Kajian pemantauan dosis neutron terhadap pekerja radiasi pada pengoperasian Linac dengan energi foton ≥ 10 MV

The use of Linac for radiotherapy is starting to use a lot of high-energy photons of 10 MV; in addition, some use 15 MV for patient therapy in routine use, there is also the use of 6 MV. The purpose of this study is to obtain an overview and information of the neutron dose that has the potential to provide additional doses for radiation workers operating the Linac 10 MV aircraft. Based on the Regulation of the Head of BAPETEN No. 3 of 2013, Article 48 paragraph (2) states that in the operation of Linac with X-ray photon energies above 10 MV, must coat the shield wall with a neutron-absorbing material. The statement follows the IAEA-TecDoc 1891 that neutrons will have the potential to have a significant radiological impact on workers if routinely operated at energies above 10 MV, so must consider protection for workers. The results of a survey from 27 hospitals, obtained information through filling out questionnaires and discussions and validated with B@LIS Pendora, it found that the trend of annual doses received by each profession in the operation of Linac 6 MV, 10 MV, and 15 MV was less than one mSv, only partially small worker dose that is above one mSv (above the 90th percentile). This study concluded that the presence of neutrons in Linacs up to 10 MV was deemed not to have a significant radiological impact on workers. The recommended criteria/mechanism for monitoring worker neutron doses in Linacs up to 10 MV could be based on if the safety study results obtained a dose received by workers 1.5 mSv/year. Then, there is no need to monitor the neutron dose. If the measurement results of exposure to neutron and gama radiation around the Linac space are 7.5 microSv/hour, there is no need for neutron monitoring. In Linacs above 10 MV, if the annual effective dose is 1.5 mSv/year, there is no need to monitor the dose of special neutron personnel. Still, routine radiation exposure monitoring may be considered every two years. Keywords: Neutron Dose, Radiation Worker, Linac, Dose Monitoring.

Common sources of radiations in a medical environment

Use of radiation is now a days so common in most of the tertiary care hospitals for diagnostic and therapeutic purpose. The ionizing radiation provides many benefits in both diagnostic as well as therapeutic interventions, but they are also potential harmful. Radiation risks, exposure and mitigation strategies should always be in mind while using to an individual (public, radiation worker, and patient) and the environment should not exceed the prescribed safe limits. Regular monitoring of hospital area and radiation workers is mandatory to assess the quality of radiation safety. This review article emphasis on radiation risks, exposure and prevention and treatment strategies.

Material Analysis of Lead Aprons Using Radiography Non-Destructive Testing

Lead Apron is a Personal Protective Equipment (PPE) against the effect of Ionizing Radiation such as X-ray. It is essential for the radiation worker to wear Radiation Protection Equipment during commissions involving ionizing radiation. In Pekanbaru, Indonesia the most common radiation worker is Radiographer which help in hospital for diagnose. In this study the Lead Apron analyzed were 6 apron which suspected to have fault due to its inappropriate tend using NDT radiography methods. Radiography methods have advantage of graphic presentation of object unlike any NDT-methods. The image produce from radiography were analyzed using Computed Radiography (CR) and measured the defection of the material. There was only 3 of 6 Lead Apron appropriate for radiation protection.

Studi Paparan Radiasi pada Pekerja Radiasi Cathlab dengan Menggunakan My Dose Mini sebagai Upaya Keselamatan Radiasi di RSUP Adam Malik Medan

Cathlab radiation workers, when performing interventional procedures, are at high risk of the effects of radiation exposure. The risk of radiation exposure is deterministic and stochastic biological effects. Therefore, radiation exposure studies of radiation workers at the cath lab were conducted to determine the value of radiation exposure received. This radiation exposure study was conducted by measuring and recording radiation exposure doses received by radiation workers. Measurements are made when the radiation officer performs the intervention procedure. The research was carried out for one month in the cath lab room of the Adam Malik General Hospital, Medan. The modalities used are GE Medical System Interventional Fluoroscopy and Phillips Allura Xper FD20. The dosimeter used is “my dose mini”, which is placed inside a shield or apron worn by radiation workers. The size of the apron shield used is 0.50 mm Pb at the front and 0.25 mm Pb at the rear. Radiation officers whose radiation exposure dose was measured consisted of 10 doctors, 11 nurses, and one radiographer. Each inspection procedure of each radiation worker has a different distance, time, and shield from the radiation source. The measurement of radiation exposure dose is (1-59 μSv) for doctors, (1-58 μSv) for nurses, and 1 μSv for radiographers. To protect against radiation must pay attention to the factors of time, distance, and shielding. Ways that can do are to avoid being close to radiation sources for too long, keep a space at a safe level from radiation, and use shields such as Pb-coated aprons, use Pb gloves, Pb goggles, and thyroid protectors. The amount of radiation exposure dose received by each radiation worker at the time of measurement is still within the tolerance limit. The Nuclear Energy Regulatory Agency (BAPETEN) regulation, which the International Commission recommends on Radiological Protection (ICRP), is 20 mSv/year. The results of this study are expected to be used as input for improving the quality of service for monitoring radiation exposure doses in the Cathlab and as reference material for further research.

A Cohort Study of Korean Radiation Workers: Baseline Characteristics of Participants

The Korean Radiation Worker Study investigated the health effects of protracted low-dose radiation among nuclear-related occupations in the Nuclear Safety and Security Commission in Korea. From 2016–2017, 20,608 workers were enrolled (86.5% men and 30.7% nuclear power plant workers). The mean cumulative dose ± standard deviation between 1984 and 2017 (1st quarter) was 11.8 ± 28.8 (range 0–417) mSv. Doses below recording level (≤0.1 mSv) were reported in 7901 (38.3%) cases; 431 (2%) had cumulative doses ≥100 mSv. From 1999–2016, 212 cancers (189 men, 23 women) occurred; thyroid cancer predominated (39.2%, 72 men, 11 women). In men, the standardized incidence ratio (SIR) for all cancers was significantly decreased (SIR = 0.76, 95% CI 0.66–0.88); however, that for thyroid cancer was significantly increased (SIR = 1.94, 95% CI 1.54–2.44). Compared to the non-exposed group (≤0.1 mSv), the relative risk (RR) in the exposed group (>0.1 mSv) after adjusting for sex, attained age, smoking status, and duration of employment was 0.82 (95% CI 0.60–1.12) for all cancers and 0.83 (95% CI 0.49–1.83) for thyroid cancer. The preliminary findings from this baseline study with a shorter follow-up than the latency period for solid cancer cannot exclude possible associations between radiation doses and cancer risk.

MONTE CARLO N PARTICLE EXTENDED (MCNPX) RADIATION SHIELD MODELLING ON BORON NEUTRON CAPTURE THERAPY FACILITY USING D-D NEUTRON GENERATOR 2.4 MeV

Based Studies were carried out to analyze the internal dose of radiation for workers at Boron Neutron Capture Therapy (BNCT) facility base on Cyclotron 30 MeV with BSA and a room that was actually designed before. This internal dose analyzation included interaction between neutrons and air. The air contained N2 (72%), O2 (20%), Ar (0.93%), CO2, Neon, Kripton, Xenon, Helium and Methane. That internal dose to the worker should be below the dose limit for radiation workers which is an amount of 20 mSv/years. From the particles that are present in the air, only Nitrogen and Argon can change into radioactive element. Nitrogen-14 activated to Carbon-14, Nitrogen-15 activated to Nitrogen-16, and Argon-40 activated to Argon-41. Calculation using tally facility in Monte Carlo N Particle version Extended (MCNPX) program for calculated Neutron flux in the air 3.16x107 Neutron/cm2s. The room design in the cancer facility has a measurement of 200 cm in length, 200 cm in width, and 166.40 cm in height. Neutron flux can be used to calculate the reaction rate which is 80.1x10-2 reaction/cm3s for carbon-14 and 8.75x10-5 reaction/cm3s. The internal dose exposed to the radiation worker is 9.08E-9 µSv.