Predictive methods for assessing corrosion damage to BWR piping and PWR steam generators =

BWR haikan oyobi PWR jokihasseiki ni okeru fushoku sonsho to yosoku

Publisher: National Association of Corrosion Engineers in Houston, Tex

Written in English
Published: Pages: 393 Downloads: 749
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Subjects:

  • Boiling water reactors -- Corrosion -- Congresses.,
  • Pressurized water reactors -- Corrosion -- Congresses.
  • Edition Notes

    Other titlesBWR haikan oyobi PWR jokihasseiki ni okeru fushoku sonsho to yosoku.
    Statementco-editors : Hideya Okada, Roger Staehle.
    ContributionsOkada, Hideya., Staehle, R. W., 1934-, National Association of Corrosion Engineers., Electric Power Research Institute.
    Classifications
    LC ClassificationsTK9202 .P743 1978, TK9202 .P743 1982
    The Physical Object
    Paginationxxix, 393, 6 p. :
    Number of Pages393
    ID Numbers
    Open LibraryOL18128692M

(e.g., PWR steam generator tube leaks, BWR recirculation pipe cracking, BWR reactor vessel internals issues, PWR reactor pressure vessel head penetration cracking and leaks). Problems associated with actual degradation pose reliability, regulatory and in some cases safety concerns, and as plants. Abstract. Flow-assisted corrosion (FAC) often has caused serious damage to carbon steel piping in nuclear power plants. As a first stage of experiments to determine the effects of water chemistry parameters on FAC, corrosion rates of carbon steel were measured in °C pure water with an online corrosion rate monitor based on electrical resistance measurement as [O 2] and flow velocity were. The Electric Power Research Institute (EPRI) conducts research, development, and demonstration projects for the benefit of the public in the United States and internationally. As an independent, nonprofit organization for public interest energy and environmental research, we focus on electricity generation, delivery, and use in collaboration with the electricity sector, its stakeholders and. 9 PWR and BWR chemistry control limits and methods e Stress corrosion cracking mitigation methods NEI Guideline Applies to all programs involving primary system materials. - Defines expectations for management of " Replaced steam generators =

  Assessment of NDE Methods on Inspection of HDPE Butt Fusion Piping Joints for Lack of Fusion NUREG/CR Stress Corrosion Cracking in Nickel-Base Alloys and Weld in Simulated PWR Environment - NUCLEAR SECTOR ROADMAPS JANUARY United States Nuclear Regulatory Commission Official Hearing Exhibit In the Matter of: Entergy Nuclear Operations, Inc. s (e.g., PWR steam generator tube leaks, BWR recirculation pipe cracking, BWR reactor vessel internals issues, PWR reactor pressure vessel head penetration cracking and leaks). Problems associated with actual degradation pose reliability, regulatory and in some cases safety concerns, and as plants age and move into. Corrosion Engineering Assignments. The recent Corrosion Engineering publication by McGraw-Hill was designed as a textbook to accompany the education of undergraduates, graduates and other technical trainees during their discovery of the modern world of corrosion engineering. The book contains many equations and principles that lend themselves to the production of assignments and questions that.

  The boiling water reactor (BWR) is a type of light water nuclear reactor used for the generation of electrical power. Lower risk (probability) of a rupture causing loss of coolant compared to a PWR, and lower risk of core damage should such a rupture occur. This is due to fewer pipes, fewer large diameter pipes, fewer welds and no steam. Leaks in your plant’s buried piping are a serious concern – and the risks only grow as equipment ages. Structural Integrity’s multi-disciplined, turnkey offer for buried piping grew out of NEI and is underpinned by our MAPPro TM software, a risk quantification and data management/visualization tool, currently being used by over 50% of US nuclear plant sites. A boiling water reactor (BWR) uses demineralized water as a coolant and neutron is produced by nuclear fission in the reactor core, and this causes the cooling water to boil, producing steam. The steam is directly used to drive a turbine, after which it is cooled in a condenser and converted back to liquid water. This water is then returned to the reactor core, completing the loop. This book offers solutions to problems in water-cooled nuclear power plants of today and the future, assists in the understanding of materials presents new insights into materials, methods, and techniques from an international,multidiscliplinary community.

Predictive methods for assessing corrosion damage to BWR piping and PWR steam generators = Download PDF EPUB FB2

Predictive methods for assessing corrosion damage to BWR piping and PWR steam generators. Houston, Tex.: National Association of Corrosion Engineers, © (OCoLC) Material Type: Conference publication: Document Type: Book: All Authors / Contributors: Hideya Okada; R W Staehle; Electric Power Research Institute.

Figure Average scores of the panel members for SCC, corrosion fatigue and toughness of PWR pressure vessel steels. 42 Figure Schematic variation of “damage. J.A. Gorman, in Steam Generators for Nuclear Power Plants, Abstract.

This chapter reviews the main corrosion problems that have affected tubes in commercial PWR steam generators. For each of these main corrosion problems the types of steam generators that.

Topics: Pipes, Pressure pipes, Failure mechanisms, ASME Standards, Compression, Fatigue, Fittings, Mechanical testing, Piping systems, Rupture High Temperature Codes and Standards Elimination of BWR Mark I Program Primary Containment Drywell-to-Wetwell Differential Pressure.

Speidel, M. O., “Overview of Methods for Corrosion Testing as Related to PWR Steam Generator and BWR Piping Problems,” in “Predictive Methods for Assessing Corrosion Damage to BWR Piping and PWR Steam Generators” (Eds.

Okada and R. Staehle), NACE,pp. 31–44 Google ScholarCited by: 2. Predictive Methods for Assessing Corrosion Damage to BWR Piping and PWR for 24 hours, which was assumed to simulate a long-term Steam Generators operation at ЊC.

Components (BWR), Predictive Methods of Assessing Corrosion Damage to BWR Piping and PWR Steam Generators, NACE Publication, Houston, pp – Zucchi, F. and Angelini, E. ( Corrosion Control and Lay‐up of the Crystal River‐3 Steam Generators and Secondary Plant during an Extended Outage (Pages: ) Rocky H.

Thompson W. (Bill) Kassen. Iwasaki, in Predictive Methods for Assessing Corrosion Damage to BWR Piping and PWR Steam Generators, H. Okada and R. Staehle (Eds.), NACE, Houston, TX,p. Intergranular stress corrosion cracking behavior of types and stainless steel weld metals in a simulated boiling water reactor environment.

Predictive Methods for Assessing Corrosion Damage to BWR Piping and PWR Steam Generators, NACE, Houston, TX,pp. –   CANDU steam generators have some small differences compared to PWR (pressurized water reactor) designs, and typically operate at lower temperatures, but with similar secondary side chemistries, but the small differences are mostly related to the steam generator tubing, and in the use of an integral pre-heater, and this will be discussed further.

Eds.), Predictive Methods for Assessing Corrosion Damage to BWR Piping and PWR steam Generators, NACE, Houston,pp. – Predictive Methods for Assessing Corrosion Damage to BWR Piping.

Weight loss method was adopted to assess the corrosion effect on the material. This research discovered that Ammonia/water solution does not completely stop corrosion but has significant effect to. (95)Q Get rights and content. Flow-Accelerated Corrosion (FAC) is a phenomenon that results in metal loss from piping, vessels, and equipment made of carbon steel.

FAC occurs only under certain conditions of flow, chemistry, geometry, and material. Unfortunately, those conditions are common in much of the high-energy piping in nuclear and fossil-fueled power plants. The main steam piping at fittings and discontinuities is susceptible to erosion-corrosion damage.

assessment methods are discussed for seven major light water reactor components: pressurized water reactor (PWR) and boiling water reactor (BWR) pressure vessels, PWR containment structures, PWR reactor coolant piping, PWR steam generators, BWR.

Outer Diameter Stress Corrosion Cracking (ODSCC) and IGA have been the main damage affecting the tube bundle of steam generators worldwide. Until the early s, ODSCC was the primary cause for SG tube plugging, literally plaguing the components, as can be seen on the graph below, from reference (Diercks et al., ).

Type weld metal overlaid on a low-alloy steel by a submerged arc welding (SAW) process using wide strip electrodes can become sensitized by not only postweld heat treatments (PWHTs) at around °C but also low-temperature aging (LTA) at °C for 24 hours after PWHT. A slow strain-rate test (SSRT) in oxygenated, high-temperature pure water revealed that type SAW weld metals after.

Some of the components that have experienced mechanical damage are reactor coolant pump shafts, PWR and BWR reactor vessel internals, PWR instrument tubes, thermal sleeves in piping, and steam generator tubes.

Various mitigation methods can be implemented to reduce or eliminate these problems. 18 refs., 5 figs., 1 tab. Visual Examination. Some visual examination is also suggested to evaluate localized corrosion like pitting or rmore, optical or scanning electron microscopes; elemental and compositional analyses such as energy dispersive X-ray spectroscopy (EDX); X-ray diffraction (XRD); and energy dispersive X-ray spectroscopy (XPS) are useful techniques to evaluate the corroded.

The accumulation of corrosion products on the tube bundle, tube-support structure, and tubesheet of nuclear steam generators (SGs) can severely degrade the SG thermal performance, increase the risk of material degradation by providing sites for localized corrosion, and degrade SG operation by increasing the hydraulic resistance to flow on the secondary side.

A predictive model for pitting corrosion in buried pipelines is proposed. The model takes into consideration the chemical and physical properties of the soil and pipe to predict the time dependence of pitting depth and rate.

Maximum pit depths were collected together with soil and pipe data at more than excavation sites over a three-year. of NPP. RCS cooling system pipes include hot leg and cold leg pipes and steam generator tubes for PWR plants and steam, feed water and recirculation pipes for BWR plants, all of which are critical for the overall safety of the reactors.

Stress corrosion cracking is a major issue for RCS. Following the ex-Surry steam generator decontamination with CAN-DECON and LOMI in November (ref.

6), OZOX-A has been used on 2 steam generators at Millstone-2 (ref. 7), CAN-DECON on 6 steam generators at 3 plants, CITROX on 2 steam generators, again at Millstone-2, and LOMI on 4 steam generators at Indian Point-3 (table 3).

Abstract. Pitting corrosion is treated as a time-dependent stochastic damage process characterized by an exponential or logarithmic pit growth.

Data from water injection pipeline systems and from the published literature are used to simulate the sample functions of pit growth on metal surfaces. PWR exposures remain high by comparison with international experience (Figs. 2&3) but exposures in newer U.S. plants are more comparable to equivalent plants in other countries.

BWR radiation exposures have been dominated by the major piping repair and replacement programs made necessary by intergranular stress corrosion cracking.

Time-dependent degradation of the boiling water reactor (BWR) primary coolant piping, recirculation piping main feedwater piping, and steam lines has a significant impact on the safe operation and license renewal of aging plants. This paper discusses the dominant degradation mechanisms active in these piping systems.

Environmentally Assisted Cracking: Predictive Methods for Risk Assessment and Evaluation of Materials, Equipment, and Structures (ASTM Special Technical Publication, ) Russell D.

Kane. This comprehensive volume features international developments on the generation of relevant materials and properties data based on laboratory tests, and.

Thus, effective decontamination methods are desirable, and may become mandatory as the nuclear plant matures. Decontamination of BWR and PWR Primary Surfaces 4.

Activated corrosion products represent the dominant radiation sources on primary surfaces. The principal radionuclides are Mn, Cr, Fe, Co and Co   This article reviews the types of steam generators, the materials used for their construction (with a special focus on tubing alloys), the corrosion processes that may occur under operation and the strategies to preserve steam generators and other components of the secondary cooling system from corrosion during plant outages.

Corrosion is a form of damage that has accompanied mankind since the very introduction of metals thousands of years ago.

Corrosion is often insidious and hidden until striking at the worst moment of a system operation. While there are many ways to try and prevent such damage, the optimum control method relies on an early diagnosis of the [email protected]{osti_, title = {Review of industry efforts to manage pressurized water reactor feedwater nozzle, piping, and feedring cracking and wall thinning}, author = {Shah, V N and Ware, A G and Porter, A M}, abstractNote = {This report presents a review of nuclear industry efforts to manage thermal fatigue, flow-accelerated corrosion, and water hammer damage to pressurized water reactor.piping is estimated using corrosion rate, present thickness and retirement thickness of the piping.

Key Words: Process piping, Corrosion rate, Retirement thickness, remaining life estimation, API standards, ASME standards. 1. INTRODUCTION The transportation of oil and gas from the reservoir to.