Curtiss-Wright Advanced NozzleCheck Valves for Generation III+ Nuclear Power Plants

2017 ◽  
Author(s):  
Haykaz Mkrtchyan ◽  
Ararat Torosyan ◽  
Tsolag Apelian
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Reverse Flow ◽  
Flow Direction ◽  
Fluid Dynamic ◽  
Check Valve ◽  
Operating Conditions ◽  
Low Flow ◽  
Flow Conditions ◽  
Valve Design

Curtiss Wright introduced the first Normally Open NozzleCheck valves to the nuclear power industry nearly 20 years ago. This passive valve design was developed to address reoccurring maintenance and reliability issues often experienced by various check valve types due to low flow conditions. Specifically, premature wear on the hinge pins, bushings and severe seat impact damage had been discovered in several applications while the systems were in steady state operating conditions. Over the last two decades, Curtiss Wright has continued to improve upon the design of the valve, with the latest generation coming most recently in support of the Westinghouse AP1000 design and similar Generation III+ passive reactor requirements. This entirely new valve is designed with minimal stroke, ensuring quick closure under specified reverse flow conditions which no other check valve design could support. Additionally, features such as first in kind test ports, visual inspection points, and the ability to stroke the valve manually or with system fluid in line have resolved many of the shortcomings of previous inline welded flow check valves. Most importantly, advanced test based methodologies and models developed by Curtiss Wright, allow for accurate prediction of NozzleCheck valve performance. This paper presents the development of Curtiss Wright’s advanced Normally Open NozzleCheck Valve for Generation III and III+ nuclear reactor designs. The Valve performance was initially determined by using verified and validated computational fluid dynamic (CFD) methods. The results obtained from the CFD model were then compared to the data gathered from a prototype valve that was built and tested to confirm the performance predictions. Curtiss Wright has fully tested and qualified the Normally Open NozzleCheck valve, which is specifically designed for applications that require a high capacity in the forward flow direction and a closure at low flow rates with short stroke to minimize the hydraulic impact on the system. Paper published with permission.

Development of Enertech’s Advanced Normally Open NozzleCheck Valves for Generation III+ Nuclear Power Plants

2014 ◽  
Author(s):  
Haykaz Mkrtchyan
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Reverse Flow ◽  
Flow Direction ◽  
Fluid Dynamic ◽  
High Capacity ◽  
Check Valve ◽  
Operating Conditions ◽  
Flow Conditions ◽  
Valve Design

Enertech introduced the first Normally Open NozzleCheck valves to the nuclear power industry nearly 20 years ago. This passive valve design was developed to address reoccurring maintenance and reliability issues often experienced by various check valve types due to low or turbulent flow conditions. Specifically, premature wear on the hinge pins, bushings and severe seat impact damage had been discovered in several applications while the systems were in steady state operating conditions. Over the last two decades, Enertech has continued to improve upon the design of the valve, with the culmination coming most recently in support of Generation III+ passive reactor requirements. This entirely new valve is designed with minimal stroke, ensuring quick closure under low reverse flow conditions which no other check valve design could support. Additionally, features such as first in kind test ports, visual inspection points, and the ability to manually stroke the valve in line have resolved many of the short comings of previous inline welded flow check valves. Most importantly, advanced test based methodologies and models developed by Enertech, allow for accurate prediction of NozzleCheck valve performance. This paper presents the development of Enertech’s advanced Normally Open NozzleCheck Valve for Generation III and III+ nuclear reactor designs. The Valve performance was initially determined by using verified and validated computational fluid dynamic (CFD) methods. The results obtained from the CFD model were then compared to the data gathered from a prototype valve that was built and tested to confirm the performance predictions. Enertech has fully tested and qualified the Normally Open NozzleCheck valve which is specifically designed for applications that require a high capacity in the forward flow direction and a quick closure during low reverse flow condition with short stroke to minimize the hydraulic impact on the system.

Recent Patents on Valve Mechanism Device

2020 ◽  
Vol 13 (3) ◽  
pp. 230-241
Author(s):  
Ye Dai ◽  
Hui-Bing Zhang ◽  
Yun-Shan Qi
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Flow Direction ◽  
Research Progress ◽  
Valve Mechanism ◽  
Advantages And Disadvantages ◽  
Valve Structure ◽  
Valve Leakage ◽  
Safety And Reliability

Background: Valves are an important part of nuclear power plants and are the control equipment used in nuclear power plants. It can change the cross-section of the passage and the flow direction of the medium and has the functions of diversion, cutoff, overflow, and the like. Due to the earthquake, the valve leaks, which will cause a major nuclear accident, endangering people's lives and safety. Objective: The purpose of this study is to synthesize the existing valve devices, summarize and analyze the advantages and disadvantages of various devices from many literatures and patents, and solve some problems of existing valves. Methods: This article summarizes various patents of nuclear-grade valve devices and recent research progress. From the valve structure device, transmission device, a detection device, and finally to the valve test, the advantages and disadvantages of the valve are comprehensively analyzed. Results: By summarizing the characteristics of a large number of valve devices, and analyzing some problems existing in the valves, the outlook for the research and design of nuclear power valves was made, and the planning of the national nuclear power strategic goals and energy security were planned. Conclusion: Valve damage can cause serious safety accidents. The most common is valve leakage. Therefore, the safety and reliability of valves must be taken seriously. By improving the transmission of the valve, the problems of complicated valve structure and high cost are solved.

Measurement of Flow-Induced Forces: A Single Flexible Tube in a Rigid Triangular-Pitch Bundle Subjected to Two-Phase Cross-Flow

2015 ◽  
Author(s):  
Enrico Deri ◽  
Joël Nibas ◽  
Olivier Ries ◽  
André Adobes
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Tube Bundle ◽  
Fluid Dynamic ◽  
Cross Flow ◽  
Maintenance Policy ◽  
Excitation Spectra ◽  
Two Phase ◽  
Force Coefficients ◽  
Out Of Plane

Flow-induced vibrations of Steam Generator tube bundles are a major concern for the operators of nuclear power plants. In order to predict damages due to such vibrations, EDF has developed the numerical tool GeViBus, which allows one to asses risk and thereafter to optimize the SG maintenance policy. The software is based on a semi analytical model of fluid-dynamic forces and dimensionless fluid force coefficients which need to be assessed by experiment. The database of dimensionless coefficients is updated in order to cover all existing tube bundle configurations. Within this framework, a new test rig was presented in a previous conference with the aim of assessing parallel triangular tube arrangement submitted to a two-phase cross-flow. This paper presents the result of the first phase of the associated experiments in terms of force coefficients and two-phase flow excitation spectra for both in-plane and out-of-plane vibration.

Design, Analysis and System-Level Modelling of a Single Axis Capacitive Accelerometer

2016 ◽  
Author(s):  
Zakriya Mohammed ◽  
Owais Talaat Waheed ◽  
Ibrahim (Abe) M. Elfadel ◽  
Aveek Chatterjee ◽  
Mahmoud Rasras
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Noise Analysis ◽  
Linear Displacement ◽  
Operating Conditions ◽  
System Level ◽  
Complete Analysis ◽  
Capacitive Accelerometer ◽  
Static Capacitance ◽  
Cross Axis

The paper demonstrates the design and complete analysis of 1-axis MEMS capacitive accelerometer. The design is optimized for high linearity, high sensitivity, and low cross-axis sensitivity. The noise analysis is done to assure satisfactory performance under operating conditions. This includes the mechanical noise of accelerometer, noise due to interface electronics and noise caused by radiation. The latter noise will arise when such accelerometer is deployed in radioactive (e.g., nuclear power plants) or space environments. The static capacitance is calculated to be 4.58 pF/side. A linear displacement sensitivity of 0.012μm/g (g = 9.8m/s2) is observed in the range of ±15g. The differential capacitive sensitivity of the device is 90fF/g. Furthermore, a low cross-axis sensitivity of 0.024fF/g is computed. The effect of radiation is mathematically modelled and possibility of using these devices in radioactive environment is explored. The simulated noise floor of the device with electronic circuit is 0.165mg/Hz1/2.

Thermal Design Options Using Uranium Carbide and Uranium Dicarbide in SCWR Uniformly-Heated Fuel Channel

2009 ◽  
Author(s):  
Leyland J. Allison ◽  
Lisa Grande ◽  
Sally Mikhael ◽  
Adrianexy Rodriguez Prado ◽  
Bryan Villamere ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Operating Conditions ◽  
Thermal Design ◽  
Nuclear Fuels ◽  
Viable Option ◽  
Bulk Fluid ◽  
Fuel Channel ◽  
Uranium Carbide

SuperCritical Water-cooled nuclear Reactor (SCWR) options are one of the six reactor options identified in Generation IV International Forum (GIF). In these reactors the light-water coolant is pressurized to supercritical pressures (up to approximately 25 MPa). This allows the coolant to remain as a single-phase fluid even under supercritical temperatures (up to approximately 625°C). SCW Nuclear Power Plants (NPPs) are of such great interest, because their operating conditions allow for a significant increase in thermal efficiency when compared to that of modern conventional water-cooled NPPs. Direct-cycle SCW NPPs do not require the use of steam generators, steam dryers, etc. allowing for a simplified NPP design. This paper shows that new nuclear fuels such as Uranium Carbide (UC) and Uranium Dicarbide (UC2) are viable option for the SCWRs. It is believed they have great potential due to their higher thermal conductivity and corresponding to that lower fuel centerline temperature compared to those of conventional nuclear fuels such as uranium dioxide, thoria and MOX. Two conditions that must be met are: 1) keep the fuel centreline temperature below 1850°C (industry accepted limit), and 2) keep the sheath temperature below 850°C (design limit). These conditions ensure that SCWRs will operate efficiently and safely. It has been determined that Inconel-600 is a viable option for a sheath material. A generic SCWR fuel channel was considered with a 43-element bundle. Therefore, bulk-fluid, sheath and fuel centreline and HTC profiles were calculated along the heated length of a fuel channel.

Environmental Fatigue Test of CF8M With a Small Autoclave Simulating PWR Conditions

2008 ◽  
Author(s):  
Il-Seok Jeong ◽  
Gag-Hyeon Ha ◽  
Tae-Ryoung Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Low Cycle Fatigue ◽  
Test Procedure ◽  
Cyclic Strain ◽  
Fatigue Tests ◽  
Operating Conditions ◽  
Measuring Instruments ◽  
Pressurized Water Reactor ◽  
Pressurized Water

To develop a fatigue design curve of cast stainless steel CF8M used in primary piping material of nuclear power plants, low-cycle fatigue tests have been conducted by Korea Electric Power Research Institute (KEPRI). A small autoclave simulated the environment of a pressurized water reactor (PWR), 15 MPa and 315 °C. Fatigue life was measured in terms of the number of cycles with the variation of strain amplitudes at 0.04%/s strain rate. A small autoclave of 1 liter and cylindrical solid fatigue specimens were used for the strain-controlled low cycle environmental fatigue tests to make the experiments convenient. However, it was difficult to install displacement measuring instruments at the target length of the specimens inside the autoclave. To mitigate the difficulty displacement data measured at the shoulders of the specimen were calibrated based on the data relation of the target and shoulder length of the specimen during hot air test conditions. KEPRI developed a test procedure to perform low cycle environmental fatigue tests in the small autoclave. The procedure corrects the cyclic strain hardening effect by performing additional tests in high temperature air condition. KEPRI verified that the corrected test result agreed well with that of finite element method analysis. The process of correcting environmental fatigue data would be useful for producing reliable fatigue curves using a small autoclave simulating the operating conditions of a PWR.

Study of the Influence of Microstructure and Intergranular Carbides on the Oxidation Behavior of a Nickel Base Alloy 690 TT in Supercritical Water Nuclear Reactor Conditions

2020 ◽  
Author(s):  
Alberto Sáez-Maderuelo ◽  
María Luisa Ruiz-Lorenzo ◽  
Francisco Javier Perosanz ◽  
Patricie Halodová ◽  
Jan Prochazka ◽  
...  
Keyword(s):  
Supercritical Water ◽  
Nuclear Power ◽  
Power Plants ◽  
Oxidation Behavior ◽  
Nuclear Reactor ◽  
Base Alloy ◽  
Nuclear Industry ◽  
Operating Conditions ◽  
Alloy 690 ◽  
Reactor Operating

Abstract Alloy 690, which was designed as a replacement for the Alloy 600, is widely used in the nuclear industry due to its optimum behavior to stress corrosion cracking (SCC) under nuclear reactor operating conditions. Because of this superior resistance, alloy 690 has been proposed as a candidate structural material for the Supercritical Water Reactor (SCWR), which is one of the designs of the next generation of nuclear power plants (Gen IV). In spite of this, striking results were found [1] when alloy 690 was tested without intergranular carbides. These results showed that, contrary to expectations, the crack growth rate is lower in samples without intergranular carbides than in samples with intergranular carbides. Therefore, the role of the carbides in the corrosion behavior of Alloy 690 is not yet well understood. Considering these observations, the aim of this work is to study the effect of intergranular carbides in the oxidation behavior (as a preliminary stage of degenerative processes SCC) of Alloy 690 in supercritical water (SCW) at two temperatures: 400 °C and 500 °C and 25 MPa. Oxide layers of selected specimens were studied by different techniques like Scanning Electron Microscope (SEM) and Auger Electron Spectroscopy (AES).

Diagnosis of Corrosion Processes in Nuclear Power Plants Secondary Piping Structures

2020 ◽  
Author(s):  
Koushik A. Manjunatha ◽  
Andrea Mack ◽  
Vivek Agarwal ◽  
David Koester ◽  
Douglas Adams
Keyword(s):  
Experimental Test ◽  
Classification Accuracy ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Conditions ◽  
Condition Based Maintenance ◽  
Support Vector ◽  
Maintenance Strategy ◽  
Test Bed

Abstract The current aging management plans of passive structures in nuclear power plants (NPPs) are based on preventative maintenance strategies. These strategies involve periodic, manual inspection of passive structures using nondestructive examination (NDE) techniques. This manual approach is prone to errors and contributes to high operation and maintenance costs, making it cost prohibitive. To address these concerns, a transition from the current preventive maintenance strategy to a condition-based maintenance strategy is needed. The research presented in this paper develops a condition-based maintenance capability to detect corrosion in secondary piping structures in NPPs. To achieve this, a data-driven methodology is developed and validated for detecting surrogate corrosion processes in piping structures. A scaled-down experimental test bed is developed to evaluate the corrosion process in secondary piping in NPPs. The experimental test bed is instrumented with tri-axial accelerometers. The data collected under different operating conditions is processed using the Hilbert-Huang Transformation. Distributional features of phase information among the accelerometers were used as features in support vector machine (SVM) and least absolute shrinkage and selection operator (LASSO) logistic regression methodologies to detect changes in the pipe condition from its baseline state. SVM classification accuracy averaged 99% for all models. LASSO classification accuracy averaged 99% for all models using the accelerometer data from the X-direction.

An LDV Study of the Influence of the Radiator on a Cooling Module Fan

2002 ◽  
Author(s):  
Robert J. Martinuzzi ◽  
Gregory A. Kopp ◽  
Brian Havel
Keyword(s):  
Laser Doppler Velocimetry ◽  
Reverse Flow ◽  
Low Flow ◽  
Doppler Velocimetry ◽  
Flow Conditions ◽  
Flow Rates ◽  
Cooling Fan ◽  
Outlet Flow ◽  
Flow Through ◽  
Cooling Module

The influence of the radiator on the flow through an automotive cooling fan module was investigated using Laser Doppler Velocimetry for three different flow conditions. It is found that at the nominal design point, the radiator acts as an effective flow straightener. At low flow rates, fan induced pre-swirl is significant, but the radiator helps reduce reverse flow through the fan. Under ram air conditions the upstream inlet distortions persist through the module resulting in a highly distorted outlet flow.

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