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.

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 Reactor Vessel Internals Segmentation Experience using Mechanical Cutting Tools

2013 ◽  
Vol 10 (2) ◽  
pp. 6-10 ◽  
Author(s):  
Petr Pospíšil
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cutting Tools ◽  
Reactor Vessel ◽  
Boiling Water Reactors ◽  
Abrasive Waterjet ◽  
Pressurized Water ◽  
Water Reactors ◽  
Mechanical Cutting ◽  
Reactor Internals

Abstract Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Mechanical cutting has been used by Westinghouse since 1999 for both Pressurized Water Reactors (PWR’s) and Boiling Water Reactors (BWR’s) and its process has been continuously improved over the years. The complexity of the work requires well designed and reliable tools. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. Westinghouse experience in mechanical cutting has demonstrated that it is an excellent technique for segmentation of internals. In summary, the purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date.

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.

Development of a Thermal-Hydraulic Analysis Code for Helical Coiled Once Through Steam Generator

2018 ◽  
Author(s):  
Jun Huang ◽  
Junli Gou ◽  
Haifu Ma ◽  
Jie Fan ◽  
Jianqiang Shan
Keyword(s):  
Heat Transfer ◽  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
High Efficiency ◽  
Fluid Model ◽  
Computer Code ◽  
Medium Size ◽  
Hydraulic Characteristics ◽  
Pressurized Water

Due to their advantages, such as compactness and high efficiency in heat transfer, helically coiled heat exchangers have been widely used by different type of nuclear power plants, especially by small and medium size reactors (SMRs). In order to analyze the thermal-hydraulic characteristics of a helical coiled once through steam generator (OTSG) for a small integral pressurized water reactor, a computer code is developed in this paper. The code is based on two-fluid model. The constitutive correlations are recommended based on the assessments with the compiled databases from the reviewed literatures. NUSOL SG is validated and verified against heat transfer in helical coiled tubes, and the calculation results agree well with the experiment data. The present study could provide references for the investigators to perform further investigations on the thermal hydraulic characteristics of helical coiled OTSGs.

Assessment of Remaining Life for Class 1 Piping Components Experiencing Wall Thinning

2004 ◽  
Author(s):  
Rajnish Kumar
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Life Extension ◽  
Remaining Life ◽  
Original Design ◽  
Accelerated Corrosion ◽  
Wall Thinning ◽  
Class 1 ◽  
Plant Components

Assessment of remaining life of power plant components is important in light of plant life management and life extension studies. This information helps in planning and minimizing plant outages for repairs and refurbishments. Such studies are specifically important for nuclear power plants. Nuclear Safety Solutions Limited (NSS) is involved in conducting such studies for plant operators and utilities. Thickness measurements of certain piping components carrying fluids at high temperature and high pressure have indicated higher than anticipated wall thinning rates. Flow accelerated corrosion (FAC) has been identified as the primary mechanism for this degradation. The effect of FAC was generally not accounted for in the original design of the plants. Carbon steel piping components such as elbows, tees and reducers are prone to FAC. In such cases, it is important to establish the remaining life of the components and assess their adequacy for continued service. Section XI of the ASME Boiler and Pressure Vessel Code is applicable for evaluation of nuclear power plant components in service. This Section of the Code does not specifically deal with wall thinning of the piping components. Code Case N-597 provides guidelines for evaluation for continued service for Class 2 and Class 3 piping components. For Class 1 piping components, this Code Case suggests that the plant owner should develop the methodology and criteria for evaluation. This paper presents methodology and procedure for establishing the remaining life and assessment of Class 1 piping components experiencing wall thinning effects. In this paper, the rules of NB-3600 and NB-3220 and Code Case N-597 have been utilized for assessment of the components for continued service. Details of various considerations, criteria and methodology for assessment of the remaining life and adequacy for continued service are provided.

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.

Comparison of Analytical Methods to Study the Impact of a Postulated Closure Head Drop Accident on the Structural Integrity of the Bottom-Mounted Instrumentation System

2010 ◽  
Author(s):  
William C. Castillo ◽  
Geoffrey M. Loy ◽  
Joseph M. Remic ◽  
David P. Molitoris ◽  
George J. Demetri ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
The United States ◽  
Reactor Vessel ◽  
Primary Coolant ◽  
Heavy Load ◽  
Pressurized Water ◽  
System Pressure ◽  
The Impact

During typical nuclear power plant refueling activities for a pressurized water reactor (PWR), the reactor vessel closure head assembly must be removed from the reactor vessel (RV), transported for storage, and returned to the RV after refueling. This is categorized as a critical heavy load lift in NUREG-0612 [1] because a drop accident could result in damage to the components required to cool the fuel in the RV core. In order to mitigate the potentially severe consequences of a closure head drop, the United States Nuclear Regulatory Commission (USNRC) has mandated that nuclear power plants upgrade to a single failure-proof crane, show single failure-proof crane equivalence, or perform a head drop analysis to demonstrate that the core remains covered with coolant and sufficient cooling is available after the head drop accident. The primary coolant-retaining components associated with the RV are the inlet and outlet nozzles and the hot and cold leg main loop piping. Typical head drop analyses have considered these components to ensure that their structural integrity is maintained. One coolant-retaining component that has not been included in head drop evaluations on a consistent basis is the bottom-mounted instrumentation (BMI) system. In a typical Westinghouse PWR, 50 to 60 BMI nozzles are connected through the bottom hemisphere of the RV to one-inch diameter guide tubes which run under the vessel to a seal table above. Failure of the BMI system has the potential to adversely affect core coolability, especially if multiple failures are postulated within the system. A study was performed to compare static and dynamic methods of analyzing the effects of a head drop accident on the structural integrity of the BMI system. This paper presents the results of that study and assesses the adequacy of each method. Acceptability of the BMI system pressure boundary is based on the Nuclear Energy Institute Initiative (NEI 08–05 [2]) criteria for coolant-retaining components, which are based on Section III, Appendix F of the ASME Code [3].

Measurement of Steam Generator or Reactor Vessel Moisture Carryover Using a Non-Radioactive Chemical Tracer

2009 ◽  
Author(s):  
Robert Fournier ◽  
Michael Thibodeau ◽  
Charles T. French
Keyword(s):  
Moisture Content ◽  
Steam Generator ◽  
Power Plants ◽  
Power Level ◽  
Thermal Power ◽  
Reactor Vessel ◽  
Saturated Steam ◽  
Power Calculation ◽  
Pressurized Water ◽  
Chemical Tracer

All nuclear power plants are licensed to operate at a specific thermal power. Verification of this power level is done by measuring the change in the enthalpy of the feedwater times the mass flow of the feedwater through each steam generator for pressurized water reactor or through the reactor vessel for boiling water reactors. When performing this measurement, it is necessary to determine the enthalpy of the steam exiting each steam generator or the reactor vessel. Since the exit enthalpy of most units is saturated steam, the state of the steam must be defined in terms of the enthalpy of dry stream minus the fraction of moisture being carried over with the steam times the heat of vaporization. The quality of the steam can be measured using a throttling calorimeter, but a much more accurate means of measurement is with the use of a chemical tracer salt that is only carried over with the main steam by droplets of water. Thus, the fraction of moisture in the steam can be defined as the ratio of concentration of the tracer in the steam divided by concentration of the tracer in the riser of the generator or reactor vessel. If the moisture content of the steam is not measured, one must assume that it is zero, which is conservative, since it provides an upper limit for the thermal power calculation. However, from an operational standpoint, if moisture carryover is present but is not accounted for, the thermal power of the unit will be below its licensed power level causing a loss in electrical output due both a lower thermal power and a reduction in turbine cycle efficiency. Moreover, if there is significant carryover, there is the potential for additional damage to the turbine blading and housing from the impingement of the water droplets. This paper describes how a chemical tracer can be used to measure the moisture content of the steam, considerations for the injection and sample points, plant alignment recommendations and the support required from a utility, when performing this test. The paper will also discuss the potential for long term degradation, the effects of extended power uprates on carryover and the limitations of chemistry guidelines.

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