Neutron spectrometry in a PET cyclotron with a Bonner sphere system

Decommissioning of a Medical Cyclotron Vault: the Case Study of the National Cancer Institute of Milano

ABSTRACT

In the widespread use of medical cyclotrons for isotope production, radiological and economic consequences related to the decommissioning of particle accelerators are often neglected. However, decommissioning regulation and its related procedures always demand efforts and costs that can unexpectedly impact on budgets. The magnitude of this impact depends strongly on the residual radioactivity of the accelerator and of the vault, and more specifically on the kind and activity concentration of residual radionuclides. This work reports and discusses a case study that analyzes in detail the characterization activities needed for optimized management of the decommissioning of a medical cyclotron vault. In particular, this paper presents the activities carried out for assessing the activity concentrations and for guiding the disposal of the cyclotron vault of the Italian National Cancer Institute of Milano (INT). An unshielded 17 MeV cyclotron vault was characterized by high resolution gamma-ray spectrometry both in-situ and in-laboratory on extracted samples. Monte Carlo simulations were also carried out to assess the overall distribution of activation in the vault. After a few months from the final shutdown of the accelerator, activity concentrations in the concrete walls due to neutron activation exceeded the clearance levels in many regions, especially close to the cyclotron target. Due to the relatively long half-lives of some radionuclides, a time interval of about 20 y after the end of bombardment is necessary for achieving clearance in some critical positions. Far from the target or in positions shielded by the cyclotron, activation levels were below the clearance level. The comparison between Monte Carlo simulations and experimental results shows a good agreement. The in-situ measurements, simpler and economically advantageous, cannot completely replace the destructive measurements, but they may limit the number of required samples and consequently the decommissioning costs. The methodology described and the results obtained demonstrated that it is possible to obtain accurate estimations of activity concentrations with cheap and quick in-situ measurements if the concentration profile in-depth inside the wall is well known. This profile can be obtained either experimentally or numerically through suitably validated Monte Carlo simulations.

Identification of Activation Isotopes in a CS-30 Cyclotron Vault

A CS-30 cyclotron has been in operation at King Faisal Specialist Hospital and Research Center (KFSHRC) since 1982. The CS-30 cyclotron has been used to produce medical radioisotopes for positron emission tomography (PET) and single-photon emission computed tomography (SPECT). Some of the nuclear reactions of radionuclide production are associated with the intense release of a wide range of fast neutrons. In this work, we investigated the radionuclides produced from neutron interactions with the cyclotron facility walls. Activation isotopes were determined by performing gamma ray spectrometry utilizing a high-purity germanium (HPGe) detector. The major radionuclides found were ¹⁵²Eu, ¹⁵⁴Eu, ¹³⁴Cs, ⁶⁵Zn and ⁶⁰Co. Activation isotope accumulation had increased the dose rate inside the facility. The surface dose rates were measured at all of the surrounding walls. The maximum surface dose rate was found to be 1.2 µSv/h, which is much lower than the permissible occupational exposure of 15 µSv/h based daily 5 work hours.

Calculation and Validation of the Response Matrix for a Neutron Multisphere Spectrometer with an Indium Central Detector

Expansion of accelerator technologies into different application areas brings the need for neutron-field measurements in their vicinity. The measurements of neutron energy spectra and neutron dose rates can be done with the aid of neutron multisphere spectrometers with passive neutron detectors. In addition to the usually used thermoluminescent detectors and gold activation foils, another activation material-indium-can be considered. This paper presents a design of such an indium-based neutron multisphere spectrometer. Its response matrix was calculated with Monte Carlo code MCNP5 in the energy interval from 0.001 eV to about 20 MeV. The response matrix was validated with an Am-Be standard neutron source. MAXED code was used for neutron energy spectra unfolding. The values of the neutron fluence rate and neutron ambient dose equivalent rate in a reference point were calculated. The properties and advantages of indium for use in neutron multisphere spectrometry are discussed.

Decontamination of the Activation Product Based on a Legal Revision of the Cyclotron Vault Room on the Non-self-shield Compact Medical Cyclotron

The non-self-shield compact medical cyclotron and the cyclotron vault room were in operation for 27 years. They have now been decommissioned. We efficiently implemented a technique to identify an activation product in the cyclotron vault room. Firstly, the distribution of radioactive concentrations in the concrete of the cyclotron vault room was estimated by calculation from the record of the cyclotron operation. Secondly, the comparison of calculated results with an actual measurement was performed using a NaI scintillation survey meter and a high-purity germanium detector. The calculated values were overestimated as compared to the values measured using the NaI scintillation survey meter and the high-purity germanium detector. However, it could limit the decontamination area. By simulating the activation range, we were able to minimize the concrete core sampling. Finally, the appropriate range of radioactivated area in the cyclotron vault room was decontaminated based on the results of the calculation. After decontamination, the radioactive concentration was below the detection limit value in all areas inside the cyclotron vault room. By these procedures, the decommissioning process of the cyclotron vault room was more efficiently performed.

Neutron spectra around a tandem linear accelerator in the generation of (18)F with a bonner sphere spectrometer

A Bonner sphere spectrometer was used to measure the neutron spectra produced at the collision of protons with an H2(18)O target at different angles. A unique H2(18)O target to produce (18)F was designed and placed in a Tandem linear particle accelerator which produces 8.5MeV protons. The neutron count rates measured with the Bonner spheres were unfolded with the MAXED code. With the GEANT4 Monte Carlo code the neutron spectrum induced in the (p, n) reaction was estimated, this spectrum was used as initial guess during unfolding. Although the cross section of the reaction (18)O(p,n)(18)F is well known, the neutron energy spectra is not correctly defined and it is necessary to verify the simulation with measurements. For this reason, the sensitivity of the unfolding method to the initial spectrum was analyzed applying small variation to the fast neutron peak.

Use of the MCNPX to calculate the neutron spectra around the GE-PETtrace 8 cyclotron of the CDTN/CNEN, Brazil

The Monte Carlo code MCNPX was used to calculate the neutron spectra in 4 points around the targets of the CDTN/CNEN cyclotron, Belo Horizonte, Brazil, during the production of the (18)FDG. Simulated data were compared with experimental data obtained with a Bonner multisphere spectrometry system (BSS) using TLD-600 and TLD-700 and the unfolding codes BUNKIUT, BUMS and NSDUAZ. In general, simulated spectra disagreed with those obtained by experimental means by a factor as high as 14. Measurements performed with a doserate meter in other 3 more shielded points, showed also an overestimation of the ambient dose equivalent rate by a factor as high as 20 in comparison with simulated results. Results are not conclusive and a more refined study is necessary. However, neutron emission rate of the source-term of radiation must be investigated and an special caution must be taken in the experimental measurements, by discriminating of the target selected for the irradiations and utilizing a matrix response suitable for the passive detectors (e.g. TLD) utilized in the experiments, instead of a matrix response (e.g. UTA4) developed for scintillation detectors.

Use of a TLD-based multisphere spectrometry system to measure the neutron spectra around a not-self-shielded PET cyclotron: Preliminary results

In the present work, we utilized the BSS system with TLD-600 and TLD-700 to measure the neutron spectra around the GE-PETtrace 8 cyclotron of the Development Centre of Nuclear Technology (CDTN/CNEN) in Belo Horizonte, Brazil. The cyclotron is capable of accelerating protons up to 16.5 MeV, to production of fluorine-18. Four points inside the bunker of the cyclotron were studied. Two points in front of the primary radiation beam and other two opposed to the primary radiation beam. The measurements were unfolded with the BUMS and the NSDUAZ computer codes. The dosimetric quantities obtained were in agreement with the other published data and were coherent with the expected from theoretical estimates obtained from source term informed by the manufacturer of the cyclotron.

Radiologic assessment of a self-shield with boron-containing water for a compact medical cyclotron

The cyclotron at our hospital has a self-shield of boron-containing water. The amount of induced radioactivity in the boron-containing water shield of a compact medical cyclotron has not yet been reported. In this study, we measured the photon and neutron dose rates outside the self-shield during cyclotron operation. We estimated the induced radioactivities of the boron-containing water used for the self-shield and then measured them. We estimated the activation of concrete outside the self-shield in the cyclotron laboratory. The thermal neutron flux during cyclotron operation was estimated to be 4.72 × 10² cm⁻² s⁻¹, and the activation of concrete in a cyclotron laboratory was about three orders of magnitude lower than the clearance level of RS-G-1.7 (IAEA). The activity concentration of the boron-containing water did not exceed the concentration limit for radioactive isotopes in drainage in Japan and the exemption level for Basic Safety Standards. Consequently, the boron-containing water is treatable as non-radioactive waste. Neutrons were effectively shielded by the self-shield during cyclotron operation.

Distribution of thermal neutron flux around a PET cyclotron

The number of positron emission tomography (PET) examinations has greatly increased world-wide. Since positron emission nuclides for the PET examinations have short half-lives, they are mainly produced using on-site cyclotrons. During the production of the nuclides, significant quantities of neutrons are generated from the cyclotrons. Neutrons have potential to activate the materials around the cyclotrons and cause exposure to the staff. To investigate quantities and distribution of the thermal neutrons, thermal neutron fluxes were measured around a PET cyclotron in a laboratory associating with a hospital. The cyclotron accelerates protons up to 18 MeV, and the mean particle current is 20 μA. The neutron fluxes were measured during both 18F production and C production. Gold foils and thermoluminescent dosimeter (TLD) badges were used to measure the neutron fluxes. The neutron fluxes in the target box averaged 9.3 × 10(6) cm(-2) s(-1) and 1.7 × 10(6) cm(-2) s(-1) during 18F and 11C production, respectively. Those in the cyclotron room averaged 4.1 × 10(5) cm(-2) s(-1) and 1.2 × 10(5) cm(-2) s(-1), respectively. Those outside the concrete wall shielding were estimated as being equal to or less than ∼3 cm s, which corresponded to 0.1 μSv h(-1) in effective dose. The neutron fluxes outside the concrete shielding were confirmed to be quite low compared to the legal limit.