scholarly journals Development and verification of methods for predicting the frequency of self-oscillations in swirling coolant flows of nuclear power plants

2021 ◽  
Vol 2088 (1) ◽  
pp. 012037
Author(s):  
KN Proskuryakov ◽  
AV Anikeev
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Power Level ◽  
Primary Circuit ◽  
Frequency Bandwidth ◽  
Mechanical Vibrations ◽  
Oscillation Frequencies ◽  
Necessary And Sufficient ◽  
Nuclear Power Plant Unit

Abstract Methods and algorithms for calculating the frequency of self-oscillations in swirling coolant flows of nuclear power plants with VVER reactors have been developed. The frequency of self-oscillations occurring in the acoustic sections of the primary circuit and their connections in the starting modes of the power unit and when operating at the rated power level is predicted. It is established that the self-oscillation frequencies can fall into the frequency bandwidth of mechanical vibrations and vibrations of the reactor plant’s internal devices. It is shown that in order to prevent the resonance of self-oscillations of the coolant with the vibration frequencies of internal devices, it is necessary and sufficient to take the frequency of self-oscillations outside the bandwidth of the vibration frequency of structures. The results of verification of the results of forecasting the frequency of self-oscillations in swirling coolant flows at a nuclear power plant unit with VVER are presented. The application of the developed technique shows that the pendulum oscillations of the VVER - 1200 body are caused by an increase in the parameters of the coolant and the geometric dimensions of the VVER - 1200 reactor compared to VVER-1000.

An Efficient, Flexible Arrangement to Generate High Quality Process or Domestic Heat in Recuperative Gas Cycle Power Plants: Application to Helium Direct Cycle Nuclear Power Plants

1979 ◽  
Vol 101 (1) ◽  
pp. 130-140 ◽  
Author(s):  
Z. P. Tilliette ◽  
B. Pierre ◽  
P. F. Jude
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Hot Water ◽  
Primary Circuit ◽  
Inlet Temperature ◽  
Total Efficiency ◽  
High Quality ◽  
Quality Process ◽  
Direct Cycle

The advantages of gas turbine power plants in general and closed cycle systems under gas pressure in particular for waste heat recovery are well known. A satisfactory efficiency for electric power generation and good conditions to obtain a significant amount of hot water above 100°C lead to a high fuel utilization. However, as in most of projects, it is not much possible to produce high temperature steam or water without significantly decreasing the electricity production. A new method for an additional generation of high quality process or domestic heat is proposed. The basic feature of this method lies in arranging one or two steam generators or preheaters in parallel with the low pressure side of the recuperator. The high total efficiency and the noteworthy flexibility of this system are emphasized. This arrangement is suitable for any kind of heat source, but the applications presented in this paper are related to helium direct cycle nuclear power plants the main features of which are a single 600 MW(e) turbomachine, a turbine inlet temperature of 775°C, no or one intermediate cooling and a primary circuit fully integrated in a pre-stressed concrete reactor vessel.

scholarly journals Optimization of the technological solutions for recharge and boron control system at the NPP for the project AES-2006

2018 ◽  
Vol 245 ◽  
pp. 07017 ◽  
Author(s):  
Anastasia Ulasen ◽  
Aleksandr Kalyutik ◽  
Anatolii Blagoveshchenskii
Keyword(s):  
Control System ◽  
Heat Exchangers ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Primary Circuit ◽  
Capital Costs ◽  
Main Pipelines ◽  
Heat Exchange Equipment ◽  
Under Construction

The article considers the possible ways to optimize the technological solutions of the recharge and boron control system of nuclear power plants under construction within the AES-2006 project. The possibilities for optimization of technological solutions of the system of recharge and boron regulation of the AES-2006 project, which will not affect the reliability and efficiency of its main functions: purge-recharge of the primary circuit and boron regulation, were studied. As a result of the analysis of technological solutions and analytical calculations carried out during the work, it was found that in the system of recharge and boron regulation of the NPP within the project AES-2006 it is possible to perform optimization basing on reduction the metal content of the heat exchange equipment by reducing the surface area of the heat exchangers of the coolant outlet, reducing the power of pumps, as well as reducing the diameter of a number of main pipelines. Implementation of the proposed optimization of technological solutions will allow a more rational arrangement of the system and reduce capital costs for the construction of nuclear power plants as a whole, while not adversely affect the safety of the system and its functions.

A Study of Power Decrease Strategy on CPR1000 Full Scope Simulator

2017 ◽  
Author(s):  
Zhenhua Zhang ◽  
Bo Zhang ◽  
Qingsong Wang ◽  
Chengzhi Wang ◽  
Dan Xing
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Power Level ◽  
Technical Specification ◽  
Reducing Power ◽  
Specific Power ◽  
Technical Problems ◽  
Control Rod ◽  
Special Cases

Normally, nuclear power plants operate at 100% full power level. However, in some special cases, nuclear power plants need to decrease power and operate at lower power level. For example, because of the typhoon, test and equipment maintenance requirements etc., nuclear power plants decrease power to specific power level and stay for several hours at this power level. Because there is inserted time limit of the power compensation rod (G rod) in the technical specification, inserting control rod to decrease power cannot always be chosen, and the strategy that decreases power by boron injection should be taken. In this paper, powers down strategies are introduced. And the similarities and differences between decreasing power by inserting control rod and decreasing power by injecting boron are also introduced. Before reducing power, reactivity equilibrium calculation should be performed and the predicted amount of boron injection can be calculated. Then according to the power down rate, Boron injection rate can be estimated. Finally, Boron injection operation can be performed. During the power reducing, R rod and some important equipment associated with boron injection should be monitored closely, especially the drain tank water level of condensate extraction system (CEX) and the response of feedwater flow control system (ARE). A key control of power down is to control the axial power deviation (delta I) that is one of the most difficult technical problems. Delta I must be kept in the functional area and it is inhibited that the operating point goes outside limits (left limit and right limit) of region II in any situations. It is performed on CPR1000 full scope simulator (FSS) to decrease power by inserting control rod and boron injection in this paper. Several curves obtained during power reducing are analyzed in detail. According to analysis results, the key procedure, operation risks and operation control during power decreasing are proposed as reference. The purpose of them is to meet the requirements of nuclear safety during power down operation.

Non-Destructive Techniques in the Tacis and Phare Nuclear Safety Programmes

2002 ◽  
Author(s):  
Michel Bie`th
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Slovak Republic ◽  
Nuclear Safety ◽  
Primary Circuit ◽  
Non Destructive ◽  
Service Inspection ◽  
Non Destructive Techniques

Decisions regarding the verification of design plant lifetime and potential license renewal periods involve a determination of the component and circuit condition. In Service Inspection of key reactor components becomes a crucial consideration for continued safe plant operation. The determination of the equipment properties by Non Destructive Techniques during periodic intervals is an important aspect of the assessment of fitness-for-service and safe operation of nuclear power plants The Tacis and Phare were established since 1991 by the European Union as support mechanisms through which projects could be identified and addressed satisfactorily. In Nuclear Safety, the countries mainly concerned are Russia, Ukraine, Armenia, and Kazakhstan for the Tacis programme, and Bulgaria, Czech Republic, Hungary, Slovak Republic, Lithuania, Romania and Slovenia for the Phare programme. The Tacis and Phare programs concerning the Nuclear Power Plants consist of: • On Site Assistance and Operational Safety, • Design Safety, • Regulatory Authorities, • Waste management, and are focused on reactor safety issues, contributing to the improvement in the safety of East European reactors and providing technology and safety culture transfer. The main parts of these programmes are related to the On-Site Assistance and to the Design Safety of VVVER and RBMK Nuclear power plants where Non Destructive Techniques for In Service Inspection of the primary circuit components are addressed.

Work structure in nuclear power plants

1983 ◽  
Author(s):  
Marjorie B. Bauman ◽  
Richard F. Pain ◽  
Harold P. Van Cott ◽  
Margery K. Davidson
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Work Structure
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