Radiological study of a wastewater treatment plant associated with radioiodine therapy at a hospital in West Java, Indonesia

This study was carried out in the NM unit of a referral hospital in West Java, Indonesia. The unit utilized two adjacent buildings in its service operations, with building one being old and the waste piping system blueprint was no longer available. Meanwhile, the second was a new six-story building that utilized the basement as the location for NM service. The study focus was on the management of radioiodine therapy patient's excreta in both places. In the NM installation of building one, liquid waste was not directed to the hospital's centralized WWTP, while in NM installation of building two, the waste was temporarily settled before entering the hospital's WWTP. I-131 Therapy is carried out in buildings one and two, adjusted for the availability of isolation beds in each building. In general, the workload of therapy places can be seen from the weekly patient data tabulation in each of the following buildings:

There are 14 isolation rooms in building one that are equipped with a variety of treatment options for patients in building two's liquid waste. The waste treatment in building one is terminated at the decay tank located in the parking lot, and it does not go on to the hospital's WWTP. Building two has four isolation rooms, each with its own advanced waste treatment system that connects to the hospital's WWTP, which ultimately empties into local rivers. Because of issues with water seepage on the walls of the NM room, the infiltration well serves the purpose of a well to manage the runoff of surface water that is caused by rain. This problem was caused by the NM room. In order to determine whether or not there is a leak in the decay tank located in building two, measurements of I-131 activity were taken in the infiltration well.

As depicted in figure 1, decay tank A receives waste from the isolation and post-injection rooms; decay tank B receives waste from the hot lab and four uptake rooms; and decay tank C may receive waste from patients who do not understand the recommendations for excretion procedures after and/or while waiting for patient discharge, causing radioiodine concentrations in the main sewage pit.

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Sampling was carried out on liquid and air waste around the NM unit. Furthermore, the number of sampling points was determined using a systematic method. The process was performed by considering the flow and storage of patients waste from points A to I and from points 1–6 for liquid and air samples, respectively, as shown in figure 1.

Liquid waste was collected using the WS750 Wastewater-Stormwater Sampler with a 1 l Marinelli beaker tube container based on the required capacity for contaminated sample analysis for radionuclide counters. Iodine-131 therapy with a dose above 1110 MBq requires 3 × 24 h until the patient can be discharged, resulting in two therapy terms per week. The weekly therapy schedule begins on Monday with a scheduled patient release on Wednesday. The second term begins on Wednesday afternoon with a scheduled patient release on Friday. At storage tanks A, B, and C, sampling was carried out in sequence every 12 h, starting at 12 h on the first day of therapy and ending at 48 h (day a term of therapy I-131). At locations D to I, sampling was performed only at 48 h to determine the settling or flow of waste containing I-131.

Air samples were taken using DFHV-1E High Volume Air Sampler (220–240 VAC) with a Charcoal filter and was counted on the high-purity germanium (HPGe) detector. The procedure of extraction efficiency determining CH3I-131 follows the American Standard for Testing and Materials (ASTM D 3803-79, 1979). Determination of CH3I-131 absorption efficiency of an activated charcoal filter is done at a temperature range of 30 °C–45 °C, the flow rate of 13 lpm to 26.3 lpm, and air humidity of 30%–90%. Sampling medium at locations marked 1–6, as shown in figure 1. These six points have the potential for the spread of I-131 in the air due to their proximity to the service unit and the temporary processing site of patient's liquid waste.

Equipment used for measuring radionuclides in wastewater was a gamma spectrometer counting system with an Ortec HPGe detector. The system consisted of an HPGe detector placed in a 10 cm thick shielding system. Furthermore, the pre-amplifier was attached to the detector body. The amplifier, HV bias supply, and multi-channel analyser were integrated into a single module called DespecPro. The gamma spectrometer system was operated with Maestro or Gamma Vision software, as schematically shown in figure 2. The device was calibrated with a standard mixed source consisting of radionuclides Co-60, Ba-133, Cs-134, Cs-137, Pb-210, and Am-241 that were traceable to the IAEA. Prepared samples were then placed in containers, such as vials or Marinelli beaker, labelled, and wrapped in plastic to prevent contamination. The radionuclide concentration in the sample was measured using a gamma spectrometer equipped with an HPGe detector calibrated using a standard source [17].

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Calibrating energy and efficiency used equations [18]:

Standard source decay,

$$\begin{equation}{A_t} = {A_0}{{\text{e}}^{ - \lambda t\,}} = \,{A_0}{{\text{e}}^{ - 0.693t/T}}.\end{equation} \tag{ 1 }$$

With:

${A_{t\,}}\,$:activity during counting (Bq)

${A_0}$: initial activity (Bq)

$t$: delay time (days)

T: half-life (days)

Energy calibration,

$$\begin{equation}Y = a + bX.\end{equation} \tag{ 2 }$$

With:

Y: gamma energy (keV)

a and b: linear constant numbers

X: channel number

Efficiency calibration,

$$\begin{equation}{\varepsilon _{\gamma \,}} = \,\frac{{({N_{s\,}}/\,{t_s}\,\, - \,{N_{{\text{BG}}\,}}/\,{t_{{\text{BG}}}}}}{{{A_{t\,.\,\,\,}}{p_\gamma }}}.\end{equation} \tag{ 3 }$$

With:

_γ : counting efficiency (%)

N_s : standard count

N_BG: background count

t_s : standard count time (seconds)

t_BG: background count time (seconds)

A_t : standard source activity at the time of count (Bq)

p_γ : gamma energy abundance (%)

To calculate the concentration of radionuclides contained in the sample (C_Sp), the following equation was used [19]:

$$\begin{equation}{C_{{\text{Sp}}\,}} = \,{C_{{\text{avg}}\,}} \pm \,uT.\end{equation} \tag{ 4 }$$

With:

C_Sp: the concentration of radioactive substances in the sample (Bq/lt or Bq/kg)

C_avg: average concentration of radioactive substances in the sample (Bq/lt or Bq/kg)

uT : measurement uncertainty (Bq lt⁻¹ or Bq kg⁻¹)

$$\begin{equation}{C_{{\text{avg}}\,}} = \,\frac{{({N_{{\text{sp}}\,}} - \,{t_{{\text{BG}}}}}}{{{\varepsilon _{\gamma \,}}\,.{p_\gamma }.\,{W_{{\text{sp}}\,\,\,}}}}.\end{equation} \tag{ 5 }$$

With:

N_Sp: sample count rate (cps)

N_BG: background count rate (cps)

_γ : efficiency at gamma energy (%)

p_γ : yield of gamma energy (%)

W_Sp: sample volume or weight (lt or kg)

$$\begin{equation}uT = \,{C_{{\text{avg}}\,}}x\,\sqrt {{{\left( {\frac{{{U_N}}}{{{N_{{\text{Sp}}}}}}} \right)}^2} + {{\left( {\frac{{{U_\varepsilon }}}{\varepsilon }} \right)}^2} + \,} {\left( {\frac{{{U_p}}}{p}} \right)^2} + \,{\left( {\frac{{{U_w}}}{{{W_{{\text{Sp}}}}}}} \right)^2}\,.\end{equation} \tag{ 6 }$$

With:

${U_N}$: sample count uncertainty (%)

$uT$: efficiency uncertainty at gamma energy (%)

${U_p}$: yield uncertainty (%)

${U_w}$: sample volume uncertainty (%)

k: 1

The minimum detectable concentration (MDC) for a gamma spectrometer system was influenced by counting efficiency, background count rate, and sample weight. To calculate the MDC with a 95% confidence level, the following equation [18] was used

$$\begin{equation}{\text{MDC}} = 4.66\,\frac{{\sqrt {\frac{{{N_{{\text{BG}}}}}}{{t_{{\text{BG}}}^2}}} }}{{{\varepsilon _{\gamma \,}}\,.\,\,{p_\gamma }\,.\,\,{F_{k\,\,\,}}.\,\,w}}.\end{equation} \tag{ 7 }$$

With:

MDC: MDC (Bq/lt or Bq/kg)

N_B : background count rate (cps)

t_B : background count time (seconds)

_γ : count efficiency (%)

p_γ : gamma energy abundance (%)

F_k : self-absorption correction factor (when sample ρ is different from the standard ρ)

w: sample volume or weight (lt or kg)

The analysis method used for liquid waste samples was Gamma Spectrometry based on SOP 004.003/KL, referring to IAEA Technical Report Series No. 295 and SNI ISO 10703:2009 with the lowest detectable concentration of I-131 at 2.6 × 10⁴ Bq m⁻³.

As part of the I-131 therapy, three decay tanks (A, B and C) had their liquid waste concentrations read every 12 h for three days. Starting with the first 12 h the patient received I-131 therapy and continuing through the 48th hour with the annotations A1, A2, A3 and A4, as well as tanks B and C, the results are shown in table 1. The fluctuations in the readings of the concentration of I-131 in the waste samples from each sampling point are explained in figure 3(b).

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Point (D) was a temporary waste collector for the subsequent flow to point G and ends at H. The absence of deposition/retention of liquid waste at point D indicated that the exposure spread along the path. What is being referred to here is the high concentration of I-131 that may be found in tank D due to the fact that it receives a direct flow from tanks A, B, and C. However, the concentration at point H is higher than the amount that is considered acceptable for radioactivity in the environment. As a result, it is necessary to take action in order to lower the value of the concentration of I-31. The percentage of I-131 removal from liquid waste at point H (WWTP outlet) with the existing NM patients waste treatment system gave a high efficiency. The efficiency from decay tanks was 94% for A to D, 34% for B to D, and 99% for the D processor to the WWTP outlet.

The measurement results of I-131 concentration in the air at six points around NM unit are presented in table 2. Air sampling was carried out at six points around NM buildings one and two, as indicated by numbers 1–6 in figure 1. The I-131 concentration was above the environmental radioactivity standard at points 4, 5 and 6 as shown at figure 4. Furthermore, point 4 was the Hotlab where radiopharmaceutical preparation was performed before therapy services, and 5 was patients isolation room post-therapy. Point 6 was an outdoor location used as a parking lot located between the radiopharmaceutical storage room and the final waste from radioiodine therapy patients isolation room in building one.

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Similar studies have been carried out in other countries to analyse the safety of NM waste management, particularly from patients excretions. Spain was one of the countries that have implemented reporting requirements for permission values and environmental radioactivity monitoring. Furthermore, monitoring was carried out by requiring every NM waste-generating facility to report the measurement results of radionuclide concentration values before entering the municipal WWTP system. Other controls were also carried out by taking samples in seven hospitals WWTP in the Barcelona Metropolitan Area from effluents and dried sludge.

The process was then continued with measurement of the concentration of I-131, Tc-99 m, In-111, Ga-67, and I-123 using gamma spectroscopy, which showed values of 0.2–4.4 Bq l⁻¹ for I-131 and 5–50 Bq l⁻¹ for Tc-99 m. Other samples were taken from municipal WWTP in the metropolitan Barcelona area, which received wastewater from several hospitals in the city. The results showed the maximum concentration found for each radionuclide in wastewater and sludge, and the activity levels were still below the specified limit [20]. Another method was carried out in Malaysia, which has also implemented reporting requirements for monitoring the results of NM waste management in waste-generating facilities. This method was carried out by measuring the exposure rate around the radio-pharmaceutical waste processor at a distance of 1.2–3 m from the collection tanks. The highest exposure rate was 177.00 µR h⁻¹ (1552 µSv h⁻¹) in a full ward condition for two consecutive weeks, which exceeded the community dose limit. An increase in I-131 activity from the samples used caused fluctuations of 10–25 mCi or 37 × 10⁷–92.5 × 10⁷ Bq [21]. A previous study showed differences in the effectiveness of using waste processing systems in reducing the activity concentration of radionuclides originating from NM. Furthermore, the number of patients affected the effectiveness of the processing system. NM building one is a three-story building built in 1970 with simple maintenance, both for the building and the WWTP which uses the overflow method in three holding tanks for patients waste treatment.

From the measurement results of I-131 concentration at point H, which was the WWTP outflow towards the urban drainage system and/or municipal water supply, a scenario model of exposure to the public through consumed groundwater, vegetables/fruits, and inhalation was performed. Due to its physicochemical properties, I-131 can easily enter the environment due to problems in standard operating procedure treatment or waste management. The exposure pathways can be described by the following scenario in figure 5.

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The concentration at point H, which was the discharge point of the hospital wastewater treatment, showed a value above the environmental radioactivity standard. The scenario was based on the assumption that WWTP effluent entered the urban drainage system flowing into a river with a distance of 0.5–2 km from point H.

The river became a source for the water company and provided irrigation water for the surrounding communities. Therefore, the community, directly and indirectly, consumed materials containing I-131. The risk assessment was performed assuming a 'single' release of I-131 through the WWTP effluent of the hospital for a 24 hour period, with data gathered from measurements of I-131 concentrations in 4 decay tanks in the NM department. This highlights the significance of I-131 monitoring in hospital WWTPs prior to the discharge of fully treated effluent.

The initial activity concentration value at point H was 0.04 Bq l⁻¹ or 37 670 Bq m⁻³. The graph enlargement of the cumulative dose analysis through RESRAD ONSITE is shown in figure 6.

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The I-131 dose was at its highest value in the middle of the 2nd and 3rd months and continued to decline. The cancer risk analysis results with the assumption that the community consumed water, vegetables, and possibly ingested materials containing I-131 are shown in figure 7. Figure 7 shows a decline in cancer risk for workers and the community after the 2nd month. The risk calculation performed by RESRAD was accomplished in the low-dose-rate program and risk factor. It was assumed that there was a linear relationship between dose and risk extrapolated from high-dose and dose-rate data.

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Another extrapolation method into the low-dose range can produce higher or lower estimates of cancer likelihood. However, studies on humans exposed to low doses were not sufficient to show the actual level of threat. According to the Committee on Interagency Radiation Research and Policy Coordination (CIRRPC), there was a scientific uncertainty on cancer risk in the low-dose range below epidemiological knowledge and the possibility that the likelihood cannot be ruled out (CIRRPC 1992). One radiological risk assessment method used a dose conversion factor (DCF) to estimate the cost-effective dose equivalent.

The measurement results in figure 5 showed that point six was above the environmental radioactivity standard value for air, namely 2.7 × 10⁻² Bq m⁻³. Considering that I-131 was classified as a radionuclide in the high radiotoxicity group due to its ability to be inhaled as a gas or ingested through food or water, and its solubility in water/alcohol, it was directly absorbed with an effective dose per unit intake by adult age at 2.0 × 10⁻³ Sv/Bq. Risk analysis of exposure to workers using air sample test data was carried out by calculating the dosage received by workers annually using the conversion factor for I-131, with the results shown in table 3. It was taken into consideration that the places where concentrations of iodine-131 were found that exceeded the standard level of environmental radioactivity were hot lab rooms and patient isolation rooms where medical workers work full time 8 h day⁻¹, and parking areas where parking workers work an average of 4 h day⁻¹ considering the fact that parking attendants move around or even work in shifts. Based on this information, scenarios were created for two distinct work periods. The dose received at the 3 far air sampling points was below the limit value determined for radiation workers and the public namely 20 mSv y⁻¹ and 1 mSv y⁻¹ respectively as shown at table 4.