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.

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.

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.

scholarly journals Small nuclear power plants for power supply in arctic regions: assessment of spent nuclear fuel radioactivity

2018 ◽  
Vol 4 (2) ◽  
pp. 119-125
Author(s):  
Vadim Naumov ◽  
Sergey Gusak ◽  
Andrey Naumov
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fuel Cycle ◽  
Mass Composition ◽  
Published Data ◽  
Pressurized Water ◽  
Fuel Cycles ◽  
Reactor Cores

The purpose of the present study is the investigation of mass composition of long-lived radionuclides accumulated in the fuel cycle of small nuclear power plants (SNPP) as well as long-lived radioactivity of spent fuel of such reactors. Analysis was performed of the published data on the projects of SNPP with pressurized water-cooled reactors (LWR) and reactors cooled with Pb-Bi eutectics (SVBR). Information was obtained on the parameters of fuel cycle, design and materials of reactor cores, thermodynamic characteristics of coolants of the primary cooling circuit for reactor facilities of different types. Mathematical models of fuel cycles of the cores of reactors of ABV, KLT-40S, RITM-200M, UNITERM, SVBR-10 and SVBR-100 types were developed. The KRATER software was applied for mathematical modeling of the fuel cycles where spatial-energy distribution of neutron flux density is determined within multi-group diffusion approximation and heterogeneity of reactor cores is taken into account using albedo method within the reactor cell model. Calculation studies of kinetics of burnup of isotopes in the initial fuel load (235U, 238U) and accumulation of long-lived fission products (85Kr, 90Sr, 137Cs, 151Sm) and actinoids (238,239,240,241,242Pu, 236U, 237Np, 241Am, 244Cm) in the cores of the examined SNPP reactor facilities were performed. The obtained information allowed estimating radiation characteristics of irradiated nuclear fuel and implementing comparison of long-lived radioactivity of spent reactor fuel of the SNPPs under study and of their prototypes (nuclear propulsion reactors). The comparison performed allowed formulating the conclusion on the possibility in principle (from the viewpoint of radiation safety) of application of SNF handling technology used in prototype reactors in the transportation and technological process layouts of handling SNF of SNPP reactors.

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.

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.

Design Issues and Considerations During Reactor Vessel Head Replacement, Steam Generator Replacement, and Steam Generator Snubber Elimination at Nuclear Power Plants

2004 ◽  
Author(s):  
Dilip Bhavnani ◽  
James Annett
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Supply System ◽  
Reactor Vessel ◽  
Original Design ◽  
Pressurized Water ◽  
Class 1 ◽  
Current Design ◽  
Successful Design

One of the key maintenance activities in a nuclear power plant is the replacement of major components in the Nuclear Steam Supply System. In order to achieve significant operational improvements, the replacement components are not an exact replacement of the existing components. The replacement of components in the nuclear steam supply system in many Pressurized Water Reactor plants may include steam generators, replacement of reactor vessel heads with integrated head assemblies, and elimination of steam generator snubbers. The replacement components may not be supplied and/or designed by the original supplier. The changes in the components have to be compared to a plant’s current design and licensing bases and regulatory commitments. The qualification of these components involves non-linear, Nuclear Class 1 analyses, where portions of the configuration and analyses are proprietary, and there is a coupling of the response between the containment structure and the components. Ultimately, the qualification of the reactor coolant system and reactor vessel internals must be demonstrated, not just the qualification of the replacement components. A key element for the successful completion of these component replacements is the method by which the design and licensing bases is maintained and the work of the various groups involved in the design coordinated. This paper outlines how in a typical two unit PWR plant, major component replacements can impact original design bases and issues that should be considered in creating successful design and configuration documents. Design interface issues, configuration combinations, and coordination requirements are identified.

Parametric Study on Determining Optimal Eddy Current Testing Method for Inspection of Steam Generator Tubes in SMART Plant

2010 ◽  
Author(s):  
Dae-Kwang Kim ◽  
Sung-Jin Han ◽  
Hak-Joon Kim ◽  
Sung-Jin Song ◽  
Yun-hang Choung
Keyword(s):  
Steam Generator ◽  
Eddy Current ◽  
Nuclear Power ◽  
Power Plants ◽  
Eddy Current Testing ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Current Testing ◽  
Current Test ◽  
Testing Condition

The SMART (System-integrated Modular Advanced ReacTor) is small sized integral type pressurized water reactor designed by KAERI (Korea Atomic Energy Research Institute), Korea. But, shape of steam generator (SG) in SMART plant differs from those in operated nuclear power plants (NPPs). Especially, SG tubes in SAMRT plant is helical type with around 600 mm of innermost diameter and thickness of 2.5 mm which is thicker than general NPPs one. For providing integrity of SG tube in SMART plant, new types of ECT method are needed because eddy current testing (ECT) is one of widely adopted method for inspection of SG tubes in NPPs. Therefore, in this study, we investigate optimal conditions or parameters for detecting and evaluating of flaws in the SG tubes in SMART plant by simulation of ECT signals with various testing condition or parameter such as frequency, coil gap and etc. From the simulated ECT signals optimal eddy current test condition or parameters are proposed.

A Modeling Method of the Branch Probability of the Temperature-Induced SGTR Phenomenon

2017 ◽  
Author(s):  
Zibin Liu ◽  
Dingqing Guo ◽  
Bing Zhang ◽  
Jinkai Wang
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Modeling Method ◽  
Radioactive Material ◽  
Severe Accident ◽  
Steam Generator Tube ◽  
Modeling Methods ◽  
Radioactive Release ◽  
Level 2

The phenomenon of temperature-induced steam generator tube rupture (TI-SGTR) is a typical phenomenon in the severe accident process of nuclear power plants. The occurrence of the phenomenon may result in the radioactive material bypass the containment, causing a large radioactive release. This paper investigates modeling methods of the phenomenon of temperature-induced SGTR in level 2 PSA and presents an optimizing modeling method to calculate the probability of branching probability of TI-SGTR, aiming at improving the rationality and veracity of level 2 PSA.

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.

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