Structure Topology Optimization Design and Shock Resistance Study on Nuclear Power Safety DCS Cast Aluminum Cabinet

2020 ◽  
Author(s):  
Wang Dongwei ◽  
Liu Mingxing ◽  
Wu Xiao ◽  
Yan Hao ◽  
Wu Zhiqiang
Keyword(s):  
Topology Optimization ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Optimization Design ◽  
High Stress ◽  
Shock Load ◽  
Normal Operation ◽  
Cast Aluminum ◽  
Structure Topology

Abstract Offshore floating nuclear power plant (FNPP) is characterized by its small and mobility, which is not only able to provide safe and efficient electric energy to remote islands, but to the oil and gas platforms. The safety digital control system (DCS) cabinet, as a carrier for the electronic devices, plays a significant role in ensuring the normal operation of the nuclear power plant. To satisfy the requirements of cabinet used in the sea environment, such as well rigidity, shock load resistance, good seal and corrosion resistance, etc, more and more attention is focused on the cast aluminum cabinet. However, the cast aluminum structure may cause larger weight of cabinet, which inevitability affects the mobility of cabinet, and increases the carried load of ship as well. Therefore, seeking for an effective approach to design a light weight cast aluminum cabinet for the offshore FNPP is definitely necessary. In this work, a frame of cast aluminum cabinet with lightweight is obtained successfully via structure topology optimization design, it is found that the weight of the frame can be reduced to 50% after optimization iterations. Subsequently, the natural frequency of the optimized cast aluminum cabinet is calculated by using ABAQUS, it is seen that the first mode frequency of the frame is beyond 30 Hz, which can meet the basic stiffness requirement. Accordingly, dynamic design analysis method (DDAM) is performed to verify the ability of the optimized cast aluminum cabinet in resisting sudden shock load, and the shock response characteristics of the cabinet are determined. Numerical results support that the optimized frame of cabinet possesses good resistance to high level shock. However, for the assembled cast aluminum cabinet, the vertical shock circumstance turns out to be the most critical condition, high stress and deformation regions occurs at the bracket and column. Reinforcements are proposed to make the bracket stiffer in this shock loading condition.

scholarly journals Efficiency of Using the Auxiliary Boiler for Peak Power Generation by a Steam Turbine of NPP

2021 ◽  
Vol 64 (1) ◽  
pp. 78-90
Author(s):  
V. A. Khrustalev ◽  
M. V. Garievskii
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Turbine ◽  
Nuclear Power ◽  
Power Plants ◽  
Normal Operation ◽  
Heat Output ◽  
Generation Process ◽  
System Effect ◽  
Industrial Enterprises

The article presents the technique of an estimation of efficiency of use of potential heat output of an auxiliary boiler (AB) to improve electric capacity and manoeuvrability of a steam turbine unit of a power unit of a nuclear power plant (NPP) equipped with a water-cooled water-moderated power reactor (WWER). An analysis of the technical characteristics of the AB of Balakovo NPP (of Saratov oblast) was carried out and hydrocarbon deposits near the NPP were determined. It is shown that in WWER nuclear power plants in Russia, auxiliary boilers are mainly used only until the normal operation after start-up whereas auxiliary boiler equipment is maintained in cold standby mode and does not participate in the generation process at power plants. The results of research aimed to improve the systems of regulation and power management of power units; general principles of increasing the efficiency of production, transmission and distribution of electric energy, as well as the issues of attracting the potential of energy technology sources of industrial enterprises to provide load schedules have been analyzed. The possibility of using the power complex NPP and the AB as a single object of regulation is substantiated. The authors’ priority scheme-parametric developments on the possibility of using the thermal power of the auxiliary boilers to increase the power of the steam turbine of a nuclear power plant unit equipped with WWER reactors unit during peak periods, as well as the enthalpy balance method for calculating heat flows, were applied. The surface area of the additional heater of the regeneration “deaerator – high pressure heaters” system and its cost were calculated. On the basis of calculations, it was shown that the additional power that can be obtained in the steam turbine of the NPP with a capacity of 1200 MW due to the use of heat of the modernized auxiliary boiler in the additional heat exchanger is 40.5 MW. The additional costs for the implementation of the heat recovery scheme of the auxiliary boiler at different prices for gas fuel and the resulting system effect were estimated in an enlarged way. Calculations have shown the acceptability of the payback period of the proposed modernization.

Water Hammer Characteristics for Parallel Pumps Water Supply Systems

2008 ◽  
Author(s):  
Wenxi Tian ◽  
Guanghui Su ◽  
Suizheng Qiu ◽  
Gaopeng Wang ◽  
Qing Lu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Water Supply ◽  
High Speed ◽  
Nuclear Power ◽  
Optimization Design ◽  
Check Valve ◽  
Water Hammer ◽  
Characteristic Line ◽  
Water Supply Systems

The water hammer induced by abrupt velocity change of fluid flow is inevitable for nuclear power plant systems because of the sudden opening or closing of valves, the sudden startup or shutdown of the pumps and the rupture of pipes. The water hammer pressure wave can damage the pipes and cause the abnormal shutdown of Nuclear Power Plant (NPP). The object of this study is a Parallel Pumps Water Supply system (PPWS) adopted in a NPP. The PPWS is composed of two parallel mixed-flow pumps connected with a check valve separately, a container, a throttle flap and pipe lines. The Method of Characteristic line (MOC) was adopted to evaluate the water hammer behaviors of the PPWS during the alternate startup and shutoff conditions of two parallel pumps. A code was developed using Fortran language to compute the transient behaviors including he peak pressure, the flow velocity and the movement of the valve plate. The results indicate that the water hammer behaviors under low speed startup condition differ from that of high speed startup condition. The maximum pressure vibration amplitude is up to 5.0MPa occurring under high-high speed startup condition. The computation results are instructive for the optimization design of the PPWS so as to minimize the damage potential induced by water hammer.

Numerical Simulation of Hydrogen Dispersion Inside a Compartment Using HYDRAGON Code

2016 ◽  
Author(s):  
M. Saeed ◽  
Yu Jiyang ◽  
B. X. Hou ◽  
Aniseh A. A. Abdalla ◽  
Zhang Chunhui
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Concentration Distribution ◽  
Nuclear Power ◽  
Turbulence Models ◽  
Hydrogen Combustion ◽  
Severe Accident ◽  
Normal Operation ◽  
Accurate Estimation ◽  
Nuclear Power Plant Accident

During severe accident in the nuclear power plant, a considerable amount of hydrogen can be generated by an active reaction of the fuel-cladding with steam within the pressure vessel which may be released into the containment of nuclear power plant. Hydrogen combustion may occur where there is sufficient oxygen, and the hydrogen release rates exceed 10% of the containment. During hydrogen combustion, detonation force and short term pressure may be produced. The production of these gas species can be detrimental to the structural integrity of the safety systems of the reactor and the containment. In 1979, the Three Mile Island (1979) accident occurred. This accident compelled experts and researchers to focus on the study of distribution of hydrogen inside the containment of nuclear power plant. However after the Fukushima Dai-ichi nuclear power plant accident (2011), the modeling of the gas behavior became important topic for scientists. For the stable and normal operation of the containment, it is essential to understand the behavior of hydrogen inside the containment of nuclear power plant in order to mitigate the occurrence of these types of accidents in the future. For this purpose, it is important to identify how burnable hydrogen clouds are produced in the containment of nuclear power plant. The combustion of hydrogen may occur in different modes based on geometrical complexity and gas composition. Reliable turbulence models must be used in order to obtain an accurate estimation of the concentration distribution as a function of time and other physical phenomena of the gas mixture. In this study, a small scale hydrogen-dispersion case is selected as a benchmark to address turbulence models. The computations are performed using HYDRAGON code developed by Department of Engineering Physics, Tsinghua University, China. HYDRAGON code is a three dimensional thermal-hydraulics analysis code. The purpose of this code is to predict the behavior of hydrogen gas and multiple gas species inside the containment of nuclear power plant during severe accident. This code mainly adopts CFD models and structural correlations used for wall flow resistance instead of using boundary layer at a wall. HYDROGAN code analyzes many processes such as hydrogen diffusion condensation, combustion, gas stratification, evaporation, mixing process. The main purpose of this research is to study the influence of turbulence models to the concentration distribution and to demonstrate the code thermal-hydraulic simulation capability during nuclear power plant accident. The calculated results of various turbulence models have different prediction values in different compartments. The results of k–ε turbulence model are in reasonable agreement as compared to the benchmark experimental data.

Ductile Tearing Simulations to Support Design of Large Scale Tests on Ferritic Pipes to Be Performed in the European Project ATLAS+

2019 ◽  
Author(s):  
Tomas Nicak
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Fracture Behavior ◽  
Large Scale ◽  
High Stress ◽  
Stress Triaxiality ◽  
Local Approach ◽  
Integrity Assessment ◽  
European Project

Abstract The safety and reliability of all systems has to be maintained throughout the lifetime of a nuclear power plant (NPP). This requires a systematic ageing management procedure for justifying their safe long term operation. One fundamental part in this process is to demonstrate the integrity of the nuclear power plant components. The European project ATLAS+ aims to develop and validate advanced methods of structural integrity assessment applicable in the ageing and lifetime management of primary and secondary circuit components. To support development and validation of those methods, a large scale test program was developed with the aim to investigate fracture behavior of relevant piping material at the component level. Three of planned large scale experiments focus on the fracture behavior of ferritic piping made of material WB 36 (15 NiCuMoNb 5), that is representative of secondary feedwater lines installed in German NPPs. In order to verify design calculations conducted by means of the classical fracture mechanics approach based on J-Integral [1], detailed local approach analyses are performed for three mock-ups with different initial defects. The local approach analyses presented in this paper are based on the local micromechanical model proposed by Gurson and further modified by Tvergaard. Calibration of required material parameters and prediction of the mock-up behavior during the test is discussed. In order to support constraint investigations stress triaxiality ahead of the crack front during crack propagation in the mock-ups is evaluated and compared to the stress triaxiality in CT20 and SENT specimens. As high stress triaxiality generally limits plastic deformation and increases the crack tip constraint, it is a good parameter to look at if constraint effects are considered.

Development and Verification of Severe Accident Simulation Capability for a NPP Simulator

2013 ◽  
Author(s):  
Zhifei Yang ◽  
Xiaofei Xie ◽  
Xing Chen ◽  
Shishun Zhang ◽  
Yehong Liao ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Simulation System ◽  
Severe Accident ◽  
Normal Operation ◽  
Feed Water ◽  
Accident Simulation ◽  
Accident Conditions ◽  
Relap5 Code

It is reflected in the severe accident in Fukushima Daiichi that the emergency capacity of nuclear power plant needs to be enhanced. A nuclear plant simulator that can model the severe accident is the most effective means to promote this capacity. Until now, there is not a simulator which can model the severe accident in China. In order to enhance the emergency capacity in China, we focus on developing a full scope simulator that can model the severe accident and verify it in this study. The development of severe accident simulation system mainly includes three steps. Firstly, the integral severe accident code MELCOR is transplanted to the simulation platform. Secondly, the interface program must be developed to switch calculating code from RELAP5 code to MELCOR code automatically when meeting the severe accident conditions because the RELAP5 code can only simulate the nuclear power plant normal operation state and design basis accident but the severe accident. So RELAP5 code will be stopped when severe accident conditions happen and the current nuclear power plant state parameters of it should be transported to MELCOR code, and MELCOR code will run. Finally, the CPR1000 nuclear power plant MELCOR model is developed to analyze the nuclear power plant behavior in severe accident. After the three steps, the severe accident simulation system is tested by a scenario that is initiated by the station black out with reactor cooling pump seal leakage, HHSI, LHSI and auxiliary feed water system do not work. The simulation result is verified by qualitative analysis and comparison with the results in severe accident analysis report of the same NPP. More severe accident scenarios initiated by LBLOCA, MBLOCA, SBLOCA, SBO, ATWS, SGTR, MSLB will be tested in the future. The results show that the severe accident simulation system can model the severe accident correctly; it meets the demand of emergency capacity promotion.

Test Facility Design for Integrated Digital Nuclear Reactor Protection System

2010 ◽  
Author(s):  
Huasheng Xiong ◽  
Duo Li ◽  
Liangju Zhang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Verification And Validation ◽  
Protection System ◽  
Normal Operation ◽  
Test Facility ◽  
Acceptance Test ◽  
Independent Verification ◽  
Reactor Protection System

Reactor protection system is one of the most important safety systems in nuclear power plant and shall be designed with very high reliability. Digital computer-based Reactor Protection System (RPS) takes great advantages over its conventional counterpart based on analog technique and faces the issues how to effectively demonstrate and confirm the completeness and correctness of the software that performs reactor safety functions in the same time. It is commonly accepted that the essential way to solve safety software issues in a digital RPS is to pass a strict and independent Verification and Validation (V&V) process, in which integrated RPS testing play an important role to form a part of the overall system validation. Integrated RPS testing must be carried out rigorously before the system is delivered to nuclear power plant. The integrated testing are often combined with the factory acceptance test (FAT) to form a single testing activity, during which the RPS is excited by emulated static and dynamic input signals. The integration testing should simulate normal operation, anticipated operational occurrences and accident conditions, as well as anticipated faults on the inputs to the DRPS such as sensors out of range or ambiguous input readings. All safety function requirements of digital RPS should be confirmed by representative testing. The design and development of a test facility to carry out the integrated RPS testing are covered in this paper, which is merged in the research on a digital RPS engineering prototype for a nuclear power plant. The test facility is based on PXI platform and LabVIEW software development environment and its architecture design also takes into account the test functions future extensions such as hardware upgrades and software modules enhancement. The test facility provides the digital RPS with redundant, synchronized and multi-channel emulated signals that are produced to emulate all protection signals from 1E class sensors and transmitters with time varied value within their possible ranges, which would put integrated RPS testing into practice to confirm the digital RPS has fully met its predefined safety functionality requirements. The designed test facility can provide an independent verification and validation process for the research of digital RPS with scientific methods and authentic data to evaluate the RPS performance thoroughly and effectively, such as measuring threshold precision and trip response time, analyzing system statistical reliability and so on.

Study on Water Hammer in the Overpressure Discharge Pipe of Nuclear Power Plant

2017 ◽  
Author(s):  
Qingmu Xu ◽  
Kun Cai ◽  
Jie Qin ◽  
Junkai Yuan ◽  
Juan Li
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Water Hammer ◽  
Liquid Column ◽  
Normal Operation ◽  
Pipeline System ◽  
Spent Fuel ◽  
Vapor Cavity ◽  
Spent Fuel Pool

Water hammer phenomenon is a significant pressure wave in pipe system caused by momentum change when the moving fluid is forced to stop or change direction instantaneously. Common causes of water hammer are sudden valve closing at the end of a pipeline system, pump failure, check valve slam etc. The steam transportation pipeline system may also be vulnerable to water hammer when it confronts with the situation where liquid and steam co-exist. Water hammer often occurs when steam condenses into water in a horizontal section of steam piping. Then steam “picks up” water to form a high-velocity “slug” and create extra stress to pipe. When steam is trapped into sub-cooled water, the collapse of vapor cavity can lead to collision of two columns of liquid, resulting in a large rise in pressure which will damage pipes, supporting structures and hydraulic machinery. Nuclear power plant is composed of complex equipments and piping systems, lots of which contain both liquid and steam. Hence, there is a potential threat of occurrence of water hammer to the normal operation of systems. Thus, this phenomenon needs to be well investigated and prevented with some effective methods. For the purpose of overpressure relief under severe accidents, the spent fuel pool cooling system of CAP1000 series nuclear power plant provides a discharge passage from containment to spent fuel pool. When the containment pressure exceeds the control value, valve is opened to discharge high-temperature and high-pressure steam until the pressure drops to a safety value. During this process, serious water hammer happens, causing pressure rise beyond the design pressure and further leading to damages to pipes and structures. Therefore, water hammer of overpressure discharge pipeline in CAP1000 plant is studied in this work. On the basis of verification of the capabilities of computational code RELAP5/MOD3.3, hydraulic transient of water hammer is simulated under different conditions. It is indicated that after steam discharge stops, residual steam in pipe condenses because of contact with sub-cooled water in spent fuel pool. Subsequently, the rapid backflow and vapor cavity lead to a severe water hammer. The detailed analysis has shown that water temperature of spent fuel pool has a decisive influence on the mechanism of water hammer phenomenon, including collision of liquid column to valve disc and cavity collapse in the horizontal pipe. The collision and separation of liquid column result in relatively lower pressure amplitude.

scholarly journals Simulation and analysis of a WWER-1000 reactor under normal and transient conditions

2016 ◽  
Vol 31 (3) ◽  
pp. 207-217
Author(s):  
Ghonche Baghban ◽  
Mohsen Shayesteh ◽  
Majid Bahonar ◽  
Reza Sayareh
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Power Distribution ◽  
Nuclear Power ◽  
Full Range ◽  
Kinetic Equations ◽  
Normal Operation ◽  
Operation Condition ◽  
Transient Conditions ◽  
Drift Flux Model

An accurate analysis of the flow transient is very important in safety evaluation of a nuclear power plant. In this study, analysis of a WWER-1000 reactor is investigated. In order to perform this analysis, a model is developed to simulate the coupled kinetics and thermal-hydraulics of the reactor with a simple and accurate numerical algorithm. For thermal-hydraulic calculations, the four-equation drift-flux model is applied. Based on a multi-channel approach, core is divided into some regions. Each region has different characteristics as represented in a single fuel pin with its associated coolant channel. To obtain the core power distribution, point kinetic equations with different feedback effects are utilized. The appropriate initial and boundary conditions are considered and two situations of decreasing the coolant flow rate in a protected and unprotected core are analyzed. In addition to analysis of normal operation condition, a full range of thermal-hydraulic parameters is obtained for transients too. Finally, the data obtained from the model are compared with the calculations conducted using RELAP5/MOD3 code and Bushehr nuclear power plant data. It is shown that the model can provide accurate predictions for both steady-state and transient conditions.

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