Assessment of accidental release of 131 I from Egyptian radioisotopes production factory (original) (raw)
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The present work gives a methodology for assessing radiological concentration of 131 I, 132 I, 133 I, 134 I, and 135 I due to a hypothetical accident of TRIGA Mark-II research Reactor at AERE, Savar, Bangladesh. The concentrations were estimated through different pathways like ingestion of vegetation, milk, and meat from air and ground deposition. The maximum air concentrations for all 16 directions were found at 110 m distance from the core of the reactor and it was found to be highest in the southern (S) direction. The maximum ground concentration occurred immediately just after the accident in different directions. In all pathways, the most concentration was found to be in S-direction. The concentrations in vegetation of 131 I, 133 I, 135 I were significant, while no concentrations of 132 I and 134 I were observed. The concentration in vegetation for 131 I was found to be highest than all other isotopes of iodine. The concentrations of 133 I were found to be higher and concentrations of 134 I were observed to be lower in both milk and meat compared to other radio isotopes of iodine. In the case of a radiological accident, the results of the present study will be a valuable guide for adopting radiological safety measures for radiation protection against the ingestion of vegetables, milk and meat from around the research reactor at AERE, Savar, Bangladesh.
International Journal of Environmental Protection and Policy, 2021
Inshas site is very important site, it is considered as a first nuclear site in Egypt. It contains the first and the second research reactors and other important facilities such as the Egyptian Fuel Manufacturing Pilot Plant (FMPP) which submits the fuel to the second research reactors. The goals of this paper are to assess different emergency plan scenarios, determine the effects of metrological conditions on the dispersion of the radioactive plume, and calculate the overall effective dose equivalent values (TEDE). The meteorological parameters for one complete year 2020 (hourly) meteorological data were analysed in details such as wind direction, wind speed and temperature. The HotSpot code was used to model atmospheric dispersion and its application resulted in a radiation dose profile around the site using meteorological parameters specific to the area under study. This study used more than one scenario to investigate the role of various meteorological parameters. The radionuclide source term of Cs-137 was assumed to be 2.10E+15Bq. The results show that the maximum dose form all postulated scenarios is about 1.1E+4 Sv is observed at 10 m from the release source for weather stability class F, which is greater than the IAEA occupational exposure limit of 0.02 Sv per year. The results can also show that the time of accident is a major effect on the impact of accident and then on the consequences of emergency plan.
MONSTROUS HAZARDS PRODUCED BY HIGH RADIOACTIVITY LEVELS AROUND ASSIUT THERMAL POWER PLANT
The natural radioactivity level of heavy oil, ash and soil samples around Assiut Thermal Power Plant (ATPP) in Egypt was determined using gamma ray spectrometry. The average concentrations of 226Ra, 232Th and 40K in fly ash were found to be 2307±143, 1281±80 and 1218±129 Bq kg-1, respectively, while the corresponding values in soil samples were 2670±107, 1401±78 and 1495±100 Bq kg-1, respectively. These are extremely high and higher by several orders of magnitude than the worldwide population-weighted average values in soil. The radium equivalent activity, the air absorbed dose rate, external hazard index and the annual effective dose rate were calculated and compared with the international recommended values. All averages of these parameters are much higher by several orders of magnitude than the international recommended values, indicating significant radiological health hazards around ATPP due to the radionuclides in the soil. Moreover, the water samples investigated have high activity concentrations indicating that the water is highly contaminated with radioactive materials. The results of the current study highlight the severity of this radioactive pollution on the population in the vicinity of ATPP.
Journal of Al-Nahrain University Science
In order to study assessment of the annual doses and cancer risk lifetime related in IRPL laboratory, the field measurements were performed on a systematic grid .Soil samples were collected and prepared for laboratory analyzed by using gamma spectrometry system (HPGe). RESRAD software version (6.5) was used to assess the annual doses and probability of excess cancer risk lifetime incurred by workers exposed to radioactive material. The results indicated that the Total Effected Dose Equivalent (TEDE) received from artificial radionuclides IP-139, IP-79, and IP-184 are (2.116, 0.394, 10.97) mSv/y respectively are exceed the regulatory dose limit adopt by NRC (Nuclear Regulatory Commission) (0.25mSv/y) and (0.3mSv/y) of IAEA criteria, while TEDE from natural radionuclides sample (IP-15) is (0.03459mSv/y) less than the basic dose limit. The artificial radionuclides cells must be treated to keep IRPL, workers and surrounding environment within accepted limit.
Assessment of Radiological Dose around a 3-MW TRIGA Mark-II Research Reactor
International Letters of Chemistry, Physics and Astronomy, 2013
A hypothetical accidental case of a 3-MW TRIGA Mark-II research reactor has been assumed to assess the radiological consequences due to the deposition of 137Cs and 90Sr on ground, vegetation, milk and meat. The air concentrations in sixteen cardinal directions have been estimated where the maximum concentration has been found to be at 110 m distance from the core of the reactor for all the directions. Calculated maximum doses of 137Cs, 90Sr and both 137Cs and 90Sr have been found to be within the ranges of 0.005-0.014 μSv hr–1, 0.013-0.036 μSv hr–1 and 0.018-0.05 μSv hr–1, respectively for all the directions, which are below the measured background dose limit 0.25 μSv hr–1 and also within the IAEA acceptable dose rate limit of 0.5 μSv hr–1. The calculated low doses due to the aforementioned radionuclides can be considered negligible with regard to the radiation hazards. The relationship between total effective dose rate for various pathways (i.e. immersion, inhalation, ground deposi...
TRANSACTIONS of the VŠB – Technical University of Ostrava, Safety Engineering Series
In the Czech Republic, under normal conditions, the radioactive sources are used transported and stored in accordance with the relevant regulations and instructions of the State Office for Nuclear Safety (SONS), which is the regulatory authority of the state administration. Its main task is to ensure adequate safety and protection of persons in accordance with current international standards and guidelines. However, if there is an incidence, accident, or other emergency, these circumstances require taking some specific measures aimed at minimizing the impact of such situations on human health and the environment. During a fire in the workplace, the radioactive sources can get out of control or radioactive substances may begin to leak into the environment. In these cases, it is necessary to evaluate the radiation situation by means of measurement and monitoring both the external radiation and radioactive contamination of the air and the surrounding environment. For this purpose, a sy...
• Atmospheric dispersion modeling for two credible accidents of the TRIGA Mark II research reactor in Kinshasa (TRICO II) was performed. • Radiological safety analysis after the postulated initiating events (PIE) was also carried out. • The Karlsruhe KORIGEN and the HotSpot Health Physics codes were used to achieve the objectives of this study. • All the values of effective dose obtained following the accident scenarios were below the regulatory limits for reactor staff members and the public, respectively. a b s t r a c t The source term from the 1 MW TRIGA Mark II research reactor core of the Democratic Republic of the Congo was derived in this study. An atmospheric dispersion modeling followed by radiation dose calculation were performed based on two possible postulated accident scenarios. This derivation was made from an inventory of peak radioisotope activities released in the core by using the Karlsruhe version of isotope generation code KORIGEN. The atmospheric dispersion modeling was performed with HotSpot code, and its application yielded to radiation dose profile around the site using meteorological parameters specific to the area under study. The two accident scenarios were picked from possible accident analyses for TRIGA and TRIGA-fueled reactors, involving the case of destruction of the fuel element with highest activity release and a plane crash on the reactor building as the worst case scenario. Deterministic effects of these scenarios are used to update the Safety Analysis Report (SAR) of the reactor, and for its current version, these scenarios are not yet incorporated. Site-specific meteorological conditions were collected from two meteorological stations: one installed within the Atomic Energy Commission and another at the National Meteorological Agency (METTELSAT), which is not far from the site. Results show that in both accident scenarios, radiation doses remain within the limits, far below the recommended maximum effective (whole body) dose of 20 mSv/year for workers and 1 mSv/year for the general public in the IAEA Basic Safety Standards 115 and demonstrate the radiation safety of this reactor. This guarantees the safety of workers and the population around the plant site.
Assessment of the radiological control at the Ipen radioisotope production facility
Brazilian Journal of Radiation Sciences, 2015
The main objective of this work is to evaluate the 2013 annual radiological control results in the radiopharmaceuticals areas of the Nuclear and Energy Research Institute, IPEN/SP, and the environmental radiological impact, resulting from the practices there performed. The current evaluation was performed through the analysis of the results obtained from occupational and environmental monitoring with air samplers and TL dosimeters. All monitoring results were compared with the limits established by national standards. The radionuclides detected by air sampling (in activated carbon cartridges and filter paper) at the workplace during radioisotope production were 131 I, 99m Tc and 99 Mo, with activities concentrations values below the annual limits values. For the radioactive gaseous releases (Bq/m 3), the activities concentrations also remained below the maximum admissible values, excepting to 125 I release due to an unusual event occurred in a researcher laboratory, but the radiological impact to environmental was no significant. The occupational monitoring assessment was confirmed by the Environmental Radiological Monitoring Program results with air samplers and TL dosimeters. The mean annual background radiation at IPEN in 2013, according to the Environmental Radiological Monitoring Program results was 1.06 mSv. y-1 , below the ICRP 103 recommended limit of 20 mSv.y-1 for workers.
Radiological safety analysis of the SAMOP reactor experimental facility
IOP Conference Series: Materials Science and Engineering
The reactor subcritical assembly for 99Mo production (SAMOP) experimental facility is a nuclear reactor operating under subcritical conditions. SAMOP reactor is fueled by uranyl nitrate inside an annular tube surrounded by the TRIGA reactor fuels. The external neutron source required for the SAMOP reactor operation comes from the Kartini reactor radial beamport. The aim of the analysis is to determine the estimated personnel radiation dose for normal conditions and severe postulated accident. The analysis methods is a calculation of source term, radiation propagation to the surrounding area, and estimation of personnel radiation dose rate for normal and severe accident conditions. The analysis result shows that the effective radiation dose received by personnel in the working area of the SAMOP reactor is below the dose limit for radiation worker at the Kartini reactor, i.e. 7 μSv/h. Whereas, in the postulated severe accident condition, the highest exposure rate at 1 m above and beside the reactor cooling tank are 4.19 mSv/h and 1.35 mSv/h respectively.
Bangladesh Journal of Nuclear Medicine
Measuring the external surface dose rate of the radioisotope I-131in order to reduce the radiation hazard during the transportation of this is to perform at National Institute of Nuclear Medicine and Allied Sciences (NINMAS) to the hospitals and clinics is a very basic but important requirement. Measured the external surface dose rate of I-131 after receiving the isotope send to NINMAS in shielded packages by Monrol N.P. Company, Turkey and Radioisotope Production Division (RIPD). The authors also measured the external surface dose rate of Tc-99m generator supplied by RIPD, AERE, Saver, Dhaka, Bangladesh, to check whether the radioisotopes package shielding are within the dose rate limit. From this study, the dose rate was found relatively higher for I-131 transportation by the smaller vehicle than large vehicle, although dose rate in both categories of vehicles are within the permissible limit. The external surface dose rates for Tc-99m generator were also found to be within the pe...