RADIOLOGICAL RISK ASSESSMENT FOR HOT CELL DECONTAMINATION

Transcription

1 RDIOLOGICL RISK SSESSMENT FOR HOT CELL DECONTMINTION C. TUC, R. DEJU,. ZORLIU 1 Horia Hulubei National Institute for Nuclear Physics and Engineering, P.O.Box MG-6, RO Bucharest-Magurele, Romania, tuca@nipne.ro; radudeju@nipne.ro; zorliu@nipne.ro Received January 31, 2017 bstract. The paper describes the radiological impact on the workers who perform the decontamination of a hot cell from the VVR-S nuclear research reactor (NR), used for production of radioisotopes during the operation of the NR. The assessment of dose equivalent takes into account the used manual procedure to make the floor decontamination. The identification of the high radiation areas is done by measurements of the ambient dose equivalent by direct measurements using: (i) a portable digital survey meter with a gamma dose rate probe placed at less 1 cm above the surface and (ii) thermo-luminescent dosimeters placed directly on surface and indirectly, by (iii) gamma ray spectrometry measurements of the samples taken from each high radiation area of the contaminated floor. lthough the measurement conditions were difficult the results are in a satisfactory agreement, validating the measurement. The dose rate assessment is done for the worker that performs the direct measurements of the ambient dose equivalent and for worker B who performs the decontamination of the contaminated area. The external penetrant dose equivalent for worker is less than 0.28 msv (for an operation at about 5 minutes) 3.39 msv/h. For the worker B the dose from external irradiation is 1.59 msv (for an operation 12 minutes), a dose rate of 7.97 msv/h and the internal committed effective dose, E (50) received by worker due to air inhalation is 1.63 µsv. Key words: nuclear reactor, hot cell, worker, dose rate, radiological risk. 1. INTRODUCTION The paper presents studies regarding the assessment of the radiological risk for workers who perform decontamination of a hot cell from VVR-S nuclear research reactor owned by Horia Hulubei National Institute for R&D in Physics and Nuclear Engineering (IFIN-HH) Bucharest-Magurele, Romania. The reactor was operated between 1957 and 1997 at a nominal thermal power of 2 MW with a maximum fluence rate of neutrons of n/cm 2 s, producing 9.59 GWd. The plant was designed and used for research in physics, biophysics, biochemistry and radiochemistry, using the external neutron flux extracted from the reactor. One of the main purposes of the reactor was the radioisotope production for industrial and medicine applications using five hot cells. Romanian Journal of Physics 62, 812 (2017)

2 rticle no. 812 C. Tuca, R. Deju,. Zorliu 2 The hot cell no. 1 was used for the receipt of the capsules from reactor and extraction of the irradiated materials using a milling device and placement of the sources into dedicated shelves. Hot cells no. 2, 3 and 4 were used for the processing of the irradiated materials and the fifth one for containers storage (with radioactive sources) and transfer. Each hot cell is a parallelepiped having the following dimensions: length 2 m, width 1.2 m and height 2 m. The biological protection of the hot cells in the upper part consists of a layer of concrete with 2 m a thickness and 2.3 g/cm 3 density; on the sides is a layer of concrete with 75 cm thickness and 4.2 g/cm 3 density, and between them is a hard concrete wall with 0.65 m thickness. Inside of these are lined with a layer (3 cm thickness) of chromium nickel austenitic stainless steel 1H18N9T type. Between the hot cells the irradiated materials are transferred using a trolley equipped with two trays (one clean and the other dirty ). This trolley is moving on the rails located into the transfer channel of the hot cells. The channel is a parallelepiped with length 0.6 m, width 0.3 m and height 0.3 m situated under the hot cell entrance. In front of each hot cell an operator room is located, for the handling of the irradiated sources with mechanical devices. The visualisation of the operations is done through lead glass portholes with 0.72 m thickness. The access inside of hot cell for decontamination or maintenance purposes is done by sealed iron doors located on the opposite wall of the mechanical devices. During the operating period there were minor events such as: hot cell floor contamination as a result of the spreading of irradiated samples content by accidently detachment of the capsule's cap or from the spilling of the irradiated liquid from vials and the breakdown of the mechanical devices from hot cell no. 4 [1]. Now, the reactor is in the last phase of the decommissioning process. In this phase is scheduled to be done the decommissioning of the hot cells, consisting on the evacuation of the equipment and irradiated materials resulted from research activities and radioisotope production and clean-up of contaminated surfaces. The radioactive waste from hot cells nos. 3, 2 and 1 are evacuated using mechanical devices and trolley, but those from hot cell no. 4 are evacuated on the iron door due to the malfunction of the mechanical devices. Related to the decontamination process of the hot cells the radiological risks for the workers who perform this task are estimated and discussed in the paper for the hot cell no. 4. The assessment of dose equivalent takes into account the used manual procedure to perform the floor decontamination. The estimation of radiological risks for the workers consists in the scanning of hot cell floor in order to determine the high radiation areas and ambient dose equivalent measurements using: portable digital survey meter with a gamma dose rate probe placed less 1 cm above the surface and thermoluminescent dosimeters placed directly on floor surface. Then, samples are taken from a 100 cm 2 square surface from each high radiation area and measured by gamma ray spectrometry in

3 3 Radiological risk assessment for hot cell decontamination rticle no. 812 order to estimate the radionuclide content and associated activities of irradiation sources. Radiological risk assessment is done for two exposure situations: i) the worker who performs direct measurement of the ambient dose equivalent (worker located about 70 cm from the point of measurement); ii) the worker who performs floor decontamination (worker B located about 45 cm from the high radiation area). The external penetrating equivalent dose is calculated for both scenarios, taking into account that exposure time is about 5 minutes for the worker and 12 minutes for worker B. In addition, for the decontamination process, the internal committed effective dose, E (50), due to the dust in air inhalation, is calculated. 2. MTERILS ND EQUIPMENT The measurements were carried out in two laboratories from IFIN-HH notified by CNCN, in compliance with EN/ISO IEC 17025:2005, according to the specific work procedures [2 4]. The laboratories and the dedicated equipment used for measurements are presented in Table 1. Laboratory Radiologic Characterization Laboratory (LCR) from Reactor Decommissioning Department (DDR) Personal Dosimetry and Environment Laboratory (LDPM) from Life and Environmental Physics Department (DFVM) Table 1 Laboratories and equipment used Equipment I. Portable digital survey meter Thermo Scientific FH 40 G type with FHZ gamma dose rate probe, type: GM C300 ZP The main characteristics: Display range (10 nsv/h H * (10) 10 Sv/h); Measuring range (0.5µSv/h H * (10) 10 Sv/h); Energy range: 60 kev 1.3 MeV; Sensitivity for 137 Cs (s -1 /(μsv/h)): (Low 1.7; High 0.017); Photon radiation overload (100); Dimensions: ø 35; length 185; weight II. Gamma-ray spectrometry system consists of a GEM60P4-95 high-purity germanium coaxial detector (HPGe), a DSPEC jr.2.0 digital signal processing and a low-background shield. The main performance specifications of the detector warranted by the producer are: relative efficiency 60%, resolution (FWHM) 1.95 kev, peak-to-compton ratio 70:1, peak shape (FWFM/FWHM) 3.0, all evaluated at 1.33 MeV peak of 60 Co. The support stand and the shield jacket are made from low carbon steel and a graded lined of copper and tin layers is provided for the suppression of lead X-rays. Mastero-32 and GammaVision-32 software are used for spectra acquisition and analysis. I. Thermoluminiscent dosimetry systems (TLDS) [4] for the estimation of personal ambient dose equivalent, H*(10), consists of: a) dosimeter with thermo-luminescent detectors type LiF: Mg, Cu, P (GR- 200) having an equivalent high sensitivity for tissue, provided with metallic filters to compensate and assure a quasi-constant response on the whole energy interval, 30 kev 3 MeV; b) REDER-NLYSER R-94 TLD system for measurement, analysis and evaluation of thermo-luminescent materials. ccuracy: +2% for multiple detector reading; Stability: better than +2% during 8 hours of work; Duration of measuring: mode REDER: s; standard 22 s; mode XREDER: s; standard-22 s; mode NLYSER: 25 s 4000 s. Three stage heating from 40 o to 400 o. Duration of each stage from 1s to 60 s; Line heating programmable in the range: o C/s; High PMT tension/voltage: automatically regulated;

4 rticle no. 812 C. Tuca, R. Deju,. Zorliu 4 Laboratory Equipment Highly stable Platinum: alloy heater; Digital input/output: LCD Display, 6 programming keys, parallel CENTRONICS interface, RS-232C serial interface; Power supply: 230V+10%, 0.8, 50Hz; Operating temperature range: 10 C 35 C; c) oven for heat treatment and regeneration of the detectors, PTW type TLDO. The treatment is carried out at a temperature of 240 C for 10 minutes; d) auxiliary pieces Vacuum tweezers for detectors handling, storage boxes for detectors, nitrogen gas generator. 3. METHODS 3.1. THE MBIENT DOSE EQUIVLENT H*(10) MESUREMENTS Scanning of the floor surface The first step in the risk assessment for the workers who perform the decontamination activities of the hot cell no. 4 consists in the floor surface contamination scanning, using a portable digital survey meter Thermo Scientific FH 40 G type with FHZ gamma dose rate probe, in order to identify the areas with high activity due to emitting alpha, beta and gamma radionuclides and for qualitative estimation of the degree of spatial uniformity of contamination/activation. Fig. 1 Floor surface scanning.

5 5 Radiological risk assessment for hot cell decontamination rticle no. 812 The direct measurements of the ambient dose equivalent are performed by the worker situated at about 70 cm from the measurement point (Fig. 1) which scans the contaminated surface, maintaining the detector at a distance less than 1 cm above it, and moving it with a speed at about 10 cm/sec according to the specific procedure [2]. Where the warning threshold is surpassed and the audible alarm is triggered, the hot areas are identified and marked, by a paint dot applied in its center H*(10) estimation with thermoluminescent dosimetry system (TLDS) On each hot area the dosimeters with thermo-luminescent detectors type GR-200 are placed for an hour in order to make parallel measurements of the ambient dose equivalent (Fig. 2). The measurement, analysis and evaluation of the detectors are done according to the specific procedure [3] with the REDER-NLYSER R-94 TLD system [5] in the Personal Dosimetry and Environment Laboratory (LDPM) from Life and Environmental Physics Department (DFVM). For this purpose each detector is removed from the dosimeter using tweezers, placed on the heating tray from the sample chamber of the apparatus and read. The thermoluminiscent signal is displayed in number of pulses. The calculation of the ambient dose equivalent H*(10) [3] is done according to eq. 1. H * (10) I R (1) where: * H (10) (msv) the ambient dose equivalent; I (imp) the average reading of irradiated detectors; R ( imp msv 1 ) the dosimeter response to the photonic radiations. The measurement uncertainty is calculated according to eq. 2: U ku c, (2) where: k the coverage factor, to establish the confidence level; u c the combined standard uncertainty. The combined standard uncertainty is calculated according eq. 3. u c 2 2 B u u, (3) where: u = the uncertainty of type, due to statistical fluctuation, calculated according to eq. 4;

6 rticle no. 812 C. Tuca, R. Deju,. Zorliu 6 where: n 2 ( xi x) 1 i1 u. (4) x n( n 1) The uncertainty of type B u B due to systematic errors consists of: u standard = u certificate /k = 0.05 uncertainty due to the standard, u certificate uncertainty specified in the calibration certificate; k the extension factor; u reponse = u display = u R = the error due to the apparatus response; 3 3 u di. = the display error due last digit displayed. 3 Thus, by replacing of the uncertainties values into the eq. 2, and considering k = 2, the measurement uncertainty U is 12% U = Fig. 2 The thermoluminiscent detectors emplacement.

7 7 Radiological risk assessment for hot cell decontamination rticle no GMM-RY SPECTROMETRY The gamma-ray spectrometry method is used for the measurement of the samples taken from the areas with high activity on the floor surface of the hot cell no Sampling and analysis From the seven points with high activity identified by measurements of the ambient dose equivalent smears were taken. Thus, the dust on an area of 100 cm 2 around the hot point was wiped out with square gauze (5 cm 5 cm dimension). The samples were placed in the plastic bags, unique numbered and then transferred in the lead containers to the Radiologic Characterization Laboratory (LCR) from Reactor Decommissioning Department (DDR) in order to be measured by gammaray spectrometry method [6] and using the specific work procedure [4]. The samples were measured in plastic bags to avoid the detector contamination by placing in the lead castle directly on the end cap of the detector. In order to perform samples analysis, the energy calibration of LCR spectrometric system was performed with standard sources located on the symmetry axis of detector, covering the energy domain of interest ( kev) [7], prepared by Radionuclide Metrology Laboratory (RML) of IFIN-HH. The spectrometers calibration was done by the same laboratory, RML [8] RDIOGICL RISK SSESSMENT The radiological risk assessment for the workers is done according to the following exposure situations: i) the external exposure for the worker who performs direct measurement of the ambient dose equivalent (worker ) located at about 70 cm from the measurement point assuming that the exposure time is about 5 minutes; ii) the external exposure of the worker who performs floor decontamination (worker B) located about 45 cm from the high radiation area assuming that the exposure time is about 15 minutes; and the internal exposure as a result of dust in air inhalation, too. The dose received by a worker during an operation is calculated based on the activity concentration of the radionuclides from each hot area determined by gamma-ray spectrometry measurements of the samples [6] Risk assessment for the ambient dose equivalent measurements The calculation hypotheses are: (i) the sampling yield for the activity measurement is 10%; (ii) during the ambient dose measurements, the worker is standing for five minutes at the entrance of the hot cell at about 70 cm of the hot

8 rticle no. 812 C. Tuca, R. Deju,. Zorliu 8 area, along its axis; (iii) the whole activity is concentrated in the hottest point sample has an activity of 10 8 Bq, three orders of magnitude higher than the other 6 hot points having similar activities of 10 5 Bq. Thus, the external penetrate dose equivalent due to the contribution of a single hot point is calculated according to eq. 5 [9]... p( 10) Ck Da (5) H where: C k (Sv/Gy) the effective dose per air kerma free-in-air, for monoenergetic photons (the values are from table.2 [10]), Ḋ a the air kerma rate, calculated according to the eq. 5.. D a, (5 ) 2 r where: (msv/h Bq at 1 m) the kerma gamma ray constant for a point source of a specific radionuclide; (Bq) the sum activity supposed concentrated in the hot spot, calculated from the measured sample activities, sample, using a sampling yield of 10%, according to eq. 5 ; sample. (5 ) The distance from a hot spot to the worker r (cm) is calculated according to eq. 5 : 2 2 r ( x R ), (5 ) where: x (cm) the distance from the source center to the worker; R (cm) the distance from the source center to the spot. Thus, the external penetrant dose equivalent for the worker due to the contribution of the contaminated area is calculated by according to eq. 6.. n H p H pi ( 10). (6) i1.

9 9 Radiological risk assessment for hot cell decontamination rticle no Risk assessment for hot cell floor decontamination The decontamination process is performed in 12 minutes by the worker B in three steps (4 minutes each of them) as follows: (i) spraying of the decontamination substance (DeconGel type 1108 a liquid which performs a decontamination up to 100% of the surfaces containing the radioisotopes, particulates, and heavy metals) from 90 cm high from the floor and 45 cm distance from contaminated area on the horizontal (Fig. 3); (ii) the spreading of the hydro gel coating on the contaminated surface using a V-tooth trowel from 40 cm high and 45 cm distance on the horizontal (Fig. 4); (iii) the removing of the gel film the product which locks the contaminants into the polymer matrix after its drying (Fig. 5). The external penetrating dose equivalent is calculated based on the above mentioned assumptions and using the equations (5 6) considering that the sampling yield for the activity is 10%. During the decontamination process the contaminated particulates from the floor could be spread in the air. Thus the internal committed effective dose, E(50) received by worker due to air inhalation is calculated based on the assumption that in the hot cell air is spread just 10-4 of the total activity of the floor and the worker wears a mask having a filter with a retention efficiency of 99%. The activity concentration in the air due to the contribution of the important seven hot areas is calculated according to the eq. 7. Fig. 3 Spraying of the decontamination substance.

10 rticle no. 812 C. Tuca, R. Deju,. Zorliu 10 Fig. 4 Spreading of hydrogel coating on the contaminated surface. Fig. 5 Removing of the gel film from the floor surface. v V 10 4, (7) where: V (m 3 ) the hot cell volume (4.8 m 3 ).

11 11 Radiological risk assessment for hot cell decontamination rticle no. 812 The inhaled activity by the worker within an hour of work, without wearing of the mask, is obtained according to eq. 8. where: nomask v, (8) v inhaled v (Bq m -3 ) activity concentration in the air; v inhaled (m 3 /h) 1.2 m 3 per hour. The activity inhaled with a mask wearing is obtained by multiplying of the activity without mask with the quantum of activity none retained in the filter, according to eq (9) mask nomask Then the the internal committed effective dose, E(50), received by worker due to air inhalation is calculated according to the eq. 10. E( 50) e inh mask, (10) where: e inh the effective dose coefficient for inhalation (SvBq -1 ) [11]. 4. RESULTS ND DISCUSSIONS 4.1. THE MBIENT DOSE EQUIVLENT H*(10) MESUREMENTS The ambient dose equivalent H*(10) values obtained for the floor surface contamination scanning with the portable digital survey meter and thermoluminiscent measurements are presented in Table 2. The measurements show that there are seven hot spots distributed on a surface at about 2000 cm 2. For six hot spots the dose rate mean value is 15.6 msv/h. For the hottest point 7 situated at about 70 cm hot cell entrance (worker position) is 400 msv/h, at almost 27 times greater than the mean value. The thermolumiscent dosimeters measurements reveal a similar distribution for dose rates values. The dose rate mean value is 28.6 msv/h for six of them and 782 msv/h, also almost 27 times greater than the mean value, for the point 7. Based on the previously results we can conclude that the most important risk for the worker who performs the scanning of the floor contamination comes from the 7 hot spot. By comparing the direct measurements results (Table 2) with those calculated (Table 3) one can notice that for those six hot spots identified by scanning, the probe was placed at about 0.9 cm above and for the hottest 7 was at about 6.5 cm. Regarding the thermosluminescent dosimeters position, six of them were placed at about 0.7 cm toward hot spot and the seventh one at 4.6 cm. lthough the measurement conditions were difficult we can conclude that the dose rates values obtained by calculation are in a satisfactory agreement with the measured ones, and validate the measurements.

12 rticle no. 812 C. Tuca, R. Deju,. Zorliu 12 Hot spots Table 2 The ambient dose equivalent values in the Hot Cell no. 4 H*(10) scanning [msv/h] H*(10) TLD [msv/h] Probe position above the hot spot [cm] TLD position toward the hot spot [cm] ± ± ± ± ± ± ± ± ± ± ± ± ± ± GMM-RY SPECTROMETRY MESUREMENTS Table 3 presents the activity of the samples taken from each hot spot determined by gamma ray spectrometry measurements. The measurement uncertainties are between % ranges. Table 3 The activity of the samples from the hot cell no. 4 Hot spots Nuclides ctivity [Bq] Standard uncertainty (1σ) [%] 60 Co 9.31E Cs 3.22E Cs 4.67E Co 2.05E Cs 3.57E Cs 4.68E m g 3.96E Co 3.17E Cs 2.91E Cs 2.84E Co 3.16E Cs 6.54E Cs 4.55E Co 1.58E Cs 2.46E Cs 4.53E Co 2.38E Cs 1.73E Cs 4.72E Co 4.57E

13 13 Radiological risk assessment for hot cell decontamination rticle no. 812 The gamma ray spectrometry measurements of the samples taken from the seven hot spots show that activity of 60 Co for 7 (4.57E + 08 Bq) hot spot is 3 times order of magnitude higher than the mean value of the activity for the six hot points (3.60E + 05Bq) and confirm the results of the ambient dose rate direct measurements. lso it is noticed the presence of the fission products such as 134 Cs and 137 Cs and of the activation product 108m g, but with significant lower activities than the key activation product 60 Co, while in the point 7 it seems to be a strong 60 Co contamination THE WORKERS RISK SSESSMENT Two phases of the decontamination process for workers risk assessment were considered. In the first phase the worker, placed at the hot cell entrance, at about 70 cm from the 7 hottest point is externally irradiated for five minutes while performing the dose ambient equivalent measurements. The penetrant dose rate values calculated according to the (5 6) equations are presented in Table 4 by considering the assumption that the sampling yield for the activity measurement is 10% and that the entire activity is concentrated in the hottest point. Hot spot [Bq] Table 4 Dose rates for hot cell no. 4 contamination scanning x [cm] R [cm] r [cm] Ḋ a [mgy/h]. H p(10) [msv/h] E E E E E E E E E E E E E E E E E E E E E+00 Total 3.39E+00 Thus, the penetrant dose rate for the worker that scan the floor contamination for hot cell no. 4 is 3.39 msv/h which means 0.28 msv, for five minutes of working time. It can be notice that the risk for this worker is quiet high due to the fact that the dose limit for a year is 20 msv according to Romanian legislation [9] taking considering 2000 working hours per year (10 µsv/hour). That means that this worker should be working less than 5.9 hours/year. In the second phase the worker B, placed at about 45 cm from the hottest point 7 is externally and internally irradiated for twelve minutes while performing the floor decontamination. In the Table 5 are presented the penetrant dose rate values for the worker B, who performs floor decontamination thus the

14 rticle no. 812 C. Tuca, R. Deju,. Zorliu 14 total dose rate is 7.97 msv/h, at about 1.59 msv for 12 minutes work time. The worker risk is very high, according to the dose limit 20 msv/year and 2000 labor hours/year (10 µsv/hour). In this case the worker should be working less than 2.5 hours/year in the radiation field. Hot spot [Bq] Table 5 Dose rates for hot cell no. 4 decontamination x [cm] R [cm] r [cm] Ḋ a [mgy/h]. H p(10) [msv/h] E E E E E E E E E E E E E E E E E E E E E+00 Total 7.97E+00 Table 6 Committed effective dose for hot cell no. 4 decontamination Hot spot Total Radionuclide [Bq] e inh [SvBq -1 ] v [Bq/m 3 ] mask [Bq] E (50) [msv] 60 Co 9.31E E E E E Cs 3.22E E E E E Cs 4.67E E E E E Co 2.05E E E E E Cs 3.57E E E E E Cs 4.68E E E E E m g 3.96E E E E E Co 3.17E E E E E Cs 2.91E E E E E Cs 2.84E E E E E Co 3.16E E E E E Cs 6.54E E E E E Cs 4.55E E E E E Co 1.58E E E E E Cs 2.46E E E E E Cs 4.53E E E E E Co 2.38E E E E E Cs 1.73E E E E E Cs 4.72E E E E E Co 4.57E E E E E E-03

15 15 Radiological risk assessment for hot cell decontamination rticle no. 812 The internal committed effective dose E(50) for decontamination process is calculated considering that inside of the hot cell is spread just 10-4 of the total activity and filter has the retention efficiency of 99%. The committed effective dose values for the internal irradiation are presented in Table 6. The main contribution for the internal irradiation of the worker B is from the 7 the hottest point about 1.62 µsv and the highest values are from key radionuclide for activation products 60 Co. This value is low enough thus that it could be stated that the internal irradiation of the worker is reduced due to the high efficiency of the filter. 5. CONCLUSIONS working methodology for contamination scanning and for radioactive inventory identification is developed in order to assess the radiological risks for the workers involved in decontamination process of a hot cell which is in the decommissioning process. The results of penetrant dose rate obtained by calculation reveal that the risk is moderate (3.39 msv/h, 0.28 msv for working time 5 minutes) for worker who scans surface contamination and for the worker B that performs floor decontamination, the risk is quite high (7.97 msv/h, 1.59 msv for working time 12 minutes). The risk of internal contamination is insignificant compared with that for the external irradiation. The results of penetrant dose rate obtained by calculation are in good agreement with the direct measurements of the ambient dose rate equivalent. Thus, we can conclude that the method used for radiological risks assessment of the workers is adequate and could be used for other similar situations, when the mechanic devices are not available and the decontamination process is performed manually. cknowledgments. The authors offer many thanks to Dr. Maria Sahagia for her suggestions and feedback when clarifications of issues were required and to Dr. Daniela Gurau for her support for samples measurements. REFERENCES 1. V. Popa, M. Dragusin, C. Petran, R. Deju, Decommissioning Plan of VVR-S Nuclear Research Reactor, Revision 12 Mai 2014, E. Ionescu, Direct measurement of the surface contamination, control procedure PC-DEZ-401, Stochioiu, The monitoring of the environmental radioactivity using thermo-luminescent dosimetry system, operational procedure, PL-UMRM-01, 2015.

16 rticle no. 812 C. Tuca, R. Deju,. Zorliu D. Gurau, Gamma-ray spectrometry analysis with HPGe detector, model GEM60P4-95, control procedure, PC-DEZ-408, Stochioiu, M. Sahagia, I. Tudor, TLD System for the monitoring of the environmental radioactivity, Rom. J. Phys. 54, , D. Radu, D. Stanga, O. Sima, method of efficiency calibration for disk sources in gamma-ray spectrometry, Rom. Rep. Phys 61, 2, , M. Sahagia,.C. Razdolescu,. Luca, R. Macrin, Volume standard sources in soil matrix, Rom. J. Phys. 42, 9 10, , Luca, B. Neacsu,. ntohe, M. Sahagia, Calibration of the high and low resolution Gammaray spectrometers, Rom. Rep. Phys. 64, 4, , J.E. Martin, Physics for Radiation Protection, Handbook, second Edition, Completely Revised and Enlarged, Ed. Wiley-WCH Verlag GmbH & Co. KGa, ISBN , ICRP 2010, Conversion Coefficients for Radiological Protection Quantities for External Radiation Exposures. ICRP Publication 116 (Elsevier: ICRP) nn. ICRP 40(2 5). 11. CNCN, Fundamental Norms for Radiological Safety, NSR-01, pproved through the 14/ CNCN President Order, published in the 404/ Official Bulletin (part I).