Ghana Research Reactor-1 is a miniature neutron source reactor (MNSR) which is currently fuelled with highly-enriched uranium (HEU) aluminium alloy fuel. Efforts are underway to convert the research reactor fuel to low-enriched uranium (LEU) oxide fuel. The project is coordinated research work funded by the International Atomic Energy Agency (IAEA) through its Coordinated Research Project (CRP) on Core Conversion. The research project was started with thermal hydraulic and neutronic calculation on both fuels. Radiological dose assessment as part of a safety assessment requirement needs to be carried out before the commencement of the core conversion project.
Accidental release of gaseous or liquid effluents is a critical issue and of a greater concern to the nuclear industry when it comes to the protection of the public and the environment. The emphasis becomes paramount when the release involves particulate of radiation particles. This paper provides a comprehensive insight report on an account of a research investigation carried out in addressing a radiological safety issue of Ghana’s Miniature Neutron Source Reactor (MNSR) during its core conversion project. The amounts of Strontium-90 (Sr-90) and Krypton-85 (Kr-85) effluents presumably released from the reactor hall to the surroundings and the consequential emission radiation to the working area within a 200 m radius were analyzed for a six-month working period. The objective was to estimate specifically the approximate total effective dose equivalent (TEDE) of Sr-90 and Kr-85 by considering a conjectural accident scenario using a well-recognized and user-friendly known atmospheric dispersion model before the preparatory period. The maximum TEDE value recorded at a ground deposition value of 4.6E − 01 kBq/m2 was approximately 1.80E − 02 mSv and 4.90E − 4 mSv for Sr-90 and Kr-85, respectively, at a maximum distance of 0.1 km from the source. The estimated dose values recorded were found to be within the recommended regulatory safety limits of 50 mSv for onsite workers and 1 mSv for the general public. No adverse effect was experienced with respect to human health and the environment.
The International Atomic Energy Agency defines a nuclear and radiation accident as an occurrence that leads to the release of radiation causing significant consequences to people, the environment, or the facility. During such an event involving a nuclear reactor, the reactor core is a critical component which when damaged, will lead to the release of significant amounts of radionuclides. Assessment of the radiation effect that emanates from reactor accidents is very paramount when it comes to the safety of people and the environment; whether or not the released radiation causes an exposure rate above the recommended threshold nuclear reactor safety. During safety analysis in the nuclear industry, radiological accident analyses are usually carried out based on hypothetical scenarios. Such assessments mostly define the effect associated with the accident and when and how to apply the appropriate safety measures. In this study, a typical radiological assessment was carried out on the Ghana Research Reactor-1. The study considered the available reactor core inventory, released radionuclides, radiation doses and detailed process of achieving all the aforementioned parameters. Oak Ridge isotope generation-2 was used for core inventory calculations and Hotspot 3.01 was also used to model radionuclides dispersion trajectory and calculate the released doses. Some of the radionuclides that were considered include I-131, Sr-90, Cs-137, and Xe-137. Total effective doses equivalent to released radionuclides, the ground deposition activity and the respiratory time-integrated air concentration were estimated. The maximum total effective doses equivalent value of 5.6 × 10−9 Sv was estimated to occur at 0.1 km from the point of release. The maximum ground deposition activity was estimated to be 2.5 × 10−3 kBq/m3 at a distance of 0.1 km from the release point. All the estimated values were found to be far below the annual regulatory limits of 1 mSv for the general public as stated in IAEA BSS GSR part 3.
The Pneumatic Transfer System (PTS) is an auxiliary system of Ghana Research Reactor-1 (GHARR-1) used to transfer the sample capsule in and out of the reactor irradiation sites. The PTS′ controller unit design and construction was carried out because the original transfer system was not designed to operate in Cyclic Neutron Activation Analysis (CNAA). To address these situations, a Programmable Logic Controller (PLC) has been used to design and construct a control unit to facilitate a CNAA application for GHARR-1. The design has been simulated successfully using LOGO! Soft Comfort software, version 8. A Failure Mode and Effect Analysis (FMEA) was conducted on the PTS Control Unit (PTSCU) to evaluate and document, by item failure mode analysis, the potential impact of each functional or hardware failure of the control unit, personnel and system safety, system performance, maintainability, and maintenance requirements. Each potential failure is ranked by the severity of its effect so that appropriate corrective actions can be taken to eliminate or control the high-risk items. The result obtained upon the analysis shows that the likelihood of occurrence of failures, detection, and severity on the control unit is low per the risk priority number. The paper outlines the severity classification and description used in FMEA, the likelihood of detecting various failures of components, and failure causes and effect.
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