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Year : 2012  |  Volume : 35  |  Issue : 3  |  Page : 126-134  

Radiation safety issues relevant to proton therapy and radioisotope production medical cyclotrons

West German Proton Therapy Centre Essen (WPE gGmbH), Hufelandsrasse 55, D-45147 Essen, Germany

Date of Web Publication5-Sep-2013

Correspondence Address:
Bhaskar Mukherjee
West German Proton Therapy Centre Essen (WPE gGmbH), Hufelandsrasse 55, D-45147 Essen
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Source of Support: None, Conflict of Interest: None

DOI: 10.4103/0972-0464.117668

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Medical cyclotrons are now constructed as turnkey facilities at nuclear medicine clinics, specialised particle therapy facilities and radioisotope production centres. Most medical cyclotrons usually accelerate protons to high energies and could be divided mainly in two categories: (a) Low energy (E P = 15-30 MeV) machines, dedicated for medical positron emission tomography and single photon emission computer tomography radioisotope production and (b) High energy (E P = 100-250 MeV) machines, predominantly used for radiotherapy of malignant tumours. Parasitic gamma and neutron radiation are produced during the operation of medical cyclotrons. Furthermore, high level of gamma radiation produced by the activated cyclotron components could impose radiation exposure to maintenance crew. Hence, radiation safety is imperative to safe and reliable operation of medical cyclotron facilities. A sound operational health physics procedure assures the minimisation of radiation exposure to patients and members of the public abiding the regulatory guidelines. This paper highlights the important radiation safety aspects related to safe operation of proton therapy and radioisotope production medical cyclotrons.

Keywords: Medical cyclotron, proton therapy, radiation safety, radiation shielding, radioisotope production

How to cite this article:
Mukherjee B. Radiation safety issues relevant to proton therapy and radioisotope production medical cyclotrons. Radiat Prot Environ 2012;35:126-34

How to cite this URL:
Mukherjee B. Radiation safety issues relevant to proton therapy and radioisotope production medical cyclotrons. Radiat Prot Environ [serial online] 2012 [cited 2022 Nov 29];35:126-34. Available from: https://www.rpe.org.in/text.asp?2012/35/3/126/117668

  Introduction Top

Cyclotrons have been vital to experimental nuclear and particle physics research following its invention by Ernest Orlando Lawrence in 1938. [1] These machines were used in biophysical and medical fields since early days as well, however, merely as "piggy back" applications. The main priority focussed on the nuclear physics applications. Since 1980s, a new breed of compact cyclotrons called "Medical Cyclotron," primarily aimed for medical applications; mainly the production of radionuclides for nuclear medicine became available. [2] Medical cyclotrons accelerating intense beams of protons to high energy, usually 100-250 MeV are now being extensively used in radiotherapy of malignant cancers. [3] A comprehensive list of radioisotope production cyclotrons worldwide in operation is published elsewhere. [4]

At radioisotope production cyclotrons, thick copper substrates electroplated with selected target material are bombarded with intense beam of 15-30 MeV protons. [5]

On the other hand, at proton therapy cyclotrons, high energy 120-235 MeV proton beams of selected shapes and profiles are delivered to the tumour volume (target) via the nozzle and sophisticated beam shaping components made of brass and polystyrene. [3] In routine operation (as above) and during accidental beam losses intense fields of neutrons and gamma rays are produced as elucidated in [Figure 1].
Figure 1: Nuclear interaction and radiation exposure pathways during routine operation of medical cyclotrons as well as accidental beam loss cases

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Prompt neutrons (color coded: Yellow) and gamma rays (color coded: Blue) are exclusively generated [Figure 1] in the target vault or in the treatment rooms of proton therapy facilities. The neutrons activate beam line components as well as room air. The neutrons and gamma rays penetrate through the roof shielding and eventually cause skyshine [6],[7] exposure to members of the public residing in the vicinity of the cyclotron facility. The neutron activation of air at a biomedical cyclotron facility and population dose calculations were discussed in details literature. [8]

The present paper highlights important radiation safety issues related to medical cyclotrons used in proton therapy and radioisotope production. A comprehensive database on cyclotron radiation safety and operational health physics is available elsewhere. [9]

  Radiation Shielding Principle Top

In [Figure 2], the footprint of the Australian National Medical Cyclotron (NMC) facility housing a 30 MeV (H - ) radioisotope production cyclotron is depicted. The NMC is situated above ground and includes: (a) Cyclotron vault, (b) target vault with two solid target stations ST1 and ST2 and a gas target station (GTS), (c) single photon emission computer tomography (SPECT) vault with a high-current solid target station ST3 and (d) positron emission tomography (PET) vault with rack mounted target array, (e) PET and SPECT radioisotope production hot cells, (f) cyclotron power supply and control room. The locations of wall mounted neutron (red circles) and gamma detectors (blue squares) are also indicated. The vault walls are made of 2.3 m thick concrete. The target transfer lines leading to hot cells are shown in red. In March 1992, the NMC commenced its routine production of PET radiopharmaceuticals 11 CO 2 , 13 NH 3 and 18 FDG in aqueous form as well as SPECT radioisotopes 201 Tl, 67 Ga and 123 I.
Figure 2: Footprint of the Australian national medical cyclotron operating a 30 MeV, H- ion medical cyclotron for radioisotope production

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In [Figure 3], the footprint of the West German Proton Therapy Centre Essen (WPE) is shown. The cyclotron, beam lines and all four treatment-rooms are located 8 m deep in underground. The 2 m thick concrete shielding walls serve primarily for the radiation protection of the WPE staff members, patients and external service personnel including security staff, delivery and utility technicians.
Figure 3: Footprint of the West German proton therapy centre essen (WPE) operating a 235 MeV, H+ ion medical cyclotron

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  Shielding Calculation Method Top

The primary goal of radiation shielding is to reduce the dose equivalent (DE) at the location of interest outside the shielding [10] imposed by statutory authority, usually 1 mSv/y in OECD (Organization for Economic Co-operation and Development) countries. The basic geometry of shielding calculation is depicted in [Figure 4].
Figure 4: The geometry relevant to shielding calculations for proton accelerators

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The DE H 0 at the impact point produced by the protons impinging on the target (T), known as the "source term", reduced to H x at the point of interest (POI) on the outer surface of the shielding of thickness "d". The distance between T and POI is designated as "r".

H = H 0 exp(-d/λ (E))/r 2 (1)

Where, λ (E) = energy dependent neutron attenuation length [11] as depicted in [Figure 5]:
Figure 5: Neutron attenuation length is plotted as a function of neutron energy

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The neutron attenuation lengths for low (NMC) and medium (WPE) energy neutrons were extrapolated to be 12.7 cm and 28.2 cm respectively. Normal concrete [12] was used for shielding construction of NMC and WPE.

The source term for the 30 MeV cyclotron of NMC was adapted from the reference. [13] A Monte-Carlo simulation code [14] was used to estimate the source term ( 0o H 0 ) of the WPE cyclotron [Figure 6] and [Figure 7].
Figure 6: Geometry setup for the Source term calculation using MNCPX 2.6.0 code

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Figure 7: Neutron energy fluence per proton in the forward direction evaluated using MCNPX code. The median neutron energy was calculated to be 68.2 MeV. The source term is shown inset

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  Shielding Validation Top

For the neutron shielding calculation, the footprint [Figure 8], maze [Figure 9] and vertical section [Figure 10] of the treatment room 3 [Figure 3] were taken into account.
Figure 8: Footprint of the treatment room #3 showing the shielding thicknesses and the "line of sight" distances of the validation points from the source S located at isocentre

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Figure 9: The details of maze of the treatment room 3 showing the length of the legs

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Figure 10: The vertical cross-section of the treatment room 3 with important dimensions

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The radiation shielding of the treatment room 3 was validated by calculating the neutron DEs at the locations shown in [Figure 8] under the consideration of the following: (a) A 10 nA proton pencil beam (3 mm diameter) impinging on a 30 × 30 × 30 cm 3 polystyrene phantom [Figure 6], (b) the source term [Figure 7] was calculated to be a DE rate of 13.04 μSv/h at 1 m from the point of impact (worst case scenario), (c) implementation of standard shielding concrete of a density of 2.35 gcm -3 [Figure 5], (d) an operation cycle of 500 h/year/treatment room (total duty schedule of the WPE facility with 4 treatment rooms: 2000 operation hours/year,) (e) permissible DE imposed by the statutory authority: 1.0 mSv/y, (f) shielding thickness calculated for neutrons is adequate to attenuate the secondary gamma rays produced in the treatment room and shielding itself.

Neutron DEs at the external surface of the shielding walls were calculated using equation 1. The corresponding concrete shielding thickness d (marked in red) and the line of sight distances r (blue dashed lines) are shown in [Figure 8]. The results are presented in [Table 1].
Table 1: Calculated neutron dose equivalents (H) for a 500 h/year operation cycle at selected locations of the treatment room #3 [Figures 8 and 10]. The line of sight distance (r) from the source and the OF are shown in columns 2 and 4 respectively

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The total attenuated neutron DE at treatment room entrance [Figure 8] is the sum of direct (Ed) and the dose component transmitted through the maze (Em). The fast neutrons produced at the isocentre (S) are attenuated via transmission through the 3-leg maze [Figure 9].

The neutron attenuation factors corresponding to 1 st (a1), 2 nd (a2) and 3 rd (a3) maze legs were calculated as: [15]

a1 = 0.98 exp(-0.80r1/√A) + 0.2 (2a)

a2 = 0.74 exp(-10r2/√A) + 0.21exp(-1.6r2/√A) + 0.05 exp(-0.54r2/√A) (2b)

a3 = 0.81exp(-3.8r3/√A) + 0.19exp(-0.7r3/√A) (2c)

Where, the values of leg lengths r1, r2 and r3 and cross section (A) are 3.56 m, 4.64 m, 2.92 m and 5.6 m 2 respectively. The neutron DE at the maze exit (Em + Ed) is given [Table 1].

The neutron DE at roof (R), at a distance of 11.2 m from the isocentre [Figure 10] was also calculated using equation 1 [Table 1].

  Neutron Skyshine Top

The inadequate roof shielding of high-energy particle accelerator containment causes neutron leakage. These leakage neutrons undergo multiple scattering with the air molecules, may propagate to remote distance from the accelerator facility, thereby causing radiation exposure to members of the public. This phenomenon is known as skyshine. The skyshine DE could be calculated using the empirical formula. [6]

ɸ = (a × Q/4 π r 2 ) exp(-r/λ) (1-exp(-r/μ)) (3)


ɸ (μ Sv) = Skyshine DE at a remote location

a = Empirical build up factor = 2.8

μ (m) = Effective interaction length of source neutrons in air ≈ 56 m

Q (μSv) = Neutron source strength

r (m) = Distance from the source > 50 m

λ (m) = Dose attenuation length on neutrons in air = 270 m, for En < 5 MeV.

At WPE, we have used an ultra-sensitive thermoluminescence dosimeters based neutron area monitor [16] to assess the neutron skyshine DEs. Three neutron monitors enclosed in 30 cm diameter × 30 cm long polystyrene moderator cylinders were placed at one of our office buildings Lernen und Lehrzentrum (LLZ) Room 019, about 300 m east from the WPE medical cyclotron facility. The dosimeters were evaluated after 2 (60 days), 3 (90 days), and 4 (120 days) months of exposure. The dosimetry exercise took place during March-June 2012. The results are shown in [Table 2].
Table 2: Neutron skyshine at LLZ 019 Bldg near WPE facility measured with the ultra-sensitive neutron area monitor[15]

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  Activation of Containment Air Top

Fast neutrons produced during the bombardment of target with high-energy protons, slow down to thermal energies through multiple collisions with the containment walls and floor. These thermal neutrons produce radioactive 41 Ar via the neutron capture reaction 40 Ar (n, γ) 41 Ar (σ =630 mb) with 40 Ar (a noble gas), present in the containment air with an abundance of 0.46% by volume. When released, the radioactive 41 Ar, a gamma emitter (Eγ =1.3 MeV) with a half-life of 1.8 h inflict radiation exposure to person in the immediate vicinity, i.e. the cyclotron workers. Furthermore, due to longer half-life, the effluent could reach remote distances through atmospheric dispersion, thereby causing radiation exposure to members of the public. [8]

At NMC natural cobalt ( 59 Co) pellets were used to estimate the thermal neutron fluence in the target vault via the neutron capture 59 Co(n, γ) 60 Co reaction. [17] At WPE, we have used the temperature compensated BDT (Bubble Detector Thermal) type superheated emulsion (bubble) detectors [18],[19] to assess the thermal neutron fluence in the treatment rooms.

Four pairs (one bare and other enclosed in a Cadmium shell of 2 mm wall thickness) of BDT detectors were placed at distance 1 m around the polystyrene phantom [marked as star " *" in [Figure 6]. The phantom was bombarded with a 200 MeV proton beam to an integrated dose of 60 Gy (total planned dose per day). The thermal neutron DE (H th μSvh -1 ) and the fluence (Φth cm -2 /h) rates were calculated using methods described elsewhere: [20]

Hth = s[N av (BDT)-N av (BDT) Cd ] (4a)

Φth = Hth /fth (4b)

Where, s and fth are the bubble detector sensitivity (0.45 μSv/bubble) and thermal neutron fluence to DE conversion factor (3.9 × 10 -6 μSvcm 2 ) [21] respectively.

N av (BDT) and N av (BDT Cd ) stand for the average bubble counts of the bare and cadmium shielded detectors respectively. Each bubble detector was counted 4 times, at viewing angles of 0°, 90°, 180° and 270°, corresponding digital photographs were taken each time and subsequently analyzed using a dedicated bubble counting software written in JAVA script: [22]

The specific saturation activity concentration of (Bqm -3 /Gy/h) of 41 Ar in the treatment room air is given as: [17]

A = N Φth σ exp (1-λT b ) exp (-λT d )/V (5)


N = Number of 40 Ar atoms per cubic meter of air = 2.5 × 10 22 atoms m -3

Φth = Thermal neutron fluence rate (cm -2 /Gy/h) in the treatment room (Equation 4b)

σ = Neutron capture cross section of 40 Ar = 630 mb (6.30 × 10 -25 cm 2 )

λ = Decay constant of 41 Ar = 0.38/h

V = Treatment room volume = 151 m 3

T b = Irradiation time per fraction = 3 min (0.05 h)

T d = Delay time following irradiation = 0 h (instantaneous release of the radioactive 41 Ar)

  Component Activation Top

During long-term operation of high current radioisotope production medical cyclotrons the induced radioactivity accumulated in cyclotron components and shielding concrete causes radiological risks to humans and the environment. The technicians and radiation workers could receive a high radiation exposure while carrying out repair and maintenance duties. [23] Furthermore, in order to prevent severe radiological and environmental hazards, utmost precaution and skilled health physics support are necessary for dismantling and storage of the shielding concrete [24],[25] originated from old, decommissioned cyclotron facilities.

At proton therapy cyclotrons, however, patient specific brass apertures [Figure 11] are attached and removed from the beam delivery nozzles by radiotherapy technicians many times a day.
Figure 11: In-situ dosimetry of an activated brass aperture at West German Proton Therapy facility Centre Essen

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For example, during routine operation of the facility, the brass-apertures will be exchanged 30 times a day, corresponding to: Three apertures per patient and 10 patients/day. This may evidently pose an excess of extremity radiation exposure to those workers. During tumor irradiation high-energy protons up to 230 MeV use to hit the brass (63% Cu, 37% Zn) apertures attached to the end of beam delivery nozzle [Figure 11], thereby generating a myriad of daughter species of various half-lives and gamma emissions via the (p, xn) reaction. [26] Hence, we have analyzed the decay characteristics of the brass aperture to optimize the active handling process, in order to achieve maximum dose reduction.

Instead of using a high-resolution gamma spectrometer, we have developed a simple experimental method based on gamma counting technique. The irradiated brass aperture was removed from the nozzle [Figure 11] and brought to a low background area outside the treatment room, a collimated gamma-probe (Geiger Mueller) counter placed atop, which is connected to a radiation monitor, interfaced to personal computers [Figure 12].
Figure 12: Set up for the analysis of decay characteristics of activated brass aperture

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The gamma radiation from the activated aperture was counted in every 5s for of 30 h, the results are shown in [Figure 13]a.
Figure 13: Results of the gamma counting experiments of an activated brass aperture

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The decay curve was unfolded with three optimized exponential functions using the principle of a genetic algorithm [27] and presented in [Figure 13]b. The fitting functions are shown below: [28]

f1(t) = 0.894 exp(-2.94t) (6a)

f2(t) = 0.091 exp(-0.42t) (6b)

f3(t) = 0.015 exp(-0.03t) (6c)

The remaining activity of the brass aperture after different decay time is shown in [Table 3].
Table 3: Remaining activity of the brass aperture

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  Results and Discussion Top

Radiation production and propagation

The knowledge of radiation production and propagation as well as the exposure pathways as depicted in [Figure 1] is an important prerequisite for the design of a medical cyclotron facility. It assists to set the input parameters for the shielding of the proposed facility in the light of the environmental and radiological impacts to radiation workers and members of the public.

Radiation shielding

The shielding calculation method presented in this report is mainly based on deterministic method instead of using a rigorous particle transport calculations using Monte-Carlo codes. The shielding adequacy at critical locations of the cyclotron and Target vault (NMC) and the proton therapy treatment room (WPE) were validated and later confirmed with dosimetric measurements [Table 1]. At WPE the treatment room #3 is symmetrically located between treatment rooms #2 and #4 [Figure 3], hence, considered for this calculations as an example. In fact, we already have validated the radiological shielding for all remaining treatment rooms as required by the regulatory authority. Evidently, the shielding thicknesses for neutrons are also adequate for the secondary gamma rays; hence, no exclusive shielding calculation for gammas was performed.

The source term for WPE shielding calculations, corresponding to 230 MeV protons hitting a thick polystyrene phantom was simulated using the Monte Carlo N-Particle Transport (MCNPX 2.6.0) code. We already have measured the angular (0°, 90°, and 180°) neutron yields of 235, 200, and 177 MeV protons on brass, iron and polystyrene, the results will be published shortly. At NMC, the source term (30 MeV protons bombarding a thick copper target) was taken from published data.

At NMC, the shielding calculation and measurement results relevant to existing cyclotron and target vaults were used to design the optimized shielding of the dedicated SPECT and PET caves [Figure 2]. Hence, the methods and results presented in this paper could be used as an important benchmark for the shielding design of new proton therapy and radioisotope production cyclotrons.

Neutron skyshine

The results of neutron skyshine measurement using a passive neutron monitor were inconclusive. The values shown in [Table 2] corresponds to the dose accumulated from natural background radiation; 0.055 to 0.433 μSvh -1 in Germany.

The other reasons could be (1) adequate and flawless roof shielding of the facility as shown in [Figure 10] and (2) Irregular cyclotron operation with frequent shut downs during the commissioning phase, when the measurements were made. We have planned to continue the skyshine measurements starting from April 2013, when the WPE commences its routine patient treatment.

Air activation in the treatment room

The activity concentration (A: Bqm -3 ) of radioactive noble gas 41 Ar in the treatment room # 3 (volume = 161 m 3 ) was calculated using superheated Bubble detectors was calculated as follows: The numbers of bubbles in bare and Cd-covered detectors were counted to be 52 and 45 respectively. Using equations 4a and 4b the thermal neutron fluence for a daily therapeutic dose of 60 Gy was calculated to be 8.08 × 10 5 cm -2 . By substituting the numerical values in equation 5, the activity concentration of 41 Ar was calculated to be 2.11 × 10 2 Bqm -3 . Furthermore, an hourly air exchange (ventilation) rate of 8, gives the effective 41 Ar concentration of 26.4 Bqm -3 , substantially lower than the permissible limit of 200 Bqm -3 .

Component activation

Activated cyclotron parts could cause high (gamma) exposure to radiation workers during repair and maintenance activities. At high current low energy (~30MeV) radioisotope production cyclotrons, most of the activated parts are generated near the target station and usually made of Copper, Brass, Steel and Aluminum. [23]

At proton therapy centers like WPE, the main activated parts are the brass apertures (diameter: 20 cm, thickness: 6.6 cm, weight: 17.6 kg) used to create patient specific beam profile for tumor irradiation with high-energy protons. There will be 30 such apertures (3 apertures/patient, 10 patients/day) generated per day. The decay characteristics of brass apertures were estimated experimentally. A personnel dose reduction of 50% could be achieved by allowing 15 min decay before handling, i.e., while removing it from the beam delivery nozzle. The activities of the brass apertures drop to an insignificant level after the lapse of 1 week [Table 3].

  Summary and Conclusion Top

This paper embodies important safety related aspects of modern medical cyclotrons, both high-current radioisotope production and high-energy particle therapy machines. This report analyzed the pathways of radiation leakage during routine operation as well as abnormal cases like accidental beam loss scenarios.

The deterministic method of shielding calculation and validation for newly designed cyclotron facilities and the method of the source term estimation for the 235 MeV proton therapy cyclotron at WPE using the MCNPX 2.6.0 code are highlighted.

The main environmental impact pathways relevant to medical cyclotron operation, i.e., neutron skyshine radiation and radioactive noble gas 41 Ar production within the containment air are discussed.

The radioactive decay characteristics of activated medical cyclotron parts, in particular the patient specific brass apertures used in cancer therapy are analyzed and personal radiotherapy technicians dose reduction and radioactive waste disposal strategies are proposed.

Topics discussed in this paper will prove to be valuable while planning to construct new medical cyclotron facilities for radioisotope production and radiotherapy.

  Acknowledgment Top

The author wishes to thank the IBA cyclotron operation team in particular, Dr. Ann-Katrin Nix and Dipl. Ing (FH) Dennis Mantel for technical support. The author is also thankful to MSc. Ing. Carolina Sofia Llina Fuentes of WPE Medical-Physics group for her patience and skillful assistance during dosimetry studies and neutron measurements and Dr. Med. Andreas Kaiser for reviewing the manuscript.

  References Top

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  [Figure 1], [Figure 2], [Figure 3], [Figure 4], [Figure 5], [Figure 6], [Figure 7], [Figure 8], [Figure 9], [Figure 10], [Figure 11], [Figure 12], [Figure 13]

  [Table 1], [Table 2], [Table 3]

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