An Assessment of Margin in Design Steam Hammer Forces for Combined Cycle Power Plants

2002 ◽  
Author(s):  
D. Zheng ◽  
A. T. Vieira ◽  
J. M. Jarvis
Keyword(s):  
Power Plants ◽  
Classical Method ◽  
Combined Cycle ◽  
Nuclear Industry ◽  
Piping System ◽  
Valve Closure ◽  
Velocity Correlation ◽  
Two Phase ◽  
Best Estimate ◽  
Main Steam

All combined cycle steam plants have rapid-closing stop valves in steam lines to protect the turbine. The rapid valve closure produces a steam hammer in the piping resulting in large forces for which the piping system and supporting structures need to be designed. These forces are typically calculated using the classical Method Of Characteristics (MOC) solution. An evaluation has been conducted which compares the forces computed using the classical methods with a best-estimate approach. This comparison has been done to define margin, and to benchmark and identify potential refinements in the techniques used for evaluating steam hammer loads. The best-estimate approach involves the use of the RELAP5 computer program. RELAP5 is used extensively in the Nuclear Industry to evaluate fast thermal hydraulic transients. It has the capability to analyze subcooled liquid, two-phase and saturated or superheated steam piping system. The models used in RELAP5 are best estimate results in comparison to the MOC solution which are mathematically derived from theory. The compressible flow program GAFT is used to obtain the MOC solution. The main steam line of a single Heat Recovery Steam Generator combined cycle plant is modeled with both the GAFT program and with a PC version of RELAP5. Identical piping lengths, mass flow rates, pressures are used in each model. Also, a stop valve closure time of 100 milliseconds is modeled. As RELAP5 output results are pressure, flow rate, velocity, and density, the resultant forces are generated using the R5FORCE program, a post-processor to compute associated transient forces on straight piping links. The GAFT program, which is specifically designed to compute steam hammer forces, computes the force history internally on straight piping lengths. A comparison of the peak force from GAFT and from RELAP for every piping link has been generated. Through the comparison, both RELAP5 and GAFT have been verified for the evaluation of rapid valve closure reaction loads. The comparison also shows that the classical method typically over-predicts the best-estimate solution by 15% to 20% for straight piping links. Although not confirmed, a better agreement between the two methods would be expected if a more accurate steam sonic velocity correlation and valve closure model are incorporated into the classical solution. Theis study helps to quantify the degree of conservatism inherent in the classical approach.

Comparisons of CFD and Traditional Solutions for Steam Hammer Events

2017 ◽  
Author(s):  
Alex Mayes ◽  
Kshitij P. Gawande ◽  
Dennis K. Williams
Keyword(s):  
Pressure Wave ◽  
Power Plants ◽  
Peak Load ◽  
Critical Length ◽  
Pressure Change ◽  
Piping System ◽  
Valve Closure ◽  
Pipe System ◽  
Closure Time ◽  
Pressure Changes

Sudden pressure changes in the piping system of power plants are inevitable, and thus potential serious damage to large components, piping system, and piping supports is possible. To protect valuable components from such events, abrupt valve closure is employed to restrict the flow and prevent significant incidents and the resulting plant downtime. Unfortunately, when a valve is suddenly closed to prevent damage caused by unexpected events, a pressure wave within the flow is created, which travels upstream and impacts at the pipeline elbows. These events, involving sudden changes in pressure, are known as steam hammer. This steam hammer pressure wave, traveling through the pipe system, is capable of producing significant transient loads and stresses, which can disrupt the piping supports. As such there is a need for further investigation. The pressure wave depends on the characteristics of the flow, valve closure time, the elbow-to-elbow pipe section lengths, and the piping system flexibility. The present study performs a CFD analysis of the fluid experiencing such a sudden pressure change. OpenFOAM is used for this analysis and considers all the flow parameters, valve closure time, and critical length of the straight pipe. The study intends to provide a means of calculating the transient steam hammer loads applied on the pipe elbows, which consequently allows appropriate pipe support selection based upon the resulting peak loads. This computational analysis is compared to analytical methods for peak load determination such as rigid column theory, the Joukowsky method, and the steam hammer method explained by Coccio (1967) and Goodling (1989).

Bull Run Fossil Plant: Technical Design Methods for Superheat Pendant Outlet Headers Replacement

2010 ◽  
Author(s):  
Salah E. Azzazy ◽  
Russell D. Cochran ◽  
Larry Sam Cox
Keyword(s):  
Power Plants ◽  
Point Of View ◽  
Piping System ◽  
Main Challenge ◽  
Tennessee Valley ◽  
Pipe Segment ◽  
Technical Evaluation ◽  
Engineering Mechanics ◽  
Main Steam ◽  
Unconventional Method

Bull Run Unit 1, rated at 950 MW, is the first of four fossil supercritical power plants at Tennessee Valley Authority (TVA). The unit went into commercial operation in 1967. The boiler (consists of two furnaces) built by Combustion Engineering (CE) has a radiant reheat twin divided furnace with tangential-fired coal burners. The unit’s maximum continuous rating (MCR) is 6,400,000 lbs/hr of main steam flow, with a design temperature of 1003°F and pressure of 3840 psig. Through the end of 2008, the unit had a total of approximately 670 cumulative starts and 333,185 operating hours. After years of numerous tube cracks at the Superheat Pendant Outlet Header/Tube Nozzles resulting in repetitive forced plant shutdowns, TVA decided to replace the two Outlet Headers (one for each furnace) in Fall 2008 during a reliability outage. Since the entire Main Steam piping system was installed with cold pull at almost every longitudinal pipe segment, the main challenge from the engineering mechanics point of view was how to restrain the piping system especially at the Crossover Outlet Links inside each furnace Penthouse. Further constructability reviews indicated that there were not enough adjacent steel frames inside each furnace to restrain the four Crossover Outlet Links in the three global directions during the Outlet Headers replacement inside each Penthouse. The only existing steel above the Crossover Outlet Links is embedded in asbestos insulation, and the removal of the insulation to provide access for the temporary restraints was determined to be costly and time consuming. The insulation removal would have also caused the scheduled outage to be extended significantly and unrealistically. After careful assessment, technical evaluation, and several constructability reviews; it was decided to take an unconventional approach for relieving the inherent cold pull in three global directions by cutting the four Mixing Headers outside each furnace. In addition, the concept of installing several temporary restraints was utilized for the vertical and lateral directions inside the furnace Penthouse, as well as several others outside the Boiler to control the piping configuration of the four Mixing Headers. This approach achieved two purposes: 1- relieving the inherent cold pulls in three global directions and 2- controlling the four Outlet Links pipe end positions with respect to the new Superheat Pendant Outlet Header nozzles. This unconventional method used to relieve the piping cold pull from outside the Boilers, to control the Outlet Links movements inside the Boiler Penthouses, and to restrain the entire Main Steam piping system was successfully developed and implemented in the Fall 2008 reliability outage to replace the two Superheat Pendant Outlet Headers. This unconventional method is described in this paper.

Best Estimate Plus Uncertainty Analysis of LBLOCA for a Pickering B CANDU Reactor

2010 ◽  
Author(s):  
Larry Blake ◽  
George Gavrus ◽  
Jack Vecchiarelli ◽  
J. Stoklosa
Keyword(s):  
Power Plants ◽  
Safety Margin ◽  
Reactor Power ◽  
Natural Uranium ◽  
Nuclear Industry ◽  
Initial Increase ◽  
Power Pulse ◽  
Pulse Phase ◽  
Best Estimate ◽  
Safety Margins

The Pickering B Nuclear Generating Station consists of four CANDU reactors. These reactors are horizontal pressure tube, heavy water cooled and moderated reactors fuelled with natural uranium. Under a postulated large break loss of coolant accident (LOCA), positive reactivity results from coolant void formation. The transient is terminated by the operation of the safety systems within approximately 2 seconds of the start of the transient. The initial increase in reactor power, terminated by the action of the safety system, is termed the power pulse phase of the accident. In many instances the severity of an LBLOCA can be characterized by the adiabatic energy deposited to the fuel during this phase of the accident. Historically, Limit of Operating Envelope (LOE) calculations have been used to characterize the severity of the accident. LOE analyses are conservative analyses in which the key operational and safety related parameters are set to conservative or limiting values. Limit based analyses of this type result in calculated transient responses that will differ significantly from the actual expected response of the station. As well, while the results of limit calculations are conservative, safety margins and the degree of conservatism is generally not known. As a result of these factors, the use of Best Estimate Plus Uncertainty (BEPU) analyses in safety analyses for nuclear power plants has been increasing. In Canada, the nuclear industry has been pursuing best estimate analysis through the BEAU (Best Estimate Analysis and Uncertainty) methodology in order to obtain better characterization of the safety margins. This approach is generally consistent with those used internationally. Recently, a BEAU analysis of the Pickering B NGS was completed for the power pulse phase of a postulated Large Break LOCA. The analysis comprised identification of relevant phenomena through a Phenomena Identification and Ranking (PIRT) process, assessment of the code input uncertainties, sensitivity studies to quantify the significance of the input parameters, generation of a functional response surface and its validation, and determination of the safety margin. The results of the analysis clearly demonstrate that the Limit of Operating Envelope (LOE) results are significantly conservative relative to realistic analysis even when uncertainties are considered. In addition, the extensive sensitivity analysis performed to supplement the primary result provides insight into the primary contributors to the results.

Failure Analysis of Low Pressure Evaporator Tubes in a Typical Combined Cycle Power Plant

2011 ◽  
Vol 110-116 ◽  
pp. 4607-4614
Author(s):  
M. Nematollahi ◽  
M. Rezaeian
Keyword(s):  
Power Plant ◽  
Steam Generator ◽  
Power Plants ◽  
Fluid Dynamic ◽  
Flow Distribution ◽  
Low Pressure ◽  
Combined Cycle ◽  
Two Phase ◽  
Combined Cycle Power Plant ◽  
Suitable Material

Flow-induced corrosion is one of the most prevalent tube damage mechanisms in steam generators of power plants. In this study, tube failure of a steam generator in Fars Combined Cycle Power Plant is evaluated. In addition to analysis of the measured tube thicknesses and the failure statistics data, computational fluid dynamic (CFD) methods are used to simulate flow distribution inside and outside of the tubes in one header of the low pressure circuit of the plant steam generator. The results show that regarding the created two-phase flow pattern inside the tubes, the droplet impingement erosion is the main source of tube failures in the bending areas where the extrados surface of the tubes are partially prone to the droplets. The results are useful for modifying the design of the steam generator from different viewpoints such as, optimal design for appropriate configuration of downcomer, header and footer and tube bending. Also, selecting suitable material for the steam generator tubes and implementation of protective coating in risky areas would benefit from the present results.

On-Line Monitoring System for Main Steam Piping of Power Plants

2009 ◽  
Author(s):  
Ning Wang ◽  
Zhengdong Wang ◽  
Yingqi Chen
Keyword(s):  
Power Plant ◽  
Monitoring System ◽  
Remote Monitoring ◽  
Power Plants ◽  
Residual Life ◽  
Thermal Power ◽  
Creep Damage ◽  
Piping System ◽  
On Line ◽  
Main Steam

An on-line life prediction system is developed for remote monitoring of material aging in a main steam piping system. The stress analysis of piping system is performed by using the finite element method. A sensor network is established in the monitoring system. The creep damage is evaluated from strain gages and a relationship is given based on a database between the damage and residual life. Web technologies are used for remote monitoring to predict the residual life for every part of the piping system. This system is useful for safety assessment procedures in thermal power plant, nuclear power plant and petrochemical industries.

Welding Residual Stress Solutions for Dissimilar Metal Surge Line Nozzle Welds

2010 ◽  
Vol 132 (2) ◽  
Author(s):  
D. Rudland ◽  
A. Csontos ◽  
T. Zhang ◽  
G. Wilkowski
Keyword(s):  
Residual Stress ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Industry ◽  
Welding Residual Stress ◽  
Nozzle Geometry ◽  
Material Strength ◽  
Piping System ◽  
Dissimilar Metal ◽  
Weld Material

At the end of 2006, defects were identified using ultrasonic testing in three of the pressurizer nozzle dissimilar metal (DM) welds at the Wolf Creek nuclear power plant. Understanding welding residual stress is important in the evaluation of why and how these defects occur, which in turn helps to determine the reliability of nuclear power plants. This paper presents analytical predictions of welding residual stress in the surge nozzle geometry identified at Wolf Creek. The analysis procedure in this paper includes not only the pass-by-pass welding steps, but also other essential fabrication steps of pressurizer surge nozzles. Detailed welding simulation analyses have been conducted to predict the magnitude of these stresses in the weld material. Case studies were carried out to investigate the change in the DM main weld stress fields resulting from different boundary conditions, material strength, weld sequencing, as well as simulation of the remaining piping system stiffness. A direct comparison of these analysis methodologies and results has been made in this paper. Weld residual stress results are compared directly to those calculated by the nuclear industry.

Load Definition on BWR Steam Dryer due to Turbine Stop Valve Closer

2013 ◽  
Author(s):  
Asgar Faal-Amiri ◽  
Hari Srivastava ◽  
Frederick J. Moody
Keyword(s):  
Compression Wave ◽  
Jet Impingement ◽  
Steam Line ◽  
Piping System ◽  
Valve Closure ◽  
Pressure Force ◽  
Stop Valve ◽  
Main Steam ◽  
Pipe Geometry ◽  
Reactor Pressure

In the boiling water reactor (BWR), closure of a turbine stop valve (TSV) on the main steam piping system, hereafter called main steam (MS) line, will reduce the steam flow velocity to zero in about 100 msec (0.1 Sec) and create a pressure disturbance that propagates backward through the MS line toward the reactor pressure vessel (RPV). When the compression wave reaches the RPV, it expands into a region at lower pressure, transmitting a compression wave at acoustic speed from the pipe geometry into a complex geometric region at reactor pressure, bounded by the steam dryer and RPV inner wall surfaces. The arriving compression wave expands on the dryer surface, creating a pressure force, while simultaneously steam in the steam line, compressed from the valve closure to a pressure higher than that in the reactor, flows backwards into the RPV, creating a jet impingement, or “backflow” force on the dryer. Simplified, conservative modeling is applied in this study to obtain reasonable bounding loads which allow for the dryer curvature and other complexities of the dryer-vessel geometry.

Time History Steam Hammer Analysis for Critical Hot Lines in Thermal Power Plants

2014 ◽  
Author(s):  
Ahmed H. Bayoumy ◽  
Anestis Papadopoulos
Keyword(s):  
Power Plant ◽  
Dynamic Response ◽  
Power Plants ◽  
Thermal Power ◽  
Time History ◽  
Piping System ◽  
Pressure Waves ◽  
Root Cause ◽  
Pipe Line ◽  
Main Steam

Pressure surges and fluid transients, such as steam and water hammer, are events that can occur unexpectedly in operating power plants causing significant damages. When these transients occur the power plant can be out of service for long time, until the root cause is found and the appropriate solution is implemented. In searching for root cause of transients, engineers must investigate in depth the fluid conditions in the pipe line and the mechanism that initiated the transients. The steam hammer normally occurs when one or more valves suddenly close or open. In a power plant, the steam hammer could be an inevitable phenomenon during turbine trip, since valves (e.g., main steam valves) must be closed very quickly to protect the turbine from further damage. When a valve suddenly stops at a very short time, the flow pressure builds up at the valve, starting to create pressure waves along the pipe runs which travel between elbows. Furthermore, these pressure waves may cause large dynamic response on the pipeline and large loads on the pipe restraints. The response and vibrations on the pipeline depend on the pressure waves amplitudes, frequencies, the natural frequencies and the dynamic characteristics of the pipeline itself. The piping flexibility or rigidity of the pipe line, determine how the pipes will respond to these waves and the magnitude of loads on the pipe supports. Consequently, the design of the piping system must consider the pipeline response to the steam hammer loads. In this paper, a design and analysis method is proposed to analyze the steam hammer in the critical hot lines due to the turbine trip using both PIPENET transient module and CAESAR II programs. The method offered in this paper aims to assist the design engineer in the power plant industry to perform dynamic analysis of the piping system considering the dynamic response of the system using the PIPENET and CAESAR II programs. Furthermore, the dynamic approach is validated with a static method by considering the appropriate dynamic load and transmissibility factors. A case study is analyzed for a typical hot reheat line in a power plant and the results of the transient analysis are validated using the theoretical static approach.

Evaluation of Potential to Air Ingestion From RWT Following RAS

2006 ◽  
Author(s):  
Young S. Bang ◽  
Ingoo Kim ◽  
Sweng W. Woo
Keyword(s):  
Water Level ◽  
Power Plants ◽  
Transient Flow ◽  
Cooling System ◽  
Check Valve ◽  
Transient Behavior ◽  
Piping System ◽  
Water Tank ◽  
Two Phase ◽  
Pipe Line

At the Recirculation Actuation Signal (RAS) when the Refueling Water Tank (RWT) water level decreased to a certain value following Loss-of-Coolant Accident (LOCA), the isolation valves of Containment Recirculation Sump (CRS) of the Korean Standard Nuclear Power Plants (KSNP) are open automatically while the RWT isolation valves would be closed manually. It was concerned whether the design has a potential to air ingestion to Emergency Core Cooling System (ECCS) pumps before completion of the manual action to close RWT isolation valves. To support the safety evaluation on this issue including the evaluation of design adequacy, an analysis of the hydraulic transient within the ECCS piping following the RAS in KSNP is performed. RELAP5/MOD3.3 code is used to calculate the transient behavior of the piping network. The code was known to have capability to calculate one-dimensional two-phase transient flow with noncondensible gas in the complex piping. Substantial portion of ECCS are modeled including RWT, CRS, each pipe line from RWT and CRS to connection point with its own isolation valve and check valve, a common pipe line to ECCS header, each pipe line from the header to High Pressure Safety Injection (HPSI) pump, Low Pressure Safety Injection (LPSI) pump, and Containment Spray (CS) pump. Transient hydraulic behavior in the piping system following RAS after LOCA is calculated. It is found that the RWT water level was always higher than the elevation of the check valve at the connecting point by more than 15 ft. It indicates the air intrusion to the check valve can be sufficiently prevented by this amount of water head.

Bull Run Fossil Plant Main Steam Piping Creep Evaluation

2007 ◽  
Author(s):  
Darryl A. Rosario ◽  
Blaine W. Roberts ◽  
M. Scott Turnbow ◽  
Salah E. Azzazy
Keyword(s):  
Power Plants ◽  
Structural Integrity ◽  
Piping System ◽  
Fossil Plant ◽  
Contributing Factors ◽  
Engineering Consulting ◽  
Commercial Operation ◽  
Burning Coal ◽  
Girth Welds ◽  
Main Steam

Bull Run Unit 1, rated at 950 MW, is the first of four fossil supercritical power plants at TVA; the unit went into commercial operation in 1967. The boiler, built by Combustion Engineering (CE), has a radiant reheat twin divided furnace with tangential-fired burners for burning coal. The unit’s maximum continuous rating (MCR) is 6,400,000 lbs/hr of main steam flow, with a design temperature of 1003°F and pressure of 3840 psig. Through the end of November 2003, the unit had a total of 589 cumulative starts and 253,343 operating hours. In 1986 TVA located and repaired extensive cracking in the mixing link headers (27 of 32 saddle welds cracked) downstream of the superheater outlet headers. Visible sag was also noted at the mid-span of the mixing headers. Since that time through 2003, additional cracking of girth welds in the mixing link headers was discovered, followed by cracking in the main piping girth welds at the connections to the mixing headers and at one of the connections to the turbine. From 1988 through 2003 several elastic analyses which were performed were unable to explain the observed girth weld cracking and sagging in the piping. In October 2003, TVA contracted with Structural Integrity Associates (SI) and BW Roberts Engineering Consulting to perform elastic and creep analyses of the Bull Run main steam piping system to determine the most likely contributing factors to noticeable creep sagging and cracking problems in the mixing header link piping and main steam piping girth welds, and, to develop recommendations to mitigate additional cracking and creep/sagging. The evaluations concluded that improper hanger sizing along with longer-term hanger operational problems (non-ideal loads/travel, topped/bottomed out hangers) contributed to the observable creep sagging and girth weld cracking. The elastic and creep piping analyses performed to address these issues are described in this paper.

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