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Gadolinium-153 as a brachytherapy isotope
This content has been downloaded from IOPscience. Please scroll down to see the full text. 2013 Phys. Med. Biol. 58 957 (http://iopscience.iop.org/0031-9155/58/4/957) View the table of contents for this issue, or go to the journal homepage for more
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IOP PUBLISHING
PHYSICS IN MEDICINE AND BIOLOGY
Phys. Med. Biol. 58 (2013) 957–964
doi:10.1088/0031-9155/58/4/957
Gadolinium-153 as a brachytherapy isotope Shirin A Enger 1 , Darrell R Fisher 2 and Ryan T Flynn 3 1 Section of Oncology, Department of Radiology, Oncology and Radiation Science, Uppsala University, Uppsala, Sweden 2 Isotope Sciences Program, Pacific Northwest National Laboratory, 902 Battelle Blvd, Richland, WA 99352, USA 3 Department of Radiation Oncology, University of Iowa, 200 Hawkins Drive, Iowa City, IA 52242, USA
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Received 8 September 2012, in final form 5 December 2012 Published 23 January 2013 Online at stacks.iop.org/PMB/58/957 Abstract The purpose of this work was to present the fundamental dosimetric characteristics of a hypothetical 153Gd brachytherapy source using the AAPM TG-43U1 dose-calculation formalism. Gadolinium-153 is an intermediateenergy isotope that emits 40–100 keV photons with a half-life of 242 days. The rationale for considering 153Gd as a brachytherapy source is for its potential of patient specific shielding and to enable reduced personnel shielding requirements relative to 192Ir, and as an isotope for interstitial rotating shield brachytherapy (I-RSBT). A hypothetical 153Gd brachytherapy source with an active core of 0.84 mm diameter, 10 mm length and specific activity of 5.55 TBq of 153Gd per gram of Gd was simulated with Geant4. The encapsulation material was stainless steel with a thickness of 0.08 mm. The radial dose function, anisotropy function and photon spectrum in water were calculated for the 153Gd source. The simulated 153Gd source had an activity of 242 GBq and a dose rate in water 1 cm off axis of 13.12 Gy h−1, indicating that it would be suitable as a low-dose-rate or pulsed-dose-rate brachytherapy source. The beta particles emitted have low enough energies to be absorbed in the source encapsulation. Gadolinium-153 has an increasing radial dose function due to multiple scatter of low-energy photons. Scattered photon dose takes over with distance from the source and contributes to the majority of the absorbed dose. The anisotropy function of the 153Gd source decreases at low polar angles, as a result of the long active core. The source is less anisotropic at polar angles away from the longitudinal axes. The anisotropy function increases with increasing distance. The 153Gd source considered would be suitable as an intermediate-energy lowdose-rate or pulsed-dose-rate brachytherapy source. The source could provide a means for I-RSBT delivery and enable brachytherapy treatments with patient specific shielding and reduced personnel shielding requirements relative to 192 Ir. (Some figures may appear in colour only in the online journal) 0031-9155/13/040957+08$33.00
© 2013 Institute of Physics and Engineering in Medicine
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1. Introduction 153
Gd is currently used in line sources and calibration phantoms for quality assurance applications, specifically for nuclear medicine imaging systems (Frey and Tsui 1995, DiFilippo 2008). The 40 to 103 keV photons with a mean energy of 60.9 keV emitted by the 153Gd radioisotope make it a potential intermediate-energy (50 keV < E < 150 keV) (Beaulieu et al 2012) brachytherapy source. Multiple opportunities justify the evaluation of 153Gd for brachytherapy. The widelyused (Nath et al 1995, Rivard et al 2004) high-energy (E > 200 keV) (Beaulieu et al 2012) 192 Ir source has substantially higher shielding requirements than 153Gd, as its gamma ray emissions average 360 keV and range in energy from 201 to 884 keV. The typical shielding wall thickness for an HDR 192Ir brachytherapy treatment room is 54 cm of concrete, 40 cm of barite concrete or 5 cm of lead (Papagiannis et al 2008), assuming a desired transmission of 2.7 × 10−4 for an uncontrolled area. (McGinley 2010). The mean linear attenuation coefficient of 153Gd emissions in lead is 79.273 cm−1 (Unger and Trubey 1982); therefore the HDR lead shielding requirement is only 1 mm. Thus with adequate patient shielding, 153Gd brachytherapy could enable the treatment of patients without substantial radiation exposure to medical staff remaining inside the brachytherapy treatment room. Gadolinium-153 would also enable patient specific shielding due to less shielding requirement from the emitted low-energy photons. Perhaps the most promising opportunity associated with 153Gd is its potential for use as an isotope for interstitial rotating shield brachytherapy (I-RSBT). Since the tenth-value layer of 153Gd is 0.37 mm of platinum, a 153Gd brachytherapy source could be used in combination with an interstitial shielding system to deliver I-RSBT. I-RSBT was previously conceptualized by Ebert (2006), but practical sources were not defined, and Lin et al (2008) later proposed such a system for breast cancer brachytherapy based on 125I. 153Gd may be superior for I-RSBT compared to low-energy (E < 50 keV) (Beaulieu et al 2012) isotopes such as 125I, 103Pd and 131 Cs, due to its longer half-life and more ideal photon emission energies. The low-energy isotopes have radial dose functions that decrease rapidly as the inverse of distance, whereas 153 Gd, with its intermediate energy, would be expected to have a radial dose function similar to that of 192Ir, which is near unity. Intermediate-energy isotopes for brachytherapy have been previously studied, including 169 Yb (Granero et al 2005, Mainegra et al 1998), 170Tm (Granero et al 2005, Munro et al 2008, Ballester et al 2010, Enger et al 2011) and 57Co (Enger et al 2012). AAPM TG-43U1 brachytherapy dosimetry parameters for each hypothetical source have been published. The isotopes considered have high enough energies for photoelectric interactions in soft tissue to be minimal and have similar radial dose functions as 192Ir, yet the emitted energies are low enough that the shielding thickness requirements are substantially lower than that of 192Ir. However, each isotope has a disadvantage relative to 192Ir. 169Yb has a short half-life of 32 days. 170Tm has non-negligible bremsstrahlung and beta contamination at depths less than 5 mm (Enger et al 2011), and a gamma ray dose constant, hereafter referred to as a photon dose constant (since it includes x-rays and gamma rays), about 95 times lower than that of 192Ir (1.674 versus 160 cGy cm2 GBq−1 h−1) (Unger and Trubey 1982), requiring a 170Tm source to be quite large relative to an 192Ir source if the dose rate of a fresh 370 GBq 192Ir source is to be maintained. Moreover, the bremsstrahlung photons emanating from the core and encapsulation are of high enough energy to make the feasibility of partial shielding for I-RSBT questionable. 57Co can be purchased in relatively low quantities (RITVERC Isotope Solutions, St Petersburg, Russia) with an excellent specific activity of 222 TBq g–1 or higher, and has a favorable half-life of 271 days, and a photon dose constant in air of 40.87 cGy cm2 GBq−1 h−1 (Unger and Trubey
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1982). 57Co production in sufficient quantities for brachytherapy is prohibitively expensive, as the current technique of irradiating enriched 57Ni targets with a 1–2 mA cyclotron would require many thousands of hours to generate a 57Co source with a dose rate at 1 cm equivalent to 370 GBq of 192Ir (Case et al 1969, Enger et al 2012). In the late 1960s, Oak Ridge National Laboratory (ORNL) published a study describing the advantages of 153Gd for the generation of a backscatter signal from photons of around 100 keV in an atmospheric density study to be conducted with a Mars probe (Case et al 1969). The 57Co isotope was also considered, along with 19 other candidate isotopes, and 153Gd was selected due to its energy characteristics, output, half-life of 242 days, and lower cost per Curie than 57Co (Case et al 1969). It is possible that 153Gd may be a more favorable brachytherapy source than 57Co for similar reasons. The disadvantage of 153 Gd relative to 57Co is that the achievable specific activity of 153Gd is lower than 57Co due to the realities inherent in reactor production of 153Gd and subsequent radiochemical processing. 153Gd is produced by thermal neutron irradiation of enriched 152Gd or 151Eu (in natural europium oxide), with the latter production method being the most commonly used. Irradiating 151Eu with neutrons produces 152Eu, some of which undergoes β-decay to 152Gd, which can then absorb a neutron to produce 153Gd. Since much of the 152Eu will absorb another neutron prior to undergoing β-decay to 152Gd, and some 152Eu will undergo β-decay to 152Sm, the irradiated Eu sample will contain multiple isotopes of Eu and Sm. The irradiation of natural europium targets produces 152Eu, 154Eu and 155Eu among other lesser contaminants. These contaminants can be removed using electroreduction techniques (Ramey 1988), reduction with zinc metal (Bray and Corneillie 2001) and reduction with zinc followed by ion exchange purification of 153Gd. Since the 153Gd generated has a much higher thermal neutron absorption cross section than 152Gd, and since 154Gd is produced by decay of 154 Eu, the separated 153Gd will only have a small fraction of the maximum theoretical specific activity of 153Gd. Production of high-specific activity 153Gd can be optimized by increasing irradiation times to about 240 effective full-power days, and by separating out the 154Eu soon after irradiation. The highest claimed specific activity in the literature is 5.55 TBq of 153Gd per gram of total Gd (Karelin et al 2000). Specific activities greater than 5.55 TBq g–1 can be achieved by optimizing the target geometry to reduce self-shielding and flux depression, and by processing targets soon after end of irradiation. The goals of this work were to (1) determine a feasible 153Gd source design and dose rate at 1 cm off axis under the specific activity constraint inherent in the current 153Gd production process, and (2) calculate the AAPM TG-43U1 brachytherapy dosimetry parameters for the hypothetical source, accounting for the encapsulation and separating the primary and scatter doses. 2. Materials and methods 2.1. Definition of source parameters The source and encapsulation dimensions for the hypothetical 153Gd source are shown in figure 1. The dimensions of the active 153Gd core were defined such that the source would be small enough to fit inside a 16 gauge needle with an inner diameter of 1.19 mm, which would be suitable for brachytherapy of the prostate (Tiong et al 2010). Assuming an encapsulation thickness of 0.08 mm and that an empty space of 0.045 mm about the source is sufficient to enable movement inside the catheter, the active source diameter is 0.84 mm. The 153Gd encapsulation material was assumed to be stainless steel composed of 67.92% Fe, 19% Cr, 10% Ni, 2% Mn, 1% Si and 0.08% C, with a density of 8.0 g cm−3. The active 153Gd source
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Figure 1. Hypothetical 153Gd source model.
length was restricted to 1 cm in order to be consistent with an older model of the Varian (Palo Alto, CA) VariSourceTM 192Ir source, which has been clinically utilized (Angelopoulos et al 2000). Due to the length of the source, applications in which the source is required to traverse a sharp bend in the catheter, such as bile duct treatments, may be limited. It was assumed that the active source was cylindrical in shape and that a solid shape was reasonable, since the exact form of the 153Gd material (pellets versus solid) and other design constraints are not yet known. The dose rate in water at 1 cm lateral to the 153Gd source was determined relative to that of a 370 GBq 192Ir source at 1 cm as follows. The dose rate, D˙ Ir , in water of a 370 GBq 192Ir source, assumed to be a Varian VariSourceTM (Varian Medical Systems, Palo Alto, CA), at 1 cm from the source axis was calculated using TG-43 parameters (Angelopoulos et al 2000) to be D˙ Ir = AIrCIr Ir = 4.18 × 104 cGy h−1. The activity, AIr, is 370 GBq, the conversion factor, CIr, is 102.7 U GBq−1 and converts 192Ir activity to air kerma strength units (U), and the TG-43 dose rate constant, Ir , is 1.101 cGy U−1 h−1 (Angelopoulos et al 2000). A validated TG-43-type dose rate constant for a 153Gd source is not yet available; therefore an alternative approach was used to calculate the 153Gd dose rate. The dose rate in water of the 153Gd source, D˙ Gd (cGy h−1), was calculated at a point a distance r (cm) perpendicular to the central axis of the source as follows: w β Tenc , (1) D˙ Gd = AGd Gd r w (cGy cm2 GBq−1 h−1) is the photon dose constant where (cm) is the length of the source, Gd (Enger et al 2011) of the source in water, β = 2 arctan (/2r) is the angle (radians) subtended by the active source with respect to the point, AGd (GBq) is the activity of the source and Tenc is the transmission factor of the source encapsulation. The photon dose constant for 153Gd in water was calculated as ∞ μ w −μw en (E )r 0 dE ρ (E ) (E ) e w a a , (2) Gd = Gd ∞ μ −μaen (E )r 0 dE ρ (E ) (E ) e where E is the photon ray energy, (E) is the energy fluence spectrum of 153Gd, r is the distance from the source in the medium (set to 1 cm), μen/ρ is the mass-energy absorption coefficient of a the medium, with the ‘a’ and ‘w’ superscripts indicating air and water, respectively. The Gd 2 −1 −1 value is known to be 46.6 cGy cm GBq h (Unger and Trubey 1982). The encapsulation transmission, Tenc, was calculated as follows: ∞ −μenc en (E )t 0 dE(E ) e ∞ Tenc = , (3) 0 dE(E ) where μenc en (E ) is the energy absorption coefficient of the encapsulation material and t is the encapsulation thickness.
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The active cylindrical 153Gd source has a radius, R, of 0.042 cm, density, ρ, of 7.9 g cm−3, mass, m, of 0.0438 g and specific activity, SA, of 5.55 TBq g–1 (Karelin et al 2000). 2.2. Monte Carlo simulations Calculations were performed with the Monte Carlo code Geant4.9.4 patch 2 using the Lawrence Livermore National Laboratory (LLNL) low-energy electromagnetic models (Agostinelli et al 2003). The LLNL models extend the range for the simulation of electromagnetic interactions down to 250 eV, which was set to the production cutoff for all particles. The LLNL approach exploits evaluated libraries EPDL97 (Cullen et al 1997), EEDL (Perkins et al 1997b) and EADL (Perkins et al 1997a). The decay of 153Gd was simulated using Geant4 radioactive decay data from the Evaluated Nuclear Structure Data File (Tuli 1987). The emitted particles were tracked and their interactions simulated. The 153Gd source was positioned at the center of a 40 cm radius spherical water phantom to simulate an unbounded phantom. The active core of the source was a pure gadolinium cylinder. The length of the simulated source guide was 5 mm. 3. Results and discussion w , is 6.45 cGy cm2 GBq−1 h−1. The The photon dose constant in water for 153Gd, Gd encapsulation transmission factor, Tenc, is 0.90 for the stainless-steel source encapsulation material considered. The activity of the 153Gd source, AGd, is 243 GBq. The dose rate in water of the 153Gd source at 1 cm off axis, calculated using (1), is 13.1 Gy h−1, or 3.1% of the dose rate of a 370 GBq 192Ir VariSourceTM . The lower dose rate of the 153Gd source relative to that of a 370 GBq 192Ir source implies that the 153Gd source would be restricted to low-dose-rate and pulsed-dose-rate brachytherapy applications. In clinical situations where the benefits of 153Gd can be exploited, such as for IRSBT, the resulting longer treatment times may be acceptable. Increasing the specific activity of the source through improved 153Gd production would translate linearly to increased dose rate. The source dimensions could also be increased to reduce treatment times, subject to the constraints of the application. For example, one could design a 153Gd source with an active diameter of 1–2 mm for intracavitary applications, which could be useful for intracavitary rotating shield brachytherapy. The photon spectrum from 153Gd decay obtained with Geant4 was compared with published data from the Nudat database in figure 2(a). The results agreed well for photon energies above 10 keV, although the Nudat database lacks the spectrum for lower energies. Photons with energies below 10 keV will be absorbed by the source encapsulation and therefore will not affect the overall dose distribution. Figure 2(b) shows the beta particle emission spectrum of 153Gd obtained with Geant4. The beta particles emitted from the source have low enough energies to be absorbed in the source encapsulation. The radiation yield of the electrons in the encapsulation material is less than 0.4% (Berger et al 2012); therefore the resulting bremsstrahlung will deliver a negligible dose to the tissue surrounding the source. Figure 2(c) shows the photon spectrum for all escaping photons in the water just after leaving the stainless-steel capsule. Figure 2(d) shows the photon spectrum at 1 cm depth in water lateral to the source. Figure 3 shows the radial dose function, g(r), of the hypothetical source, separated into primary, first scatter, multiple scatter and total scatter components. The radial dose function is dominated by dose deposition resulting from primary photon interactions at radii below about 1.5 cm from the source, and total scatter dose dominates at radii greater than 1.5 cm. The
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(a)
(b)
(c)
(d)
Figure 2. (a) Photon spectrum from 153Gd decay obtained with Geant4 compared with published data from the Nudat database. (b) Beta spectrum of 153Gd decay obtained with Geant4. (c) Photon spectrum for all escaping photons in the water just after leaving the stainless-steel encapsulation. (d) Photon spectrum at 1 cm depth in the water phantom.
Figure 3. Radial dose function, g(r), of the 153Gd source in a water phantom with 40 cm radius.
photon scatter component of the radial dose function is equal parts first and multiple scatter at radii below 1.5 cm, and multiple-scattered photons account for the majority of the photon scatter component at greater radii. Figure 4 shows the 2D anisotropy function of the hypothetical source, F(r, θ ), for polar angles, θ , between 0◦ and 180◦ , at several radial distances, r, ranging from 0.5 to 15 cm.
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Figure 4. 2D anisotropy function F(r,θ ), of the 153Gd source at several radial distances.
The anisotropy function at a fixed radius of 1 cm ranges from about 0.2 to 1 over all polar angles, indicating greater anisotropy than an 192Ir source, which may range from 0.5 to 1 at a radius of 1 cm over all polar angles (Angelopoulos et al 2000). The increase in anisotropy variation relative to an 192Ir source is likely due to the relatively long, 10 mm active core and the thick distal end of the encapsulation of the 153Gd source design. The anisotropy function increases with increasing distance. The anisotropy could potentially be reduced by reducing the encapsulation thickness at the distal end of the active source. It is not expected that a significant reduction in anisotropy could be achieved in this manner, however, since the anisotropy is dominated by self-attenuation in the active source. 4. Conclusions The 153Gd source considered would be suitable as an intermediate-energy low-dose-rate or pulsed-dose-rate brachytherapy source. The 153Gd source could provide a means for I-RSBT delivery and enable brachytherapy treatments with patient specific shielding and reduced personnel shielding requirements relative to 192Ir. References Agostinelli S et al 2003 GEANT4—a simulation toolkit Nucl. Instrum. Methods Phys. Res. A 506 250–303 Angelopoulos A, Baras P, Sakelliou L, Karaiskos P and Sandilos P 2000 Monte Carlo dosimetry of a new 192Ir high dose rate brachytherapy source Med. Phys. 27 2521–7 Ballester F, Granero D, P´erez-Calatayud J, Venselaar M and Rivard M J 2010 Study of encapsulated 170Tm sources for their potential use in brachytherapy Med. Phys. 37 1629–37 Beaulieu L, Carlsson Tedgren A, Carrier J F, Davis S D, Mourtada F, Rivard M J, Thomson R M, Verhaegen F, Wareing T A and Williamson J F 2012 Report of the task group 186 on model-based dose calculation methods in brachytherapy beyond the TG-43 formalism: current status and recommendations for clinical implementation Med. Phys. 39 6208–36 Berger M J, Coursey J S and Zucker D S 2012 ESTAR, PSTAR, and ASTAR: computer programs for calculating stopping-power and range tables for electrons, protons, helium ions (Version 1.2.2) (Gaithersburg, MD: National Institute of Standards and Technology) Bray L A and Corneillie T M 2001 Method of separating and purifying gadolinium-153 US Patent Specification 6245305B1 (issued 12 June 2001) Case F N, Acree E H and Cutshall N H 1969 Production study of gadolinium-153 Report TM-2632 (Oak Ridge, TN: Oak Ridge National Laboratory)
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Cullen D E, Hubbell J H and Kissel L 1997 EPDL97: the evaluated photon data library, ’97 version LLNL Report No UCRL-50400 vol 6 Rev 5 (Livermore, CA: Lawrence Livermore National Laboratory) DiFilippo F P 2008 Geometric characterization of multi-axis multi-pinhole SPECT Med. Phys. 35 181–94 Ebert M A 2006 Potential dose-conformity advantages with multi-source intensity-modulated brachytherapy (IMBT) Australas. Phys. Eng. Sci. Med. 29 165–71 Enger S A, D’Amours M and Beaulieu L 2011 Modeling a hypothetical 170Tm source for brachytherapy applications Med. Phys. 38 5307–10 Enger S A, Lundkvist H, D’Amours M and Beaulieu L 2012 Exploring 57Co as a new isotope for brachytherapy applications Med. Phys. 39 2342–5 Frey E C and Tsui B M W 1995 A comparison of Gd-153 and Co-57 as transmission sources for simultaneous TCT and Tl-201 SPECT IEEE Trans. Nucl. Sci. 42 1201–6 Granero D, Per´ez-Calatayud J and Ballester F 2005 Broad-beam transmission data for new brachytherapy sources, Tm-170 and Yb-169 Radiat. Prot. Dosim. 118 11–5 Karelin Y A, Efimov V N, Filimonov V T, Kuznetsov R A, Revyakin L, Andreev O I, Zhemkov I Y, Bukh V G, Lebedev V M and Spiridonov Y N 2000 Radionuclide production using a fast flux reactor Appl. Radiat. Isot. 53 825–7 Lin L, Patel R R, Thomadsen B R and Henderson D L 2008 The use of directional interstitial sources to improve dosimetry in breast brachytherapy Med. Phys. 35 240–7 Mainegra E, Capote R and L´opez E 1998 Evaluation of dose rate constant for Pd-103, Ir-192, Yb-169, and I-125 brachytherapy sources: an EGS4 Monte Carlo study Phys. Med. Biol. 43 1557–66 McGinley P H 2010 Shielding Techniques for Radiation Oncology Facilities (Madison, WI: Medical Physics Publishing) Munro J J, Medich D and Mutyala S 2008 Intraoperative high dose rate brachytherapy using 170thullium radiation sources Brachytherapy 7 160 Nath R, Anderson L L, Luxton G, Weaver K A, Williamson J F and Meigooni A S 1995 Dosimetry of interstitial brachytherapy sources: recommendations of the AAPM Radiation Therapy Committee Task Group No. 43. American Association of Physicists in Medicine Med. Phys. 22 209—34 Nath R, Anderson L L, Luxton G, Weaver K A, Williamson J F and Meigooni A S 1996 Med. Phys. 23 1579 (erratum) Papagiannis P, Baltas D, Granero D, Perez-Calatayud J, Gimeno J, Ballester F and Venselaar J L 2008 Radiation transmission data for radionuclides and materials relevant to brachytherapy facility shielding Med. Phys. 35 4898–906 Perkins S T, Cullen D E, Chen M H, Hubbell J H, Rathkopf J and Scofeld J 1997a Tables and graphs of atomic subshell and relaxation data derived from the LLNL Evaluated Atomic Data Library (EADL), Z = 1–100 LLNL Report No UCRL-50400-v-30 (Livermore, CA: Lawrence Livermore National Laboratory) Perkins S T, Cullen D E and Seltzer S M 1997b Table and graphs of electron-interaction cross sections from 10 eV to 100 GeV derived from the LLNL Evaluated Electron Data Library (EEDL) LLNL Report No UCRL-50400-v-31 (Livermore, CA: Lawrence Livermore National Laboratory) Ramey D W 1988 Gadolinium-153 production at the Oak Ridge National Laboratory Report TM-10641 (Oak Ridge, TN: Oak Ridge National Laboratory) Rivard M J, Coursey B M, DeWerd L A, Hanson W F, Huq M S, Ibbott G S, Mitch M G, Nath R and Williamson J F 2004 Update of AAPM Task Group No. 43 Report: a revised AAPM protocol for brachytherapy dose calculations Med. Phys. 31 633—74 Rivard M J, Coursey B M, DeWerd L A, Hanson W F, Huq M S, Ibbott G S, Mitch M G, Nath R and Williamson J F 2004a Med. Phys. 31 3532–3 (erratum) Tiong A, Bydder S, Ebert M, Caswell N, Waterhouse D, Spry N, Camille P and Joseph D 2010 A small tolerance for catheter displacement in high-dose rate prostate brachytherapy is necessary and feasible Int. J. Radiat. Oncol. Biol. Phys. 76 1066–72 Tuli J K 1987 Evaluated nuclear structure data file, a manual for preparation of data sets (Berkeley, CA: Lawrence Berkeley National Laboratory) available at http://ie.lbl.gov/isoexpl/ensdfman.pdf Unger L M and Trubey D K 1982 Specific gamma-ray dose constants for nuclides important to dosimetry and radiological assessment Report RSIC-45 (Oak Ridge, TN: Oak Ridge National Laboratory)
