Twenty Years of Operation in NPP Krsko

2002 ◽  
Author(s):  
Kresimir Nemcic ◽  
Robert Brems
Keyword(s):  
Water Chemistry ◽  
Steam Generator ◽  
Nuclear Power ◽  
Steam Generators ◽  
Main Design ◽  
Counter Measures ◽  
Before And After ◽  
Useful Life ◽  
Generator Performance ◽  
Generator Replacement

This paper presents the main design and fabrication improvements that were included in the new steam generators that were installed at Nuclear Power Plant Krsko in Slovenia in 2000. These improvements were a result of an on-going R & D effort associated with steam generator technology, aimed at increased reliability and better maintainability of new steam generators. The paper also provides basic information related to 20 years of operation of NPP Krsko with an emphasis on subjects related to steam generator performance and degradation. These include inservice inspection results on steam generators up to replacement, corrective actions, corrosion-erosion counter-measures, and replacement of condenser and moisture spearator reheaters with improved material. The paper also provides data regarding changes in feedwater and steam generator water chemistry, with water chemistry results before and after steam generator replacement. This paper shows that materials issues are very important for steam generator reliability, and it gives the reasons why specific materials were selected for replaced components. Finally, the paper demonstrates how the strategy used at NPP Krsko prolonged the useful life of the old steam generators until replacement steam generators could be designed, fabricated, and installed.

Steam Generator Replacement Project at TVA’s Sequoyah Unit 1 Nuclear Power Plant

2004 ◽  
Author(s):  
Myron R. Anderson
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Power ◽  
Original Design ◽  
Pressurized Water ◽  
Repair Work ◽  
Building Roof ◽  
Tennessee Valley ◽  
Generator Replacement

Pressurized Water Reactor Power Plants have at times required that large components be replaced (steam generators weighing 750,000 lbs) which have necessitated performing first time modifications to the plant that were unintended during the original design. The steam generator replacement project at Tennessee Valley Authority (TVA’s) Sequoyah Nuclear Power Station necessitated (1) two large temporary openings (21’×45’) in the plant’s Shield Building roof (2’ thick concrete) by hydro-blasting to allow the removal of the old generators and installation of the new, (2) removal and repair of the concrete steam generator enclosure roofs (20’ diameter, 3’ thick) which were removed by wire saw cutting and (3) the seismic qualification of; the design and construction of an extensive ring foundation for; the use of one of the world largest cranes to remove these components through the roof. This removal and replacement process had to be performed in an expeditious manner to minimize the amount of time the plant is shutdown so the plant could return to providing power to the grid. This paper will address some of the many technical and construction considerations required to perform this demolition and repair work safely, efficiently and in a short as possible duration.

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.

Fundamental Mechanisms of Component Degradation by Corrosion in Nuclear Power Plants of Russian Design

2008 ◽  
Author(s):  
G. Saji ◽  
V. A. Yurmanov ◽  
V. I. Baranenko ◽  
V. A. Fedorova ◽  
G. Karzov ◽  
...  
Keyword(s):  
Power Systems ◽  
Water Chemistry ◽  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Electrochemical Characteristics ◽  
Secondary System ◽  
Technological Basis ◽  
Hydrogen Water

By focusing on NPPs of Western design (e.g. PWR and BWR), the first author (G. Saji) has established that ‘long cell action’ corrosion plays a pivotal role in practically all unresolved corrosion issues for all types of nuclear power plants as presented in a series of papers already published (1–9). The authors believe that a similar study of NPPs of Russian design, with their unique scientific and technological basis compared to Western plants, are important to illustrate that this mechanism can occur even with different materials, welding technology or operation (e.g. water chemistry control). Among all the differences, it is important to note that PWSCC per se does not seem to be occurring in VVER plants, although no specific reason has yet to be identified. In this paper, a detailed electrochemical assessment is first made on the behavior of ammonia-potassium water chemistry and structural materials at the normal operational temperature in the primary water of VVERs. The chemical and electrochemical characteristics of the ammonia in VVERS were found to be significantly different from those of PWRs which use the hydrogen water chemistry. However, the water chemistry of RBMK is not fundamentally different from that of the Western BWR and therefore the previous studies on SCC of BWRs are generally applicable. On the bases of these studies, various corrosion issues commonly experienced in NPPs of Russian design (VVER and RBMK) are briefly reviewed. They include: (i) pitting corrosion in un-clad VVER-440 RV; (ii) corrosion cracking at the transition welding joints of RV nozzles and piping; (iii) corrosion issues in PGV-440 steam generator collectors; (iv) steam generator tube and collector corrosion; (v) IGSCC in RBMK with austenitic steel piping; (vi) FAC (E-C) in the secondary system of VVERs; and (vii) Anomalous corrosion products sedimentation in the core region in some VVERs. Since the long cell action hypothesis does not seem to contradict the various corrosion activities being experienced in NPPs of Russian design, the first author invites further study on the potential involvement of this mechanism since this hypothesis provides new insight into many of the unresolved corrosion issues. More specifically, the VVERs’ ammonia-potassium water chemistry has theoretically been identified as playing a key role in the prevention of PWSCC, which is one of the most troublesome mechanism of corrosion degradation in many Western PWRs. In view of this significance, the authors proposed an urgent international joint initiative to prove or disprove this mechanism’s existence in nuclear power systems.

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.

Assessment of a Steam Generator Circulation Ratio by a Thermal Balance Based on External Wall Temperature Measurements

2010 ◽  
Author(s):  
Geoffrey Deotto ◽  
Olivier Brunin ◽  
Alexandre Nicoli ◽  
Franck David
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Thermal Balance ◽  
Hot Water ◽  
Temperature Measurements ◽  
Circulation Loop ◽  
Hydraulic Pressure ◽  
Steam Generators ◽  
External Wall ◽  
Circulation Ratio

During operation, sludge steadily appears at a slow pace on the secondary side of nuclear power plant steam generators. This leads to clogging of the tube bundle support plates, and consequently to a change in the thermal-hydraulic flow conditions. The circulation ratio of a steam generator is defined as the ratio between the total flowrate circulating in the riser and the steam flowrate at the outlet of the steam generator. This is a good indicator of the hydraulic pressure losses in the circulation loop. In particular, the increase in hydraulic resistance due to the tube support plate clogging leads to a drop in this parameter. For this reason, in order to check that clogging does not reach too high a level, the circulation ratio is regularly evaluated on steam generators of French nuclear power plants, and then compared to established safety limits. The purpose of this paper is to present an accurate method to determine the circulation ratio of a steam generator based on temperature measurements taken around the wall of the steam generator. This method consists of carrying out a thermal balance of the flow circulating in the downcomer. In order to accomplish this, the temperature of the water circulating in the downcomer is evaluated using thermocouple belts put on the external wall of the appliance. However, additional hypotheses in the calculation method are considered in order to take into account for the heat transfer between hot water inside the downcomer and the sensors. The steam generator circulation loop and the clogging of the tube support plates are presented in §1. Then §2 and §3 describe in detail the method and the associated hypotheses as well as the required instrumentation. Finally, §4 presents an application of this method to real cases of clogged steam generators.

Some Consideration in the Design and Application of Heat Recovery Steam Generators

1969 ◽  
Author(s):  
E. C. Hunt
Keyword(s):  
Steam Generator ◽  
Heat Recovery ◽  
Control Design ◽  
Steam Generation ◽  
Steam Generators ◽  
Forced Circulation ◽  
Construction Methods ◽  
Parallel Control ◽  
Generator Performance ◽  
Design And Application

The characteristics of heat recovery steam generation are compared to fully fired steam generators. Methods for stating performance are discussed. A compact forced circulation design is presented in some detail with comments on possible arrangements, construction methods, materials, and the use of supplementary firing. The importance of parallel control design to the ultimate success of the steam generator performance and operation is presented.

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.

Evaluation of single heated channel and subchannel modeling of a nuclear once through steam generator (OTSG)

Kerntechnik ◽  
2020 ◽  
Vol 85 (1) ◽  
pp. 54-67
Author(s):  
A. Hamedani ◽  
O. Noori-Kalkhoran ◽  
R. Ahangari ◽  
M. Gei
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Axial Direction ◽  
Pressurized Water Reactors ◽  
Steam Generators ◽  
Pressurized Water ◽  
Heated Channel ◽  
Water Reactors ◽  
Secondary Side

Abstract Steam generators are one of the most important components of pressurized-water reactors. This component plays the role of heat transfer and pressure boundary between primary and secondary side fluids. The Once Through Steam Generator (OTSG) is an essential component of the integrated nuclear power system. In this paper, steady-state analysis of primary and secondary fluids in the Integral Economizer Once Through Steam Generator (IEOTSG) have been presented by Single Heated Channel (SHC) and subchannel modelling. Models have been programmed by MATLAB and FORTRAN. First, SHC model has been used for this purpose (changes are considered only in the axial direction in this model). Second, the subchannel approach that considers changes in the axial and also radial directions has been applied. Results have been compared with Babcock and Wilcox (B&W) 19- tube once through steam generator experimental data. Thermal- hydraulic profiles have been presented for steam generator using both of models. Accuracy and simplicity of SHC model and importance of localization of thermal-hydraulic profiles in subchannel approach have been proved.

Main Generator Life Cycle Management

2012 ◽  
Author(s):  
Arun Puri ◽  
John DiBiase
Keyword(s):  
Life Cycle ◽  
Nuclear Power ◽  
Life Cycle Management ◽  
Management Plan ◽  
Original Design ◽  
Reliable Operation ◽  
Counter Measures ◽  
Age Related ◽  
The Impact ◽  
Generator Replacement

The Ginna Nuclear Power Plant began commercial operation in 1970. Two power uprates have been completed during the 40+ years of operation and the original main generator continues to provide highly reliable operation with no significant changes to the original design. The life cycle management plan for the generator identified several age-related issues which could challenge reliable operation of the unit if left unattended. This paper outlines the approach taken to assess and quantify specific component degradation, the impact on generator reliability and the counter measures (planned and implemented) to ensure continued reliable operation until a generator replacement is performed in a future refueling outage.

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.

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