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Quantitative Assessment of Gaseous Effluents during Routine Operation: A Comparative Study of Planned Nuclear Power Plants at Lubiatowo-Kopalino and Pątnów Sites in Poland Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2024-3-12 Edyta Agata Macieja, Juyoul Kim
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
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Time-Series Forecasting of a Typical PWR Undergoing Large Break LOCA Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2024-3-8 Michal Kaminski, Aya Diab
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
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Enhancing Resilience through Nuclear Emergency Preparedness at El Dabaa Site Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2024-2-13 Waad Saleh, Juyoul Kim
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
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Evaluation of Worker Radiation Exposure during the Kori Unit 1 Steam Generator Dismantling Process Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2024-1-30 Hyeon-Ki Kim, Sang-Hwa Shin, Chang-Sig Kong, Chang-Lak Kim
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
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Detecting Unauthorized Movement of Radioactive Material Packages in Transport with an Adam-Optimized BP Neural Network Model Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2023-12-18 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
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ELSMOR European Project: Experimental Results on an Innovative Decay Heat Removal System Based on a Plate-Type Heat Exchanger Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2023-12-14 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
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Shaking Table Testing of a Scaled Nuclear Power Plant Structure with Base Isolation Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2023-10-31 Linlin Song, Xueming Zhang, Mingyang Wei, Yunlun Sun, Shicai Chen, Yan Chen
To investigate the seismic performance and isolation effect of a high-temperature gas-cooled reactor, a 1/20 scale model including a reactor, a spent-fuel plant, and a nuclear auxiliary plant was fabricated. In addition, 220 mm lead-rubber bearings were designed and produced for use in the shaking table test, which included both isolated and nonisolated conditions. Two historical earthquake records
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Advancing Small Modular Reactor Technology Assessment in the Czech Republic, Egypt, and Poland Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2023-10-21 Waad Saleh, Dalibor Kojecky, Edyta Agata Macieja, Juyoul Kim
This paper introduces the utilization of the International Atomic Energy Agency’s toolkit for reactor technology assessment (RTA) application to deploy small modular reactors (SMRs) in the Czech Republic, Egypt, and Poland. The increasing demand for clean energy has led to the prominence of small modular reactors (SMRs) in addressing global energy challenges. The successful integration of SMRs into
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Effect of Inhomogeneous Mechanical Properties on the Stress-Strain Field at the Crack Tip and Crack Growth Direction in Dissimilar Metal Welded Joints Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2023-8-9 Shuang Wang, Hongkui Zhang, Zhe Ju, Bing Li, Fandong Chen, Fei Han
In the failure analysis and safety assessment of dissimilar metal welded joints, the mechanical heterogeneity of local regions is usually ignored and limited sampling locations are selected. The mechanical behavior of the crack tip region is the main variables affecting the environmentally assisted cracking behavior, and it is crucial for understanding the impact of mechanical heterogeneity on the
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A Digital Controller for Reactivity Monitoring and Power Control Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2023-8-7 Van-Tai Vo, Van-Kien Nguyen, Van-Diep Le, Quoc-Minh Phan, La-Son Phan, Huy-Bach Nguyen, Nhi-Dien Nguyen
This paper introduces a controller unit for reactivity monitoring and automatic power control that was designed and constructed for the 500 kW Dalat Nuclear Research Reactor (DNRR). For power control and reactivity calculations, frequency signals from neutron measurement channels of starting and working ranges of the reactor are used. Two abovementioned independent functions were combined in an Artix-7
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A Reduced Order Model Based on ANN-POD Algorithm for Steady-State Neutronics and Thermal-Hydraulics Coupling Problem Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2023-7-10 Hanxing Liu, Han Zhang
The neutronics and thermal-hydraulics (N/TH) coupling behavior analysis is a key issue for nuclear power plant design and safety analysis. Due to the high-dimensional partial differential equations (PDEs) derived from the N/TH system, it is usually time consuming to solve such a large-scale nonlinear equation by the traditional numerical solution method of PDEs. To solve this problem, this work develops
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Design of a Nanosecond Voltage Comparator with PECL Logic for a Photon-Counting Radiation Imaging System Application Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2023-7-8 Huaxia Zhang, Yuewen Sun, Zijia Chen, Zhifang Wu
In this paper, a nanosecond voltage comparator with PECL logic for a photon-counting radiation imaging system is presented. To realize a high-speed comparison of four gamma detector channels in a limited board space, quad comparators MAX9602 with PECL logic are chosen. Each of the four channels is coupled with a PECL to CMOS converter ICS508, which exports CMOS logic data for later use in an FPGA.
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Methods for Predicting the Minimum Temperature of the Outage Loop and the Maximum Power Caused by the Low-Temperature Coolant Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2023-6-19 Xinxin Liu, Lei Yu, Jianli Hao, Xiaolong Wang
When the feedwater valve at the outage loop of the floating nuclear power plant leaks, thermal stratification occurs in the steam generator. It causes lower water temperature in the outage loop. The extent of hazard of this phenomenon cannot be directly determined by the existing measurement parameters, which poses a threat to the operational safety of the reactor. Therefore, this study adopts two
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Accident Sequence Precursor Analysis of an Incident in a Japanese Nuclear Power Plant Based on Dynamic Probabilistic Risk Assessment Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2023-6-7 Kotaro Kubo
Probabilistic risk assessment (PRA) is an effective methodology that could be used to improve the safety of nuclear power plants in a reasonable manner. Dynamic PRA, as an advanced PRA, allows for more realistic and detailed analyses by handling time-dependent information. However, the applications of this method to practical problems are limited because it remains in the research and development stage
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A New Experimental Method for the Nonlinear Modal Parameter Identification of a Pressurized Water Reactor Fuel Assembly Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2023-5-25 Chen Yang, Yan Guo, Xiao Hu, Yanhong Zhang
Establishing a dynamic model that accurately describes a realistic pressurized water reactor (PWR) fuel assembly is crucial to precisely evaluate the mechanical properties of the fuel assembly in seismic or loss of coolant accidents (LOCAs). The pluck test combined with the logarithmic decrement method has been widely applied in previous studies to extract fundamental modal parameters to calibrate
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Analysis of the Effect of Applied Load on Crevice Corrosion Behavior Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2023-5-3 Fuqiang Yang, Yue Zhang, Jianzhou Zhang
A two-dimensional numerical model incorporating solid mechanics, electrochemistry, mass diffusion, and ion migration processes is developed to investigate the load effect on the crevice corrosion. The model is a transient model of crevice corrosion occurring in cracks of 304 stainless steel in a dilute NaCl solution, and the interaction between stress and electrochemical corrosion was considered. By
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Acceptable Level of Acceptance and the Affecting Factors: What Is the Acceptable Public Acceptance of Building a Nuclear Power Plant Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2023-3-11 Drajat Tri Kartono, Sri Hastjarjo, undefined Sajidan, Bob Soelaiman Effendi, Dhita Karunia Ashari, Purbayakti Kusuma Wijayanto, Zahra Nadhila Saraswati, Alexander Yonathan Christy
This research determines the Acceptable Level of Acceptance (ALA) based on the countries with active Nuclear Power Plant (NPP). The ALA is a particular value of public acceptance of NPP, indicating public support and participation in the program. If the public acceptance level is lower than the ALA, then the probability of public resistance against the program is relatively high and would harm the
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Modelling and Validation of CANDU Shim Operation Using Coupled TRACE/PARCS with Regulating System Response Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2023-3-8 Simon Younan, David R. Novog
In CANDU reactors, shim operation is used when the online refuelling capability becomes temporarily unavailable. Adjuster rods, normally in-core to provide flux flattening, are withdrawn in sequence to provide additional reactivity as the fuel is depleted. In a CANDU 900 reactor, up to three of the eight adjuster banks may be withdrawn, with the power derated accordingly. In this study, the shim operation
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Steam-Jet Evaluation for Predicting Leakage Behavior and Interpretation of Experimental Verification Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2023-3-7 Dae Kyung Choi, Won Man Park, Woo-Shik Kim, Dong-Jin Euh, Tae-Soon Kwon, Choengryul Choi
Owing to pipe thinning, fatigue damage, and aging, pipes, valves, and devices installed in the primary and secondary systems of nuclear power plants may leak high-temperature/high-pressure reactor coolant. Thus, a system must be developed to determine if the leakage is exceeding the operating limit of the nuclear power plant, thereby mitigating any loss of life or economic loss in such cases. In this
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Assessing the Impact of Common Cause Failures on Site Risk within Level 1 Multi-Unit PSA Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2023-3-3 James F. Coleman, Emmanuel K. Boafo, S. Yamoah, F. Ameyaw
Common cause failures (CCFs) may lead to the simultaneous unavailability or failure of numerous components in the nuclear power plant because of the existence of a shared cause when an initiating event disrupts the normal functioning of nuclear power plants. The presence of common cause failures (intra-unit and inter-unit) can be recognized in a multi-unit probabilistic safety assessment (MUPSA) as
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Prediction of Automatic Scram during Abnormal Conditions of Nuclear Power Plants Based on Long Short-Term Memory (LSTM) and Dropout Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2023-3-3 Hanying Chen, Puzhen Gao, Sichao Tan, Hongsheng Yuan, Mingxiang Guan
A deep-learning model was proposed for predicting the remaining time to automatic scram during abnormal conditions of nuclear power plants (NPPs) based on long short-term memory (LSTM) and dropout. The proposed model was trained by simulated condition data of abnormal conditions; the input of the model was the deviation of the monitoring parameters from the normal operating state, and the output was
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Simulations of Core Damage Progression for TMI-2 Severe Accident Using CINEMA Computer Code Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2023-2-27 Rae-Joon Park, Donggun Son, Jun Ho Bae, Sung Won Bae, Bub-Dong Chung, Kwang Soon Ha
As an integrated computer code development for severe accident sequence analysis in Korea, CINEMA has been developing from an initiation event to a containment failure. The CINEMA computer code is composed of CSPACE, SACAP, and SIRIUS, which are capable of simulating core melt progression with thermal hydraulic analysis of the RCS (reactor coolant system), severe accident analysis of the containment
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Effectiveness of Serpentine Concrete as Shielding Material for Neutron Source Facility Using Monte Carlo Code Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2023-2-18 R. G. Abrefah, K. Tuffour-Achampong, P. Amoah
In recent years, much attention has been dedicated to finding techniques to reduce exposure doses. This work examines the effectiveness of using serpentine concrete to shield a neutron source using a 241Am-Be neutron source facility at the National Nuclear Research Institute (NNRI) as a case study. The results obtained for both neutrons and gamma indicate that serpentine concrete provides better shielding
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Development of an Integrated Human Error Simulation Model in Nuclear Power Plant Decommissioning Activities Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2023-1-12 Chang-Su Nam, Byung-Sik Lee
In this study, an integrated human error simulation model in nuclear power plant (NPP) decommissioning activities (HEISM-DA) that can integrate and manage various factors affecting human errors is developed. In the HEISM-DA, an error probability input method suitable for the characteristics of each performance shaping factors (PSFs) was presented. Because each PSF has different importance on human
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Application of EDG AOT Extension Based on the Risk-Informed Method in NPPs Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2023-1-5 Yunxin Feng, Wei Hu
At present, the allowed outage time (AOT) of an M310 unit emergency diesel generator (EDG) was 3 days, which can be extended to 14 days through replacement of additional diesel units; although it provides a certain online maintenance time, it cannot meet the needs of ten-years overhauls. In order to avoid stopping the reactor for maintenance of NPP due to insufficient of EDG AOT, based on risk-informed
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High-Temperature Corrosion Behavior of Incoloy 800H Alloy in the Impure Helium Environment Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-12-16 Wei Zheng, Huang Zhang, Bin Du, Haoxiang Li, Huaqiang Yin, Xuedong He, Tao Ma
The helium coolant in the primary circuit of the high-temperature gas-cooled reactor (HTGR) contains traces of impurities, which can induce the corrosion of superalloys when exposed to elevated temperatures. The superalloy damage caused by the corrosion could threaten the safe operation of the reactor. In this work, the corrosion behavior of a representative superalloy (chromium-rich iron base alloy
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A Data-Driven Fault Prediction Method for Nuclear Power Systems Based on End-to-End Deep Learning Framework Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-11-29 Lu Chao, Chunbing Wang, Shuai Chen, Qizhi Duan, Hongyun Xie
With the increase in system complexity and operational performance requirements, nuclear energy systems are developing in the direction of intelligence and unmanned, which also requires a higher demand for its safety so that intelligent fault diagnosis and prediction have become a technology that nuclear power plants need to develop at present. At the same time, due to the rapid development of deep
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Nuclear Power Sustainability Path for China from the Perspective of Operations Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-11-28 Tao Zhang, Shengzhi Liu, Weiwei Pan, Tian Wan, Chenhui Dong
Nuclear power, as a low-carbon, stable, and efficient energy, plays an important role in replacing fossil fuels in the development of a globally sustainable energy system. However, nuclear power has deviated from the path to achieve the Sustainable Development Goals of the United Nations. The path of sustainable nuclear power for China was proposed based on an analysis of the development of global
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Investigation of Early Corrosion Behavior of Canister Candidate Materials in Oxic Groundwater by the EQCM Method Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-11-17 Gha-Young Kim, Sung-Wook Kim, Junhyuk Jang, Seok Yoon, Jin-Seop Kim
This study investigated the corrosion mass changes of canister candidate materials (Cu, Ni, Ti, SS304) in an oxic groundwater solution using the electrochemical quartz crystal microbalance method in order to estimate corrosion thickness. The materials were immersed in naturally aerated groundwater with and without the addition of chloride ions to observe the mass changes as well as the open-circuit
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Development of an MPS Code for Corium Behavior Analysis: 3D Alloy Melting Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-11-3 Lijun Jian, Peng Yu, Jie Pei, Xiao Zeng, Yidan Yuan
The moving particle semi-implicit (MPS) method as a Lagrangian method is attracting increasing attention in severe accident analysis. In this paper, we developed an MPS code for the corium behavior analysis with several additional models added: an improved heat transfer model to improve the calculation between different materials, an enthalpy-based viscosity model to realize a smooth transition of
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Feasibility Study on the Initial Kartini Reactor Core Using Plate Type Fuel Elements Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-10-14 Argo Satrio Wicaksono, Satoshi Takeda, Takanori Kitada
The plate type fuel element conversion is proposed to solve a supply problem of TRIGA standard rod type fresh fuel in the long term and to extend the lifetime by reducing the dependence of buying imported elements. The plate type fuel is an alternative since the Indonesian industry has been able to produce such fuel elements. The change of core configuration is expected to improve the reactor performance
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The Study of Dosimetric Characteristics of the XHA600D Medical Linear Accelerator Based on a Monte Carlo Code Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-9-29 Ningyu Wang, Fengjie Cui, Shaoxian Gu, Chuou Yin, Shengyuan Zhang, Jinyou Hu, Yunzhu Cai, Zhangwen Wu, Jun Wang, Chengjun Gou
By investigating the influence of initial electrons on dosimetric characteristics, reasonable incident electron parameters for the nominal 6 MV photon beam of the XHA600D accelerator are finally established, i.e., a 6 MeV monoenergetic electron beam with a radial intensity FWHM of 2.5 mm and an angular divergency of 0.15°. Based on reasonable initial parameters, Percentage Depth Doses (PDDs), Off-Axis
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Research on Radionuclide Diffusion Mechanism in the Ocean and Emergency Response under Oceanic Radioactive Events Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-9-24 Zichao Li, Rongchang Chen, Chen Liu, Qingqing Xue, Zhixia Wang, Tao Zhou
On March 11, 2011, a serious radionuclide leakage accident occurred at Fukushima Daiichi nuclear power plant, and a large number of radionuclides were released, causing serious pollution to the ocean environment. On August 25, 2021, Japan announced the overall plan for the discharge of radioactive sewage from the Fukushima Daiichi nuclear power plant into the ocean, and the discharge will begin around
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Experimental and Numerical Studies of AP1000 Shield Building considering Fluid-Structure Interaction Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-09-02 Zhi Zhang, Chenning Song, Zhining Duan, Zhi Cheng
The gravity cooling water tank is a remarkable structural feature of third-generation pressurized water reactor nuclear power plant. To investigate the influence of fluid-structure interaction (FSI) on the seismic response of the structure, this study designed two 1 : 50 simplified models of the AP1000 shield building. A series of shaking table tests were conducted to study the seismic responses with
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Machine Learning-Based Approach for Hydrogen Economic Evaluation of Small Modular Reactors Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-09-01 Juyoul Kim, Mujuni Rweyemamu, Boldsaikhan Purevsuren
In this study, we evaluate hydrogen production costs using small modular reactors (SMRs). Furthermore, we employ a machine learning-based approach to predict important parameters that affect the hydrogen production cost. Additionally, we use a hydrogen economic evaluation program to calculate the hydrogen production cost when using the two types of SMRs: system-integrated modular advanced reactor (SMART)
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Numerical Study of Unsteady Pressure Fluctuation at Impeller Outlet of a Centrifugal Pump Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-08-30 Xiaojie Ma, Lulu Zheng, Jinglei Qu, Mengmeng Wang
Intense fluid-dynamic interaction at the impeller outlet strongly affects the unsteady flow and pressure stability within the centrifugal pump. In order to have a better understanding of the pressure fluctuation of centrifugal pumps, a numerical calculation is carried out by using the RNG k-epsilon turbulence model under various flow rates. The numerical calculation results are compared with the experimental
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A Hybrid Method to Predict the Remaining Useful Life of Scroll Wheel of Control Rod Drive Mechanism Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-08-29 Kang Zhu, Xinwen Zhao, Liming Zhang, Hang Yu
As one of the rotating components in the reluctance motor type control rod drive mechanism (CRDM), the life of the scroll wheel is closely related to the service life of the CRDM. In addition, the prediction of the remaining useful life of the scroll wheel helps to optimize the maintenance process of the CRDM. This paper proposes a hybrid method to predict its remaining useful life when the available
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Numerical Study of Natural Circulation Flow in Reactor Coolant System during a Severe Accident Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-08-29 Dae Kyung Choi, Won Man Park, Sung Man Son, Kukhee Lim, Yong Jin Cho, Choengryul Choi
The rupturing of steam generator tubes leads to serious accidents in nuclear power plants. It causes radioactive materials to leak into the secondary system and release outside the reactor containment region. Therefore, it is important to model a technique to determine whether the natural circulation within a reactor coolant system (RCS) can cause rupture. In this study, a computational fluid dynamics
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Debris Bed Self-Leveling Mechanism and Characteristics for Core Disruptive Accident of Sodium-Cooled Fast Reactor: Review of Experimental and Modeling Investigations Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-08-22 Ruicong Xu, Songbai Cheng
Evaluations of the Core Disruptive Accident (CDA) are significantly important for safety analysis of Sodium-cooled Fast Reactor (SFR) despite the very low probability of occurrence for CDA. During the material-relocation phase in CDA of SFR, the molten materials are possibly released from the core region into subcooled sodium, subsequently forming the debris bed on the lower part of the reactor vessel
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Investigation of Oxidation and Counter-Oxidation in a One-Quarter Circular Geometry due to Shadow Corrosion Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-08-09 Doctor Enivweru, Qingyu Wang, Abiodun Ayodeji, Yu Zhou
To optimize fission fuel and protect cladding integrity, this work investigates shadow corrosion in a one-fourth circular electrode geometry. The anodic corrosion of Zircaloy-2 (Zry-2) was investigated in a circular geometry electrode configuration under reactor operating conditions. The impact of gamma and neutron radiations on water conductivity and shadow corrosion was examined under two different
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Coincidence Summing Factor Calculation for Volumetric γ-ray Sources Using Geant4 Simulation Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-08-08 Dalal Abdullah Aloraini, Mohamed Elsafi, Aljawhara H. Almuqrin, M.I. Sayyed
Geant4 simulation was applied to correct the coincidence summing (CS) effect in detecting a volumetric γ-ray sources, and this technique was applied to a152Eu standard sources. The radioactive sources were a liquid cylindrical, rectangular, and Marinelli beaker shapes of different volume for each one. Radionuclide track (RT) including coincidence summing and monoenergetic track without coincidence
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Effect of Chloride Ions on Electrochemical Behavior of Canister Materials Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-08-03 Gha-Young Kim, Junhyuk Jang, Minsoo Lee, Jin-Seop Kim
Various canister candidate materials (SS316L, Ti-Gr.2, Alloy 22, and Cu) were studied using groundwater at the Korea Atomic Energy Research Institute (KAERI) underground research tunnel (KURT), with the addition of chloride ions using different electrochemical techniques. The corrosion potential and corrosion current of test materials were obtained by the polarization measurement. The polarization
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Characterization of Waste Generated from Nuclide Management Process in Waste Burden Minimization Technology for Spent Nuclear Fuel Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-07-30 Jung-Hoon Choi, Byeonggwan Lee, Ki-Rak Lee, Hyun Woo Kang, Hyeon Jin Eom, Seong-Sik Shin, Ga-Yeong Kim, Hwan-Seo Park
To reduce the environmental burden caused by the disposal of spent nuclear fuel, waste burden minimization technology is currently being developed at the Korea Atomic Energy Research Institute. The technology includes a nuclide management process that can maximize disposal efficiency by selectively separating and collecting major nuclides in spent nuclear fuel. To manufacture a waste form of high durability
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An Improved Steady-State and Transient Analysis of the RSG-GAS Reactor Core under RIA Conditions Using MTR-DYN and EUREKA-2/RR Codes Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-07-30 Surian Pinem, Sukmanto Dibyo, Wahid Luthfi, Veronica Indriati Sri Wardhani, Donny Hartanto
Steady-state and transient analysis of reactor core under Reactivity-Initiated Accident (RIA) conditions are important for reactor operation safety. The reactor dynamics are influenced by neutronic and thermal-hydraulic aspects of the core. In this study, steady-state and transient analysis under RIA conditions of the RSG-GAS multipurpose reactor was carried out using MTR-DYN and EUREKA-2/RR programs
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Remaining Useful Life Prediction of Nuclear Power Machinery Based on an Exponential Degradation Model Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-06-16 Gaojun Liu, Weijie Fan, Fenglei Li, Gaixia Wang, Dongdong You
Aiming at solving the problems of small fault data samples and insufficient remaining useful life (RUL) prediction accuracy of nuclear power machinery, a method based on an exponential degradation model is proposed to predict the RUL of equipment after the failure warning system alarm. After data preprocessing, time-domain feature extraction, selection, and dimensionality reduction fusion of multiple
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Study on the Dispersion of Radionuclides under Different Hydrological Conditions of Spent Fuel Shipping in Daya Bay Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-06-10 Liwei Chen, Weihua Chen, Jiazhen Lin, Chunhua Chen, Yalin Luo, Longlong Tao
The radionuclide dispersion in coastal water is mainly controlled by the water flow and tidal effect. Tracing and analysis of radioactive pollutant dispersion in coastal water can predict distribution of radionuclide under marine transportation accident of spent fuel. In this work, factors such as continuous emission, radioactive decay, and water depth are considered, and a hydrodynamic model of radionuclide
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Uranium Recovery from Phosphates for Self-Sufficient Nuclear Power in the Eastern Mediterranean Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-06-09 Nassar H. S. Haidar
Production of phosphate fertilizers (PF), without uranium recovery, amounts to dispersing uranium compounds on agricultural fields. These compounds are naturally hidden in phosphate rock deposits prior to processing. Such a dispersion is a cumulative environmental damage, that may become rather catastrophic in few hundred years, under the current rates & impurities of phosphate fertilization of agricultural
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Cold Atmospheric Plasma Inhibits the Proliferation of CAL-62 Cells through the ROS-Mediated PI3K/Akt/mTOR Signaling Pathway Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-06-08 Fang Liu, Yuanyuan Zhou, Wencheng Song, Hongzhi Wang
This study aimed to investigate the inhibitory effects of cold atmospheric plasma (CAP) on anaplastic thyroid cancer cells (CAL-62 cells) and to reveal the molecular mechanism. The effects of CAP on CAL-62 cells were evaluated by cell viability, superoxide dismutase activity, apoptosis, cell cycle, and protein expression level, and the role of reactive oxygen species (ROS) produced by plasma was also
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An Information Granulated Based SVM Approach for Anomaly Detection of Main Transformers in Nuclear Power Plants Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-06-03 Wenmin Yu, Ren Yu, Cheng Li
The main transformer is critical equipment for economically generating electricity in nuclear power plants (NPPs). Dissolved gas analysis (DGA) is an effective means of monitoring the transformer condition, and its parameters can reflect the transformer operating condition. This study introduces a framework for main transformer predictive-based maintenance management. A condition prediction method
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Investigation of Loss of Feedwater (LOFW) Accident in the APR-1400 Using Fault Tree Analysis Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-05-26 Muhammad Zubair
Nuclear power plants play a significant role in the contribution of electricity generation on a global scale. Various reactor designs have advantages over others in different aspects. APR-1400 is a pressurized water reactor that is deemed safe due to the redundancy and independence of the multiple safety systems. Probabilistic safety assessment (PSA) is well known for its effectiveness in the representation
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Influence of Weak Compressibility on the Hydrodynamic Performance Evaluation of Pump Turbines in the Pump Mode Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-05-20 Fangfang Zhang, Na Li, Di Zhu, Ruofu Xiao, Weichao Liu, Ran Tao
In general, weak compressibility is one of the properties of liquids. That is, in actual operation of hydraulic machinery, the flow is weakly compressible. However, the influence of weak compressibility is often neglected in usual numerical simulation, which makes the simulation results different from the experimental results. Based on the Computational Fluid Dynamics (CFD) solver and model test rig
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Design of Control System of Once-Through Steam Generator Based on Proximal Policy Optimization Algorithm Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-05-20 Cheng Li, Ren Yu, Wenmin Yu, Tianshu Wang
Because of the characteristics of the small water volume of OTSG, it is hard to control the outlet steam pressure when the load is changed or disturbed. This study is devoted to the control of the once-through steam generator (OTSG). A double-layer controller based on the PPO algorithm is proposed to control the outlet steam pressure of OTSG. The bottom layer is the PID controller; it directly regulates
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Pressure Distribution on the Inner Wall of the Volute Casing of a Centrifugal Pump Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-05-19 Yu-Liang Zhang, Jin-Fu Li, Tao Wang, Jun-Jian Xiao, Xiao-Qi Jia, Li Zhang
In order to grasp the distribution characteristics of the pressure field on the inner wall of the volute casing of an atypical open impeller centrifugal pump, the instantaneous pressure at different operating conditions was experimentally measured under four operating rotational speeds to obtain the distribution characteristics of the average static pressure field in the volute casing of this pump
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Determination of the Activity Inventory in the Structural Components of the Dalat Nuclear Research Reactor for Its Decommissioning Planning Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-05-18 Quoc-Duong Tran, Kien-Cuong Nguyen, Ba-Vien Luong, Ton-Nghiem Huynh, Quang-Huy Pham, Doan-Hai-Dang Vo, Nhi-Dien Nguyen
This report presents the methods and calculated results of the activity inventory in the structural components of the Dalat Nuclear Research Reactor (DNRR). These components include the shielding concrete, the reactor tank, and its inside irradiated facilities; the thermal and thermalizing columns; and the horizontal channels. The MCNP5 code with a three-dimensional neutron transport model was used
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Preliminary Study on Risk Identification and Assessment Framework for Fusion Radioactive Waste Management Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-05-17 Dingqing Guo, Jinkai Wang, Chao Chen, Dongqin Xia, Nuo Yong, Daochuan Ge
Fusion reactors are expected to be safer, more environmentally friendly, and to have a lower nuclear proliferation risk, compared with other nuclear energy systems. However, it is widely recognized that a large amount of radioactive materials will be produced by a fusion reactor. Therefore, it is important to fully understand the overall radiation risk level of fusion radioactive wastes (radwaste)
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The Concept of the Heat Removal System of a High-Flux Research Reactor Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-04-27 Vitaly Uzikov, Ildar Suleimanov, Irina Uzikova
Achieving high neutron fluxes in research pressurized water reactors is directly related to the intensity of the coolant flow through the core and the pressure in it, which provides an increased saturation temperature and a margin to critical heat flux. Therefore, it is practically impossible to provide very high neutron fluxes in pool-type reactors, especially in the case of downward movement of the
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Design, Experiment, and Commissioning of the Spent Fuel Conveying and Loading System of HTR-PM Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-04-23 Bin Wu, Jinhua Wang, Yue Li, Haitao Wang, Tao Ma
The Chinese high-temperature gas-cooled reactor pebble-bed module, HTR-PM, began fuel loading in August 2021. The reactor refuels continuously, while the spent fuel is discharged from the core. The spent fuel conveying and loading system was designed to convey the spent fuel pebbles to the spent fuel building and load them into dry canisters for on-site interim storage. This study describes the operating
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Oceanic Radionuclide Dispersion Method Investigation for Nonfixed Source from Marine Reactor Accident Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-04-18 Dingqing Guo, Jinkai Wang, Daochuan Ge, Chunhua Chen, Liwei Chen
Radionuclide dispersion model, which is of critical importance to the emergency response of severe nuclear accident, is used to estimate the consequences arising from accidental or routine releases and to predict areas of high contamination. It is difficult to evaluate the radioactive consequence accurately and rapidly for the accidental release of radionuclides from marine reactor because of the complex
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Development and Testing of TRACE/PARCS ECI Capability for Modelling CANDU Reactors with Reactor Regulating System Response Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-03-27 Simon Younan, David R. Novog
The use of the USNRC codes TRACE and PARCS has been considered for the coupled safety analysis of CANDU reactors. A key element of CANDU simulations is the interactions between thermal-hydraulic and physic phenomena with the CANDU reactor regulating system (RRS). To date, no or limited development has taken place in TRACE-PARCS in this area. In this work, the system thermal-hydraulic code TRACE_Mac1
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Investigation of the Flow and Heat Transfer Characteristics and Erosion Law of Particulate in LBE on the Subchannel Sci. Technol. Nuclear Install. (IF 1.1) Pub Date : 2022-03-22 Bingheng Zhu, Qi Xu, Pengxiang Li
A triangle subchannel model was established to study the flow and heat transfer characteristics of lead-bismuth eutectic (LBE) alloy and the erosion rate of the core channel by the particulate in LBE. Under different inlet velocities, particle types, particle diameters, and particle concentrations, the erosion law of the channel wall in LEB was investigated by using a discrete phase model (DPM). The