroscopic and microscopic cross sections of V, Co, Cu, In, Dy and Au were measured using neutron self-absorption properties. Pure foils of these elements with various thicknesses were irradiated using a 5 Ci 238Pu–Be source. After the irradiation, the gamma-spectra of their radionuclides were recorded by an HpGe detector. The gamma-photopeak areas of interest were determined by evaluating the gamma-spectra obtained from the foils. They were plotted as a function of foil thickness for each element. Then a non-linear least-squares fitting method was applied to the functions, and the total thermal neutron macroscopic and microscopic cross sections of V, Co, Cu, In, Dy and Au were obtained. 相似文献
18.
This report presents results from the application of the Monte Carlo N-Particle (MCNP) computer code to the 252Cf 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. 相似文献
19.
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 10 15 n s ?1, the facility provides the thermal neutron flux 2.0 × 10 11 n cm ?2 s ?1 and a slightly higher fast neutron flux 2.3 × 10 11 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. 相似文献
20.
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 the 52V-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 Ci 227Ac—Be isotope neutron source. After a decay of 30 s, the 52V-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. 相似文献
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