Science and Technology of Nuclear Installations
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Acceptance rate24%
Submission to final decision110 days
Acceptance to publication14 days
CiteScore1.500
Journal Citation Indicator0.380
Impact Factor1.1

An Association Rule Mining-Based Method for Revealing the Impact of Operational Sequence on Nuclear Power Plants Operating

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 Journal profile

Science and Technology of Nuclear Installations publishes research on issues related to the nuclear industry, particularly the installations of nuclear technology, and aims to promote development in the area of nuclear sciences and technologies.

 Editor spotlight

Professor Michael I. Ojovan is the Chief Editor of the journal, and is currently based at the University of Sheffield, UK. He is known for many innovations in nuclear research, including metallic and glass-composite materials for nuclear waste immobilisation.

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Research Article

Quantitative Assessment of Gaseous Effluents during Routine Operation: A Comparative Study of Planned Nuclear Power Plants at Lubiatowo-Kopalino and Pątnów Sites in Poland

On September 2021, Polish government declared that six pressurized water reactors with combined capacity of 6–9 GWe will be built by 2040 to reduce Poland’s reliance on coal. Due to the adopted schedule, construction of the first nuclear power plant will begin in 2026, with the first reactor capacity of 1–1.6 GWe to be operational in 2033. The Polish authorities announced in 2022 the selection of two locations and technologies for Poland’s first commercial nuclear power plants. Westinghouse AP1000 reactor was selected by Polish government to be built as the first plant in the location of Lubiatowo-Kopalino, on the coast of the country. In the meantime, Poland’s ZE PAK, Polska Grupa Energetyczna, and Korea Hydro & Nuclear Power have signed the letter of intent to collaborate on the project that evaluates the feasibility of building South Korean APR1400 on Pątnów site in central Poland. The objective of this study was to acquire and examine the gaseous effluents released during the standard operation of the AP1000 and APR1400 reactor technologies, with the primary goal of focusing on estimating the potential exposure of the general public. The effluents were calculated by using the GALE code based on each nuclear reactor technical specification. The obtained results were compared with those included in the Design Control Document for each reactor. Subsequently, the HotSpot software was used to calculate the radiation risk for downwind areas by utilizing GALE code results as source terms together with specific meteorological data corresponding for each localization. The results for AP1000 at Lubiatowo-Kopalino site and for APR1400 at Pątnów site were analysed and compared in the study. As the study findings were evaluated with the Polish radiation limits for the general public, all doses remained below the legal thresholds. With no previous alike studies conducted, this research begins the analysis of radiation impacts associated with the planned nuclear power plant in Poland during normal operation.

Research Article

Time-Series Forecasting of a Typical PWR Undergoing Large Break LOCA

In this work, a machine learning (ML) metamodel is developed for the time-series forecasting of a typical nuclear power plant response undergoing a loss of coolant accident (LOCA). The plant model of choice is based on the APR1400 nuclear reactor. The key systems and components of APR1400 relevant to the investigated scenario are modelled using the thermal-hydraulic code, RELAP5/MOD3.4, following the description published in the design control document. The model is tested under a spectrum of initial and boundary conditions via propagation of key uncertain parameters (UPs) which are derived from the phenomena identification and ranking table (PIRT). This is achieved by loosely coupling RELAP5/MOD3.4 with the statistical tool, Dakota. The most probable nuclear power plant (NPP) response was calculated using the best estimate plus uncertainty (BEPU) approach. Next, the database generated from the NPP system response was used as an input for the ML model. The NPP system response was represented by peak cladding temperature (PCT), safety injection system (SIT), mass flow rate, reactor power, and primary system pressure. In this research, two regression models were tested with reasonably good performance, namely, the gated recurrent unit (GRU) and the long short-term memory (LSTM).

Research Article

Enhancing Resilience through Nuclear Emergency Preparedness at El Dabaa Site

The research utilized advanced PCTRAN and RASCAL software to evaluate the potential radiological impacts of hypothetical accidents, specifically loss-of-coolant accident (LOCA) and long-term station blackout (LTSBO), at the El Dabaa Nuclear Power Plant. Over a span of ten years, comprehensive meteorological data were meticulously analyzed to assess the dispersion of radioactive substances within a 40-kilometer radius across all four seasons. The outcomes revealed that only in the case of LTSBO did the radiological levels surpass the limits set by the Environmental Protection Agency (EPA). Notably, during spring, LTSBO exhibited a maximum total effective dose equivalent (TEDE) value of 13 millisieverts (mSv) at a distance of 3.2 kilometers, and the highest thyroid dose (TD) recorded was 63 mSv at 8 kilometers. These significant findings play a crucial role in shaping strategies related to the distribution of potassium iodide (KI) and further enhance the overall preparedness and evacuation planning protocols.

Research Article

Evaluation of Worker Radiation Exposure during the Kori Unit 1 Steam Generator Dismantling Process

Kori Unit 1 was permanently shut down on June 18, 2017. Since then, Korea is actively preparing for the decommissioning of the nuclear power plant. Because decommissioning work is performed in a radioactive environment, worker radiation exposure is a significant consideration. In this study, worker radiation exposure is evaluated during the steam generator, one of the heavy components of nuclear power plant, dismantling process. A radiation evaluation for the dismantling process is performed using the code RESRAD-BUILD. A steam generator dismantling scenario and optimal cutting method are designed to evaluate worker radiation exposure, considering pipe dimensions, cutting tool speed, and experience in steam generator replacement. The evaluation results are derived for each work type and year. As a result of the evaluation, worker radiation exposure is 7.5 man-mSv at the year of planned decommissioning.

Research Article

Detecting Unauthorized Movement of Radioactive Material Packages in Transport with an Adam-Optimized BP Neural Network Model

The rapid expansion of nuclear technology across various sectors due to global economic growth has led to a substantial rise in the transportation of radioactive materials. The International Atomic Energy Agency (IAEA) estimates that approximately 20 million shipments of radioactive materials occur annually. In this context, ensuring the safety and security of radioactive material transportation is of significant importance. IAEA’s “Security of Radioactive Materials in Transport” (Nuclear Security Series No. 9-G) mandates that an effective transport security system should provide immediate detection of any unauthorized removal of the packages. In the present study, an innovative Adam-optimized BP neural network model is developed for detecting unauthorized movements of radioactive material packages. To analyze the performance of the proposed algorithm, numerous experiments were conducted. The results demonstrate that the proposed method achieves a 99.17% accuracy rate in detecting unauthorized movements of radioactive materials, with a missed alarm rate of 0.72% and a false alarm rate of 0.1%. This method also enables real-time detection of unauthorized removal of radioactive materials and effectively enhances the security of radioactive materials during transport to reduce the risks of theft, loss, diversion, or sabotage.

Research Article

ELSMOR European Project: Experimental Results on an Innovative Decay Heat Removal System Based on a Plate-Type Heat Exchanger

This paper summarises the results of an experimental campaign carried out at SIET on the ELSMOR facility built in 2022 to validate a decay heat removal system for the E-SMR. Based on the passive mechanisms of natural circulation, the system aims to dissipate the reactor decay heat to a water pool, using two heat exchangers: a plate-type one coupling the primary side to the secondary side, and a vertical tube one coupling the secondary side to the water pool. Such a system is considered to be the most effective passive system, capable of safely managing the SMR accident and accidental situations, and achieving long-term decay heat removal without the need for electricity or external inputs. A description of the primary and secondary loops of the plant is given, together with the installed instrumentation and data acquisition system. In addition, the paper summarises the tests performed in terms of test procedures, test type and associated objectives, test matrix, test results, achievements, and open issues.

Science and Technology of Nuclear Installations
 Journal metrics
See full report
Acceptance rate24%
Submission to final decision110 days
Acceptance to publication14 days
CiteScore1.500
Journal Citation Indicator0.380
Impact Factor1.1
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