Calibration of an HPGe detector and self-attenuation correction for Pb-210: Verification by alpha spectrometry of Po-210 in environmental samples

In this work the calibration of an HPGe detector for Pb-210 measurement is realised by a liquid standard source and the determination of this radionuclide in solid environmental samples by gamma spectrometry takes into account a correction factor for self-attenuation of its 46.5 keV line. Experimental, theoretical and Monte Carlo investigations are undertaken to evaluate self-attenuation for cylindrical sample geometry. To validate this correction factor (at equilibrium with Po-210 Pb-210) alpha spectrometry procedure using microwave acid digestion under pressure is developed and proposed. The different self-attenuation correction methods are in coherence, and corrected Pb-210 activities are in good agreement with the results of Po-210. Finally, self-attenuation corrections are proposed for environmental solid samples whose density ranges between 0.8 and 1.4 g/cm(3) and whose mass attenuation coefficient is around 0.4 cm(2)/g. (C) 2007 Elsevier B.V. All rights reserved

Characterisation of the PSI whole body counter by radiographic imaging.

A joint project between the Paul Scherrer Institut (PSI) and the Institute of Radiation Physics was initiated to characterise the PSI whole body counter in detail through measurements and Monte Carlo simulation. Accurate knowledge of the detector geometry is essential for reliable simulations of human body phantoms filled with known activity concentrations. Unfortunately, the technical drawings provided by the manufacturer are often not detailed enough and sometimes the specifications do not agree with the actual set-up. Therefore, the exact detector geometry and the position of the detector crystal inside the housing were determined through radiographic images. X-rays were used to analyse the structure of the detector, and (60)Co radiography was employed to measure the core of the germanium crystal. Moreover, the precise axial alignment of the detector within its housing was determined through a series of radiographic images with different incident angles. The hence obtained information enables us to optimise the Monte Carlo geometry model and to perform much more accurate and reliable simulations

Monte Carlo simulation of a whole-body counter using IGOR phantoms.

Whole-body counting is a technique of choice for assessing the intake of gamma-emitting radionuclides. An appropriate calibration is necessary, which is done either by experimental measurement or by Monte Carlo (MC) calculation. The aim of this work was to validate a MC model for calibrating whole-body counters (WBCs) by comparing the results of computations with measurements performed on an anthropomorphic phantom and to investigate the effect of a change in phantom's position on the WBC counting sensitivity. GEANT MC code was used for the calculations, and an IGOR phantom loaded with several types of radionuclides was used for the experimental measurements. The results show a reasonable agreement between measurements and MC computation. A 1-cm error in phantom positioning changes the activity estimation by >2%. Considering that a 5-cm deviation of the positioning of the phantom may occur in a realistic counting scenario, this implies that the uncertainty of the activity measured by a WBC is ∼10-20%

Uncertainty associated with Monte Carlo radiation transport in radionuclide metrology

In radionuclide metrology, Monte Carlo (MC) simulation is widely used to compute parameters associated with primary measurements or calibration factors. Although MC methods are used to estimate uncertainties, the uncertainty associated with radiation transport in MC calculations is usually difficult to estimate. Counting statistics is the most obvious component of MC uncertainty and has to be checked carefully, particularly when variance reduction is used. However, in most cases fluctuations associated with counting statistics can be reduced using sufficient computing power. Cross-section data have intrinsic uncertainties that induce correlations when apparently independent codes are compared. Their effect on the uncertainty of the estimated parameter is difficult to determine and varies widely from case to case. Finally, the most significant uncertainty component for radionuclide applications is usually that associated with the detector geometry. Recent 2D and 3D x-ray imaging tools may be utilized, but comparison with experimental data as well as adjustments of parameters are usually inevitable

CONVERTING SPECIFIC ACTIVITY INTO AMBIENT DOSE EQUIVALENT: UPDATED COEFFICIENTS FOR IN SITU GAMMA SPECTROMETRY.

In situ gamma spectrometry is a valuable tool to assess the radionuclides released in the environment and the associated dose. This requires prior establishment of coefficients allowing the conversion of the specific activity into ambient equivalent dose. The aim of this work is to calculate updated conversion factors for monoenergetic photons and for a series of radionuclides of interest. The calculation was performed using the Monte Carlo (MC) method, the GEANT4 MC code, various activity distribution models and up-to-date nuclear decay data. A new set of conversion factors is established in the energy range extending from  <100 keV to 8.5 MeV. The coefficients calculated in this work were compared to the data published in the literature

Monte Carlo simulation of a clearance box monitor used for nuclear power plant decommissioning.

When decommissioning a nuclear facility it is important to be able to estimate activity levels of potentially radioactive samples and compare with clearance values defined by regulatory authorities. This paper presents a method of calibrating a clearance box monitor based on practical experimental measurements and Monte Carlo simulations. Adjusting the simulation for experimental data obtained using a simple point source permits the computation of absolute calibration factors for more complex geometries with an accuracy of a bit more than 20%. The uncertainty of the calibration factor can be improved to about 10% when the simulation is used relatively, in direct comparison with a measurement performed in the same geometry but with another nuclide. The simulation can also be used to validate the experimental calibration procedure when the sample is supposed to be homogeneous but the calibration factor is derived from a plate phantom. For more realistic geometries, like a small gravel dumpster, Monte Carlo simulation shows that the calibration factor obtained with a larger homogeneous phantom is correct within about 20%, if sample density is taken as the influencing parameter. Finally, simulation can be used to estimate the effect of a contamination hotspot. The research supporting this paper shows that activity could be largely underestimated in the event of a centrally-located hotspot and overestimated for a peripherally-located hotspot if the sample is assumed to be homogeneously contaminated. This demonstrates the usefulness of being able to complement experimental methods with Monte Carlo simulations in order to estimate calibration factors that cannot be directly measured because of a lack of available material or specific geometries

On the reverse micelle effect in liquid scintillation counting.

This work looks into the tracks of electrons in nanoemulsive scintillating media using the Monte Carlo Geant4-DNA code which simulates event-by-event interactions of electrons in liquid water down to the eV, without resorting to the condensed history method. It demonstrates that the average number of micelles in which electrons deposit energy is quite large, increasing with their emission energy, decreasing with micelle size, and rising with micelle concentration. The probability of an electron ending its track in a micelle is found to be rather large and micelle size-dependent below 1keV, and approximating the aqueous fraction at higher energies. Analyses of the Monte Carlo estimated energy depositions in the aqueous phase and in the scintillant tell of a micelle quenching effect, with the micelle size shaping the quenching at low energy and the micelle concentration governing it at higher energies. The micelle effect on the (3)H and (63)Ni beta spectra is discussed for a range of micelle sizes and concentrations. This paper also computes the ionisation quenching function using Birk's law whilst considering the full energy losses in the micelles bisecting the electron pathway, and not just that incurred in the primary micelle enclosing the decaying nuclide. The ionisation quenching function is then used to calculate the detection efficiencies for (3)H, (63)Ni, (54)Mn and (55)Fe. The effect of the micelle size is found to be small for beta emitters but significant for the electron capture nuclides. TDCR measurements of (63)Ni samples covering 8 aqueous fractions are analysed with and without explicit treatment of the micelle effect. Activities in the two representations agree within 0.02%. The ratios of the corresponding figures of merit are found to coincide with the scintillant fractions

A comparison of alpha and gamma spectrometry for environmental natural radioactivity surveys.

An alpha-spectrometry, using automated borate fusion and sequential extraction and exchange chromatography, was used to determine the uranium and thorium based on environmental radioactivity of 20 soil samples. The same set of the samples was analysed using gamma-spectrometry with an HPGe detector. The two data sets were checked for coherence using Z-score and chi2 statistical tests. We show that gamma-spectrometry is a valid alternative to time-consuming alpha-spectrometry for the determination of natural uranium and thorium activity in soil (activity range: 12.5-58.2 Bq/kg). The measured activities were compared with the theoretical activities to ensure secular equilibrium in the 238U and 232Th series. For 226Ra, a special study was made on deconvolution of the 186 keV multiplet with the Levenberg-Marquardt algorithm. Finally, the combined use of Z-score and chi2-tests was found to be a powerful tool for comparing the results obtained with two different methods

Dose assessment following an overexposure of a worker at a Swiss nuclear power plant.

The aim of this work was to assess the doses received by a diver exposed to a radiation source during maintenance work in the fuel transfer pool at a Swiss nuclear power plant, and to define whether the statutory limit was breached or not. Onsite measurements were carried out and different scenarios were simulated using the MicroShield Software and the MCNPX Monte Carlo radiation transport code to estimate the activity of the irradiating object as well as the doses to the limbs and the effective dose delivered to the operator. The activity of the object was estimated to 1.8 TBq. From the various dose estimations, a conservative value of 7.5 Sv was proposed for the equivalent dose to the skin on the hands and an effective dose of 28 mSv. The use of different experimental and calculation methods allowed us to accurately estimate the activity of the object and the dose delivered to the diver, useful information for making a decision on the most appropriate scheme of follow up for the patient