Ambient Dose Equivalent Rate Research Articles

Determination of the ambient dose equivalent H*(10) at high-altitude cities using a NaI(Tl) scintillator.

The ambient dose equivalent rate H*(10) was determined at various locations in the metropolitan area of the cities of La Paz and El Alto in Bolivia. A NaI(Tl) scintillation detector was employed to infer the gamma-ray fluence, and the fluence was transformed in dose rates by means of the appropriate coefficients. In this paper, a full methodology for the estimation of dose rates associated with gamma rays emitted from the ground (terrestrial radiation) and the atmosphere is developed from measurements made with the scintillator instrument. The energy calibration and resolution of the apparatus are presented, followed by the estimation of its response and efficiency by means of Monte Carlo simulations. Lastly, the definition of H*(10) is used to determine the conversion coefficients needed to transform gamma-ray fluence in dose rates. With this methodology, rates for H*(10) around 123 ± 8 nSv/h were estimated using data collected in three city districts: Mallasa (3320ma.s.l.), San Antonio (3620ma.s.l.), and Zona Sur (3400ma.s.l.). The dose rates are believed to be representative of the terrestrial radiation fields encountered in the highlands of Bolivia.

Radiological protection of Young Scientist and Scientific Researcher trainees in controlled areas of the Argonaut reactor installation: analytical approach

Since the increasing use of ionizing radiations in medical and industrial proceedings, recommendations and actions have been necessaries to reduce the individual radiological risks. In the Nuclear Engineering Institute (IEN), there is the Argonauta research nuclear reactor (RARG) used for producing radionuclides, neutrongraphy studies and training young scientists (YST) and scientific researchers (SRT). During their trainings, these trainees are exposed to ionizing radiations meanly during the neutrongraphyc experiments when they need to adjust neutrongraphy plates in front the J9 channel with the RARG in critical condition. At this moment, they are exposed to neutron and gamma radiations in the RARG hall. This paper evaluated the gamma and neutron expositions that YST and SRT are submitted during the neutrongraphy experiments, by using area monitors to measure the ambient dose equivalent rates in two points near to J9 channel. Considering only this experiment, there was found that the annual individual effective dose could reach up to 2.5 mSv, which is more than 150% of the annual dose to member of the public (MP), but below the annual effective dose to occupationally exposed individual (OEI) as a worker. The question is if YST and SRT should be considered an OEI or a MP. This question is answered through an analytical approach of the regulations and recommendations. The conclusion is that YST and SRT shall be controlled as OEI, but must be submitted to MP’ restriction doses.

Performances of halide scintillators for the dosimetry based on gamma-ray spectrometry for environmental monitoring systems

High pressure ion chambers (HPIC) and NaI(Tl) scintillation detectors are widely used to monitor the ambient dose equivalent rate H*(10) within and around the Korean nuclear facilities. However, HPIC cannot provide spectrometric information and NaI(Tl) detector is limited in identifying nuclides, such as 131I, 134Cs, and 137Cs, released from nuclear facilities owing to its insufficient energy resolution. This study employed four halide scintillators – LaBr3(Ce), CeBr3, and SrI2(Eu) – to measure the ambient dose equivalent rate and detect gamma nuclides from measured energy spectrum. First, the pulse–shaping time in the signal processing unit was optimized for each scintillator. Second, energy resolution and counting efficiency were estimated for 137Cs and 60Co. Finally, an irradiation test was performed to estimate the dose rate. Based on these results, LaBr3(Ce) and NaI(Tl) were selected as in situ gamma spectrometry system for measuring environmental radiation, and field experiments were conducted near the Fukushima Daiichi nuclear power plant to measure the dose rate.

Indicators of Radioactive Contamination by Radionuclides for Samples of Plant Fertilizers and Pesticides

Abstract: This study aims to evaluate the radiation levels resulting from plant fertilizers and pesticides in Babylon City, Iraq. The specific activity of the studied radionuclides (e.g. 238U, 232Th, and 40K) was identified using the spectroscopy approach of the sodium-activated iodide thallium (NaI (Tl)) with dimensions of 3" × 3". The resulting data revealed that the specific activities for the aforementioned nuclides were 6.254 Bq/kg, 4.3694 Bq/kg, and 112.751 Bq/kg, respectively. The results indicated that the average radiation hazard metrics for fertilizers (i.e., gamma index, alpha index, absorbed dose rate, ambient dose equivalent rate, annual gonadal dose equivalent, and excess lifetime cancer risk) were lower than the reference levels. Keywords: Radiation hazard, Fertilizer, Pesticides, Risk parameters, Gamma activity.

Characterization of Neutron Irradiator as a Source of Reference Radiation Field for Calibrating Dosemeters and Doserate Meters at PSI Laboratory

A new neutron calibration laboratory has been installed at the Paul Scherrer Institute in Switzerland. The calibration laboratory has been equipped with a newly developed neutron irradiator with two calibration benches, remote control instruments and related ancillary equipment for calibrating dosemeters and doserate meters. The actual dimensions of the calibration laboratory are 2.0 m (H) X 6.8 m (W) X 12.6 m (L). On the steel grid at a height of 2 m from the basement floor, two calibration benches are installed. A transport channel is installed in the middle of the benches for the transfer of the radiation sources. The radiation sources are moved from the safe position of the irradiator to the reference irradiation position using an air pressure unit. The reference position is at a height of 1.2 m from the grid floor. The neutron irradiator is equipped with two Am-Be sources. Characterization of neutron radiation field has been calculated in detail by the MCNPX neutron transport code as well as energy distribution of neutrons and contribution of scattered neutrons. Experimentally, it was measured with a neutron/gamma digital spectrometric system equipped with a stilbene detector. The quality of the neutron radiation field has been characterized by ambient dose equivalent rate and the shape of the Am-Be spectra has been verified with the ISO-8529 standard.

Evaluation of fluence-to-dose response function of neutron survey meter using singular value decomposition method

This paper presents the evaluation of the fluence-to-ambient dose equivalent response function (referred to as F2D response function) of the Aloka TPS-451C neutron dose equivalent rate meter (hereafter referred to as a neutron meter) based on a singular value decomposition (SVD) approach. Measurements of neutron ambient dose equivalent rates, Ḣ∗(10), using the neutron meter were conducted in various neutron standard fields of 241Am-Be source with different neutron fluence rate spectra. A series of Ḣ∗(10) values and corresponding neutron fluence rate spectra were used as the input data for unfolding the F2D response function of the neutron meter. To verify the SVD-based method, the fluence response functions of a well-known Bonner sphere spectrometer system were unfolded using the SVD method, and the results were compared with those provided in the IAEA technical report No. 403. The SVD method was then applied to determine the F2D response function of the Aloka TPS-451C neutron meter with the use of the ICRP-74 F2D response function and that predicted in a previous work as initial guesses. The consistency between initial guesses and the SVD unfolded F2D response function of the Aloka TPS-451C neutron meter implies the reliability of the SVD method for determining the response functions of neutron meters.

Impacts of vertical variations in soil properties on H*(10) simulations for 137Cs deposition

Photon transport simulations based on the Monte Carlo method have played a crucial role in assessing and estimating the ambient dose equivalent rates H*(10), resulting from the deposition of 137Cs in soil following the nuclear power plant accident in Fukushima. However, a comprehensive examination of the effect of vertical variations in soil properties on the simulation outcomes has not yet been performed. Disregarding the vertical distribution of soil properties not only leads to potential inaccuracies in the shielding responses of soil layers but also in the determination of the radioactive source inventory, particularly when using the concentration data in Bq/kg. These oversights diminish the reliability of the simulation results. This study addresses several soil property factors that could potentially influence the simulation results, including variations in chemical composition induced by water content, bulk density profile, and estimated inventory profile, all evaluated through an examined simulation model. The results show that inappropriate assignment of the soil density profile can cause considerable errors in the H*(10) simulation outcomes. Furthermore, the sensitivity of H*(10) to variations in soil vertical density is analyzed, with the results indicating that H*(10) can be highly sensitive to changes in the bulk density of the top 0–5 cm soil layers. These results should facilitate the establishment of appropriate simulation strategies and support the reassessment of past simulation results.

Evaluation of radiation leakage in X-ray security inspection machine using a CZT spectrometer

Leakage radiation from X-ray security inspection machines is important, and measurement based on ionization chamber or scintillator detector is widely used. The leakage radiation is closely related to the size and the passing time of the luggage, the lead equivalent, the opening angle of the lead curtain, and the response time of the measuring instrument. To characterize the distribution of leakage radiation from the X-ray security inspection machine accurately, a small-volume CZT(CdZnTe) spectrometer was used to measure the energy spectra at a distance of 5 cm from the surface of the inspection machine. By designing and controlling the opening angle of the lead curtain according to the size of the luggage passing through the entrance and exit, the radiation dose was determined based on the measured energy spectra combined with the G(E)-function method. The results show that the maximum relative deviation between the air kerma rate and the ambient dose equivalent rate calculated by the G(E) function method with the standard dose rate does not exceed ±5%. The maximum relative deviation of the dose rate linear verification in the 137Cs radiation field is less than 2.5%. A calibrated CZT detector was utilized to measure the radiation leakage on the surface of the X-ray security inspection machine. It was discovered that the presence of the luggage items and the opening angle of the lead curtain will increase the leakage radiation dose on the surface of the security inspection machine system. This study provides a new approach for measuring scattered radiation of X-ray security inspection machines.

Neutron spectrum measurements near KAMINI reactor south beam vault door using nested neutron spectrometer.

KAlpakkam MINI reactor (KAMINI) is a 233U fuelled research reactor has various neutron irradiation locations for experimental purposes. The pit at the south beam end of KAMINI reactor is being extensively utilised for neutron attenuation experiments in prospective shielding materials as well as for neutron radiography. During reactor operation, it will be closed by a movable shield. A vault door is located above the shield and the movable shield is used to attenuate streaming neutrons and gamma-rays during reactor operation. Even with the shield, there exists significant dose because of streaming neutrons and gamma rays. Its variation depends on the power of the reactor. The neutron and gamma dose rates close to the south beam vault door have recently been found to be 275-300μSv/h and 175-200μSv/h, respectively, when the reactor is operating at 10kW. In order to characterise the streaming neutron spectra of vault door place for the first time, measurements are done using the Nested Neutron Spectrometer. Along with the neutron flux, neutron mean energy and ambient dose-equivalent rate are also measured and compared with earlier measurements carried out inside the south beam pit. It is observed that the presence of paraffin shield reduces the neutron average energy from 370 to 178keV. Apart from energy reduction, 10kW normalised neutron flux of south beam pit is also attenuated by the shield by 25000 times and it is found that the neutron spectrum of the measured location is also more thermalized. Neutron reference data of the location are generated.

New guidelines on radiation survey and sanitary assessment of residential, public and industrial buildings and facilities in terms of radiation safety indicators. Part 1

The paper (in two parts) presents an overview of the new guidelines MR 2.6.1.0333-23 (approved on 01 December 2023) that supersede guidelines MU 2.6.1.2838-11, which were used for organizing radiation surveys of residential, public and industrial buildings and facilities and their sanitary assessment in terms of radiation safety indicators over the past 12 years. Due to a large number of critical comments on the document received during this period, a need for significant revision emerged. The scope of the revised document was expanded and now covers all stages of the life cycle of buildings and facilities: commissioning; operation period; overhaul and reconstruction; demolition. In the first part of the paper, the structure and legal status of the new guidelines are considered, as well as the main innovations in terms of measuring the ambient dose equivalent rate of gamma radiation, searching and identifying local radiation anomalies and sites of radioactive contamination, measuring surface contamination levels, and determining the minimum number of premises to be surveyed. The reasons for the introduction of certain changes are outlined. The issues of evaluation of measurement uncertainty, reporting of measurement results and registration of test reports are covered in detail. In the second part of the paper, numerous innovations will be considered in terms of estimating the indoor average annual equilibrium equivalent concentration of radon isotopes.

Radiation survey of the territory of Teriberka village and Kildin Island of Murmansk Region

The Arctic zone is a strategically important territory of the Russian Federation from the point of view of socio-economic, scientific, and technological components, as well as in terms of planning the development of a system for ensuring national defense, state, and public security. Intensive exploration and development of Arctic territories, combined with a rapidly changing environment due to global climate change, carries a few risks for the environment. This determines the need to assess potential environmental damage and determine the significance of factors of potential negative impact on unique Arctic ecosystems now and in the future. The article presents current data from a radiation survey of the coastal zone of residential areas close to the site of the sinking of the nuclear submarine ‘K-159’ – Teriberka village and Kildin Island, Murmansk region, as well as the sea area to assess the primary levels of man-made and natural radionuclides before starting work on the potential lifting of the submerged object. This work was carried out as part of the implementation of the State Program of the Russian Federation ‘Health Development’ (approved by Decree of the Government of the Russian Federation dated December 26, 2017, No. 1640). The results of measuring the ambient dose equivalent rate of gamma radiation in open areas and determining the content of man-made and natural radionuclides in samples of environmental media (soil, aquatic vegetation, marine aquatic organisms) and food products of local origin and production (mushrooms, berries, fish, milk). The research results showed that the state of the study area is generally stable in terms of the radiation factor, the content of radionuclides is at the background level. As part of this work, electronic database modules have been developed based on the results of radiation monitoring in Teriberka village and on Kildin Island. Certificate No. 2022621890 was received on registration of a database with the results of the radiation situation of the above-mentioned territories.

A high-throughput focused collimator for OAR-sparing preclinical proton FLASH studies: commissioning and validation

Objective. To fabricate and validate a novel focused collimator designed to spare normal tissue in a murine hemithoracic irradiation model using 250 MeV protons delivered at ultra-high dose rates (UHDRs) for preclinical FLASH radiation therapy (FLASH-RT) studies. Approach. A brass collimator was developed to shape 250 MeV UHDR protons from our Varian ProBeam. Six 13 mm apertures, of equivalent size to kV x-ray fields historically used to perform hemithorax irradiations, were precisely machined to match beam divergence, allowing concurrent hemithoracic irradiation of six mice while sparing the contralateral lung and abdominal organs. The collimated field profiles were characterized by film dosimetry, and a radiation survey of neutron activation was performed to ensure the safety of staff positioning animals. Main results. The brass collimator produced 1.2 mm penumbrae radiation fields comparable to kV x-rays used in preclinical studies. The penumbrae in the six apertures are similar, with full-width half-maxima of 13.3 mm and 13.5 mm for the central and peripheral apertures, respectively. The collimator delivered a similar dose at an average rate of 52 Gy s−1 for all apertures. While neutron activation produces a high (0.2 mSv h−1) initial ambient equivalent dose rate, a parallel work-flow in which imaging and setup are performed without the collimator ensures safety to staff. Significance. Scanned protons have the greatest potential for future translation of FLASH-RT in clinical treatments due to their ability to treat deep-seated tumors with high conformality. However, the Gaussian distribution of dose in proton spots produces wider lateral penumbrae compared to other modalities. This presents a challenge in small animal pre-clinical studies, where millimeter-scale penumbrae are required to precisely target the intended volume. Offering high-throughput irradiation of mice with sharp penumbrae, our novel collimator-based platform serves as an important benchmark for enabling large-scale, cost-effective radiobiological studies of the FLASH effect in murine models.

Radioactivity estimation of radioactive hotspots using a Compton camera and derivation of dose rates in the surrounding environment

At the Fukushima Daiichi Nuclear Power Station, radiation sources released in the accident were deposited on various equipment and building structures. During decommissioning, it is crucial to understand the distribution of radiation sources and ambient dose equivalent rates to reduce worker exposure and implement detailed work planning. In this study, the author introduces a method for visualizing radiation sources, estimates their radioactivity using a Compton camera, and derives the dose rate around the radiation sources. In the demonstration test, the Compton camera was used to visualize radioactive hotspots caused by 137Cs radiation sources deposited in the outdoor environment and estimated the radioactivity. Furthermore, the dose rate around the hotspots was calculated from the estimated radioactivity, which confirmed that the calculated dose rate correlated with the dose rate measured using a survey meter. This approach is novel, where a series of analyses were conducted using the Compton camera to visualize radioactive hotspots, estimate the radioactivity, and derive the dose rate in the surrounding environment.

Оценка функционирования дозиметрического оборудования при проведении измерений в полях импульсного микросекундного фотонного излучения с известными характеристиками

This paper continues the series of articles [1,2,3] devoted to the determination of the main dosimetric characteristics of the fields of high-energy pulsed microsecond photon radiation generated by medical linear electron accelerator (LINAC for short) and assessment of the possibility of using LINAC as a source of reference radiation field for calibration of dosimetric equipment. This publication presents the results of evaluation of the portable dosimeter DKSAT1123 and detection unit BDKG-206 when measuring in such fields under different modes of LINAC operation. The possibility of this dosimetric equipment operation in the fields was assessed by comparing dosimeter readings with reference values of dosimetric characteristics of these fields, namely: average ambient dose equivalent rate H*(10), measured with the help of high-precision reference dosimeter DKS-AT5350. It was found that the deviation of H*(10) measurements conducted with the help of the DKS‑AT1123 and BDKG-206 from the reference values strongly depends on the radiation properties, namely the H*(10) value in the pulse. It is shown that at the H*(10) value in the pulse less than 0,2 mSv/s the deviation does not exceed 10%, at the H*(10) value in the pulse 0,2–0,4 mSv/s the deviation does not exceed 33%, and at the H*(10) value in the pulse more than 0,5 mSv/s the deviation can reach 72%.

Aircraft radiation exposure assessment in a radioactive environment: Ambient dose equivalents and effective doses

The release of radioactive plumes can occur due to nuclear power plant accidents, terrorist attacks, or scenarios in the chemical, biological, radiological, and nuclear defense (CBRN) field. Rescue aircraft may be exposed to ionizing radiation in such contaminated zones. Previous experimental assessments of these scenarios to estimate the potential dangers crew and passengers face are usually intangible. Using predictive software, it is possible to extract information from an external radiation field to which an aircraft would be subject. However, such codes cannot internally estimate the impact of this field on the aircraft. Using simulations based on the Monte Carlo method, it was possible to reproduce the scenario of plumes in the external environment (atmosphere) and, from this scenario, generate relevant information about their impact on the aircraft's internal environment. These assessments comprise the interaction of radiation from a radioactive plume with the aircraft structures and fuel tanks to estimate the radiation doses received by the crew and onboard electronics. This work evaluates the fluence rate, ambient dose equivalent rate (Ḣ*(10)), and effective dose rate (Ė) inside an aircraft flying through a radioactive plume resulting from a nuclear power plant accident. The simulation results suggest that radiation dose rates vary widely depending on the position within the aircraft. The ambient dose equivalent rate for photons varies by approximately 80% depending on the position within the aircraft. These differences reach around 98% for the ambient dose equivalent and effective dose rates for electrons. The data obtained may also be incorporated into risk assessments and support the development of protective measures to counter CBRN events.

Present and new operational quantities evaluated from photon spectrum measurements at workplaces in the research reactor and accelerator facility at the JAEA

The International Commission on Radiation Units and Measurements (ICRU) has proposed to change the definitions of the operational quantities used for the area and individual monitoring of external exposure in ICRU Report 95. Since introducing new operational quantities into the radiation monitoring may affect the dose assessment using the present personal dosimeters, precise assessments of the influence on monitoring are necessary by characterizing the energy spectrum in the workplace and the energy dependency of the dosimeters to be used. This study measured photon spectra using a NaI(Tl) scintillation detector or a LaBr3(Ce) scintillation detector at the Japan Research Reactor No. 3 (JRR-3) and Japan Proton Accelerator Research Complex (J-PARC) workplaces at the Japan Atomic Energy Agency (JAEA). The photon energy spectra were obtained by unfolding the pulse height distributions measured at each workplace. The present and new operational quantities were then evaluated and compared using the spectra measured at the workplaces.In JRR-3, the photon energy spectra were measured at the workplaces in the reactor room and thermal neutron beam hall while the reactor was operated. The ambient dose equivalent rates H˙*(10) at the workplaces were 0.2 μSv·h−1 – 13 μSv·h−1. High energy gamma-rays of 6–7 MeV from 16N were observed near the fuel failure detection system in the reactor room. Those of 7–8 MeV produced by the thermal neutron capture of iron were observed in the reactor room and the beam hall. In J-PARC, the photon energy spectra were measured at the workplaces in the beam tunnel of the 3 GeV synchrotron beam ring five days after the operation. The ambient dose equivalent rates H˙*(10) at the workplaces were 35 μSv·h−1 –430 μSv·h−1.It was found that the new personal doses, Hp, were 10%–20% smaller than the present personal dose equivalents, Hp(10), at the workplaces.

Design and testing of an irradiation room with low room-scattering for neutron calibration

IntroductionThis laboratory plans to establish neutron reference radiation fields with three neutron sources to calibrate neutron-measuring devices. To perform calibration at multiple dose rates, neutron ambient dose equivalent rate H˙*(10) needs to range 1 μSv/h to 10 mSv/h. The lower limit requires that the maximum available calibration distance should be at least 4.5 m. MethodsTo reduce room-scattered neutrons and extend the available calibration distance, MC simulations were conducted to determine the material and size of the irradiation room. A 3″ Bonner sphere and a LB6411 environmental neutron dosimeter were used to characterize the irradiation room. ResultsA 14.32 × 14.32 × 12.00 m3 irradiation room was built based on simulation results. Floor, roof, and walls are made of 75 cm concrete covered by a coating layer of 2 cm BPE and 3 cm PE. Experimental maximum available calibration distance reaches 4.65 m. The range of H˙*(10) for calibration covers 1 μSv/h to 10 mSv/h. Neutron and photon H˙*(10) outside the room are within 0.19 μSv/h and 0.22 μSv/h, respectively.

A large and feasible national survey representative of population exposure to outdoor gamma radiation in urban areas.

Although data on outdoor gamma radiation are available for many countries, they have generally been obtained with measurements performed in undisturbed environments instead of in urban areas where most of the population lives. Only one large national survey, with on-site measurements in urban areas, has been identified worldwide, probably due to high costs (e.g., personnel and instrumentation) and difficulties in selecting measuring points. A campaign of outdoor gamma radiation measurements has been carried out in the entire Italian territory. All measurement points were selected at the infrastructures of an Italian telecommunications company as representatives of all the possible situations of outdoor exposure to gamma radiation for population in urban areas. Ten replicates of portable gamma (X) detectors carried out all the measurements. Approximately 4,000 measurements have been performed. They are distributed across 2,901 Italian municipalities, accounting for 75% of the Italian population. The national population-weighted mean of the gamma ambient dose equivalent rate (ADER) is 117 nSv h-1, and it ranges from 62 to 208 nSv h-1 and from 40 to 227 nSv h-1 for 21 regions and 107 provinces, respectively. The average variability at the municipal level, in terms of the coefficient of variation (CV) is 21%, ranging from 3 to 84%. The impact of land coverage and the distance from a building on the outdoor gamma radiation level was assessed with complementary measurements, leading to differences ranging from -40 to 50% and to 50%, respectively. A representative campaign of outdoor gamma dose rate measurements has been performed in Italy, only in urban areas, to assess the exposure effect due to outdoor gamma radiation on the population. It is the largest national campaign in urban areas worldwide, with a total of 3,876 on-site measurements. The land coverage and the distance from surrounding buildings were recognized to strongly affect outdoor gamma radiation levels, leading to high variability within small areas. The collaboration with a company that owns a network of facilities on a national territory as dense as the residing population made this survey feasible and affordable. Other countries might adopt this methodology to conduct national surveys in urban environments.

Seasonal variation of background counting rates in liquid scintillation counting

A liquid scintillation background sample was measured daily in a custom-built TDCR counter for more than 17 months. The double and triple coincidence counting rates exhibit an annual sinusoidal fluctuation with a maximum in winter and a minimum in summer. Possible correlations with air temperature, air humidity, radon concentration and secondary cosmic radiation were investigated. The observation of a correlation with the ambient dose equivalent rate H˙*(10)SCR originating from the charged component of secondary cosmic radiation and an anti-correlation with the effective atmospheric temperature Teff suggest that the seasonal fluctuations in the background counting rate may be primarily driven by temporal variations in the muon flux at ground level. Additionally, a correlation was found with the indoor 222Rn concentration in air.

Neutron Activation Analysis Based on AB-BNCT Treatment Room.

Boron neutron capture therapy (BNCT) is an ideal binary targeted radiotherapy for treating refractory tumors. An accelerator-based BNCT (AB-BNCT) neutron source has attracted more and more attention due to its advantages such as higher neutron yield in the keV energy region, less gamma radiation, and higher safety. In addition to 10 B, neutrons also react with other elements in the treatment room during BNCT to produce many activation products. Due to the long half-life of some activation products, there will be residual radiation after the end of treatment and the shutdown of the accelerator, which has adverse effects on radiation workers. Therefore, the ambient dose equivalent rate in the treatment room needs to be evaluated. The AB-BNCT neutron source model proposed by Li is studied in this paper. Based on the Monte Carlo method, the Geant4 platform was used to simulate the dose induced by radionuclides near the Beam Shaping Assembly (BSA) of the source. It is concluded that the concrete wall contributed the most to the radiation dose. The dose rate of 2.45 μSv h -1 after 13 min of shutdown meets the dose rate limit of 2.5 μSv h -1 , at which point it is safe for workers to enter the treatment room area.