Subjects -> ENERGY (Total: 414 journals)
    - ELECTRICAL ENERGY (12 journals)
    - ENERGY (252 journals)
    - ENERGY: GENERAL (7 journals)
    - NUCLEAR ENERGY (40 journals)
    - PETROLEUM AND GAS (58 journals)
    - RENEWABLE ENERGY (45 journals)

ENERGY (252 journals)            First | 1 2 | Last

Showing 201 - 400 of 406 Journals sorted alphabetically
Natural Resources     Open Access  
Nature Energy     Hybrid Journal   (Followers: 27)
Nigerian Journal of Technological Research     Full-text available via subscription   (Followers: 1)
Nuclear Data Sheets     Full-text available via subscription  
Nuclear Engineering and Design     Hybrid Journal   (Followers: 16)
Oil and Energy Trends : Annual Statistical Review     Full-text available via subscription  
Oil and Gas Journal     Full-text available via subscription   (Followers: 12)
Open Journal of Energy Efficiency     Open Access   (Followers: 1)
Power Technology and Engineering     Hybrid Journal   (Followers: 3)
Proceedings of the Institution of Civil Engineers - Energy     Hybrid Journal   (Followers: 2)
Progress in Energy and Combustion Science     Hybrid Journal   (Followers: 16)
Progress in Nuclear Energy     Hybrid Journal   (Followers: 2)
Protection and Control of Modern Power Systems     Open Access   (Followers: 3)
Radioprotection     Hybrid Journal   (Followers: 1)
Science and Technology for Energy Transition     Open Access   (Followers: 3)
Science and Technology of Nuclear Installations     Open Access   (Followers: 3)
Smart Energy     Open Access  
Smart Grid and Renewable Energy     Open Access   (Followers: 9)
Solar Compass     Open Access   (Followers: 1)
Solar Energy     Hybrid Journal   (Followers: 20)
Solar Energy Advances     Open Access   (Followers: 2)
Solar Energy Materials and Solar Cells     Hybrid Journal   (Followers: 29)
South Pacific Journal of Natural and Applied Sciences     Hybrid Journal  
Strategic Planning for Energy and the Environment     Hybrid Journal   (Followers: 4)
Structural Control and Health Monitoring     Hybrid Journal   (Followers: 6)
Surface Science Reports     Full-text available via subscription   (Followers: 13)
Sustainable Energy     Open Access   (Followers: 2)
Sustainable Energy & Fuels     Hybrid Journal   (Followers: 1)
Sustainable Energy Technologies and Assessments     Full-text available via subscription  
Sustainable Energy, Grids and Networks     Hybrid Journal   (Followers: 4)
Technology and Economics of Smart Grids and Sustainable Energy     Hybrid Journal  
Technology Audit and Production Reserves     Open Access   (Followers: 1)
Turkish Journal of Energy Policy     Open Access  
Unconventional Resources     Open Access  
Universal Journal of Applied Science     Open Access  
Washington and Lee Journal of Energy, Climate, and the Environment     Open Access  
Waste Management     Hybrid Journal   (Followers: 13)
Water International     Hybrid Journal   (Followers: 19)
Wiley Interdisciplinary Reviews : Energy and Environment     Hybrid Journal   (Followers: 8)
Wind Energy     Hybrid Journal   (Followers: 3)
Wind Engineering     Hybrid Journal  
World Oil Trade     Hybrid Journal  

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Science and Technology of Nuclear Installations
Journal Prestige (SJR): 0.522
Citation Impact (citeScore): 1
Number of Followers: 3  

  This is an Open Access Journal Open Access journal
ISSN (Print) 1687-6075 - ISSN (Online) 1687-6083
Published by Hindawi Homepage  [339 journals]
  • Investigation of Oxidation and Counter-Oxidation in a One-Quarter Circular
           Geometry due to Shadow Corrosion

    • Abstract: 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 cathodes. This work also investigates the effect of current exchange density and the cathodic Tafel coefficient on the cathodic current. Using COMSOL Multiphysics 5.2, the Laplace equation was solved to obtain the electrostatic potential and current density distributions in the studied domain. When the distance d between the anode (Zry-2) and cathode (platinum/nickel) is ≤0.5 mm, it was observed that a uniform oxide layer of thickness 20 µm grew on the smooth internal surface of Zry-2 for corrosion lasting 1166 h. When d > 0.5 mm, the oxide thickness falls in a manner dictated by the degree of dissociation α of the electrolyte. At a cladding gap of 0.08 mm, a radiation-enhanced uniform corrosion rate of 2.405 10−1 mmpy was obtained for Zry-2. This value is 142 times greater than that obtained at room temperature in the absence of radiation. It was also observed that the corrosion rate falls at higher cladding gaps, and the rate of change depends on the degree of dissociation. Other phenomena such as the dynamics of shadow corrosion under varying electrode separation and electrolyte conductivities, as well as extensive evaluation of critical fuel cladding parameters, are presented in this work.
      PubDate: Tue, 09 Aug 2022 07:05:02 +000
       
  • Coincidence Summing Factor Calculation for Volumetric γ-ray Sources
           Using Geant4 Simulation

    • Abstract: 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 summing. The results obtained from two approaches compared with the experimental data and the modified KORSUM code for cylindrical γ-ray source. The comparison showed that the adopted method in this investigation is useful for coincidence summing corrections for a voluminous γ-ray sources.Moreover, this technique requires far less computation time than the techniques that depend on the calculation of total efficiency.
      PubDate: Mon, 08 Aug 2022 09:05:01 +000
       
  • Effect of Chloride Ions on Electrochemical Behavior of Canister Materials

    • Abstract: 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 measurements revealed that the addition of chloride ions was detrimental to the SS316L and Cu materials by increasing corrosion current, which is an indicator of corrosion rate. Impedance measurements and fitting analysis showed that the corrosion resistance of Cu was more than 10 times lower than that of other materials in the KURT groundwater solution containing 0.1 M of chloride ions.
      PubDate: Wed, 03 Aug 2022 03:50:01 +000
       
  • Characterization of Waste Generated from Nuclide Management Process in
           Waste Burden Minimization Technology for Spent Nuclear Fuel

    • Abstract: 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, the characteristics of the waste generated during the process should be evaluated. In this study, the physical, radiological, and thermal characteristics of the waste and waste forms for major nuclides (Cs, Sr, I, transuranic/rare earth, and Tc/Se) generated in the nuclide management process were analyzed. In the case of Cs nuclides, characterization was conducted according to the capture rate of the adsorbent in the high-temperature heat treatment process; meanwhile, in the case of Sr nuclides, characterization was performed by considering the ratio of similar nuclides in the chlorination process. For I nuclide, analysis was performed based on the available waste form, and for TRU/RE and Tc/Se nuclides, analysis was performed by considering chlorination and mid-temperature heat treatment. The radioactivity and heat generation rate of each waste and waste form were evaluated over a period of 1,000 years. The results of this study could be used to derive the centerline temperature for the thermal stability evaluation of waste forms and for the feasibility evaluation of each disposal system considered in the waste burden minimization technology.
      PubDate: Sat, 30 Jul 2022 08:20:05 +000
       
  • An Improved Steady-State and Transient Analysis of the RSG-GAS Reactor
           Core under RIA Conditions Using MTR-DYN and EUREKA-2/RR Codes

    • Abstract: 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. Neutronic calculations were performed using a few group cross-sections generated by Serpent 2 with the latest cross-section data ENDF/B-VIII.0. Steady-state conditions were carried out with a nominal power of 30 MW, while transient under RIA conditions occurred because the control rod was pulled too quickly while the reactor operated. These transient RIA conditions were performed for two cases, during start-up with an initial power of 1 W, and within power range with an initial power of 1 MW. Thermal-hydraulic parameters considered in this study are reactor power, the temperature of the fuel, cladding, and coolant. The calculated maximum fuel temperature at a steady state is 126.02°C. Meanwhile, the calculated maximum fuel temperature during RIA conditions at the initial power of 1 W and 1 MW are 64.38°C and 137.14°C, respectively. There are no significant differences in thermal-hydraulic parameters between each used program. The thermal-hydraulic parameters such as the maximum temperature of the coolant, cladding, and fuel under this postulated RIA condition are within the acceptable reactor operation safety limits.
      PubDate: Sat, 30 Jul 2022 06:50:03 +000
       
  • Remaining Useful Life Prediction of Nuclear Power Machinery Based on an
           Exponential Degradation Model

    • Abstract: 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 degradation variables, the exponential degradation model is constructed based on the Bayesian process, and prior information is used. As an application, the RUL of a nuclear power turbine was calculated based on actual monitoring data, the precision curve was used to evaluate the prediction effect, and the RUL prediction results verified the effectiveness of the proposed method.
      PubDate: Thu, 16 Jun 2022 11:50:04 +000
       
  • Study on the Dispersion of Radionuclides under Different Hydrological
           Conditions of Spent Fuel Shipping in Daya Bay

    • Abstract: 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 dispersion based on shallow water equations is established to simulate the dispersion of the radioactive pollutant in coastal waters under different hydrological conditions. As far as the characteristics of the radionuclide dispersion in coastal water are concerned, the simulation of pollutants by the hydrodynamic model is in good agreement with the work of Bailly du Bois et al., which validated the correctness of this model. The model has been applied to simulate the distribution of radionuclides in coastal water following a marine transport accident of spent fuel near Daya Bay Nuclear Power Plant in China. The simulation reveals that the distribution features are significantly affected by different hydrological conditions. In addition to limiting the diffusion range, the vortex effect can also cause the accumulation of radionuclides near the vortex, which helps to provide more practical information for nuclear emergency decision makers.
      PubDate: Fri, 10 Jun 2022 06:35:01 +000
       
  • Uranium Recovery from Phosphates for Self-Sufficient Nuclear Power in the
           Eastern Mediterranean

    • Abstract: 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 lands. It is also an avoidable irreversible waste of one of the world’s major energy resources. This study demonstrates quantitatively the low impact of U costs on the nuclear power generation costs, which happens, so far, to be a main reason for nonrecovery of uranium from the present PF industry. It reports on novel procedures for (i) estimating the required U feed to nuclear power plants (NPPs), (ii) pricing U as a function of its cumulative world production, and (iii) for quantifying U accumulation in phosphate fertilized lands. We also demonstrate that countries of the eastern Mediterranean can, in the long run, become collectively U partially self-sufficient, by recovering U from their phosphate resources, to power 13.2% of their entire electric energy generation contemporary needs.
      PubDate: Thu, 09 Jun 2022 04:20:01 +000
       
  • Cold Atmospheric Plasma Inhibits the Proliferation of CAL-62 Cells through
           the ROS-Mediated PI3K/Akt/mTOR Signaling Pathway

    • Abstract: 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 investigated. The results showed that CAP dose-dependently inhibited cell viability and promotes cell apoptosis and G2/M arrest by increasing cell ROS levels. The activity of superoxide dismutase (SOD) was enhanced by CAP which indicated that the antioxidant system of the cell was activated. Additionally, the ROS produced by CAP can inhibit CAL-62 cell proliferation by inhibiting the PI3K/Akt/mTOR signaling pathway. Therefore, these findings will provide useful support for the application of CAP for treating anaplastic thyroid cancer.
      PubDate: Wed, 08 Jun 2022 10:20:02 +000
       
  • An Information Granulated Based SVM Approach for Anomaly Detection of Main
           Transformers in Nuclear Power Plants

    • Abstract: 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 based on the online support vector machine (SVM) regression model is proposed, with the input data being preprocessed using the information granulation method, and the parameters of the model are optimized using the particle swarm optimization (PSO) algorithm. Using DGA data from the NPP data acquisition system, two experiments are designed to verify the trend tracing and prediction envelope ability of main transformers installed in NPPs with different operating ages of the proposed model. Finally, how to use this framework to benefit the maintenance plan of the main transformer is summarized.
      PubDate: Fri, 03 Jun 2022 11:50:01 +000
       
  • Investigation of Loss of Feedwater (LOFW) Accident in the APR-1400 Using
           Fault Tree Analysis

    • Abstract: 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 of risk and safety analysis of the systems in a nuclear power plant. It provides different scenarios of system failure and accident progression via fault tree analysis. A loss of feedwater (LOFW) accident may occur due to numerous reasons such as spurious closure of valves, component failure of heaters, pumps, tanks, or a loss of offsite power (LOOP) event. In the present research, a methodology has been developed that aims to investigate different factors contributing to the loss of feedwater. This research also aims to analyze LOFW accidents by developing fault tree models for the main feedwater system of the APR-1400 to identify the basic events, which may lead to a loss of feedwater accidents. The results of the top event probabilities, risk decrease factor (RDF), risk increase factor (RIF), minimal cut sets (MCS), basic event probabilities, and sensitivity analysis were compared with the WASH-1400 database. It has been found that the control valve (V04) and main feedwater isolation valve (V05) have more contribution to the LOFW accident. The common cause failure (CCF) analysis has been carried out, and it was found that the flow toward the check valve and steam generator are most critical for CCF.
      PubDate: Thu, 26 May 2022 09:20:02 +000
       
  • Influence of Weak Compressibility on the Hydrodynamic Performance
           Evaluation of Pump Turbines in the Pump Mode

    • Abstract: 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, by means of mutual verification between numerical simulation and experiment, the fitting degree between numerical results and experimental results before and after considering weak compressibility is compared and analyzed in this paper; it is obtained that the numerical results is closer to the experimental results after considering the weak compressibility. In addition, velocity field of pump turbines, head loss of main components, and the change of entropy yield are analyzed and reasons for numerical value being closer to the experimental value after considering weak compressibility of fluid are summarized and analyzed. It is proved that the consideration of weak compressibility is of great significance to improve the accuracy of results in the numerical simulation of pump turbines.
      PubDate: Fri, 20 May 2022 08:05:01 +000
       
  • Design of Control System of Once-Through Steam Generator Based on Proximal
           Policy Optimization Algorithm

    • Abstract: 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 the OTSG feed water valve and then controls the steam pressure. The top layer of the controller is the agent based on the PPO algorithm, which is responsible for optimizing the parameters of the PID in real time to obtain better control performance. The agent chooses PID parameters as actions to the environment, and then, the reward value is obtained through the reward function of the environment which enables online learning of the agent. Compared with the PID controller, the simulation experiment result shows that the method not only has a good control performance but also has a good anti-interference ability.
      PubDate: Fri, 20 May 2022 06:35:02 +000
       
  • Pressure Distribution on the Inner Wall of the Volute Casing of a
           Centrifugal Pump

    • Abstract: 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 model. The pressure pulsation amplitude and pressure pulsation intensity were also analyzed at different rotational speed cases, and the standard deviation analysis was performed. The results showed that the instantaneous pressure pulsation on the inner wall of the volute casing strongly fluctuates during the pump operating, and the closer to the volute casing outlet, the more intense the pressure pulsation was. After increasing the pump shaft speed, the fluctuation amplitude gradually decreased. The pressure pulsation on the wall of tip clearance is more intense than that on the inner wall of the volute shell. The intensity of the pressure pulsation on the wall of tip clearance decreases with the increase of the rotational speed, and the higher the speed, the less intense the pressure pulsation.
      PubDate: Thu, 19 May 2022 17:35:03 +000
       
  • Determination of the Activity Inventory in the Structural Components of
           the Dalat Nuclear Research Reactor for Its Decommissioning Planning

    • Abstract: 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 to calculate the neutron flux distribution, neutron energy spectrum at different locations, and activation cross sections of long-lived radioactive nuclides in activated major materials, including heavy concrete, reflection graphite, and aluminum of the reactor. The calculated results of the energy spectrum and activation cross sections of MCNP5 were used in the ORIGEN2.1 point depletion code to calculate the neutron-induced activity of activated materials at different time points by modeling the irradiation history and radioactive decay. Radioactivity of long-lived key activation products such as 41Ca, 60Co, 55Fe, 63Ni, and 152Eu was modeled, and volumes of radioactive waste mainly of ordinary concrete, graphite, and aluminum in the structural components of the reactor were estimated. Experimental results of neutron flux and specific activities of some typical nuclides such as 60Co, 152Eu, 55Fe, and 63Ni in activated aluminum samples showed good agreement with the calculated results. As part of the national regulation requirements, the obtained data have been used to develop the decommissioning plan for the operational DNRR, with an estimation of about 10 years before its permanent shutdown.
      PubDate: Wed, 18 May 2022 12:05:01 +000
       
  • Preliminary Study on Risk Identification and Assessment Framework for
           Fusion Radioactive Waste Management

    • Abstract: 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) compared with existing nuclear energy systems. Studies on the treatment of the fusion radwaste have been currently focused on three ultimate options: clearance, recycling, and disposal by activation assessment of radioactive materials from the operation and decommissioning of fusion reactors. However, the radiation risk in the management of fusion radwaste, especially in the final disposal, was seldom studied. Based on the comparative analysis of fusion radioactive waste with ITER and fission reactors (e.g., pressurized water reactor, PWR), this paper tries to discuss how to determine the radiation risk in the process of fusion radwaste management on the premise of the current feasible industrial technology. On this basis, a risk assessment framework for repository disposal under normal degradation and external events is proposed.
      PubDate: Tue, 17 May 2022 15:05:02 +000
       
  • The Concept of the Heat Removal System of a High-Flux Research Reactor

    • Abstract: 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 coolant in the core. At the same time, vessel-type research reactors (for example, SM-3 and HFIR) make it possible to achieve neutron flux densities up to 4 × 1015 n/(cm2 s), but at the same time, the risks of core degradation in case of violations in the heat removal system become quite high. The proposed concept of a heat removal system for a high-flux reactor facility combines the increased reliability of safe heat removal from the core and the convenience of handling irradiation cells, for example, in the production of isotopes. The concept provides for the location of a compact core in a pressurized vessel and the placement of a neutron reflector around the vessel in the reactor pool. Cooling of the reactor core in the housing and the irradiation channels in the neutron reflector is carried out by different systems of forced circulation of the coolant. At the same time, at the shutdown reactor, after opening the natural circulation valves, safe heat removal from the reactor core and the neutron reflector can be carried out by the water of the reactor pool. However, even with a complete failure of all forced circulation circuits, the evaporation of water from the surface of the pool makes it possible to safely remove the residual heat from the fuel assemblies and from the irradiation devices in the cells of the reflector.
      PubDate: Wed, 27 Apr 2022 11:50:03 +000
       
  • Design, Experiment, and Commissioning of the Spent Fuel Conveying and
           Loading System of HTR-PM

    • Abstract: 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 principles of the main functions and introduces the experiments and commissioning tests of the system. Functional tests were carried out to indicate the items of mechanical and electrical equipment are functioning in accordance with the designed requirements. Experience learned from commissioning activities was also presented as feedback for future operation and design improvement.
      PubDate: Sat, 23 Apr 2022 06:50:02 +000
       
  • Oceanic Radionuclide Dispersion Method Investigation for Nonfixed Source
           from Marine Reactor Accident

    • Abstract: 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 mobility feature in the sea. Based on CFD method, a finite-volume, three-dimensional regional oceanic dispersion model was developed in this paper to simulate the dispersion of radionuclides originating from marine reactor. The simulated dose variation of 137Cs presented good agreement with the monitoring data of marine radioactive pollution caused by Fukushima Dai-ichi nuclear accident, which demonstrated the validity of the method. A severe accident scenario of marine reactor was simulated and analyzed, which indicates that the regional oceanic dispersion model can provide dose assessment for nuclear emergency response.
      PubDate: Mon, 18 Apr 2022 11:50:02 +000
       
  • Development and Testing of TRACE/PARCS ECI Capability for Modelling CANDU
           Reactors with Reactor Regulating System Response

    • Abstract: 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_Mac1.0 is natively coupled with the core physic code PARCS_Mac1.0, and RRS control is implemented via the exterior communications interface (ECI) in TRACE. ECI is used for coupling the external codes to TRACE, including additional physical models and control system models. In this work, a Python interface to the TRACE ECI library is developed, along with an RRS model written in Python. This coupling was tested using a CANDU-6 IAEA code coupling benchmark and a 900 MW CANDU model for various transients. For the CANDU-6 benchmark, the transients did not include RRS response, however, the TRACE_Mac1.0/PARCS_Mac1.0 coupling and ECI script functionality was compared to the previous benchmark simulations, which utilized external coupling. For the 900 MW CANDU simulations, all aspects of the ECI module and RRS were included. The results from the CANDU-6 benchmark when using the built-in coupling are comparable to those previously achieved using external coupling between the two codes with coupled simulations taking 2x to 3x less execution time. The 900 MW CANDU simulations successfully demonstrate the RRS functionality for the loss of flow events, and the coupled solutions demonstrate adequate performance for figure-of-eight flow instability modeling.
      PubDate: Sun, 27 Mar 2022 09:50:01 +000
       
  • Investigation of the Flow and Heat Transfer Characteristics and Erosion
           Law of Particulate in LBE on the Subchannel

    • Abstract: 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 results of this study showed that with the increase of inlet velocity, the outlet temperature of the LEB decreases and the heat transfer capacity was strengthened. The increase of inlet velocity will lead to the increase of erosion rate on the wall, and the change is exponential. The erosion rate of particulate in the low concentration is small but cannot be ignored; with increasing concentration of particulates, the erosion of the wall by particulates becomes serious. The effect of particulate density on the wall erosion rate can be ignored. The effect of changing the particle size on the erosion rate is more significant when the particle size is small, and at the same time, the erosion rate of the particles on the wall increases with the increase of the particle size.
      PubDate: Tue, 22 Mar 2022 13:20:04 +000
       
  • Multiple Assessments on the Gamma-Ray Protection Properties of
           Niobium-Doped Borotellurite Glasses: A Wide Range Investigation Using
           Monte Carlo Simulations

    • Abstract: In this study, the monotonic effect of Ta2O5 and ZrO2 in some selected borotellurite glasses was investigated in terms of their impact on gamma-ray-shielding competencies. Accordingly, three niobium-reinforced borotellurite glasses (S1 : 75TeO2 + 15B2O3 + 10Nb2O5, S2 : 75TeO2 + 15B2O3 + 9Nb2O5 + 1Ta2O5, and S3 : 75TeO2 + 15B2O3 + 8Nb2O5 + 1Ta2O5 + 1ZrO2) were modelled in the general-purpose MCNPX Monte Carlo code. They have been defined as an attenuator sample between the point isotropic gamma-ray source and the detector in terms of determining their attenuation coefficients. To verify the MC results, attenuation coefficients were then compared with the Phy-X/PSD program data. Our findings clearly demonstrate that although some behavioral changes occurred in the shielding qualities, modest improvements occurred in the attenuation properties depending on the modifier variation and its magnitude. However, the replacement of 2% moles of Nb2O5 with 1% mole of Ta2O5 and 1% mole of ZrO2 provided significant improvements in both glass density and attenuation properties against gamma rays. Finally, the HVL values of the S3 sample were compared with some glass- and concrete-shielding materials and the S3 sample was reported for its outstanding properties. As a consequence of this investigation, it can be concluded that the indicated type of additive to be added to borotellurite glasses will provide some advantages, particularly when used in radiation fields, by increasing the shielding qualities moderately.
      PubDate: Fri, 18 Mar 2022 12:05:03 +000
       
  • Experimental Research for CHF Sensitivity of Heat Flux Distribution under
           IVR Conditions

    • Abstract: In-vessel retention (IVR) through external reactor vessel cooling (ERVC) is one of the most effective severe accident mitigation measures in the nuclear power plants. The most influential issues on the IVR strategy are in-vessel core melt evolution, the heat fluxes imposed on the lower head, and the external cooling of reactor pressurized vessel (RPV). In the molten pool research, there are mainly two different molten pool configurations: two layers and three layers. Based on the different distributions of heat flux in molten pool configurations, a new problem was raised: whether the in-vessel heat flux distribution will affect the CHF on the outer wall of RPV and further affect the effectiveness of IVR measures' A full-height external reactor vessel cooling and natural circulating facility was conducted to study the CHF sensitivity of different heat flux distributions. The experimental results show that the characteristics of natural circulation are similar and the CHF of the RPV lower head external surface is not obviously affected under the different heat flux distributions. The varying heat flux distribution during severe accident process will not threaten significantly the success of IVR strategy.
      PubDate: Wed, 16 Mar 2022 05:20:00 +000
       
  • Verification of the Efficacy of Passive Autocatalytic Recombiners in a
           Typical Pressurized Water Reactor under a Station Blackout Condition

    • Abstract: The presence of a stable stratified gas cloud inside the containment near or at the flammability limit may lead to deflagration or even detonation which may challenge the containment and cause a radioactive material release into the environment. To mitigate this risk, a number of approaches have been proposed, for example, containment inerting or venting and use of passive autocatalytic recombiners or igniters. However, for these measures to be effective, a thorough analysis of the hydrogen dispersion and associated phenomena is indispensable during the design phase as well as the mitigation phase during a severe accident. In this work, a MAAP analysis is performed to assess the hydrogen risk in a typical pressurized water reactor (PWR) containment. An extended station blackout (SBO) was chosen as an initiating event given its high contribution to the core damage frequency. RCS depressurization and external injection are mitigation techniques implemented consecutively to extend the coping capability of the plant for the extended SBO scenario. A sensitivity study is performed to select the combination of timing and flow rate that generate the most severe case for the “in-vessel phase of hydrogen generation.” Subsequently, a number of passive autocatalytic recombiners (PARs) were implemented to mitigate the hydrogen risk during the first three days of the accident. The Shapiro diagram is used to assess the flammability condition of the containment atmosphere based on MAAP analysis. The results show that the gas mixture composition is acceptable in the majority of the containment compartments and only marginally acceptable in the cavity. Even under the conservative conditions of the accident, the simulation results confirmed the sufficiency of recombiners alone without igniters in the low hydrogen concentration zones, while for compartments close to the sources, additional mitigation may be needed.
      PubDate: Sat, 12 Mar 2022 04:20:00 +000
       
  • Research on Channel Modeling and Communication Coverage of Wireless Sensor
           Networks in Barrier Area of Nuclear Power Plants

    • Abstract: In view of the multimetal barrier environment of nuclear power plant, by considering the factors such as transmission power, transmission position, and multipath interference, based on the simulation of metal pipes and equipment, this paper carries out the barrier area channel modeling in logarithmic fading mode and makes quantitative analysis on the channel transmission, path loss, channel power characteristics, and so on under the metal barrier environment. Based on the channel modeling, this paper optimizes the coverage of the network in the obstacle area by using the improved teaching and learning group intelligent algorithm. The simulation results show that the improved teaching and learning algorithm can optimize the network coverage of the obstacle area well, and under the four obstacle modules, 14 nodes can cover the whole area by more than 99%. This provides a solution to the problem of network coverage in the practical application of wireless sensor networks.
      PubDate: Mon, 07 Mar 2022 03:35:01 +000
       
  • Interaction of Mechanical Heterogeneity and Residual Stress on Mechanical
           Field at Crack Tips in DMWJs

    • Abstract: The interaction between the mechanical heterogeneity and the residual stress in dissimilar metal welded joints (DMWJs) leads to a complex mechanical field of crack tips, which strongly affects stress corrosion cracking (SCC) behaviors. A dual-field coupling model was established by using the user-defined field (USDFLD) and the predefined stress field method based on the elastoplastic finite element method in this study. Thus, the mechanical heterogeneity and the residual stress of the DMWJ are realized. The influence of the interaction between the mechanical heterogeneity and the residual stress on the mechanical field of crack tips at different locations was investigated. The results show that the mechanical heterogeneity causes the stress and strain distribution on both sides of the crack tip asymmetry. And the residual stress affects the magnitude of the stress and strain around the crack tip. The variation trend of the stress and strain along the crack propagation with crack length is basically the same as that of the residual stress. However, the stress and strain distributions are slightly lagging behind the residual stress distribution due to the redistribution of the residual stress caused by the crack propagation. In addition, the stress and strain range of cracks at different positions with crack length are also different.
      PubDate: Sat, 26 Feb 2022 16:35:01 +000
       
  • Rapid Determination of Gross Alpha/Beta Activity in Water Based on Reverse
           Osmosis Membrane Enrichment Pretreatment

    • Abstract: Radioactivity of gross alpha/beta is an index of water quality detection, which can reflect the radioactivity intensity of water. However, the traditional detection method of these parameters, thick source method, has problems of cumbersome and time consumption in sample preparation and cannot realize the rapid detection on-site. Based on this, this paper studies the enrichment method based on reverse osmosis membrane to accurately and quickly determine the gross α and gross β in water by using the reverse osmosis membrane as the carrier and enriching the radionuclides in water to the high-pressure side of the reverse osmosis membrane to replace the sample preparation process in traditional thick source method, so as to shorten the sample processing time in the detection process and avoid the cumbersome sample preparation process. The reverse osmosis membrane enrichment method for the determination of gross in 241Am and 40KCl standard solutions was used to study gross alpha/beta radioactivity, and the results showed that the average recoveries of radioactivity of gross alpha/beta were 95.0% and 93.6%, respectively. At the same time, the results of the thick source method and the reverse osmosis membrane method on the gross alpha/beta of actual water samples in 5 different regions were compared. It showed that the thick source method and the reverse osmosis membrane method had a good consistency in the detection results of total α and total β radioactivity, and the reverse osmosis membrane method had better stability than the thick source method. The average relative standard deviations (RSD) of the gross alpha and gross beta activity obtained by the thick source method are 11.9% and 7.3%, respectively, while RSD of the gross alpha and gross beta radioactivity obtained by the reverse osmosis membrane method were 6.9% and 4.7%, respectively. The preparation time of single sample was reduced by 75.7%, and the overall detection cycle time was reduced by 68.1%.
      PubDate: Tue, 15 Feb 2022 03:05:03 +000
       
  • Study on Missing Data Filling Algorithm of Nuclear Power Plant Operation
           Parameters

    • Abstract: By analyzing the recorded operation data of a nuclear power plant (NPP), its results can serve the fault detection or operation experience feedback. Data missing exists in the recorded operation data. It may lower the data quality and affect the accuracy of the analysis results. In order to improve the data quality, two parts of researches are carried on. Firstly, to locate the missing data accurately the detecting algorithm for missing data of the NPP operation parameters based on wavelet analysis. Different judging basis is proposed for discrete and continuous missing respectively. Then, the filling method based on the hot deck algorithm are studied. As the dynamic properties of the parameters are closely related to the operating state of NPP, the similarity of the operation parameter vectors are formed to express the similarity of the operating states, so as to fulfill the requirements of the hot deck algorithm. To improve the accuracy of the measuring results, taken the differences between the characteristics of the analog parameters and the switch parameters into consideration, the similarity measurements using Mahalanobis distance for the analog parameter vectors and the matching measure for the switch parameter vectors are studied respectively. Finally, the operation data is taken to build the experiment data set for the algorithm verification. The results shows that the designed algorithm performs much better than the mean interpolation method and LSTM.
      PubDate: Fri, 04 Feb 2022 04:20:01 +000
       
  • Thermomechanical Analysis of a Reactor Pressure Vessel under Pressurized
           Thermal Shock Caused by Inadvertent Actuation of the Safety Injection
           System

    • Abstract: The damage induced pressurized thermal shock (PTS) may pose to a reactor pressure vessel (RPV) is a critical safety requirement assessed as part of the ageing management programme of pressurized water reactors (PWRs). A number of researches have studied PTS initiated mainly by postulated accidents such as loss of coolant accidents (LOCAs). However, investigations on PTS-induced threat on RPV caused by inadvertent actuation of the safety injection, a frequent anticipated transient, have not been thoroughly studied. In this paper, a simplified multistep analysis method is applied to study the thermomechanical status of a two-loop PWR under PTS loads caused by inadvertent actuation of the safety injection system. A direct-coupling thermomechanical analysis is performed using a three-dimensional (3D) RPV finite element model. A 3D finite element submodel (consisting of the highiest stress concentration area in the RPV) and an assumed crack are then used to perform fracture mechanics analysis. Subsequently, the critical integrity parameter-stress intensity factor (SIF) is estimated based on FRANC3D-M-integral method coupled in the multistep simulation. The material fracture toughness of the vessel is computed based on the master curve method with experimental fracture toughness data. The results obtained from the direct coupling stress analysis in comparison with sequential coupling approach demonstrate the effectiveness of the proposed multistep method. Also, comparing SIF results obtained with that calculated based on the conventional virtual crack-closure technique (VCCT) and extended finite element method (XFEM) show good agreement. This study provides a useful basis for future studies on anticipated transient-induced crack propagation and remaining service life prediction of ageing reactor pressure vessels.
      PubDate: Wed, 02 Feb 2022 13:20:01 +000
       
  • Experimental Study of Onset of Nucleate Boiling in Vertical Rectangular
           Channels with Different Flow Path Heights

    • Abstract: A study on ONB (onset of nucleate boiling) in two vertical rectangular channels are experimentally conducted in a range of mass flux varying from 100 to 300 kg/(m2·s), inlet water temperature from 70 to 100°C, heat flux from 10 to 70 kW/m2, and local pressure of 0.145 MPa. The cross-section sizes are 1.8 mm 60 mm and 2.8 mm 60 mm, respectively. Three boiling incipience judgment methods have been used to locate ONB sites and found that (the wall superheat at ONB site) increases with the decrease of inlet temperature and increases as mass flux increases. The results also indicate that although the bubble size and behaviors in the narrow channel are different from that in the nonnarrow channel at the ONB site, the heat transfer has not been influenced evidently. In addition, in both channels can be predicted by the correlation proposed by Thom within the error range of ±30%.
      PubDate: Sun, 30 Jan 2022 12:20:05 +000
       
 
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