Evaluation of Radioactivity in Therapeutic Radiopharmaceutical Waste

Jung Ju Jo; Su Hyoung Lee; Beom Hoon Ki; Ho Jin Ryu; Tae Hwan Kim; Gi Sub Kim; Sang Kyu Lee; Dong Wook Kim; Kum Bae Kim; Sangrok Kim; Sang Hyoun Choi

doi:10.14316/pmp.2024.35.4.163

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Jo, Lee, Ki, Ryu, Kim, Kim, Lee, Kim, Kim, Kim, and Choi: Evaluation of Radioactivity in Therapeutic Radiopharmaceutical Waste

Original Article

Progress in Medical Physics 2024; 35(4): 163-171.

Published online: 31 December 2024

DOI: https://doi.org/10.14316/pmp.2024.35.4.163

Evaluation of Radioactivity in Therapeutic Radiopharmaceutical Waste

Jung Ju Jo^1,²

, Su Hyoung Lee³

, Beom Hoon Ki³

, Ho Jin Ryu³

, Tae Hwan Kim⁴

, Gi Sub Kim³

, Sang Kyu Lee⁵

, Dong Wook Kim⁶

, Kum Bae Kim^1,²

, Sangrok Kim^3,

, Sang Hyoun Choi^1,^2,

¹Radiological and Medico-Oncological Sciences, University of Science and Technology, Seoul, Korea

²Radiation Therapy Technology and Standards, Korea Institute of Radiological & Medical Sciences, Seoul, Korea

³Radiation Safety Section, Korea Institute of Radiological & Medical Sciences, Seoul, Korea

⁴Department of Medical Physics, Korea University, Sejong, Korea

⁵Department of Nuclear Medicine, Korea Institute of Radiological & Medical Sciences, Seoul, Korea

⁶Department of Radiation Oncology, Yonsei Cancer Center, Yonsei University College of Medicine, Seoul, Korea

Corresponding author Sang Hyoun Choi, (sh524mc@gmail.com), Tel: 82-2-970-1391, Fax: 82-2-978-2005, Corresponding author, Sangrok Kim, (kim@kirams.re.kr), Tel: 82-2-970-1346, Fax: 82-2-970-1963

Received 12 November 2024 Revised 22 November 2024 Accepted 11 December 2024

(open-access):

This is an open-access article distributed under the terms of the Creative Commons Attribution Non-Commercial License (http://creativecommons.org/licenses/by-nc/4.0) which permits unrestricted non-commercial use, distribution, and reproduction in any medium, provided the original work is properly cited.

Abstract

Purpose

This study aims to systematically analyze the radioactive waste generated from treatments using radioactive Iodine-131 (I-131), Lutetium-177 (Lu-177), and Actinium-225 (Ac-225) to facilitate safe waste management practices.

Methods

I-131 is primarily used in thyroid cancer treatment, while Lu-177 and Ac-225 are used to treat prostate cancer. Radioactive waste generated after these treatments was collected from patients at the Korea Cancer Center Hospital and categorized into clothing, slippers, syringes, and other items. The radioactivity concentration of each item was measured using a calibrated high-purity germanium detector. Using measurements, the self-disposal date of each waste item was calculated according to the permissible disposal levels defined by the Nuclear Safety and Security Commission (NSSC) under domestic nuclear safety regulations.

Results

For the I-131 radioactive waste, clothing, towels, and tableware exhibited high radioactivity concentrations, with most items exceeding the permissible self-disposal levels. Conversely, the type and quantity of waste generated from Lu-177 and Ac-225 that were intravenously injected were relatively minimal, with certain items below the self-disposal thresholds, enabling immediate disposal. For Ac-225, no permissible self-disposal concentration is specified by the NSSC, unlike other therapeutic nuclides. Hence, additional studies are required to establish clear guidelines.

Conclusion

These findings provide valuable data for optimizing radioactive waste management, potentially reducing disposal time and costs, minimizing radiation exposure, and enhancing hospital safety practices.

Keywords: Radiopharmaceutical waste, Self-disposal, Nuclear medicine, Radioactivity, High-purity germanium detector

Introduction

The incidence of cancer cases has increased by approximately 10% since 2020, according to the Ministry of Health and Welfare report for 2021. Notably, the incidence of thyroid and prostate cancer has steadily increased globally across both sexes since 2015. Thyroid cancer cases have a particularly high domestic incidence. Similarly, prostate cancer is a significant health threat, especially in older males [1,2]. Treatment options for these cancers include surgical resection, radiation therapy, and chemotherapy. Nuclear medicine using radioisotopes is an effective therapeutic option, specifically for managing thyroid cancer, which commonly involves the administration of radioactive Iodine-131 (I-131) to target the residual thyroid tissues and cancer cells. I-131 emits both gamma and beta rays, enabling its use for diagnostic and therapeutic purposes [3 -5]. In addition, Lutetium-177 (Lu-177), used for prostate cancer treatment, emits gamma and beta rays, while Actinium-225 (Ac-225) emits gamma and alpha rays, effectively targeting prostate cancer cells [6 -9]. Although radionuclide-based therapies are highly effective, radioactive waste is inevitably produced during these treatments [10].

In nuclear medicine, radioactive waste is generated when the administered radioactive substances are excreted or absorbed by various items in contact with the patients, making its management essential in outpatient and inpatient areas. The Nuclear Safety Act defines the classification and self-disposal standards for managing radioactive waste in South Korea. Accordingly, hospitals implement systematic protocols for waste generation and disposal and develop self-disposal procedures compliant with the guidelines of the Nuclear Safety and Security Commission [11]. Using these procedures, radioactive waste can be treated as regular waste after a specified decay period, although management practices and procedures may vary across hospitals [11,12]. Effective disposal of radioactive waste is crucial for operational efficiency and radiation safety in healthcare facilities [13,14]. Globally, several studies have assessed the self-disposability of nuclear medicine waste, following the guidelines of the International Atomic Energy Agency (IAEA), which include criteria for the measurement and storage periods to determine the self-disposal eligibility of this waste [15,16].

As the prescription methods and hospitalization requirements vary depending on the radiopharmaceutical used, the types and radioactivity of waste generated at the treatment facilities differ. To verify the permissible self-disposal concentration for each radionuclide, as per the Nuclear Safety Act, facilities should be equipped with radionuclide analysis devices such as high-purity germanium (HPGe) detectors that can measure specific radioactivity [17,18]. However, few hospitals have such devices, making precise measurements challenging.

This study aimed to analyze the types of radioactive waste generated by various facilities in South Korea using the therapeutic radioisotopes I-131, Lu-177, and Ac-225. Using a survey meter and an HPGe detector, we evaluated the contamination levels and specific radioactivity to assess the self-disposability of each radioactive waste type [3,19].

Materials and Methods

This study analyzed the specific radioactivity of waste generated from patients undergoing treatments with I-131 (number of groups [n]=3), Lu-177 (n=2), and Ac-225 (n=2), respectively, at the Korea Cancer Center Hospital to evaluate the potential for self-disposal. During I-131 inpatient treatment, the numbers of patients in each group varied, with three, eight, and seven patients in groups 1, 2, and 3, respectively. After the patients were discharged, the waste generated from each radionuclide was assessed for contamination using a RADEYE B20-ER survey meter (Thermo Fisher Scientific) and subdivided by specific waste types. The specific radioactivity of each classified waste was measured and analyzed using an HPGe coaxial p-type GC2518 model detector manufactured by Canberra (Fig. 1).

1. Collection and classification of the radioactive waste

After the patients were discharged, personnel wearing shoe covers and gloves collected the waste to minimize exposure to radioactive materials. The radioactive contamination in the waste was screened by measuring the dose rate at a 10-cm distance from the surface using a calibrated survey meter.

Waste with confirmed radioactive contamination was categorized into syringes, tableware, needles, tubes, plastics, clothing, and towels, while other waste that could not be clearly classified was categorized as combustible. Using tweezers and forceps, the classified radioactive waste was placed in 500-mL beakers, which were identical to those used during the HPGe calibration. The parameters were automatically adjusted during HPGe use.

Large radioactive waste materials, such as plastic items and slippers that could not fit into the 500-mL beaker, were cut into smaller pieces using scissors. When a large quantity of waste, such as syringes or combustibles, was present, it was divided and placed into multiple containers.

1) I-131 radioactive waste

I-131 is a radioactive iodine isotope with a half-life of 8 days. Patients undergoing treatment are isolated in the ward for 3 days following oral administration. Therefore, the radioactive waste generated can be classified into waste produced during administration or hospital stay and from items used by nurses and hospital maintenance staff, such as gloves and protective clothing. In this study, waste was categorized and evaluated as follows: clothing (underwear) worn by the patients during the hospital stay, towels, tableware (spoons and chopsticks), plastics (e.g., beverage and water bottles), and indoor slippers used in the isolation ward. Combustible items, such as tissues, gloves, and protective clothing, were classified accordingly (Fig. 2). For I-131 radioactive waste, the average specific activity was measured at different times across three groups, with an average of six patients per group.

2) Lu-177 and Ac-225 radioactive waste

The radioisotopes, Lu-177 and Ac-225, with a half-life of 6.6 and 9.9 days, respectively, are administered intravenously, and patients are discharged on the same day of treatment. Therefore, the radioactive waste generated includes waste from intravenous administration and items used by the medical staff, such as gloves and protective clothing. This waste generated was categorized and evaluated as follows: needles and lines, syringes, vials, and combustible items, such as alcohol swabs, protective clothing, and gloves (Fig. 3). For Lu-177 and Ac-225, the average specific radioactivity of the waste was measured at different times for the two patients.

2. HPGe measurement procedure

An HPGe detector, calibrated using the certified reference material from the Korea Research Institute of Standards and Science, was used, accounting for the size and height of the beaker. Before the HPGe measurement, the radioactive waste collected in a 500-mL beaker was assessed with a RADEYE B20-ER survey meter to predict the dead time during the HPGe measurement (Fig. 4a). Measurements were conducted immediately using the HPGe if the dose rate was 0.5 μSv/h or lower. For higher dose rates, additional measurements were postponed, allowing adequate decay time in a shielding container based on the half-life of the radionuclide. This indicates, based on empirical observation, that when the measured dose rate was 0.5 µSv/h, the dead time was within 3%.

If the dead time exceeded 3%, the measurement duration was extended to accurately assess the specific radioactivity considering each dead time. The net weight of the radioactive waste was measured using the tare function on an electronic scale to exclude the weight of the beaker. Additionally, a Ziplock bag was placed under the beaker before positioning it in the HPGe device to prevent contamination (Fig. 4b).

The recorded measurement parameters included the sample collection time, net sample weight, beaker volume and height, HPGe calibration details, and measurement duration.

3. Data processing and analysis

After the HPGe measurements, the predicted radionuclide information was entered into a radionuclide analysis library to analyze the energy peaks. The analysis software adjusted the data based on the radionuclide half-life and sample collection time (Fig. 5). The recorded data included the dead time, emission energy, yield per energy, and specific radioactivity (Bq/g). The average specific radioactivity was used to evaluate the potential for self-disposal by comparing the results with the half-life of each radionuclide and the permissible self-disposal concentration specified in the Nuclear Safety Act.

To calculate the self-disposability and disposal dates, information, such as the half-life of each radionuclide and the permissible concentration for self-disposal, was stored in the code using libraries, such as math and PyQt in Visual Studio Python ver. 2024.16.0. By inputting the radionuclide, sample acquisition date, and average specific radioactivity, the code calculated the disposal date according to Equation 1,

(1)

t = \frac{T_{\frac{1}{2}}}{\ln (2)} \times \ln (\frac{A_{1}}{A_{2}})

where t, T₁/₂, A₁, and A₂ represent the self-disposal date, half-life of the radionuclide, the measured specific radioactivity, and the permissible concentration for self-disposal, respectively.

Results

1. Specific radioactivity measurement results for the I-131 radioactive waste

The specific radioactivity of each group was compared with the permissible self-disposal concentration of 10 Bq/g using the average specific radioactivity for each item. The slipper pairs from six patients were cut in half, and the measured values were averaged to compare the specific radioactivity for each patient (Fig. 6).

The results indicated that all items exceeded the permissible self-disposal concentration except for Combustible 1 in Group 1. Based on the collection date, the calculated self-disposal periods were 88 days for clothing, 56 days for towels, 74 days for tableware, 36 days for plastics, and 48 days for other combustible items.

All slipper samples exceeded the permissible self-disposal concentrations except those from Patients 2 and 6. Based on the collection date, the average self-disposal period was 14 days (Fig. 7).

2. Specific radioactivity measurement results for the Lu-177|Ac-225 radioactive waste

The specific radioactivity of each patient treated with Lu-177 was compared with the permissible self-disposal concentration of 100 Bq/g using the average specific radioactivity for each item (Fig. 8).

For Lu-177, all the radioactive waste exceeded the permissible self-disposal concentration except for the combustible waste. For patient 2, the combustible waste showed no contamination. Radioactive waste that did not qualify for immediate self-disposal required an average of 100 days for needles and lines and 75 days for syringes to reach the permissible self-disposal levels. Vials could not be collected from patient 2, whereas those collected from patient 1 were found to be disposable after 147 days.

Because Ac-225 is an alpha-emitting radionuclide without a designated permissible self-disposal concentration, the dose conversion factors from IAEA Safety Standards Series No. GSR Part 3 used a dose conversion factor of 6.9 μSv/Bq, and the specific radioactivity was converted into a dose using the calculation formula from Equation 2, ensuring that it does not exceed the allowable dose of 10 µSv/year specified by the Nuclear Safety Act, allowing disposal after 274 days. In this Equation, I, A, and

e^{- λ t}

represent the dose rate, radioactivity, and radioactive decay over time, respectively. To verify this, Fig. 9 and Table 1 show the average specific radioactivity of each item and the self-disposal dates, respectively.

(2)

I = A \cdot e^{- λ t} \times D o s e C o n v e r s i o n F a c t o r

The results indicated that only the needles, tubes, and Syringe 2 were contaminated. The specific radioactivity values of the contaminated waste were converted to dose rates, which exceeded the permissible limit of 10 μSv/year. The needles and tubes reached permissible self-disposal levels after an average of 41 days, while for syringes, the average was 54 days.

Discussion

Herein, we analyzed the specific radioactivity and self-disposal dates of waste generated after treatments with I-131, Lu-177, and Ac-225. For I-131, which is administered orally and involves 3 days of inpatient treatment, the specific radioactivity was relatively high, with many items exceeding the permissible concentration for self-disposal. Clothing and towels exhibited high radioactivity, owing to the accumulation of radiopharmaceuticals excreted through the sweat of the patient. Clothing worn for a longer time had higher specific radioactivity than towels [20].

Tableware and plastics, including spoons, chopsticks, and plastic bottles, were contaminated due to contact with saliva or sweat on the hands of the patients. Specifically, tableware showed higher radioactivity than plastics due to close contact with the mouth. Combustible waste, such as tissues, gloves, and protective clothing, may have been contaminated by wiping radiopharmaceutical spills or through contact with sweat and saliva during the patient’s stay or discharge handling, although specific uses of these items were not documented [21].

The contamination level might vary depending on the purpose, resulting in a wide range of measured specific radioactivity values, from the self-disposable level of 5.4 Bq/g to a maximum recorded value of 7.8E+4 Bq/g. Slippers generally exhibited values similar to those of plastics, with an overall average of 5.62E+01 Bq/g. For Patients 2 and 6, the levels were within the self-disposable limit, possibly due to variations in sweat excretion and slipper usage durations among patients.

As Lu-177 and Ac-225 were administered intravenously, the amount of radioactive waste generated was relatively small. Radioactive waste can be classified as a low- or high-probability contamination waste. For combustible items, such as protective clothing and gloves used by medical staff or alcohol swabs used by patients, the likelihood of contamination is low, with measurements indicating either no contamination or levels suitable for immediate self-disposal. For the syringes, a significant difference in specific radioactivity was observed, with values ranging from 1.14E+02 to 2.70E+08 Bq/g. This difference is attributed to the varying amounts of residual radionuclides in the syringes used for administering radiopharmaceuticals compared with those used for saline injections or precise dosing.

In the case of Ac-225, an alpha-emitting radionuclide, the contamination levels were assessed using gamma measurements via HPGe rather than direct alpha measurements. Further research is required to examine the effects of the Ac-225 daughter nuclides, such as Francium-221 (Fr-221), Astatine-217 (At-217), and Bismuth-213 (Bi-213), and explore the alpha measurement methodologies. Additionally, as the data collected for each radionuclide in this study was limited, further data collection is necessary for a more comprehensive analysis and comparison.

Conclusions

In this study, the specific radioactivity of the waste generated after using therapeutic radiopharmaceuticals, such as I-131, Lu-177, and Ac-225, was measured, and the potential for self-disposal was evaluated. I-131-containing waste, generated following oral administration to patients, showed specific radioactivity that was within the permissible self-disposal concentration for most items. Clothing, towels, and tableware exhibited high radioactivity levels, indicating that a sufficient decay period is required for radioactive waste management.

For Lu-177 and Ac-225, administered intravenously, relatively small amounts of radioactive waste were generated, which were classified into needles with lines, syringes, vials, and combustible items (alcohol swabs and gloves). Some of this waste was within the self-disposal concentration, making immediate disposal possible. For Ac-225, the contamination levels were assessed indirectly through gamma measurements, indicating the need for additional studies on alpha measurement methodologies.

This study suggests that by identifying waste suitable for immediate self-disposal, hospitals can reduce the time and cost of radioactive waste management, enhance waste management procedures, minimize radiation exposure, and contribute to a safer healthcare environment. Future studies should focus on collecting additional data and improving the methods for measuring and analyzing alpha emissions. This is particularly useful for radionuclides such as Ac-225, which currently do not have specified allowable self-disposal concentrations in domestic guidelines, establishing more precise strategies for managing radioactive waste.

Notes

Funding

This research was supported by the National Research Council of Science & Technology (NST) grant by the Korea government (MSIT) (No. CAP22042-300) & Korea Institute of Radiological & Medical Sciences (KIRAMS) grant funded by the Korea government (Ministry of Sciences and ICT) (No. 50572-2024) & the Nuclear Safety Research Program through the Korea Foundation of Nuclear Safety (KoFONS) using the financial resource granted by the Nuclear Safety and Security Commission (NSSC) of the Republic of Korea(RS-2022-KN071220).

Conflicts of Interest

The authors have nothing to disclose.

Availability of Data and Materials

All relevant data are within the paper and its Supporting Information files.

Author Contributions

Conceptualization: Jung Ju Jo, Sang Hyoun Choi. Data curation: Jung Ju Jo, Tae Hwan Kim, Gi Sub Kim. Formal analysis: Jung Ju Jo, Su Hyoung Lee, Gi Sub Kim. Funding acquisition: Kum Bae Kim, Sang Hyoun Choi. Investigation: Dong Wook Kim, Sang Kyu Lee. Methodology: Jung Ju Jo, Sang Hyoun Choi, Sangrok Kim. Project administration: Sang Hyoun Choi, Sangrok Kim. Resources: Jung Ju Jo, Sang Hyoun Choi, Sangrok Kim. Software: Jung Ju Jo. Supervision: Kum Bae Kim, Sang Hyoun Choi, Sangrok Kim. Validation: Su Hyoung Lee, Beom Hoon Ki, Ho Jin Ryu. Visualization: Jung Ju Jo. Writing – original draft: Jung Ju Jo, Sang Hyoun Choi, Sangrok Kim. Writing – review & editing: Jung Ju Jo, Sang Hyoun Choi, Sangrok Kim.

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Fig. 1

A schematic diagram showing the measurement procedure.

Fig. 2

Types of Iodine-131-containing radioactive waste. (a) Plastic, (b) slippers, (c) tableware (spoon and chopsticks), and (d) towels.

Fig. 3

Types of waste containing Lutetium-177 and Actinium-225. (a) Combustible 1, (b) combustible 2, (c) needles and tubes, and (d) syringes.

Fig. 4

Measurement process of radiation dose rate and radioactivity. (a) Prediction of deadtime using survey meter. (b) High-purity germanium measurement using ziplock bags.

Fig. 5

HPGe peak analysis results. (a) I-131-containing combustibles, (b) Lu-177-containing vials, and (c) Ac-225-containing needles and tubes. HPGe, high-purity germanium; I-131, Iodine-131; Lu-177, Lutetium-177; Ac-225, Actinium-225.