Dose Rate Levels around Industrial Gamma Sources Héctor René Vega-Carrillo *, Ricardo Rodríguez-Juárez, Eduardo Manzanares- Acuña, Rubén Hernández-Villasana, Jaime Ramírez-González Universidad Autónoma de Zacatecas, Unidad Académica de Estudios Nucleares, Ciprés 10, Fracc. La Peñuela, 98068 Zacatecas, Zac. México Abstract. Dose rate levels around two gamma ray sources utilized in a mining corporation have been determined. Both gamma ray sources are 137 Cs and are installed in a mining corporation to measure on-line the density of mine products. Dose rate levels were calculated in several sites around the 137 Cs sources using two active and several passive thermoluminiscent dosemeters. Using the 137 Cs gamma factor dose rates were calculated in all the points. A comparison between the measured and calculated dose rate levels was carried out. Calculated dose rate levels was obtained for three cases: first, assuming the sources were bare, second, assuming the sources inside their shielding and the third, adding an extra shield to reduce the dose rate levels to those similar to local background. KEYWORDS: 137 Cs, Densitometer, Dose rate, TLD. 1. Introduction Radiation sources are widely utilized in industry, medicine, research and education. The first application of a sealed source was in 1901; until the 40 s sealed sources were made of Radium being exclusively applied in medicine. A sealed source is a small entity containing radioactive material with a high specific activity. Nowadays almost all countries have sealed sources whose activities vary from 10 3 Bq to 10 15 Bq [1]. There are several ways to define the sealed source [2-4], the common elements of all those definitions include the presence of a capsule that kept together the radioactive material, and the capsule is thick enough to avoid the contact as well as any dispersion of radioactive material into the environment when the source is utilized under normal or design conditions. With very few exceptions the materials utilized to build the capsule of a sealed source are stainless steel, titanium, platinum or any other inert metal. Each source is identified with a serial number that is written on the capsule, when the source s activity is large information about the radionuclide, activity and the fabrication date is also included. The definition given by the International Standards Organisation (ISO) for a sealed source is: Radioactive source sealed is a capsule or having a bonded cover, the capsule or cover being strong enough to prevent contact with and dispersion of radioactive material under the conditions of use and wear for which it was designed [5]. This definition gives a better understanding and includes all the elements to distinguish from other type of radioactive sources. In industrial application major concern are sealed sources utilized in sterilization and food preservation, and radiography because their activities varies from 0.1 to 400 PBq and from 0.1 to 5 TBq respectively [5]. Radioisotopes like 192 Ir, 60 Co, and sometimes 137 Cs, 170 Tm, and 169 Yb, are utilized in industrial radiography. Other industrial applications are well logging, moisture detector, conveyor gauge, static eliminators, lightning preventers, electron capture detectors, x-ray fluorescence analysers, calibration, smoke detectors, dredgers, blast furnace control, and gauges for level, thickness, and density. In these applications other radioactive sources are also utilized whose activities varies from 0.1 to 800 GBq. Source such as 137 Cs, 60 Co, 85 Kr, 90 Sr, 14 C, 147 Pm, and 241 Am, are utilized in density, level and thickness gauges. to measure the thickness or density of materials, and to measure the level of liquids in industry are utilized radioisotopes different to those utilized in industrial radiography are applied. A * Presenting author, E-mail: fermineutron@yahoo.com 1
typical radionuclide-based densitometer is shown in Fig. 1. The detector output is utilized to control other industrial processes in order to keep the material density inside a range of values. Figure 1: Density gauge In this application the measurement is carried out continuously. The device is not affected by conditions, like pressure or temperature, of material inside the duct; no special installation is required because this device can be installed in any duct in operation. However, the safety and radiological risks are the main drawback of this application. The radionuclide mostly utilized in the density gauges is 137 Cs, this has 30 y half life, emits 0.662 MeV photons, therefore is easy to shield. The radionuclide is often used as caesium chloride salt, however are also prepared in ceramic form making the radionuclide virtually insoluble in water. When taken up by the body, the highest concentrations are reached in muscle tissue. Safety concerns include the final disposal of this device or the risk to be broken or stolen. Radiological worries are related to assess the radiation levels around the source to assure radiological protection of personnel. The aim of this work was to determine the ambient equivalent dose levels inside the pit where two density gauges with 137 Cs sources are installed in a mining company. Dose levels were determined with calculations that were verified using termoluminiscent dosimeters. 2. Materials and Methods The study was carried out in the mine Proaño which belongs to Peñoles Co. The mine is located in the city of Fresnillo Zacatecas in Mexico. Silver, lead and zinc are the main products of the mine. In the process the gross mine product is molten and the valuable metals are separated, the residue is sent to large tanks whose content pass trough the density gauges to control the full industrial process. This control is made measuring continuously the density. In the mine there are 12 density gauges, two in each station. In the station 87 there are two density gauges with 137 Cs sources whose original activities were 7.4 and 1.85 GBq. The density gauges are identified as GG-5574 and GV-3293. Each source is located inside a lead shield which is fixed in the bottom of a 20.32 cm-diameter pipe. The pipe is 0.635 cm-thickness stainless steel. The detection system associated to the density gauges includes NaI(Tl) and ionizing chamber as detectors, whose signal are used as feedback to whole the mineral processing. In Fig. 2 is shown the density gauge GG-5574. Both gauges are located inside a pit where eventually the personnel must to perform corrected and preventive maintenance activities, or to take lectures from the electronics associated to the detections systems. Occasionally, the workers in charge to maintenance need to be several hours inside the pit where the proximity to the sources can not be fixed. 2
Figure 2: Details of density gauge with 237 Cs source GG-5574 (below the duct) and the NaI(Tl)-based detection system (above the pipe). The shielding quality of both sources was estimated assuming point-like sources inside its original lead casks. The ambient dose equivalent (H*(10)) was calculated inside the pit at the middle point of 137 Cs sources. The H*(10) gamma factor (Γ H*(10) ) for 137 Cs is 0.9103 msv-cm 2 -h -1 -MBq -1, the H*(10) in each point was calculated using equation (1) for Cs-137 [6, 7]. *( 10) A Exp[ µ x] = (1) r () r ΓH*( 10 2 H ) Here A is the source s activity and r is the distance between the source and a point of interest, µ is the attenuation coefficient of lead to 137 Cs photons, and x is the lead thickness. The H*(10) was calculated assuming bare sources, the sources inside their lead shielding and with an extra lead shield, thick enough to reduce the H*(10) to background levels. In order to verify the calculations 10 sites were selected inside the pit where the dose was measured using two pairs of thermoluminiscent dosimeters (TLD 100), which were left in place for 8 hours. Two pairs of TLDs were also utilized to obtain the background dose levels. At other locations the dose was measured using two active dosimeters: Berthold RO 20 and Inspector from S.E. International Inc. Background dose levels were also measured with TLD 100 and both active dosemeters. Dose values measured with TLDs were performed twice; final dose values were the results of mean values of TLDs readouts. 3. Results Las isodose plots inside the pit for both bare sources are shown in figure 3. It can be noticed that the lower dose rate is 20 µsv/h, while the larger dose rates is nearby the source with larger activity. In the area where the electronics of both detectors are located the dose rate is 100 µsv/h, this is approximately in the point (0, 112) of figure 3. 3
80,0 Figure 3: Dose rate, in µsv/h, due both bare sources inside the pit. 112 100 80 25,0 30,0 35,0 40,0 45,0 55,0 60,0 65,0 85,0 80,0 100,0 90,0 70,0 75,0 95,0 100,0 2 3 60,0 65,0 60 40 20 60,0 75,0 95,0 1 1 2 4 85,0 Length of the pit [ cm ] 0-20 -40-60 -80-100 -120-140 -160 25,0 30,0 35,0 40,0 80,0 70,0 85,0 45,0 2500,0 2000,0 1500,0 6 3600,0 5 500,0 4 400,0 9 900,0 7 1000,0 8 800,0 2 1 9000,0 10000,0 9500,0 7500,0 8000,0 8500,0 5500,0 5000,0 6000,0 4000,0 6500,0 7000,0 2000,0 1500,0 3500,0 4500,0 3000,0 9 900,0 8 800,0 2500,0 7 3 6 1000,0 600,0 5 500,0 2 400,0 4 70,0-180 55,0-200 -220-240 -260 20,0 65,0 60,0 100,0 95,0 90,0 75,0 55,0 45,0 40,0-280 35,0-300 -480-440 -400-360 -320-280 -240-200 -160-120 -80-40 0 40 80 120 160 200 240 280 320 360 400 440 Width of the pit [ cm ] In figure 4 are the plots of isodose levels considering that 137 Cs are inside their lead shield. Figure 4: Dose rate, in µsv/h, inside the pit due to both sources inside their lead shielding. Length of the pit [ cm ] 112 100 80 60 40 20 0-20 -40-60 -80-100 -120-140 -160-180 -200-220 0,40 0,70 3,00 0,70 20,00 15,00 6,00 5,50 9,50 9,005,00 7,00 8,50 8,00 7,504,50 4,00 3,00 3,00 3,00 4,50 90,00 95,00 75,00 85,00 70,00 80,00 60,00 65,00 0 45,00 55,00 40,00 20,00 35,00 30,00 15,00 9,50 9,006,00 4,50 8,00 8,505,504,00 7,50 7,005,00 3,00-240 -260-280 0,35 Width of the pit [ cm ] 0,70-300 -480-440 -400-360 -320-280 -240-200 -160-120 -80-40 0 40 80 120 160 200 240 280 320 360 400 440 4
By adding a layer of 6 cm thick of lead to the lead shielding of the sources the dose levels inside the pit are reduced to background levels 0.25 ± 0.09 µsv/h. In figure 5 are shown the dose levels inside the pit with the sources inside their lead shielding, in this figure have been included the dose rates measured with the TLDs and the active dosemeters. In this figure it can be noticed that calculated values agrees well with the measured dose rate values. Figure 5: Dose values calculated and measured. 0.53 ± 0.04 µsv/h 26.5 ± 0.7 µsv/h 0.75 ± 0.07 µsv/h 61.0 ± 1.4 µsv/h 0.88 ± 0.04 µsv/h Length of the pit [ cm ] 112 100 80 60 40 20 0-20 -40-60 -80-100 -120-140 -160-180 -200-220 0,40 0,70 3,00 0,70 20,00 15,00 6,00 5,50 9,50 9,005,00 7,00 8,50 8,00 7,504,50 4,00 3,00 3,00 3,00 4,50 95,00 90,00 85,00 80,00 75,00 70,00 65,00 60,00 0 55,00 45,00 40,00 20,00 35,00 30,00 15,00 9,50 9,006,00 4,50 8,00 8,505,504,00 7,50 7,005,00 3,00-240 -260-280 0,35 Width of the pit [ cm ] 0,70-300 -480-440 -400-360 -320-280 -240-200 -160-120 -80-40 0 40 80 120 160 200 240 280 320 360 400 440 0.62 ± 0.04 µsv/h 0.42 ± 0.05 µsv/h 4. Conclusions Worldwide sealed sources are utilized in several industrial applications. The use of radioactive sources requires safety procedures to assure proper radiological protection for personal located nearby the sources. In this work the isodose plots of two 137 Cs sources located inside a pit has been calculated. Calculations were performed assuming both sources are bare and inside their lead shielding. Dose levels were also measured using TLDs and two active dosemeters. Calculated isodose values agree in less than 10% the measured doses. The sources are utilized in the mining industry, as density gauges, to control the mining process; both are located inside a pit where, at regular basis, personal realizes preventive and corrective maintenance operations that require 5
moving around the sources. In order to reduce the dose levels to values alike the background an extra lead shielding, 6 cm thick, was designed. Acknowledgements This work is part of SYNAPSIS project, partially supported by CONACYT (Mexico) under contract SEP-2005-C01-46893. REFERENCES [1] INTERNATIONAL ATOMIC ENERGY AGENCY, Nature and magnitude of the problem of spent radiation sources, International Atomic Energy Agency Technical Report Series 620, IAEA-TECDOC-620, IAEA Vienna (1991). [2] INTERNATIONAL ATOMIC ENERGY AGENCY, Regulations for the safe transport of radioactive material, International Atomic Energy Agency Safety Series, 6, IAEA Vienna (1985). [3] INTERNATIONAL ATOMIC ENERGY AGENCY, Management of radioactive wastes produced by users of radioactive materials, International Atomic Energy Agency Safety Series, 70, IAEA Vienna (1985). [4] INTERNATIONAL ORGANIZATION FOR STANDARDIZATION, Sealed Radioactive Sources Classification, International Standard No. 2919, (1980). [5] INTERNATIONAL ATOMIC ENERGY AGENCY, Handling, conditioning and disposal of spent sealed sources, International Atomic Energy Agency Technical Report Series 548, IAEA-TECDOC-548, IAEA Vienna (1985). [6] VEGA-CARRILLO, H.R., Cálculo de los factores gamma para radioisótopos usados en medicina nuclear, Rev. Esp. Med. Nuc. 13 (1994) 43. [7] NINKOVIC, M.M., RAICEVIC, J.J. and ADROVIC, F., Air Kerma rate constants for gamma emitters used most often in practice, Radiat. Prot. Dosim. 115 (2005) 247. 6