共查询到20条相似文献,搜索用时 31 毫秒
1.
Wayne A. Taylor Robert S. Rundberg Evelyn M. Bond Francois M. Nortier David J. Vieira 《Journal of Radioanalytical and Nuclear Chemistry》2009,282(2):391-394
Lutetium has been used as a radiochemistry detector to measure neutron fluence in NTS tests. A measure of the neutron capture
cross sections on 173Lu is needed to improve the interpretation value of the Lu radiochemistry isotopic ratios. A natural hafnium target was irradiated
with protons to produce neutron poor lutetium radioisotopes. The short lived species were allowed to decay prior to chemical
processing resulting in predominantly 173Lu with a small amount of 174Lu. This material was deposited on a titanium foil for use in the neutron capture cross section measurement. 相似文献
2.
J. G. Fantidis G. E. Nicolaou C. Potolias N. Vordos D. V. Bandekas 《Journal of Radioanalytical and Nuclear Chemistry》2011,290(2):289-295
A Prompt Gamma Ray Neutron Activation Analysis (PGNAA) system, incorporating an isotopic neutron source has been simulated
using the MCNPX Monte Carlo code. In order to improve the signal to noise ratio different collimators and a filter were placed
between the neutron source and the object. The effect of the positioning of the neutron beam and the detector relative to
the object has been studied. In this work the optimisation procedure is demonstrated for boron. Monte Carlo calculations were
carried out to compare the performance of the proposed PGNAA system using four different neutron sources (241Am/Be, 252Cf, 241Am/B, and DT neutron generator). Among the different systems the 252Cf neutron based PGNAA system has the best performance. 相似文献
3.
S. E. Jarman J. Pinchin J. M. Brushwood T. McCarthy M. Bray P. A. Beeley 《Journal of Radioanalytical and Nuclear Chemistry》2007,271(1):47-53
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. 相似文献
4.
N. K. Mumba L. Vas Cs. M. Buczkó 《Journal of Radioanalytical and Nuclear Chemistry》1985,95(5):311-322
Applicability of a small neutron generator and a dual rectangular tube sample transfer system for analyses of U and Th using delayed fission neutron technique has been investigated. A way of optimizing the timing parameters is reported. At a fast neutron flux of 108 n.cm–2s–1, 0.02 w% U can be determined. For thorium determination this method is less sensitive. The Cd difference technique can be used for the simultaneous determination of U and Th but it has lower sensitivity.This work was supported in part by the IAEA. 相似文献
5.
R. Pepelnik H. -U. Fanger W. Michaelis B. Anders 《Journal of Radioanalytical and Nuclear Chemistry》1982,72(1-2):393-403
A new intense 14 MeV neutron generator with cylindrical acceleration structure has been put in operation at the GKSS Research
Center Geesthacht. The sealed neutron tube is combined with a fast pneumatic rabbit system with particular capabilities for
neutron activation analysis involving shortlived reaction products. The sample transfer time is less than 140 ms. The maximum
neutron flux available for activation is 5.2·1010 n/cm2s. Theoretical sensitivity predictions made in a previous study have been verified for some important trace elements. As a
first application, samples of freeze-dried suspended matter and fishes of the Elbe river were analyzed. 相似文献
6.
R. Ganapathy 《Journal of Radioanalytical and Nuclear Chemistry》1978,43(1):131-138
The thermal neutron activation cross-section of94Zr was found to be 49.3±0.6 millibarns. It is shown that neutron activation analysis of Zr in silicate samples with a Zr/U
ratio<10 has considerable uncertainty due to fission contribution. A correction factor for the fission contribution has been
determined experimentally. 相似文献
7.
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. 相似文献
8.
A. Tavakkoli Farsoni S. A. Mireshghi 《Journal of Radioanalytical and Nuclear Chemistry》2001,248(3):695-697
Thermal neutron analysis (TNA) technology has been used for the non-destructive detection of explosives. The system uses a relatively weak 252Cf neutron source (1.03·107 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
9.
J. G. Fantidis 《Journal of Radioanalytical and Nuclear Chemistry》2012,293(1):95-101
A transportable thermal neutron radiography system, incorporating a compact proton accelerator as neutron source has been simulated using the MCNP4B code. The neutron source will be produced via the 7Li(p,n)7Be reactions by a 2.5?MeV, 10?mA proton beam into a thick lithium target. Variable values for the collimator ratio were calculated. Thermal neutron radiography parameters are comparable to the research nuclear reactors. Sapphire filter was treated in order to improve the results. Simple and advanced neutron shielding materials considered which was further enhanced with layers of bismuth. The system was compatible with the European Union Directive on ??Restriction of Hazardous Substances?? (RoHS) 2002/95/EC, hence excluding the use of cadmium and lead. 相似文献
10.
J. H. Moon S. H. Kim Y. S. Chung J. M. Lim G. H. Ahn M. S. Koh 《Journal of Radioanalytical and Nuclear Chemistry》2009,282(1):33-35
A delayed neutron counting system has been implemented at the HANARO research reactor in 2007. Thermal neutron flux measured
at the NAA #2 irradiation hole coupled to the delayed counting system, was higher than 3 × 1013 n cm−2 s−1. The delayed neutron counting system is composed of 18 3He detectors which are divided into three groups with six detectors and the collected signals of each group are processed
to a digital signal. The count numbers were measured with the uranium mass by using NIST SRMs under fixed analytical condition
and their correlation could be determined. Finally, delayed neutron activation analysis has been carried out for the determination
of uranium mass fraction in the collected environmental samples. 相似文献
11.
Danyal Turkoglu Joshua Burke Radoslaw Lewandowski Lei R. Cao 《Journal of Radioanalytical and Nuclear Chemistry》2012,291(2):321-327
A thermal neutron beam facility has been designed and implemented at the Ohio State University Research Reactor. A project
is underway to construct a large vacuum chamber such that the facility could have neutron depth profiling and neutron radiography
capabilities as intended. The neutron beam is extracted from the reactor through a neutron collimator emplaced in Beam Port
#2. The neutron spectrum entering the neutron collimator was unfolded from foil activation analysis results and also simulated
with a full reactor core model in the MCNP Monte Carlo code. The neutron collimator uses polycrystalline bismuth as a gamma
ray filter and single-crystal sapphire as a fast neutron filter. The beam is defined by multiple 3.0 cm diameter apertures
made of borated aluminum. Characterization of the beam was performed using foil activation to find the flux and a low-budget
neutron imaging apparatus to see the beam profile. The modulation transfer function was calculated to offer insight into the
resolution of the imaging system and the collimation of the beam. The neutron collimator delivers the filtered thermal neutron
beam with a ~4 cm diameter and a thermal equivalent flux of (1.27 ± 0.03) × 107 n/(cm2s) at 450 kW power at the end of the collimator. 相似文献
12.
S. S. Nargolwalla J. Niewodniczanski J. E. Suddueth 《Journal of Radioanalytical and Nuclear Chemistry》1970,5(2):403-423
The experimental sensitivity for 72 different elements using 3 MeV neutron activation has been investigated. Using a 200 kV
Cockcroft-Walton neutron generator with a 3 MeV neutron flux of about 1.5·106n·cm−2·sec−1, γ-ray spectra of 51 elements were obtained with a sufficient number of photopeak counts for sensitivity calculations using
a photopeak integration method. A useful table summarizing the sensitivity results is given. That 3 MeV neutron activation
analysis is practical, is demonstrated by the experimental sensitivities obtained.
Guest worker from the Institute of Nuclear Techniques, Academy of Mining and Metallurgy, Krakow, Poland, at the National Bureau
of Standards, 1968–1969. 相似文献
13.
A prompt gamma neutron activation analysis (PGAA) setup installed at ANRTC has been used to analyze boron. It consists of a 22.6% REGe detector and a 740 GBq 241Am-Be neutron source moderated with water and paraffin. At the sample irradiation position, the thermal neutron fluence rate measured was 2.36·104 n·m–2· s–1 and the corresponding Cd-ratio was 22 for gold monitor. The absolute detection efficiency in the range of 120–1500 keV was determined using 152Eu standard solution. The sensitivity and detection limit for standard boric acid samples has been determined. The boron content in boric acid prepared from Turkish borate ores is measured to be 15.91±0.46% wt. 相似文献
14.
Ş. Turhan H. Yücel A. Demirbaş 《Journal of Radioanalytical and Nuclear Chemistry》2005,262(3):661-664
A prompt gamma neutron activation analysis (PGAA) setup installed at ANRTC has been used to analyze boron. It consists of a 22.6% REGe detector and a 740 GBq 241Am-Be neutron source moderated with water and paraffin. At the sample irradiation position, the thermal neutron fluence rate measured was 2.36·104 n·m–2· s–1 and the corresponding Cd-ratio was 22 for gold monitor. The absolute detection efficiency in the range of 120–1500 keV was determined using 152Eu standard solution. The sensitivity and detection limit for standard boric acid samples has been determined. The boron content in boric acid prepared from Turkish borate ores is measured to be 15.91±0.46% wt. 相似文献
15.
J. C. G. Caburo M. E. Wrenn N. P. Singh K. Crawford 《Journal of Radioanalytical and Nuclear Chemistry》1994,182(1):53-56
A routine procedure for the determination of thorium in urine of workers has been developed by the neutron activation method. The technique suggested by Dang, et. al. has been modified in order to reduce the costs involved and the sample processing time. The samples were irradiated in the MIT (Massachusetts Institute of Technology-Boston) reactor, in a thermal neutron flux of 8×1012n.cm–2.s–1 for 31/2 hours. Thorium-232 was determined by counting233Pa. 相似文献
16.
A rapid method for the determination of Al, V and Ti has been developed and is used for the analysis of these elements in different ores and alloys. An isotopic neutron source252Cf having a thermal neutron flux of the order of 8.5×107 n·cm–2 sec–1 has been used for thermal neutron bombardment. Activity measurements were performed on a HPGe detector coupled to a PC based MCA unit. Depending on the half-life of the (n, ) product, different irradiation and cooling times were employed and thus the elements of interest were analyzed sequentially. 相似文献
17.
L. De Norre J. Op De Beeck J. Hoste 《Journal of Radioanalytical and Nuclear Chemistry》1983,78(1):137-148
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. 相似文献
18.
M. Amendola S. Loreti R. Remetti M. Capogni P. L. Carconi M. L. Cozzella P. De Felice A. Fazio F. Pisacane 《Journal of Radioanalytical and Nuclear Chemistry》2014,301(1):109-115
This work deals with the absolute measurement of the neutron emission rate from a 241Am–Be source by means of the manganese sulphate bath technique, which is the principal method for the absolute determination of the neutron emission rate from radionuclide neutron sources. The facility consists of a spherical container filled with an aqueous solution of manganese sulphate with a 241Am–Be neutron source placed at the center. As well known, neutrons from the source, after having been thermalized by the aqueous solution, undergo neutron capture by hydrogen, manganese, sulphur, and oxygen nuclei, thus inducing a certain activity to the solution. Subsequent gamma spectrometry measurements of 56Mn activity generated by 55Mn neutron activation allows to determine the neutron emission rate of the source, The experimental activity has involved a variety of measurement techniques and calculation procedures, ranging from neutron reactor activation to liquid scintillation counting and Monte Carlo calculations. Neutron activations of 55Mn samples has been carried out with the TRIGA reactor of the ENEA-Casaccia Research Centre, and 56Mn activated samples were subsequently characterized by liquid scintillation counting, in order to obtain reference standards for the calibration of the NaI(Tl) scintillation detectors utilized to record gamma-ray emission from 56Mn. Monte Carlo calculations, carried out by the MCNPX code, were required to calculate neutron transport within the sulphate manganese bath, in particular to determine 55Mn neutron capture probability, and (n, α) and (n, p) concurrent reactions, as well as the neutron leakage. Such a procedure has allowed to maintaining the neutron emission rate uncertainty well below 1 %. All the measurements have been carried out at the ENEA-Casaccia Research Centre by the Italian National Institute of Ionizing Radiation Metrology. 相似文献
19.
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. 相似文献
20.
G. Perdikakis C. T. Papadopoulos M. Kokkoris R. Vlastou S. Galanopoulos A. Lagoyannis A. Spyrou G. Kalyva N. Patronis 《Journal of Radioanalytical and Nuclear Chemistry》2007,272(2):223-226
The measurement of the cross section of the reaction 241Am(n,2n)240Am has been performed at neutron energies from 8.8 to 11.1 MeV, implementing the activation technique. The neutron beam was
produced at the TANDEM accelerator of NCSR “Demokritos” by the 2H(d,n)3He reaction, using a deuterium gas target. During the 5-day long irradiation, the neutron beam fluctuations were monitored
in 100 seconds intervals by a BF3 counter connected with a multiscaling unit. The radioactive target consisted of a 37 GBq 241Am source enclosed in a Pb container. A natural Au foil, a 27Al foil and a 93Nb foil were used as reference materials for the neutron flux determination. After the end of the irradiation the activity
induced at the target and the reference foils, was measured off-line by a 56% HPGe detector. 相似文献