The Effect of Transient Operation on Gasket Contact Pressure in Steam Generator Primary Side Obround Manway

2009 ◽  
Author(s):  
Nick Idvorian ◽  
Nansheng Sun
Keyword(s):  
Contact Pressure ◽  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Complex Interaction ◽  
Time Varying ◽  
Transient Operation ◽  
Steam Generators ◽  
Start Up ◽  
Leak Tightness

The Steam Generators used in CANDU Nuclear Power Plants employ a primary-side obround manway design with two cover plates on each side of the opening. During transient operation of the units (heat up, start up, operation, shut down and cool down) a complex interaction between the manway components occurs, which not only causes time-varying gasket contact pressure but also induces a non-uniform gasket contact pressure along the gasket width. As a result, the leak tightness of the joint may be affected. The effect of different stud preloads on the gasket contact pressure during transient operation is investigated by the 3D FE Analysis of a primary-side obround manway of CANDU’s steam generator. The results are used to justify the seal performance of the manway structure under conditions of different stud pretension loads.

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 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.

Comparison of Design Requirements for Nuclear Steam Generators in the ASME and RCC-M Codes

2013 ◽  
Author(s):  
Junjie Cao ◽  
Peter Caple
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Design Criteria ◽  
Steam Generators ◽  
Class 1 ◽  
Asme Code ◽  
Historic Development ◽  
Design Requirements ◽  
Plant Components

As two major international codes for class 1 nuclear power plant components, both ASME BPVC Section III and RCC-M Code have been applied to the design of steam generators for nuclear power plants. With respect to the design requirements for steam generators, there are many similarities between the ASME and RCC-M rules as a result of the historic development of the RCC-M Code which evolved from a foundation based on the ASME Code. However, differences do exist in the two codes. In this paper, detailed requirements related to steam generator pressure boundary design in ASME Subsection NB are compared with corresponding requirements in RCC-M Subsection B to identify the differences between the two codes. Supplemental requirements to the ASME Code requirements are proposed for steam generator pressure boundary design to meet the design criteria in both codes. This approach may be also applied to other Class 1 components.

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.

scholarly journals A new method for determining allowable medium temperature during transient operation of thick-walled elements in a supercritical power plant

2010 ◽  
Vol 31 (3) ◽  
pp. 55-72
Author(s):  
Piotr Duda ◽  
Dariusz Rząsa
Keyword(s):  
Power Plant ◽  
Power Plants ◽  
New Method ◽  
Optimum Operating ◽  
Transient Operation ◽  
Medium Temperature ◽  
Heating And Cooling ◽  
Optimum Parameters ◽  
Start Up ◽  
Working Parameters

A new method for determining allowable medium temperature during transient operation of thick-walled elements in a supercritical power plantConstruction elements of supercritical power plants are subjected to high working pressures and high temperatures while operating. Under these conditions high stresses in the construction are created. In order to operate safely, it is important to monitor stresses, especially during start-up and shut-down processes. The maximum stresses in the construction elements should not exceed the allowable stress limit. The goal is to find optimum operating parameters that can assure safe heating and cooling processes [1-5]. The optimum parameters should guarantee that the allowable stresses are not exceeded and the entire process is conducted in the shortest time. In this work new numerical method for determining optimum working parameters is presented. Based on these parameters heating operations were conducted. Stresses were monitored during the entire processes. The results obtained were compared with the German boiler regulations - Technische Regeln für Dampfkessel 301.

scholarly journals A Conceptual Approach to Eliminate Bypass Release of Fission Products by In-Containment Relief Valve under SGTR Accident

2018 ◽  
Vol 2018 ◽  
pp. 1-12
Author(s):  
Taeseok Kim ◽  
Wonjun Choi ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Hoichul Jung ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Base Case ◽  
Conceptual Approach ◽  
Relief Valve ◽  
Radioactive Nuclides

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.

scholarly journals Gas Turbine Heat Recovery Steam Generators for Combined Cycles Natural or Forced Circulation Considerations

1988 ◽  
Author(s):  
Akber Pasha
Keyword(s):  
Gas Turbine ◽  
Steam Generator ◽  
Heat Recovery ◽  
Power Plants ◽  
Natural Circulation ◽  
Combined Cycle ◽  
Heat Recovery Steam Generator ◽  
Steam Generators ◽  
Forced Circulation ◽  
Gas Turbine Exhaust

In recent years the combined cycle has become a very attractive power plant arrangement because of its high cycle efficiency, short order-to-on-line time and flexibility in the sizing when compared to conventional steam power plants. However, optimization of the cycle and selection of combined cycle equipment has become more complex because the three major components, Gas Turbine, Heat Recovery Steam Generator and Steam Turbine, are often designed and built by different manufacturers. Heat Recovery Steam Generators are classified into two major categories — 1) Natural Circulation and 2) Forced Circulation. Both circulation designs have certain advantages, disadvantages and limitations. This paper analyzes various factors including; availability, start-up, gas turbine exhaust conditions, reliability, space requirements, etc., which are affected by the type of circulation and which in turn affect the design, price and performance of the Heat Recovery Steam Generator. Modern trends around the world are discussed and conclusions are drawn as to the best type of circulation for a Heat Recovery Steam Generator for combined cycle application.

The Geometry of Steam Generator Tube and Its Relevance to the Occurrence of Stress Corrosion Cracking in Operating Nuclear Power Plants

2010 ◽  
Author(s):  
Deok Hyun Lee ◽  
Do Haeng Hur ◽  
Myung Sik Choi ◽  
Kyung Mo Kim ◽  
Jung Ho Han ◽  
...  
Keyword(s):  
Stress Corrosion ◽  
Steam Generator ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Corrosion Cracking ◽  
Location Type ◽  
Steam Generator Tubes ◽  
Non Destructive

Occurrences of a stress corrosion cracking in the steam generator tubes of operating nuclear power plants are closely related to the residual stress existing in the local region of a geometric change, that is, expansion transition, u-bend, ding, dent, bulge, etc. Therefore, information on the location, type and quantitative size of a geometric anomaly existing in a tube is a prerequisite to the activity of a non destructive inspection for an alert detection of an earlier crack and the prediction of a further crack evolution [1].

Crack Growth Prediction and Leakage Potential of Steam Generator Tubes Subjected to Flow Induced Vibrations

2014 ◽  
Author(s):  
Salim El Bouzidi ◽  
Marwan Hassan ◽  
Jovica Riznic
Keyword(s):  
Crack Growth ◽  
Steam Generator ◽  
Power Plants ◽  
Mechanical Energy ◽  
Steam Generators ◽  
Two Phase ◽  
Element Analysis ◽  
Wide Range ◽  
Flow Induced Vibrations ◽  
Crack Growth Model

Nuclear steam generators are critical components of nuclear power plants. Flow-Induced Vibrations (FIV) are a major threat to the operation of nuclear steam generators. The two main manifestations of FIV in heat exchangers are turbulence and fluidelastic instability, which would add mechanical energy to the system resulting in great levels of vibrations. The consequences on the operation of steam generators are premature wear of the tubes, as well as development of cracks that may leak radioactive heavy water. This paper investigates the effect of tube support clearance on crack propagation. A crack growth model is used to simulate the growth of Surface Flaws and Through-Wall Cracks of various initial sizes due to a wide range of support clearances. Leakage rates are predicted using a two-phase flow leakage model. Non-linear finite element analysis is used to simulate a full U-bend subjected to fluidelastic and turbulence forces. Monte Carlo Simulations are then used to conduct a probabilistic assessment of steam generator life due to crack development.

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.

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