Structural Integrity Role in Plant Life Extension - A Case Study

2015 ◽  
Vol 750 ◽  
pp. 295-306
Author(s):  
Jin Hua Shi
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Operating Conditions ◽  
Life Extension ◽  
Guide Tube ◽  
Gas Temperature ◽  
Steam Generators ◽  
Steam Generator Tube ◽  
The Uk

The steam generators at Advanced Gas-Cooled Reactor (AGR) nuclear power stations in the UK are potentially life-limiting components. Enhancing the capability to monitor the steam generators has been identified as having the potential to provide key evidence in justifying the extension of the generating lifetime of the stations. It has been proposed to install new temperature measuring instrumentation to monitor reactor gas temperature and to provide additional data regarding steam generator operating conditions. The modification will be to introduce thermocouples to the bore of an intact steam generator tube to facilitate temperature measurement at or near to the locations of interest. The modified steam generator tube will be sealed at the feed header upstand. Between the upper surface of the superheater header tubeplate and the wall of the superheater header, the thermocouple bundle and sheath will be contained within a rigid stainless steel guide tube. The guide tube will be attached at both ends by welds, each forming a pressure boundary. At the tubeplate a weld will separate the bore of the sealed guide tube from the steam space within the superheater header; a weld between the guide tube and the superheater header will separate the steam space within the superheater header from atmosphere outside the header. In order to obtain a better design, three 3-dimentional finite element models have been created using ABAQUS. A series of cyclic pressure, and start-up and shutdown thermal transient stress analyses have been carried out to provide stress values for structural integrity assessments to be conducted using ASME III, Subsection NH and R5.

Experimental Investigation of Subcooled Choking Flow in a Steam Generator Tube Crack

2016 ◽  
Author(s):  
Hung Nguyen ◽  
Mark Brown ◽  
Shripad T. Revankar ◽  
Jovica Riznic
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Test Facility ◽  
Experimental Program ◽  
Steam Generators ◽  
Steam Generator Tube ◽  
Loss Of Coolant ◽  
Steam Generator Tubes

Steam generator tubes have a history of small cracks and even ruptures, which lead to a loss of coolant from the primary side to the secondary side. These tubes have an important role in reactor safety since they serve as one of the barriers between radioactive and non-radioactive materials of a nuclear power plant. A rupture then signifies the loss of the integrity of the tube itself. Therefore, choking flow plays an integral part not only in the engineered safeguards of a nuclear power plant, but also to everyday operation. There is limited data on actual steam generators tube wall cracks. Here experiments were conducted on choked flow of subcooled water through two samples of axial cracks of steam generator tubes taken from US PWR steam generators. The purpose of the experimental program was to develop database on critical flow through actual steam generator tube cracks with subcooled liquid flow at the entrance. The knowledge of this maximum flow rate through a crack in the steam generator tubes of a pressurized water nuclear reactor will allow designers to calculate leak rates and design inventory levels accordingly while limiting losses during loss of coolant accidents. The test facility design is modular so that various steam generator tube cracks can be studied. Two sets of PWR steam generators tubes were studied whose wall thickness is 1.285 mm. Tests were carried out at stagnation pressure up to 6.89 MPa and range of subcoolings 16.2–59°C. Based on these new choking flow data, the applicability of analytical models to highlight the importance of non-equilibrium effects was examined.

Crack Initiation Factor by Impact Fretting Wear at INCONEL Alloy for Steam Generator Tube in Nuclear Power Plants

2007 ◽  
Vol 26-28 ◽  
pp. 1269-1272
Author(s):  
Chi Yong Park ◽  
Jeong Kun Kim ◽  
Tae Ryong Kim ◽  
Sun Young Cho ◽  
Hyun Ik Jeon
Keyword(s):  
Steam Generator ◽  
Wear Mechanism ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Conditions ◽  
Fretting Wear ◽  
Steam Generator Tube ◽  
Inconel Alloy ◽  
The Impact

Inconel alloy such as alloy 600 and alloy 690 is widely used as the steam generator tube materials in the nuclear power plants. The impact fretting wear tests were performed to investigate wear mechanism between tube alloy and 409 stainless steel tube support plates in the simulated steam generator operating conditions, pressure of 15MPa, high temperature water of 290°C and low dissolved oxygen(<10 ppb). From investigation of wear test specimens by the SEM and EDS analysis, hammer imprint, which is known to be an actual damaged wear pattern, has been observed on the worn surface, and fretting wear mechanism was investigated. Wear progression of impact-fretting wear also has been examined. It was observed that titanium rich phase contributes to the formation of voids and cracks in sub-layer of fretting wear damage by impact fretting wear.

scholarly journals Evolved Fuzzy Control System for a Steam Generator

2010 ◽  
Vol 5 (2) ◽  
pp. 179 ◽  
Author(s):  
Daniela Hossu ◽  
Ioana Făgărășan ◽  
Andrei Hossu ◽  
Sergiu St. Iliescu
Keyword(s):  
Control System ◽  
Water Level ◽  
Steam Generator ◽  
Measurement Errors ◽  
Nuclear Power ◽  
Power Plants ◽  
Fuzzy Model ◽  
Operating Conditions ◽  
Steam Generators ◽  
Control Approach

Poor control of steam generator water level is the main cause of unexpected shutdowns in nuclear power plants. Particularly at low powers, it is a difficult task due to shrink and swell phenomena and flow measurement errors. In addition, the steam generator is a highly complex, nonlinear and time-varying system and its parameters vary with operating conditions. Therefore, there is a need to systematically investigate the problem of controlling the water level in the steam generator in order to prevent such costly reactor shutdowns. The objective of this paper is to design, evaluate and implement a water level controller for steam generators based on a fuzzy model predictive control approach. An original concept of modular evolved control system, seamless and with gradual integration into the existent control system is proposed as base of implementation of the presented system.

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.

Structural Integrity Assessment of Steam Generator Tubes Using a New EPRI Statistical Approach

2002 ◽  
Author(s):  
Carlos Alexandre de Jesus Miranda ◽  
Miguel Mattar Neto
Keyword(s):  
Steam Generator ◽  
Wall Thickness ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Tube Wall ◽  
New Approach ◽  
Inconel 600 ◽  
Integrity Assessment ◽  
Tube Wall Thickness ◽  
Uniform Tube

A fundamental step in tube plugging management of a Steam Generator (SG), in a Nuclear Power Plant (NPP), is the tube structural integrity evaluation. The degradation of SG tubes may be considered one of the most serious problems found in PWRs operation, mainly when the tube material is the Inconel 600. The first repair criterion was based on the degradation mode where a uniform tube wall thickness corrosion thinning occurred. Thus, a requirement of a maximum depth of 40% of the tube wall thickness was imposed for any type of tube damage. A new approach considers different defects arising from different degradation modes, which comes from the in-service inspections (NDE) and how to consider the involved uncertainties. It is based on experimental results, using statistics to consider the involved uncertainties, to assess structural limits of PWR SG tubes. In any case, the obtained results, critical defect dimensions, are within the regulatory limits. In this paper this new approach will be discussed and it will be applied to two cases (two defects) using typical data of SG tubes of one Westinghouse NPP. The obtained results are compared with ‘historical’ approaches and some comments are addressed from the results and their comparison.

U.S. Operating Experience With Thermally Treated Alloy 600 Steam Generator Tubes

2002 ◽  
Author(s):  
April Smith ◽  
Kenneth J. Karwoski
Keyword(s):  
Steam Generator ◽  
Operating Experience ◽  
Alloy 600 ◽  
Steam Generators ◽  
Pressurized Water ◽  
Steam Generator Tube ◽  
Water Reactors ◽  
1960S And 1970S ◽  
The 1960S ◽  
Thermally Treated

Steam generators placed in service in the 1960s and 1970s were primarily fabricated from mill-annealed Alloy 600. Over time, this material proved to be susceptible to stress corrosion cracking in the highly pure primary and secondary water chemistry environments of pressurized-water reactors. The corrosion ultimately led to the replacement of steam generators at numerous facilities, the first U.S. replacement occurring in 1980. Many of the steam generators placed into service in the 1980s used tubes fabricated from thermally treated Alloy 600. This tube material was thought to be less susceptible to corrosion. Because of the safety significance of steam generator tube integrity, this paper evaluates the operating experience of thermally treated Alloy 600 by looking at the extent to which it is used and recent results from steam generator tube examinations.

UK Technical Advisory Group on the Structural Integrity of High Integrity Plant: Overview of Recent Activities

2009 ◽  
Author(s):  
F. M. Burdekin ◽  
R. A. Ainsworth ◽  
D. P. G. Lidbury
Keyword(s):  
Recent Work ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Chemical Plant ◽  
Advisory Group ◽  
Technical Work ◽  
Technical Advisory Group ◽  
High Integrity ◽  
The Uk

TAGSI is an acronym for The UK Technical Advisory Group on the Structural Integrity of High Integrity Plant. The purpose of TAGSI is to address and make recommendations on issues raised by its sponsoring organisations in discussion with independent advisors. Collectively, these sponsoring organisations operate, maintain, assess and regulate nuclear power producing and chemical plant. Against this background, an overview is presented of the recent work of TAGSI in terms of structural integrity issues addressed, symposia organised, and output appearing in the open literature. Other papers at this conference describe some of the technical work of TAGSI in more detail.

scholarly journals Russian and foreign steam generators for NPP power units with wet steam turbines

2020 ◽  
Vol 178 ◽  
pp. 01007
Author(s):  
Mikle Egorov ◽  
Ivan Kasatkin ◽  
Ivan Kovalenko ◽  
Irina Krectunova ◽  
Nataliya Lavrovskaya ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Temperature Drop ◽  
Heating Surface ◽  
Vessel Diameter ◽  
Absorption Efficiency ◽  
Steam Turbines ◽  
Steam Generators ◽  
Horizontal Type ◽  
Generator Design

The main aim of the current study is to analyze advantages and shortcomings of horizontal and vertical types of steam generator design. Design solutions and experience of operation of steam generators of horizontal type accepted in Russia and of vertical type applied by Westinghouse, Combustion Engineering, Siemens, Mitsubishi, Doosan were analyzed within the framework of the present study. It was established that steam generator equipment of horizontal type is characterized by disadvantages of design, technological and operational nature. Thus, horizontal steam generators with dimensions permissible for railroad transportation and, for VVER-1200 with reactor vessel diameter equal to 5 m, by water transport as well, have exhausted the possibilities for further significant increase of the per unit electric power. The demonstrated advantages of vertical-type steam generators are as follows: 1) absence of stagnant zones within the second cooling circuit; 2) uniformity of heat absorption efficiency of the heating surface that ensures improved conditions for moisture separation; 3) increased temperature drop with parameters of generated steam elevated by 0.3 – 0.4 MPa. Conclusion was made on the advisability of introduction of steam generators with vertical-type layout in the Russian nuclear power generation.

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