Development of a gamma compensated boron lined ionisation chamber for reactor safety and control applications

Boron lined ionisation chambers with an overall diameter of 85 mm and maximum length of 165 mm have been developed and tested. The chamber consists of 34 numbers of parallel plate aluminium electrodes spaced at a distance of 2 mm and mounted on SS rods and radiation resistant polyetheretherketone (PEEK) spacers. One surface of the signal electrode and both the surfaces of the +HT electrodes are dip coated with boron. It is filled with nitrogen gas at a pressure of 128 cm of Hg. Tests at the ⁶⁰Co source facility at gamma fields ranging from 200 R/h to 830 kR/h showed that the chamber required 500 V to obtain 90% of the saturation current at 830 kR/h. The gamma compensation factor was measured as 0.12–7% for various gamma fields for polarising voltages of +400 and −350 V. Neutron measurements at the Apsara Thermal Column showed that the linearity of the chamber response as a function of reactor power was within 2%. The neutron sensitivity was measured as 3.9 fA/nv.     

Comparison of different methods for the assessment of the environmental gamma dose

Four different techniques for the assessment of environmental gamma dose are discussed and compared: high pressure ionisation chamber (Reuter-Stokes type), TLD dosemeters (GR-200), in situ gamma ray spectrometry and dose evaluation from the activity concentration of radionuclides in the soil. Soil samples gathered from a field near Turin (North-West Italy) have been analysed with a p-type HPGe, and their radionuclide concentrations have been used to evaluate the dose at I m above the soil surface. At the same location the dose rate was evaluated with the in situ gamma ray spectrometry (with a portable n-type HPGe) and with the ionisation chamber and the TLD dosemeters. The contribution of cosmic rays was added to the dose rate calculated from the soil samples and the in situ gamma spectrometry. The results obtained with these four techniques agree well within 20%.     

Simulation studies on a prototype ionisation chamber for measurement of personal dose equivalent, Hp(10)

A prototype ionisation chamber for direct measurement of the personal dose equivalent, Hp(10), similar to the one developed by the Physikalisch-Technische Bundesantalt (PTB), was designed and constructed by the Metrological Laboratory of Ionizing Radiation (LMRI) of Nuclear and Technological Institute (ITN). Tests already performed have shown that the behaviour of this chamber is very similar to the PTB chamber, mainly the energy dependence for the X-ray radiation qualities of the ISO 4037-1 narrow series N-30, N-40, N-60, N-80, N-100 and N-120 and also for gamma radiation of 137Cs and 60Co. However, the results obtained also show a dependence on the energy and angles of incident radiation and a low magnitude of the electrical response of the ionisation chamber. In order to optimise the performance of the chamber, the LMRI initiated numerical simulation of this ionisation chamber by Monte Carlo method using the MCNPX code.     

Remains of 137Cs contamination in the city of Goiânia, Brazil

The results of measurements, performed in 1999, of the remaining 137Cs contamination in some of the sites where fragments of a radioactive source of a teletherapy unit had been manipulated in 1987 are presented. This episode occurred in the city of Goiania, during Brazil's worst radiological accident ever reported. Using the technique of gamma ray spectrometry, analyses of both surface and profile soil and vegetable samples were made. High values of 137Cs activity per unit mass were found in soil layers at depths between 10 and 40cm from the surface. Some values exceeded by up to eight times the action level of 22.5 kBq x kg(-1) proposed by the Brazilian National Nuclear Energy Commission (CNEN) during the decontamination process at the time of the accident, for the first year after the accident. Absorbed dose rates at 1 m above the ground were calculated from the data of 137Cs concentration in the soil and compared with those obtained from in situ gamma ray spectrometry and from thermoluminescence dosimetry.     

Physical parameters and correction factors for ionization chambers for absolute measurement of air kerma in γ-ray fields

Values of physical parameters and correction factors essential for the absolute measurement of air kerma in ¹³⁷Cs and ⁶⁰Co γ-ray fields were obtained using an EGS5 program for spherical, cylindrical and pancake ionization chambers. The mean mass collision stopping power ratio for graphite and air, was found to vary depending on the cutoff energy of electrons employed in calculation. The ratio between the energies deposited in cavity air due to Compton electrons emitted from the air and those from the graphite wall increases as the chamber size is increased. It also increases as the γ-ray energy is reduced and is equal to 0.09 for ¹³⁷Cs γ-rays in a spherical ionization chamber of cavity diameter 12 cm. Correction factors for γ-ray attenuation in chamber walls and those for the contribution of scattered γ-rays to chamber responses were obtained separately. The wall correction factor, which is equal to the product of these two factors, is close to unity for pancake chambers.     

Calibration of a 7.6 cm x 7.6 cm (3 inch x 3 inch) sodium iodide gamma ray spectrometer for air kerma rate

An experimental procedure is described for converting a gamma ray spectral measurement from a 7.6 cm x 7.6 cm (3 inch x 3 inch) sodium iodide (NaI) detector to air kerma rate. The calibration procedure involves measuring the energy deposited in the detector using 10 radioactive sources of known activity covering an energy range from 60 keV to 1,836 keV. For each of the 10 sources, gamma ray spectra were measured with the source at different angles to the detector axis. The total energy deposited in the detector for the ten sources was confirmed by Monte Carlo calculations. The spectra measured at different angles were combined to produce a spectrum that would represent a homogeneous semi-infinite source of radiation. The resultant spectrum was then subdivided into 10 energy regions. Based on the known air kerma rates due to the sources, a calibration coefficient was calculated for each of the 10 energy regions. These calibration coefficients could then be used to convert the energy deposited in the 10 regions of an unknown spectrum to air kerma rate. The calibration procedure was confirmed by comparing the results from the detector with those from calibrated collimated beams of 137Cs and 60Co. A comparison of measurements using a calibrated pressurised ionisation chamber with those from a similar Nal spectrometer in Finland provided additional confirmation of the calibration procedure.     

Hp(0.07) photon irradiations at Seibersdorf for the EURADOS extremity dosemeter intercomparison 2009

In August 2009, almost 1000 passive extremity dosemeters were irradiated at the Dosimetry Laboratory Seibersdorf as part of the EURADOS intercomparison IC2009. Forty-four European individual monitoring services participated, with a total of 59 dosimetry systems (46 finger ring, 4 finger tip and 9 wrist/ankle dosemeter systems). Additionally, finger-ring dosemeters from the Dosimetry Service Seibersdorf were irradiated in a non-competitive manner. Dosemeter irradiations on rod and pillar phantoms in four photon-radiation fields complying with the ISO standard 4037 were performed with personal dose equivalent values (H(p)(0.07)) ranging from 4 to 480 mSv. Traceability was established by using an air-kerma-calibrated monitor ionisation chamber together with the X-ray facility as well as a calibrated (137)Cs gamma radiation field with a collimated beam geometry. The ISO-tabulated conversion coefficients from air kerma free-in-air to H(p)(0.07) were applied, resulting in the main contribution to the expanded measurement uncertainties.     

Thermoluminescence properties of LiF:Mg,Cu,Na,Si pellets in radiation dosimetry

Sintered LiF:Mg,Cu,Na,Si thermoluminescence (TL) pellets have been developed for application in radiation dosimetry. LiF:M,Cu,Na,Si TL pellets were made from TL powders using a sintering process, that is, pressing and heat treatment. These pellets have a diameter of 4.5 mm, and a thickness of 0.8 mm are blue in colour and have a mass of 28 mg each. After 400 pellets had been produced they were irradiated with 137Cs gamma radiation and samples having a sensitivity within a +/-5% standard deviation were selected for experimental use. In the present study, the physical and dosimetric properties of LiF:Mg,Cu,Na,Si TL pellets were investigated for their emission spectrum, dose response, energy response and fading characteristics. Photon irradiation for the experiments was carried out using X ray beams and a 137Cs gamma source at the Korea Atomic Energy Research Institute (KAERI). The average energies and the dose were in the range of 20-662 keV and 10(-6) - 10(2) Gy respectively. The glow curves were measured with a manual type thermoluminescence dosimetry reader (system 310, Teledyne) at a constant nitrogen flux and a linear heating rate. For a constant heating rate of 5 degrees C.s(-1). the main dosimetric peak of the glow curve appeared at 234 degrees C, its activation energy was 2.34 eV and the frequency factor was 1.00 x 10(23). The TL emission spectrum appeared at the blue region centred at 410 nm. A linearity of photon dose response was maintained up to 100 Gy. The photon energy responses relative to the 137Cs response were within +/-20% in the overall photon energy region. No fading of the TL sensitivity of the pellets stored at room temperature was found over the course of a year. Therefore LiF:Mg,Cu,Na,Si TL pellets can be used for personal dosimetry, but more research is needed to improve the characteristics for repeated use.     

Microdosimetric interpretation of the photon energy response of LiF:Mg,Ti detectors

Photon energy response of MTS-N (LiF:Mg,Ti) detectors (TLD Poland) and of MTS-N detectors sensitised with 200 Gy of 60Co gamma rays, followed by UV irradiation (sMTS-N), has been determined using X rays with narrow energy spectra, in the energy range from 20 to 300 keV. The over-response of LiF:Mg,Ti detectors for X rays (relative TL efficiency eta = 1.1) can be explained as an ionisation density effect. Low energy X rays produce short electron tracks, which locally deposit a high radiation dose and, consequently, lead to an enhanced (supralinear) response. This over-response has not been observed in sensitised MTS-N where supralinearity in the response after gamma ray doses above 1 Gy is not seen. Using the dose-response curves measured for MTS-N detectors after 137Cs gamma ray irradiation and local doses calculated using Monte Carlo generated electron tracks, it was possible to predict the relative TL effectiveness for different X ray energies. The calculation procedure can be applied to predict the photon energy response of LiF:Mg,Ti detectors in an arbitrary photon field.     

Goiânia: 12 years after the 137Cs radiological accident

In 1987, in the city of Goiania, Brazil, occurred one of the worst radiological accidents ever reported. The remains of 137Cs contamination in a terrain where part of a radiotherapy unit had been manipulated in 1987 were measured in 1999-2000, and some of the results are presented here. Using the technique of gamma ray spectrometry in situ and in the laboratory, the ambient dose equivalent rate at 1 m above the ground and 137Cs concentration in soil were determined. Values higher than the ones established by the Brazilian National Nuclear Energy Commission (CNEN) as action levels in 1987, namely 0.8 microGy x h and 22.5 kBq x kg(-1), were obtained in that terrain. The 137Cs distribution profile in the soil shows high values of the specific activity in a layer located at a depth of 10-40 cm from the surface, where the soil is mixed with rubble, reaching values as high as 175 kBq x kg(-1).     

Thermoluminescence response of K2YF5:Tb3+ crystals to photon radiation fields

This investigation has been performed to test the feasibility of using K2YF5:Tb3+ crystals as thermoluminescence dosemeters (TLD). K2YF5 single crystals doped with 0.2, 10.0 and 50.0 at.% of trivalent optically active Tb3+ ions as well as K2TbF5 and undoped K2YF5 crystals have been synthesized under hydrothermal conditions. Polished crystal platelets with thickness of about 1 mm have been irradiated with X and gamma rays in order to study thermoluminescent (TL) sensitivity as well as dose and energy response in terms of the Tb3+ concentration in K2YF5. Within this concentration series, K2YF5 crystals doped with 10.0 at.% Tb3+ have been found to have maximum TL response due to a broad asymmetric TL glow peak at 269 degrees C with good linearity of dose response and reproducibility of dose measurements. After deconvolution, the main dosimetric peak has been revealed to be composed of two individual peaks, both with linear TL response behaviour, centered at 210 and 269 degrees C. As it has been proved, the linear TL signal coefficient for K2Y0.9Tb0.1F5 is almost 10 times greater than that for commercial TLD-100 (LiF:Mg,Ti), irradiated with a 137Cs gamma radiation source at the same conditions. The reported results indicate that K2YF5 crystals doped with Tb3+ have potential as promising materials for radiation dosemeters.     

Effect of radiation-induced charge accumulation on build-up cap on the signal current from an ionisation chamber

The signal current from a thimble ionisation chamber with a build-up cap made of an insulator decreases by about 0.41 % after being irradiated for 17 h at an air kerma rate of 41 Gy h(-1) by a collimated (60)Co gamma-ray beam in air. In contrast, the signal current remains constant when the thimble ionisation chamber is irradiated in a water phantom. During irradiation, positive charge is considered to accumulate near the outer surface of the build-up cap where electron equilibrium is not achieved. Secondary electrons travelling in the build-up cap and the chamber wall toward the ionisation volume are decelerated by the electric field generated by the positive charge. Consequently, the signal current decreases with increasing charge accumulation because some secondary electrons are prevented from entering the ionisation volume. In the water phantom, electron equilibrium is established in and around the ionisation chamber and charge does not accumulate. To confirm this hypothesis, the signal current was measured for an ionisation chamber in air with a build-up cap wrapped with Al foil and covered with PMMA tubes. Electron equilibrium was established over the build-up cap because the tubes were thicker than the secondary electron range. The signal current decreased with increasing positive voltage applied to the Al foil. It was estimated from the results that positive charges equivalent to a voltage of over 6 kV applied to the Al foil accumulated during irradiation. The signal current was also measured for an ionisation chamber with a metal build-up cap and for an ionisation chamber with a wall and build-up cap made of conductive plastic.     

Method for determination of gamma and neutron dose components in mixed radiation fields using a high-pressure recombination chamber

A new method is proposed for the determination of dose components in mixed radiation fields (gamma + neutrons) using a recombination chamber. The method involves the determination of the ratio of ionisation currents measured at two different voltages applied to the chamber without the need of determining the saturation current, neither in the radiation field investigated nor during calibration. Therefore, the chamber can be filled with a gas under a pressure much higher than that used in presently available recombination chambers. This paper presents theoretically derived formulae supporting the method and the experimental results of dose component measurements using a high-pressure recombination chamber filled with methane. The method can be used for determining neutron and gamma dose components in the environment, especially in the vicinity of nuclear centres.     

Non-destructive method of characterisation of radioactive waste containers using gamma spectroscopy and Monte Carlo techniques

During the decommissioning of the SATURNE accelerator at CEA Saclay (France), a number of concrete containers with radioactive materials of low or very low activity had to be characterised before their final storage. In this paper, a non-destructive approach combining gamma ray spectroscopy and Monte Carlo simulations is used in order to characterise massive concrete blocks containing some radioactive waste. The limits and uncertainties of the proposed method are quantified for the source term activity estimates using 137Cs as a tracer element. A series of activity measurements with a few representative waste containers were performed before and after destruction. It has been found that neither was the distribution of radioactive materials homogeneous nor was its density unique, and this became the major source of systematic errors in this study. Nevertheless, we conclude that by combining gamma ray spectroscopy and full scale Monte Carlo simulations one can estimate the source term activity for some tracer elements such as 134Cs, 137Cs, 60Co, etc. The uncertainty of this estimation should not be bigger than a factor of 2-3.     

Large diameter radial semiconductor drift chambers for low noise radiation detection

A semiconductor drift chamber with cylindrical symmetry has been investigated theoretically for its suitability as a low noise radiation detector. The signal/noise of a silicon drift chamber at room temperature is likely to be determined by the detector dark current and the drift chamber should be superior in performance to a conventional pn junction detector of equivalent dimensions. A device of 48 mm diameter and 300 μm thickness is predicted to be capable of an energy resolution of less than 5 keV fwhm at 300 K. This is about seven times better than that predicted for its conventional counterpart.     

Accuracy of field spectrometry in estimating 137Cs contamination in high altitude alpine soils

In situ gamma ray spectrometry is an attractive method for providing information on the concentrations of radionuclides in the soil. This method requires separate knowledge of the soil density and of the radioactivity distribution relative to soil depth. In-field gamma spectrometry measurements were performed together with sampling of the soil at the same site for subsequent gamma spectrometry analysis in the laboratory. Results of 137Cs concentration were compared in order to: (i) verify the effectiveness of in situ gamma spectrometry in averaging local inhomogeneties in the soil; and (ii) quantify the sensitivity of the gamma ray spectrometry soil radioactivity evaluation with respect to the variability of the depth profiles obtained from a single site sampling. The site of study is a high altitude pasture, in the Alpine environment.     

The influence of post-exposure heating on the stability of MCP-N (LiF:Mg,Cu,P) TL detectors

Post-exposure annealing of highly sensitive LiF:Mg,Cu,P (MCP-N) detectors, at 100 degrees C over 10 or 20 min prior to readout, is usually recommended for routine dosimetry. The purpose of this anneal is to eliminate low-temperature peaks, especially peak 3, which fades at room temperature in about 3 months. However, as this annealing procedure does not entirely eliminate peak 3, 10% of its thermoluminescent (TL) signal still being readable, a fading correction must be applied. The aim of this work was to optimise the conditions of post-exposure treatment, i.e. its temperature and duration, in order to facilitate the use of MCP-N detectors in routine dosimetry. MCP-N detectors were annealed in standard conditions, i.e. at 240 degrees C over 10 min and exposed to a dose of 5 mGy (137Cs). For post-exposure annealing, six different temperatures between 100 degrees C and 150 degrees C and two time periods (10 and 20 min) were tested. TL glow curves were deconvoluted with the GCA code. A post-exposure anneal at 120 degrees C over 10 min was found to be optimal. Heating at this temperature eliminates 100% of the TL signal of peak 3, while maintaining the area and maximum intensity of the main peak 4 unchanged. In this case, no fading correction needs to be applied. Annealing at higher temperatures, up to 150 degrees C, results in a loss of peak 4 signal, and is therefore not recommended.     

Treatment of radioactive liquid waste by sorption on natural zeolite in Turkey

Liquid radioactive waste has been generated from the use of radioactive materials in industrial applications, research and medicine in Turkey. Natural zeolites (clinoptilolite) have been studied for the removal of several key radionuclides ((137)Cs, (60)Co, (90)Sr and (110m)Ag) from liquid radioactive waste. The aim of the present study is to investigate effectiveness of zeolite treatment on decontamination factor (DF) in a combined process (chemical precipitation and adsorption) at the laboratory tests and scale up to the waste treatment plant. In this study, sorption and precipitation techniques were adapted to decontamination of liquid low level waste (LLW). Effective decontamination was achieved when sorbents are used during the chemical precipitation. Natural zeolite samples were taken from different zeolite formations in Turkey. Comparison of the ion-exchange properties of zeolite minerals from different formations shows that Gordes clinoptilolite was the most suitable natural sorbent for radionuclides under dynamic treatment conditions and as an additive for chemical precipitation process. Clinoptilolite were shown to have a high selectivity for (137)Cs and (110m)Ag as sorbent. In the absence of potassium ions, native clinoptilolite removed (60)Co and (90)Sr very effectively from the liquid waste. In the end of this liquid waste treatment, decontamination factor was provided as 430 by using 0.5 mm clinoptilolite at 30 degrees C.     

The overview of internal exposure monitoring in Lithuania

This study deals with analysis of situation and results of internal exposure monitoring of radiation workers and population in Lithuania. Radiation workers are assessed for internal exposures by direct methods--whole body counting or organ counting by gamma spectrometry at the Radiation Protection Centre and Ignalina Nuclear Power Plant (INPP). Results of monitoring of INPP and nuclear medicine workers show that no significant activities were detected. The annual committed effective doses of workers are <1 mSv. The measured average activity of 40K in males and females was 3.7 +/- 1.0 and 2.5 +/- 0.7 kBq, respectively. Mixed diet sampled at hospitals in 2001-5 was analysed for (90)Sr and (137)Cs activity concentrations. Average effective dose due to 90Sr and 137Cs in mixed diet was 0.6 +/- 0.2 and 0.47 +/- 0.13 microSv, respectively. Indoor radon measurements were done in multi-storey houses. Average concentration was 15.1 +/- 1.0 Bq m(-3). The annual effective dose aused by radon was 0.38 mSv.     

CdWO4 scintillator as a compact gamma ray spectrometer for planetary lander missions

The objective of this work is to develop a gamma ray spectrometer (GRS) suitable for use on planetary rover missions. The main characteristics of this detector are low weight, small volume low power and resistance to cosmic ray radiation over a long period of time. We describe a 3 cm diameter by 3 cm thick CdWO₄ cylindrical scintillator coupled to a PMT as a GRS for the energy region 0.662–7.64 MeV. Its spectral performance and efficiency are compared to that of a CsI(Tl) scintillator 2.5 cm diameter by 6 cm thick coupled to a 28 mm×28 mm PIN photodiode. The comparison is made experimentally using ¹³⁷Cs, ⁶⁰Co, 6.13 MeV gamma rays from a ¹³C(,γn)O¹⁶* source, 7.64 MeV thermal neutron capture gamma rays emitted from iron bars using a ²⁵²Cf neutron source, and natural radioactivity 1.46 MeV ⁴⁰K and 2.61 MeV ²³²Th gamma rays. We use a Monte Carlo method to calculate the total peak efficiency of these detectors and the full energy, first and second escape peak efficiencies. The experimental and calculated results agree well. We investigated the usefulness of these detectors for a GRS on a Mars lander mission. Although both detectors meet desired specifications, it was found that CdWO₄ has advantages over CsI(Tl) being a more compact detector of higher efficiency. Using a shaping amplifier of 24 ms, CdWO₄ spectrometer exhibited a 6.8% FWHM at 662 keV. At 6.13 MeV, CdWO₄ detector possesses an intrinsic total and full energy peak efficiencies of 16.7% and 6.3%, respectively. These efficiencies are nearly a factor of 1.6 and 4 greater than the corresponding efficiencies of the CsI(Tl) detector.

A proposed gamma ray spectroscopy system to be placed on a rover, consists of a central detector surrounded by a Compton suppressor shield. The central detector is a cylindrical CdWO₄ detector and the Compton suppressor shield is made of segmented CdWO₄, coupled to PIN photodiodes. The shield also prevents thermal neutron activation of the central detector.