Validation of Analysis Models on Relocation Behavior of Molten Core Materials in Sodium-Cooled Fast Reactors Based on the Melt Discharge Experiment

2020 ◽  
Author(s):  
Kai Igarashi ◽  
Ryoji Onuki ◽  
Takaaki Sakai ◽  
Shinya Kato ◽  
Ken-ichi Matsuba ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
Heat Removal ◽  
Early Discharge ◽  
Core Material ◽  
Bottom Plate ◽  
Guide Tube ◽  
Control Rod ◽  
Analysis Models

Abstract In order to improve the safety of nuclear power plants, it is necessary to make sure measures against their severe accidents. Especially, in the case of a sodium-cooled fast reactor (SFR), there is a possibility of significant energy release due to formation of a large-scale molten fuel pool accompanied by re-criticality in the event of a core disruptive accident (CDA). It is important to ensure in-vessel retention that keeps and confines damaged core material in the reactor vessel even if the CDA occurs. CDA scenario initiated by Unprotected Loss Of Flow (ULOF), which is a typical cause of core damage, is generally categorized into four phases according to the progression of core-disruptive status, which are the initiating, early-discharge, material-relocation and heat-removal phases for the latest design in Japan. During the material-relocation phase, the molten core material flows down mainly through the control rod guide tube and is discharged into the inlet coolant plenum below the bottom of the core. The discharged molten core material collides with the bottom plate of the inlet plenum. Clarification of the accumulation behavior of molten core material with such a collision on the bottom plate is important to reduce uncertainties in the safety assessment of CDA. In present study, in order to make clear behavior of core melt materials during the CDAs of SFRs, analysis was conducted using the SIMMER-III code for a melt discharge simulation experiment [1] in which low-melting-point alloy was discharged into a shallow water pool. This report shows the validation results for the melt behavior by comparing with the experimental data.

scholarly journals Control Rod Drop Hydrodynamic Analysis Based on Numerical Simulation

2020 ◽  
Vol 8 ◽  
Author(s):  
Jie Cheng ◽  
Zhenying Wang ◽  
Chuan Lu ◽  
Jibin Zhang ◽  
Shumao Bi ◽  
...  
Keyword(s):  
Numerical Simulation ◽  
Nuclear Power ◽  
Power Plants ◽  
Degrees Of Freedom ◽  
Nuclear Reactor ◽  
Three Dimensional ◽  
Guide Tube ◽  
Control Rod ◽  
Drop Time ◽  
Maximum Impact Force

The drop time of the control rod plays an important role in judging whether a nuclear reactor can be safely shut down in an emergency condition and has become one of the most important parameters for the safety analysis of nuclear power plants. Exact assessment of the drop time is greatly dependent on the forces acting on the control rod. In this research, a three-dimensional numerical simulation of the control rod in a low-temperature heating reactor was established based on 6DOF (6 degrees of freedom) model using dynamic meshing technology, and it was used to analyze the control rod dropping experiment. The behavior of dropping the control rod was obtained, including the velocity, the displacement, and the pressure distribution on the control rod guide tube. The comparison between the simulation and the experiment results indicated that the simulation was capable of simulating the dropping characteristic of the control rod. Some important parameters can be calculated, such as the time of control rod dropping process and the maximum impact force. Based on this, useful information could be provided for the design of control rod driveline structure.

scholarly journals On the Root Causes of the Fukushima Daiichi Disaster from the Perspective of High Complexity and Tight Coupling in Large-Scale Systems

Symmetry ◽  
2021 ◽  
Vol 13 (3) ◽  
pp. 414
Author(s):  
Atsuo Murata ◽  
Waldemar Karwowski
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
High Complexity ◽  
Tight Coupling ◽  
Large Scale Systems ◽  
Safety Issues ◽  
Loss Of Coolant ◽  
Station Blackout ◽  
Fukushima Daiichi

This study explores the root causes of the Fukushima Daiichi disaster and discusses how the complexity and tight coupling in large-scale systems should be reduced under emergencies such as station blackout (SBO) to prevent future disasters. First, on the basis of a summary of the published literature on the Fukushima Daiichi disaster, we found that the direct causes (i.e., malfunctions and problems) included overlooking the loss of coolant and the nuclear reactor’s failure to cool down. Second, we verified that two characteristics proposed in “normal accident” theory—high complexity and tight coupling—underlay each of the direct causes. These two characteristics were found to have made emergency management more challenging. We discuss how such disasters in large-scale systems with high complexity and tight coupling could be prevented through an organizational and managerial approach that can remove asymmetry of authority and information and foster a climate of openly discussing critical safety issues in nuclear power plants.

Development of 1500-r/min 75-Inch Ultra-Long Last Stage Blades for Nuclear Steam Turbine

2017 ◽  
Author(s):  
Deqi Yu ◽  
Jiandao Yang ◽  
Wei Lu ◽  
Daiwei Zhou ◽  
Kai Cheng ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
Vibration Monitoring ◽  
Steam Turbines ◽  
Element Analysis ◽  
Rotating Blade ◽  
Lifetime Analysis ◽  
Computational Fluid Dynamics Cfd ◽  
Last Stage

The 1500-r/min 1905mm (75inch) ultra-long last three stage blades for half-speed large-scale nuclear steam turbines of 3rd generation nuclear power plants have been developed with the application of new design features and Computer-Aided-Engineering (CAE) technologies. The last stage rotating blade was designed with an integral shroud, snubber and fir-tree root. During operation, the adjacent blades are continuously coupled by the centrifugal force. It is designed that the adjacent shrouds and snubbers of each blade can provide additional structural damping to minimize the dynamic stress of the blade. In order to meet the blade development requirements, the quasi-3D aerodynamic method was used to obtain the preliminary flow path design for the last three stages in LP (Low-pressure) casing and the airfoil of last stage rotating blade was optimized as well to minimize its centrifugal stress. The latest CAE technologies and approaches of Computational Fluid Dynamics (CFD), Finite Element Analysis (FEA) and Fatigue Lifetime Analysis (FLA) were applied to analyze and optimize the aerodynamic performance and reliability behavior of the blade structure. The blade was well tuned to avoid any possible excitation and resonant vibration. The blades and test rotor have been manufactured and the rotating vibration test with the vibration monitoring had been carried out in the verification tests.

scholarly journals Seismic safety evaluation during site selection for the nuclear power plants in Bangladesh

2018 ◽  
Vol 4 (4) ◽  
pp. 251-256 ◽  
Author(s):  
Sergey Shcheklein ◽  
Ismail Hossain ◽  
Mohammad Akbar ◽  
Vladimir Velkin
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Nuclear Fuel ◽  
Heat Removal ◽  
Good Time ◽  
Diagnostics System ◽  
Seismic Impacts

Bangladesh lies in a tectonically active zone. Earlier geological studies show that Bangladesh and its adjoining areas are exposed to a threat of severe earthquakes. Earthquakes may have disastrous consequences for a densely populated country. This dictates the need for a detailed analysis of the situation prior to the construction of nuclear power plant as required by the IAEA standards. This study reveals the correlation between seismic acceleration and potential damage. Procedures are presented for investigating the seismic hazard within the future NPP construction area. It has been shown that the obtained values of the earthquake’s peak ground acceleration are at the level below the design basis earthquake (DBE) level and will not lead to nuclear power plant malfunctions. For the most severe among the recorded and closely located earthquake centers (Madhupur) the intensity of seismic impacts on the nuclear power plant site does not exceed eight points on the MSK-64 scale. The existing predictions as to the possibility of a super-earthquake with magnitude in excess of nine points on the Richter scale to take place on the territory of the country indicate the necessity to develop an additional efficient seismic diagnostics system and to switch nuclear power plants in good time to passive heat removal mode as stipulated by the WWER 3+ design. A conclusion is made that accounting for the predicted seismic impacts in excess of the historically recorded levels should be achieved by the establishment of an additional efficient seismic diagnostics system and by timely switching the nuclear power plants to passive heat removal mode with reliable isolation of the reactor core and spent nuclear fuel pools.

scholarly journals Simulation of Steady States of Helium Loop at Long Time Scenarios

2020 ◽  
Vol 328 ◽  
pp. 01009
Author(s):  
František Világi ◽  
Branislav Knížat ◽  
Róbert Olšiak ◽  
Marek Mlkvik ◽  
Peter Mlynár ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Natural Circulation ◽  
Heat Removal ◽  
Iteration Process ◽  
Heat Accumulation ◽  
Proposed Model ◽  
Long Time ◽  
Convection Cooling

The natural circulation helium loop is an experimental facility designed for the research of the possibility of utilizing natural convection cooling for the case of decay heat removal from a fast nuclear reactor. This concept would bring an improved automated safety system for future nuclear power plants operating a gas-cooled reactor. The article presents a new possibility of direct use of energy conservation laws in a 1D simulation of natural circulation loops. The calculation is performed by a triple iteration process, nested into each other. The results of the calculations showed good agreement with the measurements at steady state. A calculation with the proposed model at unsteady state is not yet possible, especially because of the exclusion of heat accumulation into the material.

scholarly journals Modelling of Passive Heat Removal Systems: A Review with Reference to the Framatome BWR Reactor KERENA: Part II

Energies ◽  
2019 ◽  
Vol 13 (1) ◽  
pp. 109 ◽  
Author(s):  
René Manthey ◽  
Frances Viereckl ◽  
Amirhosein Moonesi Shabestary ◽  
Yu Zhang ◽  
Wei Ding ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Gravitational Force ◽  
Heat Removal ◽  
Dimensional System ◽  
Thermal Gradients ◽  
Two Phase ◽  
Passive Safety Systems ◽  
External Power ◽  
Safety Systems

Passive safety systems are an important feature of currently designed and constructed nuclear power plants. They operate independent of external power supply and manual interventions and are solely driven by thermal gradients and gravitational force. This brings up new needs for performance and reliably assessment. This paper provides a review on fundamental approaches to model and analyze the performance of passive heat removal systems exemplified for the passive heat removal chain of the KERENA boiling water reactor concept developed by Framatome. We discuss modeling concepts for one-dimensional system codes such as ATHLET, RELAP and TRACE and furthermore for computational fluid dynamics codes. Part I dealt with numerical and experimental methods for modeling of condensation inside the emergency condenser and on the containment cooling condenser. This second part deals with boiling and two-phase flow instabilities.

Planning of Large-Scale In-Situ Gas Generation Experiment in Korean Radioactive Waste Repository

2010 ◽  
Author(s):  
Juyoul Kim ◽  
Sukhoon Kim ◽  
Jin Beak Park ◽  
Sunjoung Lee
Keyword(s):  
Radioactive Waste ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Large Scale ◽  
Radioactive Waste Disposal ◽  
Gas Generation ◽  
Radioactive Waste Repository ◽  
Waste Repository

In the Korean LILW (Low- and Intermediate-Level radioactive Waste) repository at Gyeongju city, the degradation of organic wastes and the corrosion of metallic wastes and steel containers would be important processes that affect repository geochemistry, speciation and transport of radionuclides during the lifetime of a radioactive waste disposal facility. Gas is generated in association with these processes and has the potential threat to pressurize the repository, which can promote the transport of groundwater and gas, and consequently radionuclide transport. Microbial activity plays an important role in organic degradation, corrosion and gas generation through the mediation of reduction-oxidation reactions. The Korean research project on gas generation is being performed by Korea Radioactive Waste Management Corporation (hereafter referred to as “KRMC”). A full-scale in-situ experiment will form a central part of the project, where gas generation in real radioactive low-level maintenance waste from nuclear power plants will be done as an in-depth study during ten years at least. In order to examine gas generation issues from an LILW repository which is being constructed and will be completed by the end of December, 2012, two large-scale facilities for the gas generation experiment will be established, each equipped with a concrete container carrying on 16 drums of 200 L and 9 drums of 320 L of LILW from Korean nuclear power plants. Each container will be enclosed within a gas-tight and acid-proof steel tank. The experiment facility will be fully filled with ground water that provides representative geochemical conditions and microbial inoculation in the near field of repository. In the experiment, the design includes long-term monitoring and analyses for the rate and composition of gas generated, and aqueous geochemistry and microbe populations present at various locations through on-line analyzers and manual periodical sampling. A main schedule for establishing the experiment facility is as follows: Completion of the detailed design until the second quarter of the year 2010; Completion of the manufacture and on-site installation until the second quarter of the year 2011; Start of the operation and monitoring from the third quarter of the year 2011.

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