Flow Signal Characteristics of an Ultrasonic Transducer at a High Temperature

2011 ◽  
Author(s):  
Ki Won Lim ◽  
Jaeheun Rho
Keyword(s):  
High Temperature ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Transit Time ◽  
Nuclear Power ◽  
Room Temperature ◽  
Ultrasonic Transducer ◽  
Feed Water ◽  
Pipe Line ◽  
Driving Circuit

The exact measurement of feed water flow is the major factor in nuclear power plant efficiency. However, due to the fouling problem, the venturi nozzle used in feed water measurement frequently causes a decrease in the efficiency of the nuclear power plant. To avoid this problem, ultrasonic technology is a reasonable candidate. The temperature of the feed water is about 300 °C. The commercial piezoelectric element used in an ultrasonic transducer preserves its characteristics up to a temperature of 120 °C. This problem must be overcome in order to use an ultrasonic flowmeter to measure the feed water flow. To address this issue, we designed a thermal block to insulate the high temperature from the pipe line. The method we used included a clamp-on type transducer and a driving circuit with a transit time difference method. The signals from the driving circuit were measured and the ultrasonic transducer assembly was tested at room temperature and at a high temperature of 300 °C. The test results revealed that the transit time difference was reasonably proportional to the flow velocity at room temperature, and the signals of the transducer installed on the pipe line were the same at 300 °C as those at room temperature. This result confirmed that the ultrasonic pulse was working well through the thermal block and in the high temperature fluid.

Verification of Alarm Displays for the Nuclear Power Plant With Two Modular High-Temperature Gas-Cooled Reactors

2018 ◽  
Author(s):  
Jia Qianqian ◽  
Guo Chao ◽  
Li Jianghai ◽  
Qu Ronghong
Keyword(s):  
High Temperature ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Modular Design ◽  
Early Stage ◽  
Integrated System ◽  
Alarm System ◽  
Control Room ◽  
High Temperature Gas

The nuclear power plant with two modular high-temperature gas-cooled reactors (HTR-PM) is under construction now. The control room of HTR-PM is designed. This paper introduces the alarm displays in the control room, and describes some verification and validation (V&V) activities of the alarm system, especially verification for some new human factor issues of the alarm system in the two modular design. In HTR-PM, besides the regular V&V similar to other NPPs, the interference effect of the alarm rings of the two reactor modules at the same time, and the potential discomfort of the two reactor operators after shift between them are focused. Verifications at early stage of the two issues are carried on the verification platform of the control room before the integrated system validation (ISV), and all the human machine interfaces (HMIs) in the control room, including the alarm system are validated in ISV. The test results on the verification platform show that the alarm displays and rings can support the operators understand the alarm information without confusion of the two reactors, and the shift between the two reactor operators have no adverse impact on operation. The results in ISV also show that the alarm system can support the operators well.

scholarly journals Estimation of the efficiency of combining a npp with a hydrogen facility under conditions of safe use of hydrogen in a steam turbine cycle

2021 ◽  
Vol 23 (2) ◽  
pp. 56-69
Author(s):  
R. Z. Aminov ◽  
A. N. Bairamov
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Power Unit ◽  
Nuclear Power ◽  
Peak Power ◽  
Live Steam ◽  
Feed Water ◽  
Current Period ◽  
Additional Heating ◽  
The Future

THE PURPOSE. System efficiency and competitiveness assess of a new scheme for combining a nuclear power plant with a hydrogen complex based on additional heating of feed water and superheating of live steam in front of the high-pressure cylinder of a steam turbine. METHODS. Basic laws of thermodynamics were applied when developing and substantiating a new scheme for combining a nuclear power plants (NPP) with a hydrogen facility; theoretical regularities were applied of heat engineering; basic regularity were applied of fatigue wear of power equipment and assessment of its working resourse; basic regularities were applied for the assessment of operating costs and net present value (NPV). RESULTS. A new scheme is presented of the combination of a nuclear power plant with a hydrogen facility and a description of its operating principle on the example of a two-circuit nuclear power plant with a VVER-1000 reactor and a C-1000-60 / 1500 turbine. The data are presented on an increase in the productivity of steam generators at nuclear power plants with additional heating of feed water in the range of 235-250 ° C from its nominal value of 230 ° C. The temperature was estimated of live steam superheat depending on the temperature of the additional heating of the feed water. The results are presented of the calculation of the generated peak power by the power unit and the efficiency of conversion of the night off-peak power of the NPP into peak power, as well as the efficiency of the power unit of the NPP depending on the temperature of additional heating of the feed water. Main regularities are given for taking into account the fatigue wear of the main equipment of the hydrogen facility, including the rotor of the NPP turbine in the conditions of the stress-cyclic operation. The results are presented of assessing the cost of peak electricity NPP in combination with a hydrogen facility in comparison with a pumped storage power plant (PSPP) both for the current period and for the future until 2035. CONCLUSION. Hydrogen facility efficiency and competitiveness depends significantly on the intensity of the use of the main equipment in the conditions of the intense-cyclic operation. The hydrogen facility will competitiveness noticeably increase in comparison with the PSPP in the future. Efficiency of the NPP power unit and NPV is highest when the feed water is heated to 235 ° C and superheating of live steam in front of the high-pressure cylinder of the C-1000-60/1500 turbine up to 470°C.The hydrogen facility competes with the PSPP with her specific capital investment at the level of 660 USD / kW, provided that the boosting capabilities of the turbine are used with live steam overheating at 300 ° C and additional heating of feed water to 235°C on the current period. The PSPP does not compete with the hydrogen facility both for the current period and in the future with her specific capital investment of $ 1,500 / kW and above.

Current Wall Thinning Measured on Piping System of Main or Auxiliary Boiler Plant in Ships

2008 ◽  
Author(s):  
H. Shiihara ◽  
H. Matsushita ◽  
Y. Nagayama
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Wall Thickness ◽  
Nuclear Power ◽  
Feed Water ◽  
Piping System ◽  
Thickness Gauge ◽  
Corrosion Failure ◽  
Boiler Plant ◽  
Secondary Line

A disaster happened in a nuclear power plant in Japan in August 2004, which was caused by failure of condensation water pipe in the secondary line. Shipping industries were concerned for possibility of occurrence of such a disaster in ships due to its construction similarity to marine boiler plant in steam, feed water and condensation piping for main or auxiliary boilers. Nippon Kaiji Kyokai has therefore investigated and gathered data of piping lines corrosion in ships collaborated with major Japanese ship owners right after the disaster. The results show that similar corrosion failure as in the nuclear power plant has occurred in shipboard steam/feed water/condensation water pipes for main and auxiliary boiler plants without causing severe consequences. The wall thickness measurements on actual pipe lines of steam, feed water and condensation water at bend parts, at T-junction, behind orifices, behind valves and at diffusers/reducers with a ultrasonic thickness gauge show a very definite evidence of a reduction in wall thickness of carbone steel pipes. It was confirmed that the amount of actual reduction in wall thickness could be well predicted by Kastner Equation [2–3].

Advanced Nuclear Process Heat Concepts and Applications

2006 ◽  
Author(s):  
Reiner W. Kuhr ◽  
Charles Bolthrunis ◽  
Michael Corbett ◽  
Ed Lahoda
Keyword(s):  
High Temperature ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Carbon Utilization ◽  
Nuclear Process ◽  
High Temperature Process ◽  
Process Plant ◽  
Screening Study ◽  
Process Heat

This paper presents a summary of a screening study to select the most advantageous applications for nuclear process heat. The review is focused on the application of the Pebble Bed Modular Reactor (PBMR) technology adapted for process heat applications. This technology is unique in its smaller modular size and ability to deliver high temperature process heat at conditions that allow higher value applications. The implementation of projects for nuclear process heat and hydrogen production will require collaboration between nuclear power plant operators and process plant owners who will benefit from lower costs of heat delivery. Heat and hydrogen from nuclear water splitting can be used to displace expensive fuels, extend carbon utilization for products and reduce CO2 emissions and other environmental impacts.

Fatigue Monitoring in the Context of Long-Term Operation of the Goesgen Nuclear Power Plant Using AREVA’s FAMOSi

2015 ◽  
Author(s):  
Thomas Wermelinger ◽  
Florian Bruckmüller ◽  
Benedikt Heinz
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Primary Circuit ◽  
Feed Water ◽  
Term Operation ◽  
Operation Modes ◽  
Fatigue Monitoring ◽  
Load Evaluation

In the context of long-term operation or lifetime extension most regulatory bodies demand from utilities and operators of nuclear power plants to monitor and evaluate the fatigue of system, structures and components systematically. As does the Swiss Federal Nuclear Safety Inspectorate ENSI. The nuclear power plant Goesgen started its commercial operation in 1979 and will go into long-term operation in 2019. The increased demand for monitoring and evaluating fatigue due to the pending long-term operation led the Goesgen nuclear power plant to expand the scope of their surveillance and therefore to install AREVA’s fatigue monitoring system FAMOSi in the 2014 outage. The system consists of 39 measurement sections positioned at the primary circuit and the feed-water nozzles of the steam generators. The locations were chosen due to their sensitivity for fatigue. The installed FAMOSi system consists of a total of 173 thermocouples which were mounted in order to get the necessary input data for load evaluation. The advantage of FAMOSi is the possibility to obtain real data of transients near places with highest fatigue usage factors. Examples of steam generator feed-in during heating-up and cooling-down will be given. In addition, spray events before and after the installation of closed loop controlled spray valves will be compared. The measurements and the results of the load evaluation are not only of interest for internal use e.g. in regard to optimization of operation modes (e.g. load-following), but must also be reported to ENSI annually. In addition, by evaluation of stresses and determination of usage factors combined with an optimization of operation modes an early exchange of components can be avoided.

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