- Email: [email protected]
Determination of gaseous radionuclide forms in the stack air of nuclear power plants
ARTICLE IN PRESS Applied Radiation and Isotopes 67 (2009) 950–952
Contents lists available at ScienceDirect
Applied Radiation and Isotopes journal homepage: www.elsevier.com/locate/apradiso
Determination of gaseous radionuclide forms in the stack air of nuclear power plants J. Tecl a,, I. Svetlik b a b
Department of Spectrometry, SURO, Bartoskova 48, 140 00 Prague 4, Czech Republic Department of Radiation Dosimetry, NPI AS CR, Prague, Czech Republic
a r t i c l e in f o
Keywords: Nuclear-energy facilities Radioactive noble gases 14 C 85 Kr Grab samples
a b s t r a c t The determination of gaseous forms of radionuclides in ventilation stacks utilising grab samples was performed as a part of independent monitoring of nuclear power plants in the Czech Republic. The activities of radionuclides (e.g. 41Ar, 133Xe, 135Xe, 85mKr, 87Kr, and 131mXe) were determined by semiconductor gamma-ray spectrometry in samples collected into pressure vessels. Radiochemical preparation of samples for determination of 14CO2, combustible 14C forms, and 85Kr was performed after the radioactive decay of radionuclides with short half-life. & 2009 Elsevier Ltd. All rights reserved.
1. Introduction Two nuclear power plants (NPP) with altogether six lightwater pressurized reactors (LWPR) are operating in the Czech Republic (approx. 10 300 000 citizens, 78 900 km2). Sampling and measurement of gaseous releases from ventilation stacks is performed as a part of independent monitoring of nuclear facilities in the Czech Republic. Apart from the aerosol form in gaseous releases, samples for determination of noble gases and gaseous forms of 14C from ventilation stacks are also taken. The determination of gaseous forms of radionuclides in ventilation stacks utilising grab samples is performed by National Radiation Protection Institute (Czech abbreviation SURO) in co-operation with Department of Radiation Dosimetry of Nuclear Physics Institute (NPI) of the Academy of Sciences of the Czech Republic (ASCR). Currently, the most significant artificial sources of radiocarbon in the environment are effluents from nuclear energy facilities, even though they represent a minor contribution in comparison with natural 14C production. Nevertheless, 14C is a major contributor to the collective effective dose among all radionuclides released by NPPs with LWPR during normal operation (UNSCEAR, 2000). The collective effective dose for the critical group of population around Czech NPPs is driven approx. 90% by 14 C, approx. 4% by 41Ar, other noble gases contribute less than 1% (Koc et al., 2005). 85Kr (the key nuclide for discharges from lightwater reactor, European Union, 2004) together with tritium, 14C, and 129I rank among waste products from nuclear facilities, which
Corresponding author. Tel.: +420 226 518 231; fax: +420 241 410 215.
E-mail address: [email protected] (J. Tecl). 0969-8043/$ - see front matter & 2009 Elsevier Ltd. All rights reserved. doi:10.1016/j.apradiso.2009.01.063
due to their long half-life and high mobility are important sources of environmental contamination on the global scale.
2. Sampling The grab samples of stack air with gaseous radionuclides were collected into pressure vessels which are shown in Fig. 1. These pressure vessels (the SURO construction) contain seven small vessels with inner volume about 0.35 L in common construction, with central distribution of gas with manometer and central discharge valve (Fig. 1, left). The total inner volume is approx. 2.5 L, inner diameter of vessel well, for gamma spectrometry measurement, is approx. 8 cm (Fig. 1, right). These vessels are pressurised on approx. 20 MPa by an air compressor, the sampling time is about 10 min.
3. Measuring procedure 3.1. Gamma-ray emitting radionuclides The activities of gamma-ray emitting radionuclides (e.g. 41Ar, Xe, 135Xe, 85mKr, 87Kr, and 131mXe) were determined by semiconductor gamma-ray spectrometry. During the measurement a pressure vessel is put on a detector (seven small vessels create a ring around a detector, which is positioned in the vessel’s well). The measurement can be performed without shielding (e.g. directly in the room nearby ventilation stack) or with shielding (in the gamma-ray spectrometric laboratory). In the SURO three HPGe detectors from Canberra or Ortec with relative efficiency from 10% to 75% (detector with rel. efficiency 75% is 133
ARTICLE IN PRESS J. Tecl, I. Svetlik / Applied Radiation and Isotopes 67 (2009) 950–952
951
Fig. 1. Pressure vessels for samples of gaseous radionuclides, SURO construction, top and bottom/side view.
Table 1 Detection limits for measurements in a shielding of 200 mm thick iron. Nuclide
Half-life
41
Ar Kr Kr 88 Kr 131m Xe 133 Xe 133m Xe 135 Xe
110 min 4.5 h 76 min 2.9 h 11.8 days 5.2 days 2.2 days 9.1 h
85m 87
Detection limits (Bq m3) tm ¼ 1200 sa
tm ¼ 80000 sb
130 90 400 200 240 40 55 30
– – – – 30 5 7 5
tm—Time of measurement. a Delay time between end of sampling and start of measurement is near 1 min. b Delay time between end of sampling and start of measurement is around 3 h.
portable) were used for measurement of samples of noble gases. The gamma-ray spectrometric laboratory of SURO is equipped by four iron shieldings with thickness of 200 mm and 10 lead shieldings with thickness of 100 mm (and with inner layers from cooper and tin approx. 2 mm). Detection limits, i.e. 5% probability of the first kind of observation error (Currie, 1995) for measurements in a 200 mm shielding (two different times of measurement, HPGe detector with relative efficiency 75%) are shown in Table 1. The measurement without shielding was usually performed immediately after the end of sampling (short measurement, approx. 1200 s). The laboratory measurement in the shielding was started immediately after transport of sampling vessels to the laboratory. The first measurement was short-time due to presence of radionuclides with short half-life. The second measurement was performed in the laboratory only, inside the shielding. Duration of the second measurement was in the range 60 000–90 000 s to determine radionuclides with longer half-life.
3.2.
14
C
All collected grab samples were stored for at least one month to reduce activities of short-lived radionuclides (half-life of 14C is 5730 years). In the laboratory, gas samples were sequentially transferred into a low-pressure gas storage bag, equipped by two inlet/outlet (opposite sided) tubes.
For determination of 14CO2 and combustible chemical 14C forms the gas was transferred into low-pressure bag through a gas-meter and traps with humidifying and washing solution (1% H3PO4), sorption solution (3 M NaOH), and a condensation flask with 1% H3PO4 cooled solution (+5 1C) to avoid water vapor condensation in inner parts of the system (Svetlik et al., 2007). When the transfer of gas was completed, a cycling pump was connected between the condensation flask and the gas storage input. Then the gas flows from the bag’s output into the gas-meter input and continues into flasks. The pressure inside the gas-meter and flasks is lower than the atmospheric pressure to avoid leaking of the gas sample. The total volume cycled corresponds to 13 volumes of a gas sample processed to reach a sufficiently high yield (above 98%) of CO2 sorption in the NaOH solution. Sorption yield experiments testing a single sample passage through a trap with NaOH, i.e. one sorption cycle resulted in the yield in the range 33–36% depending on the gas flow rate. Hence, the longterm cycling allows achieving a high yield with a negligible 14CO2 residue in the gas. To determine 14C in combustible compounds, the catalyst (CuO, heated on 700 1C) in a quartz tube was inserted between the cycling bag output and the gas-meter input. The solution with a radiocarbon sample was sequentially purged by nitrogen bubbling to remove residual dissolved radioactive noble gasses. After the end of sample processing, the residual gasses were vacated from low-pressure gas storage and the bag was two times flushed out with nitrogen. Finally, 3 g of sample solution were mixed with 17 mL of Hionic scintillation cocktail in a 20 mL low potassium glass vial. Blanc samples were prepared using the same routine from 3 M NaOH after fossil CO2 addition. All samples, including blank, were counted for 60 min by TriCarb 1050 or TriCarb 3170 TR/SL in normal mode. The measurement of each sample series was repeated in 2 weeks intervals, three times at minimum. A counting window derived from the optimized one was utilized, with the lower edge above the maximum energy of tritium beta emission. After each measurement, the spectra were checked in the region above the maximum energy of 14C beta emission. Completing the third measurement, count rates from each sample measurement with their combined uncertainties were compared. Significant differences of the sample measurement could indicate presence of interfering radionuclides with a shorter half-life, or instability of the sample scintillation mixture. Counting efficiency was determined individually, usually by standard addition (0.05 g of carbonate solution spiked with a known activity of 14C) into the sample after the third measure-
ARTICLE IN PRESS 952
J. Tecl, I. Svetlik / Applied Radiation and Isotopes 67 (2009) 950–952
ment. For purpose of calibration, radioactive standards (14C labeled sodium carbonate solution), type ER 25, produced by the Czech Metrology Institute-Inspectorate for Ionizing Radiation Prague, were utilized. Detection limits are about 2 Bq m3, for corresponding probability of the first kind observation error about 5% (Currie, 1995). 3.3.
85
Kr
After one month’s storage, the grab samples were processed also by a routine for 85Kr determination in the laboratory of NPI (half-life of 85Kr is 10.75 years). Before measurements the samples have to be enriched in krypton. The routine is based on cryogenic selective adsorption of air components on charcoal beds. Krypton is isolated in repeated adsorption–desorption cycles within the temperature range of 77–473 K, which are performed in two adsorption columns. Separation yield is determined by gaschromatography and mass spectrometry analysis of the krypton amount in the prepared sample. For determination of 85Kr activity a measuring system was designed in the NPI. Beta-emissions of 85Kr are detected by a CaF2(Eu) scintillation crystal. The system and procedures used are described in detail in Wilhelmova et al. (1985) and Wilhelmova et al. (1986). To determine 85Kr activity in the sampled stack air with uncertainty of typically about 10%, processing of a sample with volume about 0.2 m3 is sufficient. Detection limit are about 1 Bq m3, for corresponding probability of the first kind observation error about 5% (Currie, 1995).
4. Conclusion The methods for grab sampling in the stack and measurements presented are routinely used by SURO in the framework of Independent monitoring of NPPs or other nuclear facilities in the Czech Republic. Activity concentrations of gamma emitting noble gases, 85Kr and forms of 14C have been routinely determined in each grab sample. The determination of gamma emitting noble gases described in this article is routinely pursued in the laboratory where the HPGe detectors are put in the shielding, but in some cases they are placed directly in the room nearby a ventilation stack (without shielding, using a portable HPGe detector). Detection limits are lower in the laboratory (about ten
times), but radionuclides with short half-lives are hardly detectable due to radioactive decay during transfer. Some results from independent monitoring were presented in (Tecl, 2005); all results from the independent monitoring and from the radiation monitoring network of the Czech Republic are published annually in the ‘‘Report on Radiation Situation on the Territory of the Czech Republic’’ in Czech language. Due to relatively low detection limits the method is usable for determination of gaseous radionuclide forms (primarily gammaray emitting radionuclides) also in the air in case of nuclear accident.
Acknowledgments This work was partly supported by program of independent monitoring of nuclear power plants in the Czech Republic and by institutional funding of the Nuclear Physics Institute ASCR (AV0Z 10480505).
References Currie, L.A., 1995. Nomenclature in evaluation of analytical methods including detection and quantification capabilities (IUPAC Recommendation 1995). Pure Appl. Chem. 67 (10), 1699–1723. European Union, 2004. European Union Commission Recommendation, 2004/2/ Euratom, notified under document number C, 2003, p. 4832. Koc, J., Kulich, V., Pospichal, J., Vokalek, J., Hak, J., Fiala, L., 2005. Review of radioactive outfalls from NPPs in the Czech Republic and evaluation of impact on it’s vicinity. XXVII Days of Radiation Protection, Liptovsky Jan, Slovakia, 28.11–2.12.2005. In: Conference Proceedings, pp. 106–111 (in Czech). Svetlik, I., Rulik, P., Michalek, V., Tomaskova, L., Mizera, J., 2007. Determination of carbon-14 in grab samples of stack air from nuclear power plants. In: Chalupnik, S., Scho¨nhofer, F., Noakes, J. (Eds.), Radiocarbon, LSC 2005, Advances in liquid scintillation spectrometry, pp. 417–422. Tecl, J., 2005. Results of independent monitoring of releases of noble gases from nuclear facilities collected by the SURO Prague. XXVII. Days of Radiation Protection, Liptovsky Jan, Slovakia, 28.11–2.12.2005. In: Conference Proceedings, pp. 234–238. UNSCEAR, 2000. Report of the United Nations Scientific Committee on the Effects of Atomic Radiation to the General Assembly (UNSCEAR). Exposures from natural and man-made sources of radiation, Report 1. Wilhelmova, L., Dvorak, Z., Tomasek, M., Stukheil, K., 1986. Application of the CaF2(Eu) scintillator to 85Kr monitoring in atmospheric air samples. Appl. Radiat. Isot. 37 (5), 429–432. Wilhelmova, L., Tomasek, M., Dvorak, Z., 1985. Monitoring of the atmospheric activity of 85Kr in Prague. Radioanal. Nucl. Chem. Lett. 95 (1), 451–455.
Contents lists available at ScienceDirect
Applied Radiation and Isotopes journal homepage: www.elsevier.com/locate/apradiso
Determination of gaseous radionuclide forms in the stack air of nuclear power plants J. Tecl a,, I. Svetlik b a b
Department of Spectrometry, SURO, Bartoskova 48, 140 00 Prague 4, Czech Republic Department of Radiation Dosimetry, NPI AS CR, Prague, Czech Republic
a r t i c l e in f o
Keywords: Nuclear-energy facilities Radioactive noble gases 14 C 85 Kr Grab samples
a b s t r a c t The determination of gaseous forms of radionuclides in ventilation stacks utilising grab samples was performed as a part of independent monitoring of nuclear power plants in the Czech Republic. The activities of radionuclides (e.g. 41Ar, 133Xe, 135Xe, 85mKr, 87Kr, and 131mXe) were determined by semiconductor gamma-ray spectrometry in samples collected into pressure vessels. Radiochemical preparation of samples for determination of 14CO2, combustible 14C forms, and 85Kr was performed after the radioactive decay of radionuclides with short half-life. & 2009 Elsevier Ltd. All rights reserved.
1. Introduction Two nuclear power plants (NPP) with altogether six lightwater pressurized reactors (LWPR) are operating in the Czech Republic (approx. 10 300 000 citizens, 78 900 km2). Sampling and measurement of gaseous releases from ventilation stacks is performed as a part of independent monitoring of nuclear facilities in the Czech Republic. Apart from the aerosol form in gaseous releases, samples for determination of noble gases and gaseous forms of 14C from ventilation stacks are also taken. The determination of gaseous forms of radionuclides in ventilation stacks utilising grab samples is performed by National Radiation Protection Institute (Czech abbreviation SURO) in co-operation with Department of Radiation Dosimetry of Nuclear Physics Institute (NPI) of the Academy of Sciences of the Czech Republic (ASCR). Currently, the most significant artificial sources of radiocarbon in the environment are effluents from nuclear energy facilities, even though they represent a minor contribution in comparison with natural 14C production. Nevertheless, 14C is a major contributor to the collective effective dose among all radionuclides released by NPPs with LWPR during normal operation (UNSCEAR, 2000). The collective effective dose for the critical group of population around Czech NPPs is driven approx. 90% by 14 C, approx. 4% by 41Ar, other noble gases contribute less than 1% (Koc et al., 2005). 85Kr (the key nuclide for discharges from lightwater reactor, European Union, 2004) together with tritium, 14C, and 129I rank among waste products from nuclear facilities, which
Corresponding author. Tel.: +420 226 518 231; fax: +420 241 410 215.
E-mail address: [email protected] (J. Tecl). 0969-8043/$ - see front matter & 2009 Elsevier Ltd. All rights reserved. doi:10.1016/j.apradiso.2009.01.063
due to their long half-life and high mobility are important sources of environmental contamination on the global scale.
2. Sampling The grab samples of stack air with gaseous radionuclides were collected into pressure vessels which are shown in Fig. 1. These pressure vessels (the SURO construction) contain seven small vessels with inner volume about 0.35 L in common construction, with central distribution of gas with manometer and central discharge valve (Fig. 1, left). The total inner volume is approx. 2.5 L, inner diameter of vessel well, for gamma spectrometry measurement, is approx. 8 cm (Fig. 1, right). These vessels are pressurised on approx. 20 MPa by an air compressor, the sampling time is about 10 min.
3. Measuring procedure 3.1. Gamma-ray emitting radionuclides The activities of gamma-ray emitting radionuclides (e.g. 41Ar, Xe, 135Xe, 85mKr, 87Kr, and 131mXe) were determined by semiconductor gamma-ray spectrometry. During the measurement a pressure vessel is put on a detector (seven small vessels create a ring around a detector, which is positioned in the vessel’s well). The measurement can be performed without shielding (e.g. directly in the room nearby ventilation stack) or with shielding (in the gamma-ray spectrometric laboratory). In the SURO three HPGe detectors from Canberra or Ortec with relative efficiency from 10% to 75% (detector with rel. efficiency 75% is 133
ARTICLE IN PRESS J. Tecl, I. Svetlik / Applied Radiation and Isotopes 67 (2009) 950–952
951
Fig. 1. Pressure vessels for samples of gaseous radionuclides, SURO construction, top and bottom/side view.
Table 1 Detection limits for measurements in a shielding of 200 mm thick iron. Nuclide
Half-life
41
Ar Kr Kr 88 Kr 131m Xe 133 Xe 133m Xe 135 Xe
110 min 4.5 h 76 min 2.9 h 11.8 days 5.2 days 2.2 days 9.1 h
85m 87
Detection limits (Bq m3) tm ¼ 1200 sa
tm ¼ 80000 sb
130 90 400 200 240 40 55 30
– – – – 30 5 7 5
tm—Time of measurement. a Delay time between end of sampling and start of measurement is near 1 min. b Delay time between end of sampling and start of measurement is around 3 h.
portable) were used for measurement of samples of noble gases. The gamma-ray spectrometric laboratory of SURO is equipped by four iron shieldings with thickness of 200 mm and 10 lead shieldings with thickness of 100 mm (and with inner layers from cooper and tin approx. 2 mm). Detection limits, i.e. 5% probability of the first kind of observation error (Currie, 1995) for measurements in a 200 mm shielding (two different times of measurement, HPGe detector with relative efficiency 75%) are shown in Table 1. The measurement without shielding was usually performed immediately after the end of sampling (short measurement, approx. 1200 s). The laboratory measurement in the shielding was started immediately after transport of sampling vessels to the laboratory. The first measurement was short-time due to presence of radionuclides with short half-life. The second measurement was performed in the laboratory only, inside the shielding. Duration of the second measurement was in the range 60 000–90 000 s to determine radionuclides with longer half-life.
3.2.
14
C
All collected grab samples were stored for at least one month to reduce activities of short-lived radionuclides (half-life of 14C is 5730 years). In the laboratory, gas samples were sequentially transferred into a low-pressure gas storage bag, equipped by two inlet/outlet (opposite sided) tubes.
For determination of 14CO2 and combustible chemical 14C forms the gas was transferred into low-pressure bag through a gas-meter and traps with humidifying and washing solution (1% H3PO4), sorption solution (3 M NaOH), and a condensation flask with 1% H3PO4 cooled solution (+5 1C) to avoid water vapor condensation in inner parts of the system (Svetlik et al., 2007). When the transfer of gas was completed, a cycling pump was connected between the condensation flask and the gas storage input. Then the gas flows from the bag’s output into the gas-meter input and continues into flasks. The pressure inside the gas-meter and flasks is lower than the atmospheric pressure to avoid leaking of the gas sample. The total volume cycled corresponds to 13 volumes of a gas sample processed to reach a sufficiently high yield (above 98%) of CO2 sorption in the NaOH solution. Sorption yield experiments testing a single sample passage through a trap with NaOH, i.e. one sorption cycle resulted in the yield in the range 33–36% depending on the gas flow rate. Hence, the longterm cycling allows achieving a high yield with a negligible 14CO2 residue in the gas. To determine 14C in combustible compounds, the catalyst (CuO, heated on 700 1C) in a quartz tube was inserted between the cycling bag output and the gas-meter input. The solution with a radiocarbon sample was sequentially purged by nitrogen bubbling to remove residual dissolved radioactive noble gasses. After the end of sample processing, the residual gasses were vacated from low-pressure gas storage and the bag was two times flushed out with nitrogen. Finally, 3 g of sample solution were mixed with 17 mL of Hionic scintillation cocktail in a 20 mL low potassium glass vial. Blanc samples were prepared using the same routine from 3 M NaOH after fossil CO2 addition. All samples, including blank, were counted for 60 min by TriCarb 1050 or TriCarb 3170 TR/SL in normal mode. The measurement of each sample series was repeated in 2 weeks intervals, three times at minimum. A counting window derived from the optimized one was utilized, with the lower edge above the maximum energy of tritium beta emission. After each measurement, the spectra were checked in the region above the maximum energy of 14C beta emission. Completing the third measurement, count rates from each sample measurement with their combined uncertainties were compared. Significant differences of the sample measurement could indicate presence of interfering radionuclides with a shorter half-life, or instability of the sample scintillation mixture. Counting efficiency was determined individually, usually by standard addition (0.05 g of carbonate solution spiked with a known activity of 14C) into the sample after the third measure-
ARTICLE IN PRESS 952
J. Tecl, I. Svetlik / Applied Radiation and Isotopes 67 (2009) 950–952
ment. For purpose of calibration, radioactive standards (14C labeled sodium carbonate solution), type ER 25, produced by the Czech Metrology Institute-Inspectorate for Ionizing Radiation Prague, were utilized. Detection limits are about 2 Bq m3, for corresponding probability of the first kind observation error about 5% (Currie, 1995). 3.3.
85
Kr
After one month’s storage, the grab samples were processed also by a routine for 85Kr determination in the laboratory of NPI (half-life of 85Kr is 10.75 years). Before measurements the samples have to be enriched in krypton. The routine is based on cryogenic selective adsorption of air components on charcoal beds. Krypton is isolated in repeated adsorption–desorption cycles within the temperature range of 77–473 K, which are performed in two adsorption columns. Separation yield is determined by gaschromatography and mass spectrometry analysis of the krypton amount in the prepared sample. For determination of 85Kr activity a measuring system was designed in the NPI. Beta-emissions of 85Kr are detected by a CaF2(Eu) scintillation crystal. The system and procedures used are described in detail in Wilhelmova et al. (1985) and Wilhelmova et al. (1986). To determine 85Kr activity in the sampled stack air with uncertainty of typically about 10%, processing of a sample with volume about 0.2 m3 is sufficient. Detection limit are about 1 Bq m3, for corresponding probability of the first kind observation error about 5% (Currie, 1995).
4. Conclusion The methods for grab sampling in the stack and measurements presented are routinely used by SURO in the framework of Independent monitoring of NPPs or other nuclear facilities in the Czech Republic. Activity concentrations of gamma emitting noble gases, 85Kr and forms of 14C have been routinely determined in each grab sample. The determination of gamma emitting noble gases described in this article is routinely pursued in the laboratory where the HPGe detectors are put in the shielding, but in some cases they are placed directly in the room nearby a ventilation stack (without shielding, using a portable HPGe detector). Detection limits are lower in the laboratory (about ten
times), but radionuclides with short half-lives are hardly detectable due to radioactive decay during transfer. Some results from independent monitoring were presented in (Tecl, 2005); all results from the independent monitoring and from the radiation monitoring network of the Czech Republic are published annually in the ‘‘Report on Radiation Situation on the Territory of the Czech Republic’’ in Czech language. Due to relatively low detection limits the method is usable for determination of gaseous radionuclide forms (primarily gammaray emitting radionuclides) also in the air in case of nuclear accident.
Acknowledgments This work was partly supported by program of independent monitoring of nuclear power plants in the Czech Republic and by institutional funding of the Nuclear Physics Institute ASCR (AV0Z 10480505).
References Currie, L.A., 1995. Nomenclature in evaluation of analytical methods including detection and quantification capabilities (IUPAC Recommendation 1995). Pure Appl. Chem. 67 (10), 1699–1723. European Union, 2004. European Union Commission Recommendation, 2004/2/ Euratom, notified under document number C, 2003, p. 4832. Koc, J., Kulich, V., Pospichal, J., Vokalek, J., Hak, J., Fiala, L., 2005. Review of radioactive outfalls from NPPs in the Czech Republic and evaluation of impact on it’s vicinity. XXVII Days of Radiation Protection, Liptovsky Jan, Slovakia, 28.11–2.12.2005. In: Conference Proceedings, pp. 106–111 (in Czech). Svetlik, I., Rulik, P., Michalek, V., Tomaskova, L., Mizera, J., 2007. Determination of carbon-14 in grab samples of stack air from nuclear power plants. In: Chalupnik, S., Scho¨nhofer, F., Noakes, J. (Eds.), Radiocarbon, LSC 2005, Advances in liquid scintillation spectrometry, pp. 417–422. Tecl, J., 2005. Results of independent monitoring of releases of noble gases from nuclear facilities collected by the SURO Prague. XXVII. Days of Radiation Protection, Liptovsky Jan, Slovakia, 28.11–2.12.2005. In: Conference Proceedings, pp. 234–238. UNSCEAR, 2000. Report of the United Nations Scientific Committee on the Effects of Atomic Radiation to the General Assembly (UNSCEAR). Exposures from natural and man-made sources of radiation, Report 1. Wilhelmova, L., Dvorak, Z., Tomasek, M., Stukheil, K., 1986. Application of the CaF2(Eu) scintillator to 85Kr monitoring in atmospheric air samples. Appl. Radiat. Isot. 37 (5), 429–432. Wilhelmova, L., Tomasek, M., Dvorak, Z., 1985. Monitoring of the atmospheric activity of 85Kr in Prague. Radioanal. Nucl. Chem. Lett. 95 (1), 451–455.