DESCRIPTION OF THE AREA

The plan of the area, which comprises several dumping mounds, is presented in Fig. 1.

The main part of the dump is surrounded by the field road. North of this part of the dump, idle land is situated. On this area, next to Shaft No. 9, a small steep dumping mound with an irregular top (A) is located. The eastern part of the dumping site is bounded by a high dumping mound (B) stretching along the south-north axis. This mound is connected to two smaller ones (C) having flat tops and three steep stony dumping mounds (D) in the direction of a concrete construction (E). At the foot of the high stony dumping mound (D), a sandy flattened dumping mound covered with birch forest (F) is situated. On its west side, a circular re-entrant is seen (G). In the south-west part of the area, close to the road, a narrow sickle-shaped embankment (H) is located, and further north a flattened part (I) going into a small valley, behind which there is promontory in the form of a sandy dumping mound. A small dumping mound a few metres high (J) is located in the western part of the area. Between the concrete construction and the external road there is a long dumping mound having a plain ridge (K).

FIG. 1. Plan of disposal dumps at Radoniów 3. MEASUREMENT PROGRAMME AND METHODS

The assessment programme consisted of:

• Environmental measurements at the disposal site;

• Assessment of the usability of the materials from the disposal site for the construction of the road;

• Analysis of the risk to workers exposed to the material deposited on the dumps.

The tasks performed are described below.

3.1. Radiological mapping of the area

The whole area was measured using a mobile spectrometric laboratory based upon a Toyota Land Cruiser GX90. A scintillation detector type Exploranium GPX-256 with an NaI(Tl) crystal having a volume of 4 L (dimensions 40×10×10 cm) was mounted on the roof of the car. The detector was placed in an aluminum container and covered with polyurethane foam. Its function was to permanently measure the environmental gamma spectra, both along the measuring route and while standing. The mobile laboratory was also equipped with a 660 system mounted on the back seat of the car in a special shockproof container. The GR-660 system consisted of an on-board computer with touch screen connected to a computer with a long cable. Such a connection enabled the operator to sit on the front seat of the car.

The computer stored the collected measurement data and performed the on-line visualization.

The second element of the GR-660 system was an Exploranium GR-320 256-channel spectrum analyzer connected to the detector placed on the roof of the car. In addition, the car was equipped with a differential geographic positioning system (DGPS), enabling the position of the car to be determined very precisely (within 0.5 m). The results from the GPS were transferred to the on-board computer, enabling the measuring route to be presented on the screen and radiological maps to be prepared.

The GR-660 spectrometer enabled the concentrations of natural radionuclides to be determined down to the following levels:

• ⁴⁰K: 0.13% (40 Bq/kg),

• Uranium: 1.6 ppm (20 Bq/kg),

• Thorium: 1.0 ppm (4 Bq/kg).

The results were stored on the hard disk drive of the on-board computer. Further processing of the data was performed in the laboratory using specialized software for detailed spectra and auxiliary data analysis, e.g. the creation of radiological maps of the investigated area.

3.2. Gamma measurements at specific locations using a spectrometer

Gamma dose rates were measured at 41 measuring locations and gamma spectra were measured at 5 locations (3 in the investigated area and 2 at ‘neutral points’ representing natural background), using an Exploranium GR-130 hand-held spectrometer equipped with an NaI(Tl) detector. A GM detector was used for measurements in high dose rate areas. When the dose rate became high, the device automatically switched to the GM for dose rate measurement. The gamma spectra were stored in the memory of the device for onward transfer to the computer and analysis for radionuclide identification.

3.3. Beta and gamma measurements at specific locations using a radiometer

Using a standard RKP-1 radiometer calibrated with a ⁹⁰Sr source, the radiation above the dumped material was measured by counting the pulses from beta and gamma radiation while the detector window was open and then counting the pulses with the detector window closed. The difference gave the count rate associated with beta radiation. The gamma dose rate was measured with the detector window closed and the device switched to the dose rate mode.

3.4. Gamma dose rate measurements at specific locations using an ionization chamber Gamma dose rate was measured at 5 locations using a current-type pressurized ionization chamber. The chamber comprised a 5 L high-pressure steel tank with a wall thickness of 4 mm. The pressure of argon gas was 35 atmospheres. The tank was covered by 1.5 mm of aluminum to prevent noise. The self-counting (the background) of the chamber was 2 nGy/h. A VA-J-51 electrometer equipped with a dynamic capacitor powered by 220 volts AC or an accumulator (in the field) was used to measure the current. The calibration of the instrument was performed assuming that terrestrial background gamma radiation was 70%

equivalent to a ²²⁶Ra source filtered by 0.5 mm of platinum and 30% equivalent to a ¹³¹I source. The calibration procedure took into account the influence of scattered radiation, directional anisotropy of the ionization chamber and the attenuation of the radiation in air.

The background dose rate 1 m above ground, D (in nGy/h), was calculated from following equation:

D0

VK D= −

where: Vis the velocity of the rise of charge (mV/s), given by V =z t, z being the voltage rise on the capacitor during measurement and t being the voltage rise time (the time of measurement — average of 5 readings);

K is the calibration coefficient (nGy/h per mV/s) determined during the calibration procedure;

D0 is the self-counting of the chamber (nGy/h).

The dose rate was determined with an error of ±10%.

3.5. Measurement of radionuclide activities in material samples using gamma spectrometry

The following samples were taken for analysis in the Dosimetry Department of CLOR:

• 41 samples of the surface material of the dump (10 cm depth);

• 10 samples from the deep layers of the dump (taken by the specialized geological company Geological Services);

• 2 samples representing the local natural environment.

Each surface material sample was averaged from 5–7 sub-samples taken from the central point and from the circumference of a 2 m diameter circle around that point.

The method used for determining the activity of the samples was based on a comparison of the concentrations of the natural radionuclides ⁴⁰K, ²²⁶Ra and ²²⁸Th. It entailed the analysis of the count rate registered in three channels — separately for the examined sample and three volume calibration sources of potassium, radium and thorium. The count rate was measured using a scintillation detector with AZAR-90 and MAZAR-95 three-channel amplitude analysers. The scintillation detector was located in a lead shielding chamber. The walls of the shielding were 50 mm of lead and 2 mm of steel. The Marinelli beaker containing the sample (1.7 L) was placed inside the chamber. The function of the chamber was to reduce the external gamma background. Such geometry allowed more counts to be achieved in the defined measurement time. The detector was powered by a stabilized high voltage power supply. Energy peaks from the detector were processed by a linear amplifier and passed to one of three single-channel analyzers. In addition, inside the chamber there was a ¹³⁷Cs source

for controlling the slope of the calibration curve. Each single-channel analyzer was set to its energy range:

• Range I: 1.26–1.65 MeV covering potassium ⁴⁰K with an energy peak of 1.46 MeV;

• Range II: 1.65–2.30 MeV covering mainly the energy peak of 1.76 MeV for bismuth ²¹⁴Bi from the ²³⁸U series;

• Range III: 2.30–2.85 MeV covering mainly the energy peak of 2.62 MeV for ²⁰⁸Tl from the ²³²Th series.

The peaks from the outputs of each analyzer were counted separately, and the results were stored in three memory groups. The device was powered by a 230 V AC electrical supply.

The measurement technique for ⁴⁰K, ²²⁶Ra and ²²⁸Th is described in detail in the Guidelines on the Determination of Natural Radioactivity in Raw and Building Materials, Instruction 234/2003 of the Institute of Building Technology. Based on the measured concentrations of

40K, ²²⁶Ra and ²²⁸Th, the radioactivity indices f1 and f2, and the dose rate above a flat infinite plane of a thick layer (a few metres) of the measured material were calculated.