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.

Radiation safety analysis for the A-BNCT facility in Korea

A Proton Accelerator based Boron Neutron Capture Therapy (A-BNCT) facility is under development in Korea. Neutron beams for treatment are produced from a beryllium (Be) target and an 8 mA, 10 MeV proton beam. The purpose of the research is a radiation shielding analysis and an activation analysis for the facility design satisfying the radiation safety requirements as well as obtaining an operating license for the radiation facility according to a domestic nuclear commissioning procedure. The radiation shielding analysis was performed using the MCNPX computational particle transport code. The radiation source terms in the facility were evaluated and utilized in the shielding calculations. The minimum concrete thickness satisfying the designated dose rate of 5 μSv/h for the worker's area and 0.25 μSv/h for the public area were estimated and applied to the design. For an assessment of the radiation safety inside the facility, the dose rates were evaluated at several positions, such as behind the shielding door, around the primary barriers near the radiation sources, and in the penetrations of the ducts. The dose rate distribution was mapped for verification of the radiation safety for the entire facility. An activation analysis was carried out for the concrete walls, air, target assembly, beryllium target, and cooling water using FISPACT-2010 code. Concentrations of the activation products and dose rate induced by the radionuclides after shutdown were evaluated for the purpose of safe operation of the facility. The results were reviewed with the radiation safety regulations in Korea. As a result, it was proved that the final facility design satisfies the safety requirements.

Application of the National Ignition Facility distinguishable-from-background program to accelerator facilities at Lawrence Livermore National Laboratory

Lawrence Livermore National Laboratory must control potentially activated materials and equipment in accordance with U.S. Department of Energy (DOE) Order 458.1, Radiation Protection of the Public and the Environment, which requires DOE approval of the process used to release volumetrically contaminated personal property and establishes a dose constraint of 10 µSv y(-1) (1 mrem y(-1)) for clearance of such property. The National Ignition Facility at Lawrence Livermore National Laboratory developed a technical basis document and protocol for determining the radiological status of property that is potentially activated from exposure to neutron radiation produced via fusion of tritium and deuterium. The technical basis included assessment of the neutron energy, the type of materials potentially exposed and the likely activation products, and the sensitivity of radiation detectors used to survey the property. This paper evaluates the National Ignition Facility technical basis document for applicability to the release of property from Lawrence Livermore National Laboratory's various accelerator facilities considering the different types of particles accelerated, radiations produced, and resultant activation products. Extensive process knowledge regarding the accelerators' operations, accompanied by years of routine surveys, provides an excellent characterization of these facilities. Activation studies conducted at the Stanford Linear Accelerator and the High Energy Accelerator Research Organization in Japan corroborate that the long-lived radionuclides produced at accelerator facilities are of the same variety produced at the National Ignition Facility. Consequently, Lawrence Livermore National Laboratory concludes that the release protocol developed for the National Ignition Facility can be used appropriately at all its accelerator facilities.

AMS analysis of 36Cl induced in concrete of accelerator facilities

Accelerator mass spectrometry (AMS) was applied to the analysis of 36Cl induced in concrete samples obtained from accelerator facilities. In order to use a small amount of concrete sample and to separate chlorine as pure as possible, an improvement of separation process was developped. Chlorine was extracted from 1–5 g of concrete into 0.01 M nitric acid in a pressurized decomposition vessel. After determining chlorine using ion chromatography, a certain amount of NaCl solution was added to obtain sufficient amounts of AgCl precipitate and to dilute to a suitable isotope ratio of 36Cl to 35Cl (36Cl/35Cl) for the AMS (10-12–10-10). A careful purification procedure was applied to reduce 36S interference in AMS. Good reproducibility and small error throughout the chemical process for sample preparation was attained. Depth profiles of 36Cl/35Cl in concrete of a medium-energy cyclotron were measured by the developed method and compared with the results of γ-emitters induced by thermal neutrons. Since it was confirmed that 36Cl was produced by thermal neutron capture of 35Cl, the thermal neutron fluence irradiated during accelerator operation could be obtained using 36Cl/35Cl. In order to estimate the neutron fluences, the 36Cl/35Cl measurement by AMS is more useful than radioactivity measurements of other isotopes such as γ-emitters because AMS directly provides the isotope ratio and the half-life of 36Cl is very long.

Measurement of depth distributions of and induced in concrete shielding of an electron accelerator facility

The estimation of radioactivity induced in concrete shielding is important for the decommissioning of accelerator facilities. Concentrations of (3)H and (14)C in the concrete shielding of an electron linear accelerator were measured, and the depth distributions of (3)H and (14)C and gamma-ray emitters were discussed in relation to their formation reactions.