scholarly journals EN Analysis of modern approaches to improve the efficiency of blackout accident management at nuclear power plants

2019 ◽  
Vol 55 (4) ◽  
pp. 227-234
Author(s):  
A. Denysova ◽  
V. Skalozubov ◽  
V. Spinov ◽  
D. Spinov ◽  
D. Pirkovskiy ◽  
...  
Keyword(s):  
Steam Generator ◽  
Power Plants ◽  
Real Life ◽  
Reactor Core ◽  
Heat Removal ◽  
Integrated Approach ◽  
Necessary Condition ◽  
Multiple Failures ◽  
Feedwater Pump ◽  
Accident Management

The paper analyzes the approaches to improve the efficiency of blackout accident management taking into account the lessons of the great accident at Fukushima Daiichi NPP in 2011. It is found that the afterheat removal passive systems by natural circulation through steam generators cannot provide conditions for adequate safety functions to remove heat from the reactor and maintain the required feedwater level in the steam generator during blackout accidents and multiple failures of safety-related systems. The application of alternative approaches using auxiliary feedwater steam generator driven pumps requires additional experiment-calculated operability / reliability qualification for a blackout accident and multiple failures of NPP safety-related systems. However, implementation of alternative SDEFP system requires in-depth qualification for the conditions of blackout accidents. Safety systems of passive heat removal from the steam generator (adequately to active safety electrical systems) cannot ensure safety functions on control of required feedwater level in the steam generator and heat removal from the reactor core during blackout accidents (at least 72 hours) and multifailure accidents. The system of the steam generator driven emergency feedwater pump can be the alternative solution to ensure safety functions on heat removal through the steam generator during blackout accidents. Additional study of efficiency of steam driven pumps at the experimental facilities that meet real-life criteria of hydrodynamic similarity is a necessary condition for implementation of system of the steam driven emergency feedwater pump. Application of an integrated approach to manage blackout accidents is reasonable. At the initial stage of accident with relatively high steam pressure in the steam generator it is required supply of feedwater by the steam driven emergency pump

scholarly journals STUDI AWAL KARAKTERISTIK TEMPERATUR DAN TEKANAN PADA ALIRAN SIRKULASI ALAM DI FASILITAS UJI UNTAI FASSIP-02 MENGGUNAKAN SIMULASI CFD

POROS ◽  
2018 ◽  
Vol 16 (1) ◽  
Author(s):  
Ainur Roidi Rosidi
Keyword(s):  
Flow Rate ◽  
Power Plants ◽  
Heat Dissipation ◽  
Cfd Simulation ◽  
Reactor Core ◽  
Heat Removal ◽  
Test Facility ◽  
Working Fluid ◽  
Fluid Temperature ◽  
Heating Section

TFASSIP-02 loop is a test facility that is used for research and development of safety technologies for future nuclear power plants based on natural law. This test facility is designed to study natural circulation phenomena caused by differences in fluid density due to temperature differences in the one-phase heat dissipation system during the simulation of heat removal from the reactor core when an accident occurs. FASSIP-02 loop consists of piping components, water heating tanks, water cooling tanks and expansion tanks. The purpose of this study was to understand the conditions of temperature change and pressure of water working fluid based on temperature changes in the heater section which were simulated on the loop geometry FASSIP-02. The research method was carried out in a simulation of Computational Fluid Dynamics using FLUENT 6.3 software. The working fluid in the FASSIP-02 loop uses water with a temperature of 27°C, the flow rate is varied 0.3 m/s and 0.45 m / s, while the temperature in the heating section is 70°C. CFD simulation results show that the increase in the working fluid temperature of the water with a flow rate of 0.3 m/s after passing through the heating section is 39°C, while the temperature increase of the working fluid of the water with a flow rate of 0.45 m/s is 36.6°C. Pressure drops at flow rates of 0.3 m/s and 0.45 m/s each occur in water working fluid before entering through WHT and after passing through the heating section. 

scholarly journals Safety Analyses of VVER-1200/V491 reactor for longterm station blackout along with small LOCAs

2016 ◽  
Vol 6 (4) ◽  
pp. 8-17
Author(s):  
Thi Hoa Bui ◽  
Tan Hung Hoang ◽  
Minh Giang Hoang
Keyword(s):  
Steam Generator ◽  
Heat Removal ◽  
Safety Analysis ◽  
Severe Accident ◽  
Management Guideline ◽  
Accident Management ◽  
Station Blackout

Performance of  Passive Heat Removal through Steam Generator (PHRS-SG) of VVER-1200/V491 reactor presented in Safety Analysis Report for Ninh Thuan 1 shows that in case of long term station black out (SBO),  VVER-1200/V491 reactor can be cooldown and remained in safety mode at least 24 hours based on PHRS-SG performance. Anyway, long term station blackout along with small break in main coolant pipe of VVER-1200/V491 is assumed to be happening as an extension design condition that needs to be investigated. This study focuses on investigation of SBO along with different size of small break of LOCAs with expectation of finding the range of break size that the reactor is still kept in safety mode during 24 hours. During the investigation, some indicators for fuel damage such as the timing of HA1 actuation or mass of coolant inventory discharged are introduced as necessary information contributed to Severe Accident Management Guideline (SAMG).

Development of an Integrity Evaluation System for Steam Generator Tubes in a Nuclear Power Plant

2007 ◽  
Vol 120 ◽  
pp. 157-162
Author(s):  
J.C. Kim ◽  
Sang Min Lee ◽  
Yoon Suk Chang ◽  
Jae Boong Choi ◽  
Young Jin Kim ◽  
...  
Keyword(s):  
Steam Generator ◽  
Evaluation System ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Core ◽  
Limit Load ◽  
Evaluation Procedure ◽  
Efficient System ◽  
Reference Pressure ◽  
Steam Generator Tubes

Steam generators working in nuclear power plants convert water into steam from heat produced in the reactor core and each of them contains from 3,000 to 16,000 tubes. Since these tubes constitute one of primary barriers under radioactive and high pressure condition, the integrity should be maintained carefully during the operation. The objective of this research is to introduce an integrity evaluation system for steam generator tubes as a substitute of well-trained engineers or experts. For this purpose, a couplet examination has been carried out on the complicated evaluation procedure and an efficient system named as STiES was developed employing three representative integrity evaluation methods: fracture mechanics analysis (crack driving force diagram and J-integral/Tearing modulus method) and limit load method. Exemplary analyses for steam generator tubes with various types of flaws showed good applicability of the proposed integrity evaluation system. So, it is anticipated that the system can be used for the calculation of reference pressure to decide either the continued operation or repair until next outage.

Research on Emergency System of Injection Estimation of Reactor Core Recovery and Reactor Core Decay Heat Removal

2017 ◽  
Author(s):  
Liao Feiye ◽  
Jiang Pingting ◽  
Liu Wang ◽  
He Dongyu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Human Error ◽  
Reactor Core ◽  
Heat Removal ◽  
Lessons Learned ◽  
Fukushima Accident ◽  
Decay Heat ◽  
Core Recovery ◽  
Emergency System

One of the lessons learned from Fukushima accident is that the existing procedures used in Nuclear Power Plants (NPPs) are not executed effectively and quickly enough after such an extended accident, for the accident is complex and people are too nervous in such a situation. Thus, emergency system that helps to raise diagnosis efficiency is necessary. In the paper, a quick diagnosis system on injection estimation of reactor core recovery and decay heat removal injection estimation is developed to meet the urgent needs and strengthen requirements for the training and application among utilities and nuclear regulators. The system will assist regulators to quickly know whether the currently flow will probably recover the reactor core, or whether the current injection capacity is sufficient to quench and recover the reactor core, directly after input present parameters into the system. In the system, Matlab method is used, and intuitive insights are considered, which is propitious to give immediate graphical interface and reduce possibility of human error.

Creation of Automatic System for Monitoring, Forecasting and Control of Condition of Lead-Bismuth (Lead) Coolant and Surfaces of Circuits of Nuclear Power Plants

2009 ◽  
Author(s):  
P. N. Martynov ◽  
R. Sh. Askhadullin ◽  
A. A. Simakov ◽  
A. Yu. Chaban’ ◽  
M. E. Chernov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Reactor Core ◽  
Heat Removal ◽  
Severe Accident ◽  
Medium Size ◽  
Safety Level ◽  
Nitride Fuel ◽  
Active Safety Systems ◽  
Accident Conditions

Lead-bismuth coolant is preferable for the medium size reactors, since, in contrast to the sodium coolant, it does not interact with water and air, it is radiation resistant, insignificantly activated and it is not combustible [1]. Combination of natural properties of lead-based coolants, mono-nitride fuel, fast reactor neutronics and design approaches used for the reactor core and heat removal system brings SVBR 75/100 NPP [2] to achieve a new safety level and assures its stability without operation of active safety systems even under severe accident conditions. Analysis of possible sequences of the events even under conditions of such severe accidents as addition of total excess reactivity or all pumps trip accompanied by safety system failure leads to the conclusion on that power unit with SVBR 75/100 reactor plant (RP) has high safety level.

scholarly journals Analysis of the Processes in Spent Fuel Pools in Case of Loss of Heat Removal due to Water Leakage

2013 ◽  
Vol 2013 ◽  
pp. 1-11 ◽  
Author(s):  
Algirdas Kaliatka ◽  
Viktor Ognerubov ◽  
Virginijus Vileiniškis ◽  
Eugenijus Ušpuras
Keyword(s):  
Reactor Core ◽  
Heat Removal ◽  
Severe Accident ◽  
Accident Analysis ◽  
Spent Fuel ◽  
Decay Heat ◽  
Management Measures ◽  
Fuel Assemblies ◽  
Accident Management ◽  
Spent Fuel Pool

The safe storage of spent fuel assemblies in the spent fuel pools is very important. These facilities are not covered by leaktight containment; thus, the consequences of overheating and melting of fuel in the spent fuel pools can be very severe. On the other hand, due to low decay heat of fuel assemblies, the processes in pools are slow in comparison with processes in reactor core during LOCA accident. Thus, the accident management measures play a very important role in case of some accidents in spent fuel pools. This paper presents the analysis of possible consequences of fuel overheating due to leakage of water from spent fuel pool. Also, the accident mitigation measure, the late injection of water was evaluated. The analysis was performed for the Ignalina NPP Unit 2 spent fuel pool, using system thermal hydraulic code for severe accident analysis ATHLET-CD. The phenomena, taking place during such accident, are discussed. Also, benchmarking of results of the same accident calculation using ASTEC and RELAP/SCDAPSIM codes is presented here.

scholarly journals A decision-support approach to severe accident management in nuclear power plants

2020 ◽  
pp. 1-12
Author(s):  
Marko Bohanec ◽  
Ivan Vrbanić ◽  
Ivica Bašić ◽  
Klemen Debelak ◽  
Luka Štrubelj
Keyword(s):  
Decision Support ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Severe Accident ◽  
Accident Management

Flow and Temperature Fields Measurement Inside Rod Bundle by the Combined Use of PIV and LIF Technique

2018 ◽  
Author(s):  
Xing Li ◽  
Sichao Tan ◽  
Zhengpeng Mi ◽  
Peiyao Qi ◽  
Yunlong Huang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Great Influence ◽  
Reactor Core ◽  
Transfer Characteristic ◽  
Index Of Refraction ◽  
Temperature Fields ◽  
Spacer Grid ◽  
Rod Bundle ◽  
Combined Use

Thermal hydraulic research of reactor core is important in nuclear energy applications, the flow and heat transfer characteristics of coolant in reactor fuel assembly has a great influence on the performance and safety of nuclear power plants. Particle image velocimetry (PIV) and Laser induced fluorescence (LIF) are the instantaneous, non-intrusive, whole-field fluid mechanics measuring method. In this study, the simultaneous measurement of flow field and temperature field for a rod bundle was conducted using PIV and LIF technique. A facility system, utilizing the matching index of refraction approach, has been designed and constructed for the measurement of velocity and temperature in the rod bundle. In order for further study on complex heat and mass transfer characteristic of rod bundle, the single-phase experiments on the heating conditions are performed. One of unique characteristics of the velocity and temperature distribution downstream the spacer grid was obtained. The experimental results show that the combined use of PIV and LIF technique is applied to the measurement of multi-physical field in a rod bundle is feasible, the measuring characteristics of non-intrusive ensured accuracy of whole field data. The whole field experimental data obtained in rod bundle benefits the design of spacer grid geometry.

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