Measurement of the 238U(n,γ)239U reaction cross sections in the energy range of 2.0–5.0 MeV by using a neutron activation technique

The neutron capture cross-sections of ²³⁸U at the neutron energies of 4.38?±?0.05 MeV, 3.02?±?0.49 MeV and 2.04?±?0.26 MeV have been measured using the activation method and off-line gamma-ray spectrometric technique. The effects of neutron flux fluctuation, multiple scattering, flux self-shielding and gamma-ray self-absorption were corrected. The excitation function of the ²³⁸U(n,γ)²³⁹U reaction was also calculated using the TALYS-1.9 code. The experimental results were compared with the evaluated data, the theoretical data and the previous experimental data.

     

Determination of the 209Bi(n,γ) capture cross section at a cold neutron beam

Summary The total capture cross section of ²⁰⁹Bi was determined at the cold neutron beam PGAA-NIPS facilities at the Budapest Neutron Centre. The measurements were performed using a coaxial HPGe detector with Compton suppression. The total and partial gamma-ray production cross sections were deduced relative to the ¹⁴N(n,γ) partial gamma-ray production cross section. By using a bismuth nitrate stoichiometric compound as the sample, we excluded various systematic uncertainties. The total capture cross section is in very good agreement with the compilation of Mughaghab, but is slightly lower than the most recent value determined at the high flux reactor of the ILL in Grenoble, France. We also performed measurements using a 0.5 mm thin Bi metal disc. The relative intensities determined from the Bi disc and the compound samples are in good agreement.     

Precise determination of macro contents of nitrogen in titanium carbonitride by instrumental photon activation analysis

SYNTH, a Windows^TM based software package developed for generating synthetic gamma-ray spectra, has been updated and extended to include the ability to generate gamma-ray spectra resulting from neutron activation. Along with a new gamma-ray library (based on the NNDC PCNuDat compilation), and the best available neutron cross-sections, it is now possible to simply, and quickly predict the interference effects of different bulk and trace element compositions by generating a synthetic gamma-ray spectrum that will be representative of a specific set of input parameters. The parameters include, but are not limited to: elemental composition (bulk, and trace) of the sample; irradiation, decay, and count times; thermal, and resonance neutron flux; sample to detector distance; detector specifications; and electronics configuration. Using existing data reduction codes, it is then possible to generate Minimum Detectable Activities (MDA's) for other trace elements that you may wish to detect in this type of matrix.     

Application of BGO detector to monitor in situ16N concentration in a reactor pool

A high efficiency BGO detecting system coupled to a coolant pumping device was used to monitor on-line the 6.13 MeV high energy gamma-ray, which is emitted from the 7.1 s half-life of¹⁶N activated by¹⁶O(n, p) reaction in a nuclear reactor. The system has been demonstrated effectively to monitor the real-time reactor power level as well as the space-time neutron flux distribution in reactor coolant.     

Design and evaluation of a TNA explosive-detection system to screen carry-on luggage

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.     

2-D reconstruction of elemental distribution within a sample using neutron capture prompt gamma-rays

The technique of emission tomography employing neutron capture prompt gamma-rays is described. Experiments have been carried out to demonstrate this technique employing a high flux reactor neutron beam using an HPGe detector and a scanning system which incorporates a BBC microcomputer for control, data acquisition, image reconstruction and display. Neutron tomography of the same object was also performed in order to correct the emission tomography results for the neutron flux depression within the sample. The images produced represent the intensity of the induced gamma-ray of interest, and hence the concentration of the isotope of interest.     

Simplified evaluation of214Bi/238U activity ratios in fossil bones using non-destructive activation analysis

A simple and reliable method has been developed for the evaluation of radioactive disequilibrium state in fossil bones. The fossil bone samples were irradiated with an extremely low neutron fluence, together with a standard pitchblende prevailing the secular equilibrium among the uranium series. The²³⁹Np activity induced from²³⁸U in both samples were adjusted to be gamma-ray spectrometrically nearly equivalent to the naturally occurring radioactivities by controlling the neutron flux and cooling time. Using single gamma-ray spectrometry of the irradiated samples, the determination of²¹⁴Bi/²³⁸U in a fossil bone was carried out by comparing the photopeak ratios of²¹⁴Bi /609 keV/ and²³⁹Np /278 keV/ instead of²³⁸U with the same ratios from the pitchblende standard sample.     

Evaluation of measurement uncertainty components associated with results of radiochemical neutron activation analysis for determination of uranium traces

Being aware of the importance to consider every step in the evaluation of the combined measurement uncertainty of the result, the purpose of this work was to evaluate the contribution of the radial thermal neutron flux gradient to the uncertainty budget for trace level uranium determination in biological materials by a radiochemical neutron activation analysis (RNAA). Determination of uranium via the short-lived nuclide ²³⁹U was based on solvent extraction with TBP and measurement of the chemical yield from the gamma-ray spectrum of the isolated fraction via ²³⁵U. It has been shown previously, that radial neutron flux gradient, could have a relevant effect on the final result obtained by RNAA. In the present work, radial neutron flux gradient within the irradiation assembly generally accepted in our lab (standards tapped beside the sample), varied between 93 and 108% around the mean value and contributes approximately 20% to the combined measurement uncertainty of the result.     

Analytical use of neutron-capture gamma-rays

Certain elements which are not possible to detect with conventional neutron activation analysis can be measured using thermal neutron-capture gamma-ray analysis. The use of a curved neutron guide at the High Flux Reactor, Grenoble, with a thermal neutron flux of 1.5·10¹⁰n·cm⁻²·sec⁻¹ and the advantage of a low-background counting system (Ge(Li) detector) far from the reactor core are described. Experimental detection limits of a number of elements are given for the low-energy and the high-energy regions. Some applications of the capture gamma-ray method in the whole energy range are studied and are briefly discussed.     

A field-deployable, aircraft-mounted sensor for the environmental survey of radionuclides

The Environmental Radionuclide Sensor System (ERSS)³ is an extremely sensitive sensor, which has been cooperatively developed by Pacific Northwest National Laboratory (PNNL) and Special Technologies Laboratory (STL) for environmental surveys of radionuclides. The ERSS sensors fit in an airborne pod and include twenty High-Purity Germanium (HPGe) detectors for the high-resolution measurement of gamma-ray emitting radionuclides, twenty-four³He detectors for possible neutron measurements, and two video cameras for visual correlation. These acrial HPGe sensors provide much better gamma-ray energy resolution than can be obtained with NaI(TI) detectors. The associated electronics fit into three racks. The system can be powered by the 28 V DC electrical supply of typical aircraft or 120 V AC. The data acquisition hardware is controlled by customized software and a real-time display is provided. Each gamma-ray event is time stamped and stored for later analysis. This paper will present the physical design, discuss the software used to control the system, and provide some examples of its use.     

Determination of metals in alloys by neutron capture gamma-ray spectrometry

A collimated neutron beam capable of providing a thermal neutron flux of 4.75·10⁷ n·cm⁻²·sec⁻¹ has been used to analyze alloy samples of 1–5 g during relatively short irradiation times of 30 min by the use of neutron capture gamma-ray spectrometry. The analyses were performed by using a mathematical treatment that relates the count ratio of every constituent present in the matrix with the concentration and thus it requires no standards. The technique was applied to the analysis of steel and gold alloy samples. Errors ranged from 0.8%–10%.     

The neutron activation analysis laboratory at the Ecole polytechnique de Montreal

The SLOWPOKE reactor at the Ecole Polytechnique de Montréal has been used for neutron activation analysis since 1976. The reproducible neutron flux in the five irradiation sites may be set to values between 10¹⁰ and 10¹² cm^–2s^–1. Associated equipment includes versatile pneumatic sample transfer systems and three germanium gamma-ray spectrometers with mechanical sample changers. The laboratory is used by researchers from several universities and by a variety of industries.     

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.     

Upgrade of the prompt gamma activation analysis and the neutron-induced prompt gamma spectroscopy facilities at the Budapest research reactor

The Budapest Research Reactor’s Prompt Gamma Activation Analysis (PGAA) and Neutron-Induced Prompt gamma Spectroscopy (NIPS) facilities were significantly upgraded during the last few years. The higher neutron flux, achieved by the partial replacement and realignment of the neutron guides, made feasible the automation and specialization of the two experimental stations. A new neutron flux monitor, computer-controlled beam shutters and a low-level counting chamber have been put into operation to assist with in-beam activation experiments. An automatic sample changer has been installed at the PGAA station, while the NIPS station was redesigned and upgraded with a Compton suppressor to use for the non-destructive analysis of bulky samples. In the near future the latter setup will be completed with a neutron tomograph and a moving table, to turn it into a Neutron Radiography/Tomography-driven PGAA equipment.     

Optimization of HANARO cold neutron induced prompt gamma activation analysis system by using Monte Carlo code

Prompt gamma activation analysis (PGAA) is a nuclear analytical technique for non-destructive determination of elemental and isotopic compositions. The principle of PGAA technique is based on detection of captured gamma-ray emitted from an analytical sample while being irradiated with neutrons. Use of a cold neutron beam guide greatly reduces the gamma-ray background at the analytical sample while maintaining a neutron capture rate is comparable to that of standard thermal neutron PGAA. A new cold neutron induced prompt gamma activation analysis (CN-PGAA) system has been under construction since April of 2009 at the HANARO Cold Neutron Building (KAERI, Republic of KOREA). In this study, the Compton suppression factor of the CN-PGAA system was estimated to be 5.5 using a ⁶⁰Co radioactive source in conjunction with the MCNPX simulations. Several parameters of the CN-PGAA system were studied to estimate and optimize the performance of the system: scintillation material in the guarded detector of a Compton suppression spectrometer (CSS); the relative positions of the HPGe detector and annular detector; and the distance between the HPGe detector and back catcher BGO detectors of the CSS. In addition, the neutron ray-trace simulation package, McStas, was adopted to predict the neutron flux and wavelength distribution at the end of the cold neutron beam guide. These results served as input for the MCNPX simulation of the CN-PGAA system.     

The application of NAA with irradiation at reduced temperature for the study of element losses during pretreatment of biological samples

The neutron capture gamma-ray spectroscopy facility assembled at the Institute of Radiochemistry, KfK (for analytical purposes) using a²⁵²Cf neutron source with a strength of 6·10⁷ n/s, has been used to check its applicability and sensitivity for quantitative analyses of ores. The analysis of Sm, Cd and Mn in phosphate and monazite rock samples has been carried out. The results from this study show a variation of about 25% from the values determined by RNAA method. This discrepancy could be mainly due to the low signal-to-background ratio observed which is caused by (i) scattering of the source gammarays by the target, and (ii) interference from the 2223.1 keV neutron capture hydrogen gamma-rays produced by the moderated materials and from their compton scattering in the detector. To overcome these difficulties we suggest to introduce a 2.5 cm thick polyethylene sheet between the detector⁶Li-cap shielding and the target as well as to increase the detection solid angle. Also the strength of the²⁵²Cf neutron source should be increased by an order of magnitude and the neutron beam should be collimated to obtain the optimal thermal neutron flux with a low level of²⁵²Cf gamma-rays. This can be achieved by setting up between the neutron source and the target a conical polyethylene collimator with a thickness of 10 cm containing a 1 cm thick lead sheet.     

A quick method for the determination of the Al/Si weight ratio in alumino-silicates using an Am-Be neutron source

An irradiation procedure with fast and thermal neutrons from a 5 Ci Am-Be isotopic neutron source irradiation facility in combination with a 3'3' NaI(Tl) detector system has been used to determine Al/Si weight ratios in alumino-silicates. Samples were irradiated with and without Cd cover for 10 minutes and counted for 10 minutes after a waiting time of 1 minute. The peak area analysis of the 1779 keV gamma-ray line of ²⁸Al product radionuclide produced via ²⁷Al(n,γ)²⁸Al and ²⁸Si(n,p)²⁸Al reactions in combination with the neutron flux parameter at the irradiation site and nuclear data were used to determine Al/Si ratios. Due to discrepancy in literature data, Am-Be neutron source spectrum averaged cross sections of (n,p) reactions on ²⁷Al, ²⁸Si and ⁵⁶Fe were determined by the activation technique using ¹¹⁵In(n,n')^115mIn as the fast neutron flux monitor reaction. The method was tested using mixtures of high-purity Al₂O₃ and SiO₂ with known weight ratios of Al/Si and validated by a certified reference material BCS-CRM 348 (Ball Clay). Results are presented for bentonite, kaolin, bauxite, feldspar and ball clay samples from Nigeria. The method is non-destructive, rapid and suitable for use in-situ for large-scale exploration works and industrial process control. This revised version was published online in July 2006 with corrections to the Cover Date.     

Neutron activation analysis without multielement standards

To carry out neutron activation analysis without using multielement standards requires knowledge of (i) the absolute photopeak efficiency as a function of energy of the gamma-ray detector, (ii) nuclear data for each reaction used, and (iii) neutron flux parameters for the irradiation position. The present paper discusses each of these topics and shows an example of the determination of flux parameters and improved nuclear data.     

Cross section measurements for 141Pr(n,γ)142Pr reaction at neutron energies from 13.5 to 14.8?MeV

Activation cross-sections were measured for the ¹⁴¹Pr(n,??)¹⁴²Pr 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 Pd-300 neutron generator. The natural high-purity Pr₂O₃ powder was used as target material. 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 neutron fluences were determined by the cross section of ²⁷Al(n,??)²⁴Na reaction. The neutron energy in the measurement were by the cross section ratios of ⁹⁰Zr(n,2n)^89m+gZr and ⁹³Nb(n,2n)^92mNb reactions. The data for ¹⁴¹Pr(n,??)¹⁴²Pr reaction cross sections are reported to be 3.3?±?0.2, 2.7?±?0.2 and 2.2?±?0.2 mb at 13.5?±?0.2, 14.1?±?0.2, and 14.8?±?0.2?MeV incident neutron energies, respectively. Results were discussed and compared with some corresponding values found in the literature.     

Compton suppression system at Penn State Radiation Science and Engineering Center

A Compton suppression system is used to reduce the contribution of scattered gamma-rays that originate within the HPGe detector to the gamma-ray spectrum. The HPGe detector is surrounded by an assembly of guard detectors, usually NaI(T1). The HPGe and NaI(T1) detectors are operated in anti-coincidence mode. The NaI(T1) guard detector detects the photons that Compton scatter within, and subsequently escape from the HPGe detector. Since these photons are correlated with the partial energy deposition within the detector, much of the resulting Compton continuum can be subtracted from the spectrum reducing the unwanted background in gamma-ray spectra. A commercially available Compton suppression spectrometer (CSS) was purchased from Canberra Industries and tested at the Radiation Science and Engineering Center at Penn State University. The PSU-CSS includes a reverse bias HPGe detector, four annulus NaI(T1) detectors, a NaI(T1) plug detector, detector shields, data acquisition electronics, and a data processing computer. The HPGe detector is n-type with 54% relative efficiency. The guard detectors form an annulus with 9-inch diameter and 9-inch height, and have a plug detector that goes into/out of the annulus with the help of a special lift apparatus to raise/lower. The detector assembly is placed in a shielding cave. State-of-the-art electronics and software are used. The system was tested using standard sources, neutron activated NIST SRM sample and Dendrochronologically Dated Tree Ring samples. The PSU-CSS dramatically improved the peak-to-Compton ratio, up to 1000:1 for the ¹³⁷Cs source.