The heating, ventilation, and air conditioning (HVAC) system plays a crucial role in ensuring the safety of workers and preventing the release of gaseous radioactive materials into the environment during the decommissioning of a nuclear power plant (NPP). To establish an HVAC operation strategy, decommissioning phases were divided into four stages, and the HVAC systems were reclassified. In addition, assumptions have been made regarding design modifications and maintenance for the reactor containment building (RCB) HVAC, fuel handling building (FHB) HVAC, and main control room (MCR) HVAC. Based on these, for RCB HVAC operation, natural ventilation and RCB purge operation during the transition period are proposed. In the decommissioning stage, recirculation operation, entire ventilation operation consisting of continuous operation and purge operation, and finally partial ventilation operation to purify local space were proposed. Moreover, during the transition period, the FHB HVAC was proposed to operate as normal NPP, and the MCR HVAC was suggested to operate with safety-related equipment removed.
{"title":"Design Change and Operational Consideration of the HVAC System during Nuclear Power Plant Decommissioning","authors":"Ho-Jin Jeon, Chang-Lak Kim","doi":"10.1155/2024/4701409","DOIUrl":"https://doi.org/10.1155/2024/4701409","url":null,"abstract":"The heating, ventilation, and air conditioning (HVAC) system plays a crucial role in ensuring the safety of workers and preventing the release of gaseous radioactive materials into the environment during the decommissioning of a nuclear power plant (NPP). To establish an HVAC operation strategy, decommissioning phases were divided into four stages, and the HVAC systems were reclassified. In addition, assumptions have been made regarding design modifications and maintenance for the reactor containment building (RCB) HVAC, fuel handling building (FHB) HVAC, and main control room (MCR) HVAC. Based on these, for RCB HVAC operation, natural ventilation and RCB purge operation during the transition period are proposed. In the decommissioning stage, recirculation operation, entire ventilation operation consisting of continuous operation and purge operation, and finally partial ventilation operation to purify local space were proposed. Moreover, during the transition period, the FHB HVAC was proposed to operate as normal NPP, and the MCR HVAC was suggested to operate with safety-related equipment removed.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2024-04-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140831013","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The micro heat pipe-cooled reactor is an innovative type of reactor that utilizes heat pipes to cool its core. It consists of a reactor core, an energy conversion system, shielding, and a heat removal system. This reactor shows great potential as a viable option for supplying electricity in remote areas. By incorporating a monolithic core with heat pipes and an efficient heat conversion system, this reactor design eliminates the need for a main pipeline, circulating pump, and auxiliary equipment, resulting in a cost-effective, compact, and transportable system. The monolithic reactor design has undergone significant advancements in neutronics and thermal hydraulics. This article focuses on evaluating the impact of the latest released nuclear data libraries, ENDF/B-VIII.0 and JENDL-5, on calculated neutronics and kinetics parameters. The total keff uncertainty was propagated and found to be significant for both recently evaluated nuclear data libraries (678.52 pcm for ENDF/B-VIII.0 and 525.91 pcm for JENDL-5, respectively). The total uncertainty originated from nuclear data was evaluated for total ν, reaction cross sections, and angular distributions in the case of JENDL-5, and for ENDF/B-VIII.0, uncertainty from angular distributions was not included because of the unavailability of its multigroup structure covariance matrices. The results reveal that the largest contributor for ENDF/B-VIII.0 is 235U total (409.18 pcm), while that for JENDL-5 is 56Fe capture cross section (361.93 pcm). For the kinetic parameter’s uncertainty, the impact on the total βeff, leff, and λeff simulation results was found to be not significant (about 1%).
{"title":"Accuracy Evaluation of Monte Carlo Simulation Results Using ENDF/B-VIII.0 and JENDL-5 Libraries for 10 MWth Micro Heat Pipe-Cooled Reactor","authors":"Thanh Mai Vu, Le Quang Linh Tran","doi":"10.1155/2024/5565346","DOIUrl":"https://doi.org/10.1155/2024/5565346","url":null,"abstract":"The micro heat pipe-cooled reactor is an innovative type of reactor that utilizes heat pipes to cool its core. It consists of a reactor core, an energy conversion system, shielding, and a heat removal system. This reactor shows great potential as a viable option for supplying electricity in remote areas. By incorporating a monolithic core with heat pipes and an efficient heat conversion system, this reactor design eliminates the need for a main pipeline, circulating pump, and auxiliary equipment, resulting in a cost-effective, compact, and transportable system. The monolithic reactor design has undergone significant advancements in neutronics and thermal hydraulics. This article focuses on evaluating the impact of the latest released nuclear data libraries, ENDF/B-VIII.0 and JENDL-5, on calculated neutronics and kinetics parameters. The total <i>k</i><sub>eff</sub> uncertainty was propagated and found to be significant for both recently evaluated nuclear data libraries (678.52 pcm for ENDF/B-VIII.0 and 525.91 pcm for JENDL-5, respectively). The total uncertainty originated from nuclear data was evaluated for total <i>ν</i>, reaction cross sections, and angular distributions in the case of JENDL-5, and for ENDF/B-VIII.0, uncertainty from angular distributions was not included because of the unavailability of its multigroup structure covariance matrices. The results reveal that the largest contributor for ENDF/B-VIII.0 is <sup>235</sup><i>U</i> total (409.18 pcm), while that for JENDL-5 is <sup>56</sup>Fe capture cross section (361.93 pcm). For the kinetic parameter’s uncertainty, the impact on the total <i>β</i><sub>eff</sub>, <i>l</i><sub>eff</sub>, and <i>λ</i><sub>eff</sub> simulation results was found to be not significant (about 1%).","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2024-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140577749","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Hongchi Zhou, Fang Liu, Xin Yuan, Jun Wu, Guanda Li
We explored the nonlinear characteristics of energy resolution (ER) for the sealed NaI (Tl) scintillator detector by using a gamma-ray spectroscopy system and Monte Carlo simulation. Our research focused on the two primary factors of energy resolution including the photomultiplier tube (PMT) voltage and the distance between the gamma-ray sources (137Cs and 60Co) and the scintillator detector. The experimental results showed that energy resolution decreased when the PMT voltage increased, and the energy resolution of NaI (Tl) detectors reached a smaller value (6.92%, 6.76%, and 6.56%), especially with the PMT voltage in the range of 575–595 V. In addition, a suitable distance between the gamma-ray source and the scintillator (5 cm) can also effectively reduce the energy resolution. We established the simulation models of the experimental NaI (Tl) detectors and simulated their energy spectra. The simulation results in the peak area agreed with the experimental results. A possible better PMT voltage choice has been proposed to obtain a smaller energy resolution.
{"title":"Effect of Photomultiplier Tube Voltage on the Performance of Sealed NaI (Tl) Scintillator Detectors","authors":"Hongchi Zhou, Fang Liu, Xin Yuan, Jun Wu, Guanda Li","doi":"10.1155/2024/5206668","DOIUrl":"https://doi.org/10.1155/2024/5206668","url":null,"abstract":"We explored the nonlinear characteristics of energy resolution (ER) for the sealed NaI (Tl) scintillator detector by using a gamma-ray spectroscopy system and Monte Carlo simulation. Our research focused on the two primary factors of energy resolution including the photomultiplier tube (PMT) voltage and the distance between the gamma-ray sources (<sup>137</sup>Cs and <sup>60</sup>Co) and the scintillator detector. The experimental results showed that energy resolution decreased when the PMT voltage increased, and the energy resolution of NaI (Tl) detectors reached a smaller value (6.92%, 6.76%, and 6.56%), especially with the PMT voltage in the range of 575–595 V. In addition, a suitable distance between the gamma-ray source and the scintillator (5 cm) can also effectively reduce the energy resolution. We established the simulation models of the experimental NaI (Tl) detectors and simulated their energy spectra. The simulation results in the peak area agreed with the experimental results. A possible better PMT voltage choice has been proposed to obtain a smaller energy resolution.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2024-04-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140577821","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Yanru Ren, Min Zhu, Dongyu Xu, Minghui Liu, Xuehui Dai, Shengao Wang, Longxian Li
With the rapid advancement of information technology, microelectronic devices have found widespread applications in critical sectors such as nuclear power plants, aerospace equipment, and satellites. However, these devices are frequently exposed to diverse radiation environments, presenting significant challenges in mitigating radiation-induced damage. Hence, this review aims to delve into the intricate damage mechanisms of microelectronic devices within various radiation environments and highlight the latest advancements in radiation-hardening techniques. The ultimate goal is to bolster the reliability and stability of these devices under extreme conditions. The review initiates by outlining the spectrum of radiation environments that microelectronic devices may confront, encompassing space radiation, nuclear explosion radiation, laboratory radiation, and process radiation. It also delineates the potential damage types that these environments can inflict upon microelectronic devices. Furthermore, the review elaborates on the underlying mechanisms through which different radiation environments impact the performance of microelectronic devices, which includes a detailed analysis of the characteristics and fundamental mechanisms of damage when microelectronic devices are subjected to total ionizing dose effects and single-event effects. In addition, the review delves into the promising application prospects of several key radiation-hardening techniques for enhancing the radiation tolerance of microelectronic devices.
{"title":"Overview on Radiation Damage Effects and Protection Techniques in Microelectronic Devices","authors":"Yanru Ren, Min Zhu, Dongyu Xu, Minghui Liu, Xuehui Dai, Shengao Wang, Longxian Li","doi":"10.1155/2024/3616902","DOIUrl":"https://doi.org/10.1155/2024/3616902","url":null,"abstract":"With the rapid advancement of information technology, microelectronic devices have found widespread applications in critical sectors such as nuclear power plants, aerospace equipment, and satellites. However, these devices are frequently exposed to diverse radiation environments, presenting significant challenges in mitigating radiation-induced damage. Hence, this review aims to delve into the intricate damage mechanisms of microelectronic devices within various radiation environments and highlight the latest advancements in radiation-hardening techniques. The ultimate goal is to bolster the reliability and stability of these devices under extreme conditions. The review initiates by outlining the spectrum of radiation environments that microelectronic devices may confront, encompassing space radiation, nuclear explosion radiation, laboratory radiation, and process radiation. It also delineates the potential damage types that these environments can inflict upon microelectronic devices. Furthermore, the review elaborates on the underlying mechanisms through which different radiation environments impact the performance of microelectronic devices, which includes a detailed analysis of the characteristics and fundamental mechanisms of damage when microelectronic devices are subjected to total ionizing dose effects and single-event effects. In addition, the review delves into the promising application prospects of several key radiation-hardening techniques for enhancing the radiation tolerance of microelectronic devices.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2024-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140361735","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Yuxuan He, Jian Song, Shaoke Shi, Haibo Lian, Jiangyang He, Ren Yu, Tete Liu, Bin Sun, Jiangtao Yuan, Yingbin Hu
The operations of the operators are important for nuclear safety, but conventional operating experience feedback and common data-driven methods make it difficult to explicitly find valuable information hidden in these operational sequences that can help the operator to provide advice at the operational level. During the nuclear power plant (NPP) operation, a large amount of historical operating data is accumulated, which records the operational sequences of the operators and the state parameters of equipment. Therefore, this paper proposes the use of association rule techniques to mine the NPP operating data to discover the operational characteristics of operators and reveal their possible impact on the NPP operation. This work helps to improve the operational performance of operators and prevent human-factor events. To this end, the concept of state switching values for describing the operating states of NPPs is proposed to enable the proposed method to be adapted to different practical application scenarios. A sequence segmentation method is proposed to be able to transform historical NPP operating data into a sequence data set for association rule mining. Furthermore, an ensemble algorithm based on sequence pattern mining and sequence rule mining and its postprocessing method are designed. The empirical study was carried out using 20 batches of historical operating data of the cold start-up. A total of 164 original association rules are generated using the proposed method and were analyzed by experts. The recommendations were made for 4 different cases that would improve the operational performance of the operators.
{"title":"An Association Rule Mining-Based Method for Revealing the Impact of Operational Sequence on Nuclear Power Plants Operating","authors":"Yuxuan He, Jian Song, Shaoke Shi, Haibo Lian, Jiangyang He, Ren Yu, Tete Liu, Bin Sun, Jiangtao Yuan, Yingbin Hu","doi":"10.1155/2024/6618975","DOIUrl":"https://doi.org/10.1155/2024/6618975","url":null,"abstract":"The operations of the operators are important for nuclear safety, but conventional operating experience feedback and common data-driven methods make it difficult to explicitly find valuable information hidden in these operational sequences that can help the operator to provide advice at the operational level. During the nuclear power plant (NPP) operation, a large amount of historical operating data is accumulated, which records the operational sequences of the operators and the state parameters of equipment. Therefore, this paper proposes the use of association rule techniques to mine the NPP operating data to discover the operational characteristics of operators and reveal their possible impact on the NPP operation. This work helps to improve the operational performance of operators and prevent human-factor events. To this end, the concept of state switching values for describing the operating states of NPPs is proposed to enable the proposed method to be adapted to different practical application scenarios. A sequence segmentation method is proposed to be able to transform historical NPP operating data into a sequence data set for association rule mining. Furthermore, an ensemble algorithm based on sequence pattern mining and sequence rule mining and its postprocessing method are designed. The empirical study was carried out using 20 batches of historical operating data of the cold start-up. A total of 164 original association rules are generated using the proposed method and were analyzed by experts. The recommendations were made for 4 different cases that would improve the operational performance of the operators.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2024-03-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140302983","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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.
{"title":"Quantitative Assessment of Gaseous Effluents during Routine Operation: A Comparative Study of Planned Nuclear Power Plants at Lubiatowo-Kopalino and Pątnów Sites in Poland","authors":"Edyta Agata Macieja, Juyoul Kim","doi":"10.1155/2024/8699926","DOIUrl":"https://doi.org/10.1155/2024/8699926","url":null,"abstract":"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.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2024-03-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140107611","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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).
{"title":"Time-Series Forecasting of a Typical PWR Undergoing Large Break LOCA","authors":"Michal Kaminski, Aya Diab","doi":"10.1155/2024/6162232","DOIUrl":"https://doi.org/10.1155/2024/6162232","url":null,"abstract":"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).","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2024-03-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140072267","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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.
{"title":"Enhancing Resilience through Nuclear Emergency Preparedness at El Dabaa Site","authors":"Waad Saleh, Juyoul Kim","doi":"10.1155/2024/9345812","DOIUrl":"https://doi.org/10.1155/2024/9345812","url":null,"abstract":"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.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2024-02-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139762539","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Panpan Jiang, Xiaohua Yang, Yaping Wan, Tiejun Zeng, Mingxing Nie, Chaofeng Wang, Yu Mao, Zhenghai Liu
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.
{"title":"Detecting Unauthorized Movement of Radioactive Material Packages in Transport with an Adam-Optimized BP Neural Network Model","authors":"Panpan Jiang, Xiaohua Yang, Yaping Wan, Tiejun Zeng, Mingxing Nie, Chaofeng Wang, Yu Mao, Zhenghai Liu","doi":"10.1155/2023/6363270","DOIUrl":"https://doi.org/10.1155/2023/6363270","url":null,"abstract":"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.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2023-12-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"138715483","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Roberta Ferri, Andrea Achilli, Cinzia Congiu, Stefano Marcianò, Stefano Gandolfi, Mattia Marengoni, Alberto Bersani, Alessandro Passerin D’Entreves
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.
{"title":"ELSMOR European Project: Experimental Results on an Innovative Decay Heat Removal System Based on a Plate-Type Heat Exchanger","authors":"Roberta Ferri, Andrea Achilli, Cinzia Congiu, Stefano Marcianò, Stefano Gandolfi, Mattia Marengoni, Alberto Bersani, Alessandro Passerin D’Entreves","doi":"10.1155/2023/6672504","DOIUrl":"https://doi.org/10.1155/2023/6672504","url":null,"abstract":"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.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":null,"pages":null},"PeriodicalIF":1.1,"publicationDate":"2023-12-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"138632428","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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