Experimental Investigation on Flash Evaporation Related to Pipe Leakage

2019 ◽  
Vol 6 (1) ◽  
Author(s):  
Benan Cai ◽  
Qi Zhang ◽  
Yu Weng ◽  
Hongfang Gu ◽  
Haijun Wang
Keyword(s):  
High Temperature ◽  
Thermal Fatigue ◽  
Weber Number ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Power Plants ◽  
Response Speed ◽  
Flash Evaporation ◽  
Main Pipe ◽  
Surge Line

Abstract Pipelines such as the surge line and main pipe are easily subjected to thermal stratification and thermal fatigue as a result of the nonuniform temperature distribution in the nuclear power plants. When the surge line or main pipe subjected to thermal stratification and thermal fatigue keeps operating for long time, the pipe leakage may happen due to the existence of pipeline crack. When the fluids with high temperature and pressure leak in the crack, the water will evaporate quickly, which means this process belongs to spray flash evaporation process. The flash evaporation related to pipe leak was experimentally studied in the paper. The experiment was carried out under high temperature and high pressure with low spray rate. The temperature and relative humidity (T&H) variations over time were monitored in the experiment with installing T&H detectors. The T&H variations at different measurement positions and with different spray rates were analyzed, respectively. In addition, the effect of the dimensionless parameters including the Weber number and Jakob number was also investigated. Results indicated that the response speed increased with the increase of the spray flow rate. Higher Weber number and higher Jakob number led to higher evaporation rate. The slight pipe leakage can be predicted by using the (T&H) in the hazardous areas.

Experimental Investigation on the Pipe Leakage

2017 ◽  
Author(s):  
Benan Cai ◽  
Qi Zhang ◽  
Yu Weng ◽  
Hongfang Gu ◽  
Haijun Wang
Keyword(s):  
High Temperature ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Normal Operation ◽  
Pipe System ◽  
Humidity Change ◽  
Temperature And Humidity ◽  
Surge Line

Pipelines are widely used in many fields including power industry, petroleum system etc. Pipelines such as the surge line and main pipe are easily subjected to thermal stratification as a result of the non-uniform temperature distribution in the nuclear power plants. Furthermore, pipelines can suffer from thermal fatigue in virtue of long-term uneven stress distribution. When the surge line or main pipe subjected to thermal stratification and thermal fatigue keeps operating for long time, the pipe leakage may happen because of the existence of pipeline crack. The thermal pipeline crack leakage mainly appears in the region with stress concentration. As the pipe system is always covered with thermal insulation layer in the actual nuclear power plants, it is hard for workers to observe pipeline leak, which can have a bad effect on the normal operation. Since the temperature and humidity close to the pipe crack due to leakage can change compared to the normal operation, we can infer from the temperature and humidity changes that the pipe leakage occurs. Based on this idea, the temperature and humidity near the crack of the pipe need to be measured to detect the leakage fields. As the fluids with high pressure and high temperature flow in the pipe system in an actual nuclear power plant, the pipe leakage experiment was performed in the high pressure and high temperature condition. When the fluids with high temperature and pressure leak in the crack, the water will evaporate quickly, which means this process belongs to spray flash evaporation process. The temperature and humidity variations were monitored in the experiment with temperature and humidity probes which have the advantage of responding to the change of temperature and humidity sensitively. The data collection program was mainly written based on the LABVIEW platform. The collecting time step was set 1s. As the measuring position and leakage flux are two key factors for the pipe leakage, the experiment was carried out with different measuring positions and leakage fluxes conditions. The experimental results showed that the leak flux had an important influence on the temperature and humidity near the pipe crack. The temperature and humidity started to change in a very short time with large leak flux. At the same time, the velocity of the temperature and humidity change was high with large leak flux. When the pipe leakage occurred in the location near the temperature and humidity probe, the temperature and humidity responded quickly and the velocity of temperature and humidity change was large. The experiment data can be used for the prediction of the pipe leakage in the nuclear power plants.

Stress and Fatigue Analysis of Pressurizer Surge Line under Thermal Stratification

2019 ◽  
Vol 795 ◽  
pp. 268-275
Author(s):  
Peng Tang ◽  
Zhi Wei Liu ◽  
Hong Wei Qiao ◽  
Peng Zhou Li
Keyword(s):  
Temperature Distribution ◽  
Temperature Difference ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Power Plants ◽  
Load Condition ◽  
Maximum Temperature ◽  
Large Temperature ◽  
Maximum Temperature Difference ◽  
Surge Line

Pressurizer surge line is one of the key equipments of nuclear power plants. The thermal stratification due to the intersection of hot and cold fluids inside the pressurizer surge line may affect the safe operation of nuclear power plant. In order to investigate the stress distribution and fatigue characteristics of surge line subjected to long-term thermal stratified loadings, a mechanical model of the surge line was established. And then, according to different temperature distribution assumptions, thermal stress analysis and fatigue assessment were conducted. The results show that the maximum stress appears under the load condition with maximum temperature difference, and finer temperature distribution can obtain more accurate stress and displacement results. The maximum value of fatigue cumulative coefficient appears at the junction of straight pipe and elbow with large temperature difference.

A Validation of RELAP on Predicting Nuclear Power Plant Phenomena

2018 ◽  
Author(s):  
Ji Soo Ahn ◽  
Michael Bluck ◽  
Matthew Eaton ◽  
Chris Jackson
Keyword(s):  
Thermal Stratification ◽  
Nuclear Power ◽  
Power Plants ◽  
Feed Water ◽  
Multiple Channels ◽  
Horizontal Pipe ◽  
One Dimensional ◽  
Piping Systems ◽  
Surge Line ◽  
Input Model

In this study, RELAP5’s capability to simulate thermal stratification under different conditions is assessed. In nuclear power plants (NPPs), thermal stratification can occur in the following locations: pressurizer, piping systems such as hot legs, cold legs, surge lines, and cooling tanks if available. In general, thermal stratification in a horizontal pipe could not be simulated by RELAP5 due to the inherent one-dimensional setting. Moreover, RELAP5 failed to simulate turbulent penetration which was often a pre-requisite prior to thermal stratification in a pipe. This type of situation could arise in connection between hot leg and surge line, spray lines, feed water lines, etc. It is recommended that for this type of problem CFD be used. In the literature, it was found that RELAP5 was capable of simulating thermal stratification in a pool or a tank-like component if multiple channels and crossflow junctions were used. However, due to uncertainties associated with the input model, the current RELAP5 model failed to reproduce experimental data and therefore further investigation would be required to identify the sources of error.

scholarly journals Benefits, Drawbacks, and Future Trends of Brayton Helium Gas Turbine Cycles for Gas-Cooled Fast Reactor and Very-High Temperature Reactor Generation IV Nuclear Power Plants

2020 ◽  
Vol 7 (1) ◽  
Author(s):  
Arnold Gad-Briggs ◽  
Emmanuel Osigwe ◽  
Pericles Pilidis ◽  
Theoklis Nikolaidis ◽  
Suresh Sampath ◽  
...  
Keyword(s):  
High Temperature ◽  
Gas Turbine ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Optimum Operating ◽  
Future Trends ◽  
Design Point ◽  
Gen Iv ◽  
Very High

Abstract Numerous studies are on-going on to understand the performance of generation IV (Gen IV) nuclear power plants (NPPs). The objective is to determine optimum operating conditions for efficiency and economic reasons in line with the goals of Gen IV. For Gen IV concepts such as the gas-cooled fast reactors (GFRs) and very-high temperature reactors (VHTRs), the choice of cycle configuration is influenced by component choices, the component configuration and the choice of coolant. The purpose of this paper to present and review current cycles being considered—the simple cycle recuperated (SCR) and the intercooled cycle recuperated (ICR). For both cycles, helium is considered as the coolant in a closed Brayton gas turbine configuration. Comparisons are made for design point (DP) and off-design point (ODP) analyses to emphasize the pros and cons of each cycle. This paper also discusses potential future trends, include higher reactor core outlet temperatures (COT) in excess of 1000 °C and the simplified cycle configurations.

scholarly journals Study on Probabilistic Safety Goals for Multimodule High-Temperature Gas-Cooled Reactor Based on Chinese Societal Risks

2021 ◽  
Vol 2021 ◽  
pp. 1-10
Author(s):  
Jinghan Zhang ◽  
Jun Zhao ◽  
Jiejuan Tong
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Regulatory Commission ◽  
Nuclear Safety ◽  
Energy Agency ◽  
Water Reactor ◽  
Regulatory Commission ◽  
High Temperature Gas ◽  
Probabilistic Safety

Nuclear safety goal is the basic standard for limiting the operational risks of nuclear power plants. The statistics of societal risks are the basis for nuclear safety goals. Core damage frequency (CDF) and large early release frequency (LERF) are typical probabilistic safety goals that are used in the regulation of water-cooled reactors currently. In fact, Chinese current probabilistic safety goals refer to the Nuclear Regulatory Commission (NRC) and the International Atomic Energy Agency (IAEA), and they are not based on Chinese societal risks. And the CDF and LERF proposed for water reactor are not suitable for high-temperature gas-cooled reactors (HTGR), because the design of HTGR is very different from that of water reactor. And current nuclear safety goals are established for single reactor rather than unit or site. Therefore, in this paper, the development of the safety goal of NRC was investigated firstly; then, the societal risks in China were investigated in order to establish the correlation between the probabilistic safety goal of multimodule HTGR and Chinese societal risks. In the end, some other matters about multireactor site were discussed in detail.

Numerical Estimation of Flow Penetrating Depth Into Stagnant Branch Pipes for Thermal Fatigue Prediction

2014 ◽  
Author(s):  
Yoshihiro Ishikawa ◽  
Yukihiko Okuda ◽  
Naoto Kasahara
Keyword(s):  
Thermal Stress ◽  
High Temperature ◽  
Temperature Fluctuation ◽  
Thermal Stratification ◽  
Branch Pipe ◽  
Large Temperature ◽  
Bend Pipe ◽  
Main Pipe ◽  
High Temperature Water ◽  
Temperature Water

In the nuclear power plants, there are many branch pipes with closed-end which are attached vertically to the main pipe. We consider a situation in which the high temperature water is transported in the main pipe, the branch pipe is filled with stagnant water which has lower temperature than the main flow, and the end of the branch pipe is closed. At the branch connection part, it is known that a cavity flow is induced by the shear force of the boundary layer which separates from the leading edge of the branch pipe along the main pipe wall. In cases where the high temperature water penetrates into the branch pipe, there is a possibility that a steep and large temperature gradient field, called “thermal stratification layer” is formed at the boundary between high and low temperature water in the branch pipe. If the thermal stratification layer is formed in a bend pipe, which is used for connecting the vertical branch pipe and to a horizontal pipe, at the same time, the temperature fluctuation by the thermal stratification layer motion occurs, there may cause the thermal stress in the piping material. Furthermore, keeping the piping material under the thermal stress, there might be a possibility of a crack on the surface of the bend pipe. For this reason, the evaluation of the position where the thermal stratification layer reaches is very important during early piping design process. And, deeply understanding regarding the phenomena, is also important. However, because of the complexities of the phenomena, it is difficult to immediately clarify the whole mechanisms of the thermal stress arising due to the temperature fluctuation by the thermal stratification layer change. The complete prediction method for the position of the thermal stratification layer based on the mechanisms that is able to be applied to any piping system, any temperature and any velocity conditions, is also difficult. Therefore, a practical approach is required. The authors attempt to develop the practical estimation method for the thermal stratification layer position using the three-dimensional Navier-Stokes simulation which was based on the Reynolds-average in order to reduce the computational costs. In this paper, three different configurations of the piping were simulated and the simulation results were compared with the experimental results obtained by the other research group.

Computational Study of Conjugate Heat Transfer in T-Junctions

2009 ◽  
Author(s):  
Simon Kuhn ◽  
Bojan Nicˇeno ◽  
Horst-Michael Prasser
Keyword(s):  
Thermal Fatigue ◽  
Nuclear Power ◽  
Power Plants ◽  
Cold Water ◽  
Computational Study ◽  
Solid Wall ◽  
Turbulent Viscosity ◽  
Life Time ◽  
Scale Model ◽  
Piping Systems

Thermal fatigue is a relevant problem in the context of life-time extension of nuclear power plants (NPP). In many piping systems in NPPs hot and cold water is mixed, which leads to high temperature fluctuations in the region close to the solid wall and resulting thermal loads on the pipe walls that can cause fatigue. One of the relevant geometric test cases for thermal fatigue is the mixing in T-junctions. In this study we apply large–eddy simulations (LES) to the mixing of hot and cold water in a T-junction. We perform a set of simulations by using different formulations of the LES subgrid scale model, i.e. standard Smagorinsky and dynamic procedure, to identify the influence of the modelled subgrid scales on the simulation results. The results exhibit a large difference between the models, which is caused by the use of turbulent viscosity wall–damping functions when applying the standard model.

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