Portable Ge Detector Research Articles

In situ gamma spectrometry using a portable HPGe detector. Radiological characterisation and environmental surveillance around an operating nuclear power plant. Possibilities and limits.

The in situ technique for measuring radionuclides in the soil using a portable Ge detector is a highly versatile tool for both the radiological characterisation and for the monitoring of operating nuclear power plants. The main disadvantage of this technique is related to the lack of knowledge of the geometry of the source whose activity concentration is to be determined. However, its greatest advantage is the high spatial representability of the samples and the reduced time and resource consumption compared to gamma spectrometry laboratory measurements. In this study, the possibilities and limits offered by in situ gamma spectrometry with a high-resolution gamma portable detector in two common uses are shown. First, the radiological background characterisation and its relationship with the geology of an area of 2700 km2 are assessed. Second, its potential for monitoring man-made activity concentration in soils located around an operating nuclear power plant in Spain for surveillance purposes is evaluated. Finally, high-accuracy radiation maps were prepared from the measurements that were carried out. These radiation maps are essential tools to know the radioactive background of an area, especially useful to assess artificial radioactive deposits produced after a nuclear accident or incident.

IN SITU GAMMA SPECTROMETRY MEASUREMENTS AND MONTE CARLO COMPUTATIONS FOR THE DETECTION OF RADIOACTIVE SOURCES IN SCRAP METAL

A very limited number of field experiments have been performed to assess the relative radiation detection sensitivities of commercially available equipment used to detect radioactive sources in recycled metal scrap. Such experiments require the cooperation and commitment of considerable resources on the part of vendors of the radiation detection systems and the cooperation of a steel mill or scrap processing facility. The results will unavoidably be specific to the equipment tested at the time, the characteristics of the scrap metal involved in the tests, and to the specific configurations of the scrap containers. Given these limitations, the use of computer simulation for this purpose would be a desirable alternative. With this in mind, this study sought to determine whether Monte Carlo simulation of photon flux energy distributions resulting from a radiation source in metal scrap would be realistic. In the present work, experimental and simulated photon flux energy distributions in the outer part of a truck due to the presence of embedded radioactive sources in the scrap metal load are compared. The experimental photon fluxes are deduced by in situ gamma spectrometry measurements with portable Ge detector and the calculated ones by Monte Carlo simulations with the MCNP code. The good agreement between simulated and measured photon flux energy distributions indicate that the results obtained by the Monte Carlo simulations are realistic.

Radiological maps of outdoor and indoor gamma dose rates in Greek urban areas obtained by in situ gamma spectrometry

The results obtained from 259 indoor and outdoor in situ gamma spectrometry measurements with a portable Ge detector and 707 total gamma dose rate measurements with an NaI detector in urban areas of 16 Greek islands are presented. From the in situ gamma spectra, the absorbed dose rate in air due to Uranium series, Thorium series, (40)K and (137)Cs are derived and discussed. The results obtained from the present work in conjunction with those reported previously were used for the realization of a complete indoor and outdoor gamma radiation map of Greek urban areas using in situ gamma spectrometry with portable Ge detector.

Monte Carlo based method for conversion of in-situ gamma ray spectra obtained with a portable Ge detector to an incident photon flux energy distribution.

A Monte Carlo based method for the conversion of an in-situ gamma-ray spectrum obtained with a portable Ge detector to photon flux energy distribution is proposed. The spectrum is first stripped of the partial absorption and cosmic-ray events leaving only the events corresponding to the full absorption of a gamma ray. Applying to the resulting spectrum the full absorption efficiency curve of the detector determined by calibrated point sources and Monte Carlo simulations, the photon flux energy distribution is deduced. The events corresponding to partial absorption in the detector are determined by Monte Carlo simulations for different incident photon energies and angles using the CERN's GEANT library. Using the detector's characteristics given by the manufacturer as input it is impossible to reproduce experimental spectra obtained with point sources. A transition zone of increasing charge collection efficiency has to be introduced in the simulation geometry, after the inactive Ge layer, in order to obtain good agreement between the simulated and experimental spectra. The functional form of the charge collection efficiency is deduced from a diffusion model.

Gamma Ray Spectrometry for the Determination of Absorbed Dose Rate in the Air Following Myocardial Perfusion Scintigraphy with 201Tl

The time dependence of the absorbed dose rate in air at a distance of 1 m from a patient who was undergoing myocardial perfusion scintigraphy with 201 Tl was measured. The absorbed dose rate was deduced by in situ gamma ray spectrometry using a 'calibrated' portable Ge detector which allows the direct measurement of spectral dose rate distribution in the surroundings of a source without any assumption about the source geometry. The decrease of the absorbed dose rate with time can be expressed by a single exponential decay. The effective and biological half-lives of 201 Tl were estimated to be 63 h and 20 days respectively. The total absorbed dose in air at a distance of 1 m from the patient was found to be 0.03 mGy which, converted to effective dose, gives 0.03 mSv which is small compared with the arithmetic annual mean worldwide, weighted for population, effective dose due to terrestrial gamma rays (0.46 mSv) and cosmic rays (0.38 mSv).

Fast neutron detection with germanium detectors: Unfolding the 692 keV peak response for fission neutron spectra

A new method is proposed for obtaining information on the spectral distribution of fast fission neutrons in the energy range from 0.7 MeV up to about 6 MeV from the shape of the Ge 692 keV inelastic scattering peak response e.g. in the pulse-height distribution measured with a portable Ge detector. In order to reduce photon background events, a lead shielding is put around the active volume of the Ge detector. Peak responses associated with the 692 keV peak are computed by Monte-Carlo simulations including the lead shield. A set of 80 response functions for neutron energies in the aforementioned energy range are used in an unfolding procedure based on least-squares methods for the 692 keV peak response. The procedure was used to determine shape and magnitude of a fission neutron spectrum from a 252Cf source. In a numerical example, the procedure was also applied to unfolding the calculated peak response to spectra of fission neutron sources shielded by carbon, concrete enriched with iron, and a thick lead shield. It is shown that the peak responses for such different spectra can be substantially different and, thus, used for a limited neutron spectrometry.

Measurement of localized metal removal in pipes by gamma-ray back-scattering method

Localized metal removal in pipes, especially those carrying liquid at high pressure, can creat a real problem where it may make a whole causing a jet stream of the liquid to out-side. Measuring the the localized metal removal region before it develope to a hole will prevent reaching this situation. In this method, which can be used to scan the pipe from out-side, a point gamma-ray source is attached to the center a high resolution portable Ge detector which is connected to a multi-channel analyzer. When the detector and the source are attached to the pipe wall from out-side, gamma photons will interact with the pipe material and some will back-scatter to the detector with lower energy than those of the primary photons. Therefore they will fall in a lower channel number and will not overlap with the primary ones. The amount of the back-scattered photons is proportional to the thickness of the pipe at the source position. Thicker material will scatter more photons and vice-versa. The intensity will depend also on the photon energy and pipe material. In localized metal removal region less photons will back-scatter to the detector and therefore the region can immediately be found out. Localized metal removal region down to a mm range was measured.

Part III preliminary surveillance of prospective site — Environmental considerations —

The environmental issues in the R-project are mainly considered from radiological safety point of view. Supposing a new experimental site for the program to be constructed in Toki Area, land and water surveys have been made since 1982 on the background radiation levels and radioactive material concentrations. Results of in-situ observations using instruments of various types are given and compared, such as TLD's (exposed for every three months), a pressurized ionization chamber, a NaI(Tl) detector in DBM mode, and a portable Ge detector. Also shown are the data from laboratory analyses of sampled materials in α- and γ-spectrometries, and neutron activation method. Tritium concentrations in river water samples have been quarterly measured with a liquid scintillation counter. These kinds of data will further be compiled to define the regional characteristics of the site environments. In relating to the expected radiation yields in the R-project, the importance of monitoring methods effective to pulsed radiation components is particularly emphasized. Possible use of ionization chamber and TLD's for this purpose is discussed in some detail with the preliminary application to the existing tokamak experiments on JIPP T-IIU.

Part III. Preliminary surveillance of prospective site — environmental considerations —

The environmental issues in the R-project are mainly considered from radiological safety point of view. Supposing a new experimental site for the program to be constructed in Toki Area, land and water surveys have been made since 1982 on the background radiation levels and radioactive material concentrations. Results of in-situ observations using instruments of various types are given and compared, such as TLD's (exposed for every three months), a pressurized ionization chamber, a NaI(Tl) detector in DBM mode, and a portable Ge detector. Also shown are the data from laboratory analyses of sampled materials in α- and γ-spectrometries, and neutron activation method. Tritium concentrations in river water samples have been quarterly measured with a liquid scintillation counter. These kinds of data will further be compiled to define the regional characteristics of the site environments. In relating to the expected radiation yields in the R-project, the importance of monitoring methods effective to pulsed radiation components is particularly emphasized. Possible use of ionization chamber and TLD's for this purpose is discussed in some detail with the preliminary application to the existing tokamak experiments on JIPP T-IIU.