Crevice Boiling in Steam Generators

1987 ◽  
Vol 109 (3) ◽  
pp. 761-767 ◽  
Author(s):  
S. Tieszen ◽  
H. Merte ◽  
V. S. Arpaci ◽  
S. Selamoglu
Keyword(s):  
Power Plants ◽  
Heated Surface ◽  
Nucleate Boiling ◽  
Forced Flow ◽  
High Speed Photography ◽  
Flow Conditions ◽  
Steam Generators ◽  
Pressurized Water ◽  
Support Plate ◽  
Primary Fluid

Experimental results are presented on the influence of confinement (normal to heated surface) on nucleate boiling in forced flow. The forced flow conditions and confinement geometry studied are similar to those found for boiling between a primary-fluid tube and a tube-support plate in steam generators of pressurized-water-reactor nuclear power plants. Visual observations of the boiling process within the confined region (crevice) between the tube and its support plate, obtained by high-speed photography, are related to simultaneous two-dimensional temperature maps of the hot primary-fluid-tube surface. The results demonstrate the existence of three confinement-dependent boiling regimes in forced flow conditions that are similar to those found in pool boiling conditions. These regimes are shown to be associated with the Weber number.

scholarly journals Calculation of Potential Radiation Doses Associated with Predisposal Management of Dismantled Steam Generators from Nuclear Power Plants

2020 ◽  
Vol 12 (12) ◽  
pp. 5149
Author(s):  
Ga Hyun Chun ◽  
Jin-ho Park ◽  
Jae Hak Cheong
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Radiation Doses ◽  
Steam Generators ◽  
Pressurized Water ◽  
Water Reactor ◽  
General Pathway ◽  
Heavy Water Reactor ◽  
Past Experiences

Although the generation of large components from nuclear power plants is expected to gradually increase in the future, comprehensive studies on the radiological risks of the predisposal management of large components have been rarely reported in open literature. With a view to generalizing the assessment framework for the radiological risks of the processing and transport of a representative large component—a steam generator—12 scenarios were modeled in this study based on past experiences and practices. In addition, the general pathway dose factors normalized to the unit activity concentration of radionuclides for processing and transportation were derived. Using the general pathway dose factors, as derived using the approach established in this study, a specific assessment was conducted for steam generators from a pressurized water reactor (PWR) or a pressurized heavy water reactor (PHWR) in Korea. In order to demonstrate the applicability of the developed approach, radiation doses reported from actual experiences and studies are compared to the calculated values in this study. The applicability of special arrangement transportation of steam generators assumed in this study is evaluated in accordance with international guidance. The generalized approach to assessing the radiation doses can be used to support optimizing the predisposal management of large components in terms of radiological risk.

Age-Related Degradation of Steam Generator Internals Based on Industry Responses to Generic Letter 97-06

2002 ◽  
Author(s):  
M. Subudhi ◽  
E. J. Sullivan
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressurized Water Reactor ◽  
Long Term Effects ◽  
Steam Generators ◽  
Pressurized Water ◽  
Age Related ◽  
The U.S

This paper presents the results of an aging assessment of the nuclear power industry’s responses to NRC Generic Letter 97-06 on the degradation of steam generator internals experienced at Electricite de France (EdF) plants in France and at a United States pressurized water reactor (PWR). Westinghouse (W), Combustion Engineering (CE), and Babcock & Wilcox (B & W) steam generator models, currently in service at U.S. nuclear power plants, potentially could experience degradation similar to that found at EdF plants and the U.S. plant. The steam generators in many of the U.S. PWRs have been replaced with steam generators with improved designs and materials. These replacement steam generators have been manufactured in the U.S. and abroad. During this assessment, each of the three owners groups (W, CE, and B&W) identified for its steam generator models all the potential internal components that are vulnerable to degradation while in service. Each owners group developed inspection and monitoring guidance and recommendations for its particular steam generator models. The Nuclear Energy Institute incorporated in NEI 97-06, “Steam Generator Program Guidelines,” a requirement to monitor secondary side steam generator components if their failure could prevent the steam generator from fulfilling its intended safety-related function. Licensees indicated that they implemented or planned to implement, as appropriate for their steam generators, their owners group recommendations to address the long-term effects of the potential degradation mechanisms associated with the steam generator internals.

Numerical analysis of bubble condensation behavior under high-pressure flow conditions

2020 ◽  
Vol 234 (18) ◽  
pp. 3725-3741
Author(s):  
SalaiSargunan S Paramanantham ◽  
Thanh-Hoang Phan ◽  
Warn-Gyu Park
Keyword(s):  
High Pressure ◽  
Flow Boiling ◽  
Nucleate Boiling ◽  
Sensitivity Study ◽  
Transfer Model ◽  
Condensation Rate ◽  
Flow Conditions ◽  
Two Phase ◽  
Pressurized Water ◽  
Water Reactors

Heat transfer during subcooled flow boiling has a pivotal role in pressurized water reactors; it also occurs in boiling water reactors prior to the onset of saturated nucleate boiling. We examined the condensation behavior of vapor bubbles in the subcooled liquid phase using the fully compressible two-phase homogeneous mixture method, solved by an implicit dual-time preconditioned method. The continuous surface force method was applied to determine the surface tension between the phases in the simulation. To predict the empirical coefficient, we conducted a sensitivity study using Lee’s mass transfer model. For nuclear applications, we simulated high-pressure vapor–water conditions under higher mass flow conditions. The comparison of the numerical simulation and experimental results showed that the proposed model accurately predicted the condensation behavior of the bubble. Additionally, we investigated single bubble condensation behavior at different operating pressures, subcooling temperatures, bubble diameters, and bulk velocities. We also investigated the effects of high-pressure condensation on bubble shape. At lower subcooling temperatures, the condensation rate increased as pressure increased; however, at higher subcooling temperatures, pressure had no significant impact on the condensation rate.

Dynamics of a Water Droplet on the Heated Surface of Nano- and Micro-Structures

2012 ◽  
Author(s):  
Seol Ha Kim ◽  
Ho Seon Ahn ◽  
Joonwon Kim ◽  
Moo Hwan Kim
Keyword(s):  
Water Droplet ◽  
High Speed ◽  
Zirconium Alloy ◽  
Liquid Droplet ◽  
Heated Surface ◽  
Nucleate Boiling ◽  
High Speed Photography ◽  
Complete Wetting ◽  
Structured Surfaces ◽  
Bare Surface

In this study, we investigated the dynamic behavior of a water droplet near the Leidenfrost point (LFP) of bare and modified zirconium alloy surfaces with bundles of nanotubes (∼10–100 nm) or micro mountain-like structures using high-speed photography. A deionized water droplet (6 μL) was dropped onto the sample surfaces (20 × 25 × 0.7 mm) that were heated to temperatures ranging from 250°C to the LFP. The modified zirconium alloy surfaces showed complete wetting and well-spread features at room temperature due to strong liquid spreading by the structure. The meniscus of the liquid droplet on the structured surface experienced more vigorous dynamics with intensive nucleate boiling, compared with the clean, bare surface. The cutback phenomenon was observed on the bare surface; however, the structured surfaces showed a water droplet “burst”. We observed that the LFPs were 449°C, 522°C, and 570°C, corresponding to the bare, micro-, and nano-structures, respectively.

PWR Steam Generator Tube and AVB Wear Under Perpendicular Impacting

2013 ◽  
Author(s):  
Christian Phalippou ◽  
Franck Ruffet ◽  
Emmanuel Herms ◽  
François Balestreri
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Published Data ◽  
Steam Generators ◽  
Dynamic Interactions ◽  
Pressurized Water ◽  
Steam Generator Tube ◽  
Wear Rates ◽  
Alloy 690

Flow-induced vibrations of steam generator tubes in nuclear power plants may result in wear damage at support locations. The steam generators in EPR power plants have a design life of 60 years; as wear is an identified ageing damage in steam generators, it is therefore important to collect experimental results on wear of tube and support due to dynamic interactions at EPR secondary side temperature. In this study, wear tests were performed between a steam generator tube (Alloy 690) and two flat opposite anti-vibration bars (AVB in 410s stainless steel) at different impact force levels. Tests were performed in pressurized water at 290°C in wear machines for long term repeated predominant impact motions. The worn surfaces were observed by SEM, the wear coefficients of tube and AVB were evaluated using the work rate approach. Significant scoring, due to the importance of sliding when impacts occur, was shown on wear scar patterns. There were greater wear volumes and depths on tubes than on AVBs, but dynamic forced conditions and rigid mounting of AVB in the rigs have prevailed for finally getting an upper bound of the wear rates. Alloy 690 for tubes and 410s for AVB remain a satisfactory material combination considering comparative wear results with other published data.

CFD Analysis of Subcooled Wall Boiling at Shell Side of Steam Generator With TSP

2014 ◽  
Author(s):  
Chenglong Wang ◽  
Dalin Zhang ◽  
Suizheng Qiu ◽  
Wenxi Tian ◽  
Guanghui Su
Keyword(s):  
Heat Transfer ◽  
Numerical Simulation ◽  
Steam Generator ◽  
Flow Boiling ◽  
Nucleate Boiling ◽  
Two Phase ◽  
Pressurized Water ◽  
Support Plate ◽  
Coupled Heat Transfer ◽  
Secondary Side

This paper addresses the numerical simulation of two-phase flow heat transfer among the tube bundles with tube support plate (TSP) of an integral type pressurized water reactor steam generator using RPI wall boiling model. The subcooled nucleate boiling phenomenon and the coupled heat transfer between the SG primary side and secondary side were obtained. Also, the effects of tube support plate (TSP) and the different inlet subcooling on the thermal-hydraulic characteristics of SG were studied. From the results of the present numerical simulation, it reasonably revealed the subcooled flow boiling occurred in the SG secondary side and the distributions of key parameters around TSP, elucidating that this model can provide useful information to the design of the steam generator.

An Experimental Investigation Into Potential Nuclear Reactor Vessel Flow Instabilities

2010 ◽  
Author(s):  
Jeffrey A. Brown ◽  
James W. Rowland ◽  
H. Joseph Fernando
Keyword(s):  
Physical Model ◽  
Transient Flow ◽  
Plant Protection ◽  
Reactor Vessel ◽  
Flow Instabilities ◽  
Engineering System ◽  
Flow Conditions ◽  
Steam Generators ◽  
Pressurized Water ◽  
Flow Measurements

An investigation into the increase in Plant Protection System (PPS) alarms at a three-unit US Pressurized Water Reactor (PWR) plant has determined that the alarms are the result, in part, of a hydraulic instability that has developed within the Reactor Coolant System (RCS) following the replacement of the steam generators in all three units of the Palo Verde Nuclear Generating Station (PVNGS). An experimental effort has been established by Arizona Public Service Company and Arizona State University in an attempt to determine the cause of these instabilities. Preliminary investigations have determined that the time scale of these instabilities is consistent with larger scale transient flow processes of the reactor vessel. Accordingly, the flow characteristics were assessed and localized flow measurements made using a one-fifth scale physical model of the upper plenum region of the reactor core of the Combustion Engineering System 80 reactor vessel to verify the postulation that large vortex structures referred to as “precessing” vortices [Ref. 1] affect the core exit flow conditions resulting in the noted flow instabilities. The physical model investigation was complemented by numerical analysis based on a Computational Fluid Dynamics (CFD) code performed for the same geometry. Benchmarking of the CFD model by the scaled physical model is intended to provide increased confidence in the CFD code. If verified, the CFD code may be modified so as to establish corrective actions for this condition, where physical modeling would probably be time consuming and cost prohibitive. The initial results for the physical and computational models demonstrate very good agreement between the measured and calculated flows in the upper-plenum region. The results of the complementary experimental and analytic evaluations do not support the presence of any large scale vortices of appropriate space scales that could affect flow conditions within the upper-plenum region. The elimination of the reactor vessel as the source of the instabilities suggests that the replacement steam generators may be the root cause of the flow instabilities. There is a possibility, however, that frequencies pertinent to vortices may be triggering mechanisms for flow instabilities in the entire system.

Role of Boiling Mode and Rate in Formation of Waterside Deposits in Heat Recovery Steam Generator Evaporator Tubes

2014 ◽  
Author(s):  
David S. Moelling ◽  
James Malloy
Keyword(s):  
Steam Generator ◽  
Power Plants ◽  
Nucleate Boiling ◽  
Combined Cycle ◽  
Steam Generators ◽  
Deposit Formation ◽  
Convective Boiling ◽  
Oil Gas ◽  
Steam Generator Tubes

Waterside Deposits in evaporator tubes have been an issue in steam generators as long as boilers have been used. Substantial experience in deposit formation and management has been gained in conventional goal and oil/gas boilers over time. The role of boiling modes in the steam generator tubes is very critical to areas of deposit formation. Incipient boiling, nucleate boiling and convective boiling modes all have different deposition behavior. When Gas Turbine Combined Cycle (GTCC) power plants of larger size (> 100 MW) began operation in the 1990’s, deposits in evaporator tubes were not considered a significant issue. Operating boiler pressures were low (500–900 psig) as were flue gas temperatures, use of supplemental firing was limited. Other than known problems with feedwater contamination such as operation with leaking seawater-cooled condensers, deposits were not found to be forming. The rapid increase in size and operating pressures in HRSG’s raised the likelihood of waterside deposits developing. Both Vertical and Horizontal Gas Path HRSG designs are considered. Drawing on field observations, the morphology and location of HRSG deposits are reviewed, as are changes in deposit formation with the mode and rate of boiling.

scholarly journals New method to assess tube support plate clogging phenomena in steam generators of nuclear power plants

2010 ◽  
Vol 16 (3) ◽  
pp. 257-267 ◽  
Author(s):  
D. Bouskela ◽  
V. Chip ◽  
B. El Hefni ◽  
J.M. Favennec ◽  
M. Midou ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
New Method ◽  
Steam Generators ◽  
Support Plate

Effects of Gap Geometry and Gravity on Boiling Around a Constrained Bubble in 2-Propanol/Water Mixtures

2006 ◽  
Vol 129 (2) ◽  
pp. 114-123
Author(s):  
Chen-li Sun ◽  
Van P. Carey
Keyword(s):  
Heat Flux ◽  
Heated Surface ◽  
Molar Fraction ◽  
Nucleate Boiling ◽  
Confined Space ◽  
Cold Surface ◽  
Boiling Regime ◽  
Wall Superheat ◽  
Boiling Process ◽  
Lower Heat

In this study, boiling experiments were conducted with 2-propanol/water mixtures in confined gap geometry under various levels of gravity. The temperature field created within the parallel plate gap resulted in evaporation over the portion of the vapor-liquid interface of the bubble near the heated surface, and condensation near the cold surface. Full boiling curves were obtained and two boiling regimes—nucleate boiling and pseudofilm boiling—and the transition condition, the critical heat flux (CHF), were identified. The observations indicated that the presence of the gap geometry pushed the nucleate boiling regime to a lower superheated temperature range, resulting in correspondingly lower heat flux. With further increases of wall superheat, the vapor generated by the boiling process was trapped in the gap to blanket the heated surface. This caused premature occurrence of CHF conditions and deterioration of heat transfer in the pseudo-film boiling regime. The influence of the confined space was particularly significant when greater Marangoni forces were present under reduced gravity conditions. The CHF value of x (molar fraction)=0.025, which corresponded to weaker Marangoni forces, was found to be greater than that of x=0.015 with a 6.4mm gap.

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