Measurement of activity of the predominant gamma-emitting radionuclides in activated components of a medical cyclotron plant

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1 Measurement of activity of the predominant gamma-emitting radionuclides in activated components of a medical cyclotron plant Pietro Guarino, Salvatore Rizzo, Elio Tomarchio * University of Palermo, Nuclear Engineering Department, Viale delle Scienze, Parco d Orleans, Ed. 6, I-90128, Palermo, Italy. Abstract. An identification and activity evaluation of the predominant gamma-emitting radionuclides in activated parts of a medical cyclotron plant by high resolution gamma-ray spectrometry were performed. The use of LaBr 3 :Ce scintillation and HPGe detectors was examined in order to solve the measurement problems due to high activity of some components used inside a plant provided with an IBA CYCLONE 18/9 cyclotron. For each component, the identification of some key-radionuclides allows to evaluate the surface dose behaviour as function of decay time. The largest activity values are related to radionuclides with half-life of days, while radionuclides with higher half-lives represent less than 10 per cent of the total activity. The surface dose rate will be reduced to about 1/1000 of the starting value after a decay period of approximately 3 years, with a relatively safety at product disposal work. KEYWORDS: PET; Cyclotron; induced activity; gamma-ray spectrometry. 1. Introduction The increasing importance of Positron Emission Tomography (PET) as a diagnostic tool has led to a rapid growth on dedicated site number worldwide. To conduct a PET screening, a positron-emitting radioactive substance needs to administrate to a patient. The preparation of a radiopharmaceutical dose is carried out by transferring and processing into suitable synthesis modules a radioactive product obtained by proton irradiation of a target in a cyclotron. The interactions of proton beam or secondary neutrons produced in the target induce high activity in the cyclotron components, especially into the target body parts. The specific activity varies depending on the accelerated beam, on the location of the material and on the decay time already elapsed. The most active are the target body components. From a radiation protection point of view, the high activity induced in some components is a problem because operators are exposed to high dose levels during maintenance. In fact, most of the components and target elements are periodically replaced and generally stored in a Pb-shielded container in order to wait for their radioactive decay before the disposal. The evaluation of the activities of the main radionuclides in these components is an important tool to evaluate either operator dose susceptibility or radionuclide inventory at waste packaging time. The latter item is an important legal requirement to perform the removal of radioactive waste components by an authorized company, because some national legislations no provide exemption activity limits for their disposal. Studies on radioactivity induced in a target after an irradiation cycle were already presented in [1-4]. In this work we perform a nuclide identification and activity evaluation of some activated cyclotron components by high resolution gamma-ray spectrometry with High-Purity Germanium (HPGe) detectors and a LaBr 3 :Ce scintillator. None handling or sample preparation was carried out unlike what already operated in [2]. Indeed, most of the components can be measured at high distance from the HPGe detector while to measure target parts whose activity is very high, i.e. fresh-irradiated target windows, suitable attenuators or a lead-walled collimator, as that presented elsewhere[5], have been realized. Such systems avoid problems related to high measurement dead-time and reduce maintenance service operators exposure, too. Measurements on some components used inside an IBA CYCLONE 18/9 cyclotron, allow the identification and activity evaluation for the most important radionuclides produced by activation of the materials. For each waste component, radionuclides with higher activity and main responsible of the surface dose are listed as key-radionuclides in agreement with what specified in [6]. The measurement of their activities permit to determine the surface dose through simple formulations. * Corresponding author, tomarchio@din.din.unipa.it 1

2 2. Materials and Methods All the components considered in this work have been irradiated at the Nuclear Medicine Center S. Gaetano in Bagheria (a town near Palermo, Italy). The IBA CYCLONE 18/9 cyclotron used in the S. Gaetano Center was installed in May 2002 and started its 18 F-fludeoxyglucose ([ 18 F]FDG) production in January The CYCLONE 18/9 is capable of accelerating negative hydrogen ion (H ) or deuteron (D ) to energies of 18 MeV and 9 MeV, respectively. The cyclotron has been principally used for the production of [ 18 F]FDG via 18 O(p,n) 18 F reaction into a niobium target cell filled with about 0.5 ml (small volume target) or about 2.2 ml (large volume target) of [ 18 O]enriched water. During the last period was preferred the use of a large volume target, so as to have 18 F high activity both for diagnostic and commercial use. The target was typically irradiated for min per day with 18 MeV protons and an average current of 35 µa. The main target components are generally replaced, except for failures or anomalies, after about 3-4 months of use (integrated irradiation time about 150 h). During periodically maintenance operations some damaged parts were replaced. To this goal, after a reasonable waiting period since last irradiation (at least 24 h), the operators enter the cyclotron vault room and proceed to the extraction of the target body and/or other parts. Due the high induced activities, the target body and the other parts are readily transferred to a manipulation shielded (10 cm Pb) bench, where all replacement operations are carried out. A measurement of dose rate value near the large volume body target inside the cyclotron results about 200 µsv h -1 after a 3-day decay period. The replaced parts are stored in a 5-cm thick lead-walled container, placed within a controlled area because the emerging dose level is about 100 times higher than the ambient background dose rate. Some of the components replaced and stored in the container were taken into consideration. Havar foils, carousel, stripper double forks, carbon foils and titanium vacuum window of target irradiated during a normal production cycle were analized by high resolution gamma-ray spectrometry. Most of the spectrometric measurements were performed using an HPGe detector manufactured by ORTEC model GEM18180 (relative efficiency 18%, energy resolution 1.8 kev at 1332 kev) connected to an ORTEC mod. 672 Amplifier and an ORTEC 919E MultiChannel Buffer into an Ethernet environment. Data analysis was performed by ORTEC Gamma-Vision Software (version 6.06)[7]. High specific activity Havar foils (50 µm thick, 30 mm in diameter) or titanium window (12.5 µm thick, 24 mm dia.) give origin to a high measurement dead-time even when large distances from the detector were used. Low dead-time and acceptable count-rate must be achieved in order to improve the spectrometric measurement performances. The increase of source-detector distance can be useful to reduce measurement dead-time for some activated parts but it is not enough for the most active shares. Other methods may regard the adoption of a special attenuators, or, as adopted in a previous work [5], a Pb-walled collimator device. The suitability of the methods can be assessed by measuring a source (for example a titanium window) with and without the absorber or the Pb-walled collimator. The ratio of the measurements made it possible to infer a ratio value behaviour to be used for activity calculation. However, a collimator should be adopted only when necessary because the results can be influenced by high uncertainties since calibration sources with enough high activity to be used with a Pb-device mitigation are not always available. Generally large source-detector distances, as 25, 80 and 100 cm were adopted to consider samples like a point source and neglect coincidence-summing effects. For a few components, tiny and relatively little active, smaller distances 2.5, 6 or 10 cm were used, too. Energy and efficiency calibrations were carried-out by measuring standard sources of 57 Co, 139 Ce, 113 Sn, 137 Cs, 54 Mn, 65 Zn (single line sources) and 152 Eu (multi-line source). The first sources, furnished by CEA-LMRI, were employed for close geometries (source-detector distance less than 10 cm), whereas 152 Eu source (furnished by Amersham) was used in far geometries (25 cm and more) and in attenuated (with a Pb thickness ) or collimated geometries. Energy and gamma-ray emission probability values were taken from [8], an easily usable web database, and from [9]. Fig. 1 reports the experimental efficiency values for the main gamma emissions of 152 Eu related to most used measurements geometries besides the log-log 4 th order polynomial fits. If calibrated sources are not available, a possible but less precise alternative calibration procedure involves the use of 22 Na, 137 Cs and 57 Co sources normally available in a PET plant and used for tomograph and dose measurement device calibrations, respectively. 2

3 Figure 1: Efficiency behaviour for various measurement geometries and log-log 4 th order polynomial fits. Calibration point source : 152 Eu. Efficiency (counts per gamma) cm 25 cm 80 cm Pb-collimator We want to point out that the accuracy of the measurements is high even when large distances are used. In fact, because of high activity, photopeak areas higher than counts can be obtained also with short counting times. The accuracy of activity determinations is on average less than 0.5%, with slightly higher values for nuclides characterized by lower specific activity. Some measurements were performed using a 2 2 in. lanthanum bromide (LaBr 3 :Ce) scintillation detector directly coupled to a 2-in.-diameter photomultiplier tube (Scintibloc), obtained recently by Saint-Gobain Crystals (BrilLanCe 380 ) [10]. The physical characteristics of this detector (light yield of 61,000 photons/mev, energy resolution FWHM 2.8% at 662 kev, density 5.3 g/cm 3, decay time 35 ns) make it very attractive for many application of gamma-ray spectrometry with high counting rate, particularly when the use of an HPGe detector is not easy [11-13]. A negative aspect is represented by a significant intrinsic internal contamination due to natural radioisotopes such as 138 La and 227 Ac. In particular, 138 La produces two gamma rays at 789 kev (β -decay to 138 Ce) and 1436 kev (EC-decay to 138 Ba), which features are evident in a background gamma spectrum, togheter with a significant X-ray peak at kev. For these reasons it is appropriate to limit the investigation to a kev energy range, satisfactory for the aims of this work. In any case, it shows better performances with respect to a Tennelec NaI(Tl) 3 3 in. detector, shielded with Pb bricks, employed in preliminary determinations. Both the detectors were connected to an Ortec NOMAD Plus Portable Multichannel System and the HV supply is furnished by an Ortec 276 PM Base with preamplifier. Data analysis was done using ORTEC Maestro -32 software [14]. With the LaBr 3 :Ce detector and the use of an ORTEC DigiBase connector coupled to a portable computer was also realized a portable spectrometric system that can be employed in situ measurements of the components, without other electronic devices and shielding. 3. Results and discussion Several gamma-ray spectrometric measurements were performed on each component stored in the Pbwalled container for a time period higher than 1 year (about 500 days) in order to identify long-lived radionuclides. Examples of gamma-ray spectra are reported in Figs. 2 through 6. Data analysis allows the identification of the key-radionuclides, to be considered as references for further evaluations and significant contributors to the total activity of the item [6]. For some components, in Table 1 are reported the main identified long-lived radionuclides besides the evaluation of the activities at the time of the storage into the container. We note that the total activity 3

4 of waste stored in the container is nearly entirely related to Havar foils, whereas other pieces were less important in activity. With regard to the stripper, the activity contribution derived by measuring the fixing screw and the one due to fork seal of carbon sheets were highlighted. This demonstrates the importance of the choice of materials as the total activity of a stripper holder can be attributed only to the fixing screw. Measurements carried out on sheets of carbon, placed within polyethylene bags, did not reveal significant presence of induced radioactivity. For a unidentified nuclide, the assessment of minimum detectable activity (MDA) in each measurement allows to evaluate an initial activity of some orders of magnitude lower than the one of identified nuclides and, anyhow, related to a no significant radiological risk. Table 1: Evaluation of activity values of some long-lived radionuclides in various components at the storage time into the lead-walled waste container. Isotope Halflife Activity (kbq) Havar Titanium Stripper Carousel (days) foil window Screw fork 22 Na < MDA < MDA < MDA < MDA Sc < MDA 7.0E+2 < MDA < MDA < MDA 48 V E+4 < MDA < MDA < MDA 51 Cr E+5 < MDA 1.1 < MDA Mn E Co E E E+2 57 Co E Co E+5 < MDA 1.3 < MDA Zn < MDA E+2 (MDA = Minimun detectable activity for the pertinent measurement geometry) The analysis of data of Table 1 helps to identify the key-nuclides for each component, as 54 Mn, 56 Co, 57 Co, 58 Co and 51 Cr for the Havar foil, 46 Sc, 48 V, 54 Mn, 56 Co and 57 Co for the titanium window. Not always a radioisotope with greater activity is a key-nuclide, since the long half-life can be more significant especially in determining long term dose behaviour. Moreover, the largest contribution to activity is related to radionuclides with half-life of about days. This means that after approximately 2 years (about 10 times the half-life), the activity of these radionuclides will be reduced about 1000 times. Figure 2 shows the measured gamma-ray spectrum detected on a Havar foil extracted after a 3-months cycle of irradiation with a large volume target and stored for about 400 days. After this decay time, the requisite of an acceptable dead-time value (about 3%) can be fulfilled with a source-detector distance of 80 cm. In Fig. 3 is shown the gamma-ray spectrum detected on the same Havar foil after a decay period of 500 days with the BrilLanCe-380 detector. The fast shaping property of the detector allows the realization of a low measurement dead-time without any shielding device, besides to the high source-detector distance of 100 cm. In the latter spectrum, the separation of the 122 and 136 kev photopeaks of 57 Co is noteworthy. There are also evident some peaks related to 56 Co while it needs a multiplet deconvolution to determine the 54 Mn and 58 Co contributions to the broad peak at kev. However, the evaluation of the activities carried out by a calibration with a 152 Eu source, led to values at the time of extraction of about 29 MBq for 56 Co and 20 MBq for 57 Co, close to the ones of Tab. 1 despite the higher uncertainty in the area determination. This allows us to say that the use of a LaBr 3 :Ce detector permit the evaluation of the key-radionuclide activities with sufficient accuracy. Figs. 4 and 5 show gamma-ray spectra detected on a carousel stripper device, with HPGe and LaBr 3 :Ce detectors after 20 and about 100 days decay period, respectively. A distance greater than 20 cm from the detector was adopted to assimilate the carousel to a point source (such as calibration one). The high-energy gamma absorption in the sample thickness can be considered no significant. For lower energies a self-absorption evaluation was performed by measuring suitable calibrated sources above and below the sample. For other pieces, as the thickness is very small, does not take into account any absorption correction. 4

5 Figure 2: Gamma-ray spectrum detected on an Havar foil. Decay time: 400 days, S.E.= single-escape peak; D.E.= double-escape peak Counts per channel X-rays Re Co Co Mn Co ( 56 Mn) Co Co Co Co Co K 0.35 kev/channel Co (D.E.) Co Co Co ( 56 Mn) detector : HPGe Ortec GEM18180 geometry : 80 cm counting live time: 30,000 s dead-time: 3.0% Channel Figure 3: Gamma-ray spectra detected on a Havar foil with a LaBr 3 :Ce scintillator. Decay time : 510 days. Counts per channel detector: BrilLanCe-380 geometry : 100 cm counting live time: 5,200 s dead-time: 2.6% Co Mn Co 0.76 kev/channel Channel 5

6 Figure 4: Gamma-ray spectrum detected on a carousel stripper. Decay time: 20 days. Counts per channel X-rays Cr Bi detector : HPGe Ortec GEM18180 geometry : 25 cm counting live time: 266,000 s dead-time: 2.9% Co Mn Channel Co ( 56 Mn) Bi Co Bi 0.35 kev/channel Zn Na Co Co Figure 5: Gamma-ray spectrum detected on a carousel stripper. Decay time: 100 days. Counts per channel detector: BrilLanCe-380 geometry : 20 cm counting livetime: 14,300 s dead-time: 2.7% Channel Co Mn Co Co 0.76 kev/channel Zn Na Co Confirming what highlighted in Table 1, Fig. 6 reports a comparison between the gamma-ray measurements detected on the fixing screw and one of two forks of the same stripper holder, with differences in countings of more than 10 times. A further investigation has regarded radionuclide half-life values. In Figs. 7 and 8 are reported trends of activity as a function of time for the Havar foil and titanium window key-radionuclides. A linear fit allows to determine the apparent half-life and the activity value at the time (t=0) of extraction and storage into the container. While the value of 71.3 days for the 58 Co has an error of only 0.5 per cent compared to the literature value of 70.9 days, small differences are obtained for the other isotopes. The higher error was registered for 57 Co, with a calculate value of 290 days against a value of 272 days. The difference of 6.6 per cent between the two values is principally related to the uncertainties of 122- kev photopeak area determination in the measurements performed with the Pb-collimator. 6

7 Figure 6: Gamma-ray spectra detected on a screw and a stripper fork used inside an extractor carousel. Decay time : 440 days, c.p.s.= counts per second, D.E.= double-escape peak. c.p.s. per channel c.p.s. per channel X-rays X-rays 0.35 kev/channel detector: HPGe Ortec GEM18180 geometry : 10 cm counting live time: 176,000 s dead-time: 0.2% stripper screw stripper fork 0.35 kev/channel Co Co (D.E.) Co Co Mn detector : HPGe Ortec GEM18180 geometry : 10 cm counting live time: 436,000 s dead-time: 0.03% Co Mn Co ( 56 Mn) Co ( 56 Mn) Co Channel Zn Co (S.E.) Zn Na Co Co Co K Co (S.E.) Co (D.E.) Co (D.E.) Co Co ( 56 Mn) Co Co ( 56 Mn) Other long-lived radionuclides can be recognized in the spectra detected on Havar foils, stripper forks and carousels. As pointed in Figs.2,4,6, little quantities of 60 Co, 65 Zn, 207 Bi and 183 Re and others were detected. Nevertheless, their contribution to the total activity evaluation is very small and cannot be considered among the key-radionuclides without significant errors. The behaviour of surface dose rate over time, in order to assess a dose rate value either at the extraction time (t=0) or at the time of packaging and disposal of waste, was evaluated. If we consider a minimum distance of 30 cm (about 10 times the Havar foil diameter), the samples can be considered as a point-like sources and dose rate computed by the known relation dd/dt (r) = n Γi Ai ( t) r 2 (1) i= 1 where n is the key-nuclide number, Γ i and A i (t) are the Specific Gamma Constant in air (msv h -1 per MBq at 1 m) and the activity (MBq) after a decay time t of the i-th key-radionuclide whereas r is the source distance (m). The Eq. (1) requires the fulfillment of charged particle equilibrium conditions, easily performed at the large distances adopted for the dose measurements. Assuming the values of Γ i reported in [15] (available also in [16]), we obtain the dd/dt trends reported in Fig. 9 for Havar foil and titanium window. The dose rate values that can be deduced from the decay curve match with experimental data (at 1 m) obtained with a Berthold TOL/F LB1321 ionization chamber. In fact, at a distance of 1 m from an Havar foil, dose rate was about 60 µsv h -1 after a decay period of about 25 days from the extraction. 7

8 Figure 7: Key-nuclide activity behaviour in a typical irradiated Havar foil. Activity (Bq) Havar foil Total 58 Co 57 Co 56 Co 54 Mn 51 Cr Decay time (days) Figure 8: Key-nuclide activity behaviour in a titanium window of a target Titanium window Total 48 V 46 Sc 56 Co 57 Co 54 Mn Activity (Bq) Decay time (days) Both the dose rate behaviours seem to vary almost exponentially in the first period (t < 200 days) with apparent half-life of 78 days and 16 days for Havar foil and titanium window, respectively. The effective half-life increases over time to 208 days after a period of about 800 days for Havar foil whereas to about 90 days for titanium window after about 500 days. As highlighted in Fig. 9, after a decay period of approximately 1200 days the Havar foil surface dose will be reduced to about 1/1000 of the starting value, with a relatively safety at product disposal work. 8

9 Figure 9: Surface dose rate behaviour at 1 m distance from an Havar foil and a titanium window of a target Dose rate (msv h -1 at 1 m ) Titanium window Havar foil Time (days) 4. Conclusion Gamma-ray spectrometric measurement of activated target components is an useful technique to evaluate the nuclide activities in order to fulfil radiation protection requirements. The use of HPGe detectors is optimal, although it needs a specially dedicated room and regular supply of liquid nitrogen for its operation. Alternatively, it seems promising the use of a LaBr 3 :Ce scintillator for its physical characteristics but also for its simplicity of use and the possibility to realize a portable spectrometric system. Both systems allow to verify residual activity levels of cyclotron and shielding components. Since the most important contribution to waste activity is related to radionuclides with relatively short half-life, we can confirm that after a decay time period of about 2-3 years the waste components represent a rather low radiation hazard and may make their disposal with safety. In order to fulfil the last requirement, at least two Pb-walled containers seem to be required. In the first the most recent replaced parts must be stored in order to reduce the total activity level whereas in the second the oldest components are kept. To reduce emerging dose to no significant levels, is appropriate to adopt a shielding lead wall at least 10 cm thick. The determination of activities makes it possible to assess the surface dose of the various components at different decay times. The knowledge of the surface dose behaviour is an important information for waste handling with a safety at work. Acknowledgements Work supported by Italian Ministero Ricerca Scientifica. The authors wish to thank the Nuclear Medicine Center S. Gaetano in Bagheria and particularly Mr. D. Greco for their cooperation in carrying out this work. REFERENCES [1] NCRP REPORT No. 144, Radiation Protection for particle accelerator facilities, Bethesda, Maryland, [2] O DONNEL, R.G., et al., Measurement of the residual radioactivity induced in the front foil of a target assembly in a modern cyclotron, Appl. Radiat. Isot. 60 (2004)

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