Simulation of the Acoustic Loads Generated in the Intersection of a Main Steam Line and its Safety Relief Valve Branch of a BWR Plant under Extended Power Uprate Condition

2015 ◽  
Vol 362 ◽  
pp. 190-199
Author(s):  
Arturo Ocampo-Ramirez ◽  
Luis Héctor Hernández-Gómez ◽  
Pablo Ruiz-López ◽  
Noel Moreno-Cuahquentzi ◽  
Guillermo Urriolagoitia-Calderón ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
The United States ◽  
Steam Flow ◽  
Normal Operation ◽  
Relief Valve ◽  
Steady State Condition ◽  
Main Steam ◽  
Extended Power

The structural integrity of a BWR nuclear power plant can be compromised due to severe dynamic loads. Acoustic loads coming from a Safety Relief Valve branch can be adversely amplified if the steam flow is increased at the Main Steam Piping. This phenomenon has been reported previously in one BWR nuclear power plant. Its steam dryer was fractured and loose parts were generated due to high-cycle fatigue. This event has driven the United States Nuclear Regulatory Commission to issue specific regulations to evaluate acoustic loads which would be detrimental to the BWR steam dryer. In this paper, the acoustic loads were simulated when the steam flow is incremented from normal operation conditions to an Extended Power Uprate condition. It was analyzed, when the output power was incremented 14% and 28%. The initial conditions were determined with Computational Fluid Dynamics under steady state condition. This data was used in subsequent transient analysis. The model of Large Eddy Simulation was used and the acoustic simulation was performed with the Fowcs Williams and Hawkings Method. The Power Spectral Density was obtained with Fast Fourier Transform. The frequency peaks were found between 148 Hz and 155 Hz. These results are consistent with those obtained with the Helmholtz model and other results reported in the open literature. The results show that the peak pressure can be increased up to six times in resonance conditions, corresponding to a power uprate of 28%.

Investigation of a Possible Emergency Procedure for the VVER 1000 NPP in Case of a Total Loss of Feedwater and a Main Steam Line Break

2004 ◽  
Author(s):  
Nikolaus Muellner ◽  
Walter Giannotti ◽  
Francesco D’Auria
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Total Loss ◽  
Normal Operation ◽  
Feed Water ◽  
Emergency Procedures ◽  
Design Basis ◽  
Management Procedures ◽  
Main Steam

Accident management procedures associated with nuclear power plant beyond design basis accidents should be developed with the aid of the more recent versions of advanced computational tools (best estimate codes) in order to verify the effectiveness of normal operation systems of the plant to avoid or minimize core damage. A. Madeira showed in her paper “A PWR Recovery Option for a Total Loss of Feedwater Beyond Design Basis Scenario” that it is possible to safely control a total loss of feedwater scenario in the Angra2 nuclear power plant, using two emergency procedures, namely the opening of the steam generator (SG) relief valves, and on the primary side the complete manual opening of all pressurizer relief and safety valves. This paper investigates the effectiveness of the procedure opening of the SG relief valves, followed by primary side feed and bleed for a generic VVER-1000 NPP in case of a total loss of feed water. The results indicate that the procedure is successful in reducing the primary side pressure and temperature to safe conditions, i.e. long term core cooling is achievable.

Study of Steam Hammer in the Main Steam System

2017 ◽  
Author(s):  
Yemin Dong ◽  
Cuiyun Wang ◽  
Jiaming Zhao ◽  
Pei Yu ◽  
Bin Zhao
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Normal Operation ◽  
Saturated Steam ◽  
Pressurized Water ◽  
Main Steam ◽  
Safety Operation ◽  
Steam System

The main steam system plays an important role to transfer the saturated steam generated from the steam generator to the main turbine and other steam consumed devices in a pressurized water reactor. During the normal operation, the main steam system transfers the high temperature and pressure steam generated from the steam generator in the nuclear power plant. Once there is an accident situation delivering the main steam isolation valve fast close signal or an unexpected main steam isolation valve close signal, the steam hammer phenomenon will be induced in the main steam system by the main steam isolation valve fast close incident. The steam hammer phenomenon might induce pressure rise rapidly in the main steam system and generate unintended transient load on the main steam system which might have effect on the safety operation of the nuclear power plant. Therefore the steam hammer phenomenon in the main steam system should be studied. The study creates the main steam system model which includes the main steam pipes, devices and connected system based on PIPENET software following the engineering data by a third generation nuclear power plant. The study takes the advantage of the transient mode in PIPENET to simulate the steam hammer phenomenon in the main steam system. The study simulates different boundary conditions and device parameters in order to analyze the different effects on the steam hammer phenomenon in the main steam system. The simulation model could calculate the pressure, load and other parameters in the main steam system during the main steam isolation valve fast close period. The effects of the steam hammer phenomenon could be analyzed through these characteristic parameters. The PIPENET model could simulate the main steam system action during the main system isolation valve fast close incident which helps the study to master the operation and function of the main steam system and verify the integrality of the main steam system in the steam hammer phenomenon. With the simulation and analysis of the steam hammer phenomenon in the main steam system simulated in the PIPENET, the pressure raise which induced by the steam hammer wouldn’t threat the integrality of the main steam system. And the main steam system could ensure the safety operation by steam discharge through the steam dump valves and main steam safety valves.

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.

scholarly journals Anti-Earthquake Design of Industrial Facilities

2006 ◽  
Vol 1 (2) ◽  
pp. 190-200
Author(s):  
Heki Shibata ◽  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
The United States ◽  
Potential Hazard ◽  
Oil Refineries ◽  
Probability Of Occurrence ◽  
Aseismic Design ◽  
Low Probability

In Japan two sets of guidelines pertaining to modern aseismic design are being prepared. One is the guideline for the aseismic design of petrochemical plants and oil refineries, and the other is the code of aseismic design of nuclear power plants. The International Atomic Energy Agency also established its own guideline very recently. Several other countries also provide their own codes or guidelines. Among these, the regulatory guides of the United States are well known and quoted often; however some of them seem to be too sophisticated, for example, the three dimensional input problem. The reason for this is that the requirement of safety for a nuclear power plant is so severe that all events which have even a very low probability of occurrence should be considered. Therefore, if the results of theoretical study indicate an event which may occur even in very low probability, then from the viewpoint of conservatism, the designer must consider that event in this design. Although for the design of a nuclear power plant this might be partly true, the author feels that the probability of occurrence of the event should be evaluated in relation to the potential hazard of the design object. As well as this, he believes that proper understanding of the event in relation to the actual record of failures during past destructive earthquakes should be taken into consideration.

scholarly journals Comparison and Evaluation of Domestic and Foreign Radiation Environmental Standards for Nuclear Power Plants

2021 ◽  
Vol 2083 (2) ◽  
pp. 022020
Author(s):  
Jiahuan Yu ◽  
Xiaofeng Zhang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
The United States ◽  
Discharge System ◽  
Radioactive Effluent ◽  
Effluent Discharge ◽  
The Impact

Abstract With the development of the nuclear energy industry and the increasing demand for environmental protection, the impact of nuclear power plant radiation on the environment has gradually entered the public view. This article combs the nuclear power plant radiation environmental management systems of several countries, takes the domestic and foreign management of radioactive effluent discharge from nuclear power plants as a starting point, analyses and compares the laws and standards related to radioactive effluents from nuclear power plants in France, the United States, China, and South Korea. In this paper, the management improvement of radioactive effluent discharge system of Chinese nuclear power plants has been discussed.

Study on the Liquid Seal Discharge Process in an Over-Pressurized Accident

2020 ◽  
Author(s):  
Dan Wu ◽  
Jian Deng ◽  
Sijia Du ◽  
Libo Qian
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressure Rise ◽  
Discharge Process ◽  
Relief Valve ◽  
Pressure Rise Rate ◽  
Liquid Seal ◽  
Rise Rate

Abstract In an over pressure accident, one or more pressurizer safety (or relief) valves will open due to the rapid pressure rise process. Once the safety (or relief) valves are open, the liquid seal will be discharged, and this will generate great discharge force to the downstream pipes. Multi-level protection is chosen using pressurizer safety (or relief) valves with different setpoint in most of Nuclear Power Plant, especially in the self-designed Generation-III Nuclear Power Plants. As the over pressure accident progresses, one or more safety (or relief) valves will be open. The downstream pipes will experience one or more times of impacts, which will influence the arrangement of the pipes. The whole discharge process is very complex, and the key influence factors are the pressure rise rate, safety (or relief) valve opening time, liquid seal temperature and volume, and the arrangement of the downstream discharge pipes. In present paper, liquid seal discharge process in an over pressure accident is studied. The pressure rise rate is so fast that three safety (or relief) valves will open one after another, which will generate three impacts on the downstream discharge pipes. It is found that for a specific design of Nuclear Power Plant, well design of the safety (or relief) valve setpoint is very important to the discharge force analysis results.

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.

Incorporation of Risk Insights in the Regulatory Treatment of Nuclear Power Plant Structures, Systems, and Components

2002 ◽  
Author(s):  
Eugene Imbro ◽  
Thomas G. Scarbrough
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Special Treatment ◽  
Regulatory Commission ◽  
High Level

The U.S. Nuclear Regulatory Commission (NRC) has established an initiative to risk-inform the requirements in Title 10 of the Code of Federal Regulations (10 CFR) for the regulatory treatment of structures, systems, and components (SSCs) used in commercial nuclear power plants. As discussed in several Commission papers (e.g., SECY-99-256 and SECY-00-0194), Option 2 of this initiative involves categorizing plant SSCs based on their safety significance, and specifying treatment that would provide an appropriate level of confidence in the capability of those SSCs to perform their design functions in accordance with their risk categorization. The NRC has initiated a rulemaking effort to allow licensees of nuclear power plants in the United States to implement the Option 2 approach in lieu of the “special treatment requirements” of the NRC regulations. In a proof-of-concept effort, the NRC recently granted exemptions from the special treatment requirements for safety-related SSCs categorized as having low risk significance by the licensee of the South Texas Project (STP) Units 1 and 2 nuclear power plant, based on a review of the licensee’s high-level objectives of the planned treatment for safety-related and high-risk nonsafety-related SSCs. This paper discusses the NRC staff’s views regarding the treatment of SSCs at STP described by the licensee in its updated Final Safety Analysis Report (FSAR) in support of the exemption request, and provides the status of rulemaking that would incorporate risk insights into the treatment of SSCs at nuclear power plants.

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.

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