Initiation of PWSCC in Nickel Base Alloy 182

2013 ◽  
Author(s):  
Mickaël Wehbi ◽  
Jérôme Crépin ◽  
Thierry Couvant ◽  
Cécilie Duhamel
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Base Alloy ◽  
Nickel Base Alloy ◽  
Pressurized Water ◽  
Material Environment ◽  
Primary Water ◽  
Water Reactors ◽  
Nickel Base ◽  
Actual Initiation

To date, welded nickel base alloy 182 used in Pressurized Water Reactors (PWRs) components have shown a higher susceptibility to Primary Water Stress Corrosion Cracking (PWSCC) during laboratory tests than in power plants. However, an increasing number of cracks reported in American, Swedish and Japanese nuclear power plants on Alloy 182 enlighten the need for a predictive initiation model of PWSCC. Initiation of PWSCC involves several factors such as material, environment and loading history, interacting with each other. Building such a model first requires to focus on these parameters separately, in order to have a better understanding of the involved mechanisms at a local scale in crack initiation. This study focuses on the correlation between EBSD/strain field results to improve the accuracy of the actual initiation model [1] involving local parameters.

Review of Thermo-Mechanical Testing and Micro-Characterization Studies of Ductility-Dip Cracking in Nickel-Base Alloy 52 Weld Filler Metal

2005 ◽  
Author(s):  
Steve McCracken
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Mechanical Testing ◽  
Nuclear Power ◽  
Filler Metal ◽  
Base Alloy ◽  
Nickel Base Alloy ◽  
Primary Water ◽  
Characterization Studies ◽  
Nickel Base

Nickel-base Alloy 690 wrought material and Alloy 52 (ERNiCrFe-7) weld filler metal have in recent years become the material of choice in new fabrication and repair of commercial nuclear power plant components. Alloys 690 and 52 are preferred due to improved resistance to primary water stress corrosion cracking (PWSCC) as compared to other nickel-base alloys or filler metals, such as Alloy 600 and Alloy 82 (ERNiCr-3). Nickel-base alloys are commonly used in dissimilar metal joints between quenched and tempered low alloy steel and austenitic stainless steel components in nuclear power primary water systems. Nuclear power industry experience with Alloy 52 filler metal using manual or machine gas tungsten arc welding (GTAW) in multi-pass welds and highly restrained thick section welds has been troublesome. Ultrasonic and radiographic examination of Alloy 52 welds has, in some cases, revealed multiple subsurface micro-cracks in the weld metal heat affected zone (HAZ). Recent laboratory thermo-mechanical testing using modified varestraint test methods and a newly developed Gleeble-based strain-to-fracture test method indicate nickel-base alloys are susceptible to a ductility-dip cracking (DDC) phenomenon during the on cooling cycle of welding. Thermo-mechanical testing also demonstrates that initiation of DDC is dependent on exceeding a specific strain threshold or strain rate in susceptible alloys. Microanalysis and micro-characterization studies indicate that DDC is a solid-state thermo-mechanical phenomenon that occurs most commonly along migrated grain boundaries of single-phase austenitic stainless steel and nickel-base alloys. Though not fully understood, DDC is believed to initiate in the temperature range where material ductility drops concurrent with high shrinkage strains during the on cooling weld cycle. This paper reviews the most current thermo-mechanical laboratory test results and micro-characterization studies of nickel-base alloys for susceptibility to DDC. Alloy 52 weld filler metal is discussed in detail due to its importance to the nuclear power industry. Finally, welding parameters and specific filler metal chemistry to reduce potential for DDC are presented and information for evaluation of specific heats of Alloy 52 for susceptibility to DDC are discussed.

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.

scholarly journals Plant Performance History of an Innovative Gate Valve in Critical Service Applications

2017 ◽  
Author(s):  
M. S. Kalsi ◽  
Patricio Alvarez ◽  
Thomas White ◽  
Micheal Green
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Gate Valve ◽  
Plant Performance ◽  
Flow Loop ◽  
Performance History ◽  
Pressurized Water ◽  
Maintenance Costs ◽  
History Of ◽  
Water Reactors

A previous paper [1] describes the key features of an innovative gate valve design that was developed to overcome seat leakage problems, high maintenance costs as well as issues identified in the Nuclear Regulatory Commission (NRC) Generic Letters 89-10, 95-07 and 96-05 with conventional gate valves [2,3,4]. The earlier paper was published within a year after the new design valves were installed at the Pilgrim Nuclear Plant — the plant that took the initiative to form a teaming arrangement as described in [1] which facilitated this innovative development. The current paper documents the successful performance history of 22 years at the Pilgrim plant, as well as performance history at several other nuclear power plants where these valves have been installed for many years in containment isolation service that requires operation under pipe rupture conditions and require tight shut-off in both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The performance history of the new valve has shown to provide significant performance advantage by eliminating the chronic leakage problems and high maintenance costs in these critical service applications. This paper includes a summary of the design, analysis and separate effects testing described in detail in the earlier paper. Flow loop testing was performed on these valves under normal plant operation, various thermal binding and pressure locking scenarios, and accident/pipe rupture conditions. The valve was designed, analyzed and tested to satisfy the requirements of ANSI B16.41 [9]; it also satisfies the requirements of ASME QME 1-2012 [10]. The results of the long-term performance history including any degradation observed and its root cause are summarized in the paper. Paper published with permission.

Power Maximization of Fully Passive Lead-Bismuth Eutectic Cooled Small Modular Reactor

2014 ◽  
Author(s):  
Jaehyun Cho ◽  
Yong-Hoon Shin ◽  
Il Soon Hwang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Natural Circulation ◽  
Reactor Core ◽  
Design Stage ◽  
Pressurized Water ◽  
Small Modular Reactor ◽  
Power Maximization ◽  
Lead Bismuth Eutectic ◽  
Water Reactors

Although the current Pressurized Water Reactors (PWRs) have significantly contributed to the global energy supply, PWRs have not been considered as a trustworthy energy solution owing to its several problems; spent nuclear fuels (SNFs), nuclear safety, and nuclear economy. In order to overcome these problems, lead-bismuth eutectic (LBE) fully passive cooling Small Modular Reactor (SMR) system is suggested. It is possible to not only provide the solution of the problem of SNFs through the transmutation feature of LBE coolant, but also increase the safety and economy through the concepts of the natural circulation cooling SMRs. It is necessary to maximize the advantages (safety and economy) of this type of Nuclear Power Plants for several applications in future. Accordingly, objective of the study is to maximize the reactor core power while the limitations of shipping size, materials endurance, long-burning criticality as well as safety under Beyond Design Basis Events must be satisfied. Design limitations of natural circulating LBE-cooling SMRs are researched and power maximization method is developed based on obtained design limitations. It is expected that the results are contributed to reactor design stage with providing several insights to designers as well as the methods for design optimization of other type of SMRs.

Background on Low Temperature Overpressure Protection System Setpoints for Pressure-Temperature Curves

2020 ◽  
Author(s):  
Jeffrey C. Poehler ◽  
Gary L. Stevens ◽  
Anees A. Udyawar ◽  
Amy Freed
Keyword(s):  
Low Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pressure Vessels ◽  
Pressurized Water Reactors ◽  
Documentation System ◽  
Pressurized Water ◽  
System Pressure ◽  
Water Reactors

Abstract ASME Code, Section XI, Nonmandatory Appendix G (ASME-G) provides a methodology for determining pressure and temperature (P-T) limits to prevent non-ductile failure of nuclear reactor pressure vessels (RPVs). Low-Temperature Overpressure Protection (LTOP) refers to systems in nuclear power plants that are designed to prevent inadvertent challenges to the established P-T limits due to operational events such as unexpected mass or temperature additions to the reactor coolant system (RCS). These systems were generally added to commercial nuclear power plants in the 1970s and 1980s to address regulatory concerns related to LTOP events. LTOP systems typically limit the allowable system pressure to below a certain value during plant operation below the LTOP system enabling temperature. Major overpressurization of the RCS, if combined with a critical size crack, could result in a brittle failure of the RPV. Failure of the RPV could make it impossible to provide adequate coolant to the reactor core and result in a major core damage or core melt accident. This issue affected the design and operation of all pressurized water reactors (PWRs). This paper provides a description of an investigation and technical evaluation regarding LTOP setpoints that was performed to review the basis of ASME-G, Paragraph G-2215, “Allowable Pressure,” which includes provisions to address pressure and temperature limitations in the development of P-T curves that incorporate LTOP limits. First, high-level summaries of the LTOP issue and its resolution are provided. LTOP was a significant issue for pressurized water reactors (PWRs) starting in the 1970s, and there are many reports available within the U.S. Nuclear Regulatory Commission’s (NRC’s) documentation system for this topic, including Information Notices, Generic Letters, and NUREGs. Second, a particular aspect of LTOP as related to ASME-G requirements for LTOP is discussed. Lastly, a basis is provided to update Appendix G-2215 to state that LTOP setpoints are based on isothermal (steady-state) conditions. This paper was developed as part of a larger effort to document the technical bases behind ASME-G.

Thermal Aging Assessment and Microstructural Investigations of Alloy 52 Dissimilar Metal Welds for Nuclear Components

2019 ◽  
Author(s):  
Miguel Yescas ◽  
Pierre Joly ◽  
François Roch
Keyword(s):  
Fracture Toughness ◽  
Alloy Steel ◽  
Thermal Aging ◽  
Base Alloy ◽  
Narrow Gap ◽  
Low Alloy Steel ◽  
Nickel Base Alloy ◽  
Pressurized Water ◽  
Dissimilar Metal ◽  
Nickel Base

Abstract Dissimilar Metal Welds (DMW) are commonly found between the ferritic low alloy steel heavy section components and the austenitic stainless steel piping sections in nuclear power plants. In the EPR™ design which is the latest FRAMATOME Pressurized water reactor (PWR) these DMW involve a narrow gap technology with no buttering, and only one bead per layer of a nickel base alloy weld filler metal (Alloy 52). In order to assess the thermal aging performance of this relatively new narrow gap DMW design, a significant internal R&D program was launched some years ago. Several representative mock-ups were thoroughly characterized in the initial condition as well as in the thermal aged condition, up to 50,000 hours aging at 350°C. The characterisations were focused on the fusion line between the ferritic low alloy steel (LAS) and the nickel base alloy since a particular microstructure is present in this area, especially in the carbon depleted area of the Heat Affected Zone (HAZ) which is often regarded as the weak zone of the weld joint. Metallography, hardness, nanohardness, chemical analyses, and Atom Probe Tomography, as well as fracture toughness tests were carried out on different specimens in different thermal aging conditions. The results show that the fracture toughness behaviour in the ductile-brittle domain of the low alloy steel carbon depleted HAZ at the interface with the alloy 52 weld metal of the DMWs is excellent, even for a thermal ageing equivalent to 60 years at service temperature. This was found in spite of the carbon depleted zone of the HAZ, the variations of hardness, chemical composition, particularly the carbon gradients, and the thermal aging effect induced by phosphorous segregation at grain boundaries.

Coordinated Research Projects at IAEA Concerning RPV and Structural Integrity

2006 ◽  
Author(s):  
William Server ◽  
Timothy Hardin ◽  
Milan Brumovsky´
Keyword(s):  
Ferritic Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Current Status ◽  
Research Projects ◽  
Energy Agency ◽  
Pressurized Water ◽  
Toughness Testing ◽  
Water Reactors

The International Atomic Energy Agency (IAEA) has had a series of reactor pressure vessel (RPV) structural integrity programs that started back in the 1970s. These Coordinated Research Projects most recently have focused on use of the Master Curve fracture toughness testing approach for RPV and other ferritic steel components and on the issue of pressurized thermal shock (PTS) in operating pressurized water reactors. This paper will provide the current status for these projects and discuss the implications for improved safety of key ferritic steel components in nuclear power plants (NPPs).

The French Regulatory Experience and View on Nickel-Base Alloy PWSCC Prevention and Treatment

2004 ◽  
Author(s):  
David Emond ◽  
Jacques Reuchet
Keyword(s):  
Steam Generator ◽  
Base Alloy ◽  
Hydraulic Testing ◽  
Primary Circuit ◽  
Nickel Base Alloy ◽  
Detection Systems ◽  
Initiation And Propagation ◽  
Primary Water ◽  
Nickel Base

This paper presents the experience feedback and views of the French Regulatory Authority (ASN) and of the technical support institute (IRSN) on PWSCC prevention since the initiation in 1989 of the “Inconel Zones Review” requested by ASN to Electricite´ de France (EDF), the national operator of 58 PWRs plants. This proactive requirement, launched before the discovery, in September 1991, of the only CRDM nozzle leak in France, on Bugey unit 3, was then triggered by the recurrence of many alloy 600 rapid degradations and leaks, world wide, and also in France in the late 1980s, particularly on steam generator tubes and on some pressurizer penetrations. Thus, the ASN requested that EDF, perform a comprehensive (generic) proactive assessement on all the nickel-base alloy components and parts of the main primary circuits, which of course included vessel head penetrations and bottom mounted instrumentation penetrations (BMI), and some other zones. This proactive “review” did, a minima, include the following tasks and actions: • Update and complete, by an extensive R&D program, the understanding and characterization of the Ni base alloys prone to PWSCC, • Analyze the various materials, metallurgical features, mechanical stresses, and physicochemical conditions of the parts exposed to primary water, in order to predict the occurrence of PWSCC initiation and propagation, • Provide a prioritization of the zones to be inspected, • Implement by improved NDE techniques a practical inspection program on the 58 PWRs, Prepare and implement any needed mitigation actions as a result of the components conditions assessment. The present paper relates the main features of the French regulatory experience over more than 13 years and recalls the main principles of the assessment, which were applied by ASN. These principles, which are formalized in the current regulation rules revised in 1999, are briefly listed hereunder: • It is based on avoiding and preventing any leaking on the main primary circuit. • In service inspections (ISI), including volumetric and surface NDE, have been agreed upon between ASN and EDF for all vessel head penetrations, with a re-inspection schedule. • The preexisting regulatory hydraulic testing program was carefully implemented, which implied the removal of thermal insulation on the vessel heads. • A comprehensive R&D program had to be conducted by EDF, the main progress reports and presentations had to be regularly submitted to DGSNR and IRSN staff. • The assessment and the ranking of the sensitivity of the different nickel base alloy zones, derived from R&D and empirical models, would have to be confirmed by a comprehensive ISI program, including bottom head penetrations, steam generator partition plates, and more specific weld metal zones. • ASN reviewed the various mitigations and preventive measures proposed by EDF, either temporary, such as leak detection systems, anti-ejections devices, interim repairs, or long term commitment of the French operator to replace in due time the vessel heads comprising the most affected CRDM penetrations. This paper also presents the ASN’s follow up of the domestic and international feedback, such as the occurrence of PWSCC cracking (initiation and propagation) in the weld, whose occurrence is rather limited in France.

Comparison of Nuclear Fitness for Service Codes: ASME Section XI and RSE-M AFCEN Codes

2014 ◽  
Author(s):  
Claude Faidy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Light Water Reactors ◽  
Pressurized Water ◽  
Term Operation ◽  
Mechanical Components ◽  
Water Reactors ◽  
Test Requirements ◽  
Service Inspection

Two major Codes are used for Fitness for Service of Nuclear Power Plants: one is the ASME B&PV Code Section XI and the other one is the French RSE-M Code. Both of them are largely used in many countries, partially or totally. The last 2013 RSE-M covers “Mechanical Components of Pressurized Water Reactors (PWRs): - Pre-service and In-service inspection - Surveillance in operation or during shutdown - Flaw evaluation - Repairs-Replacements parts for plant in operation - Pressure tests The last 2013 ASME Section XI covers “Mechanical components and containment of Light Water Reactors (LWRs)” and has a larger scope with similar topics: more types of plants (PWR and Boiling Water Reactor-BWR), other components like metallic and concrete containments… The paper is a first comparison covering the scope, the jurisdiction, the general organization of each section, the major principles to develop In Service Inspection, Repair-Replacement activities, the flaw evaluation rules, the pressure test requirements, the surveillance procedures (monitoring…) and the connections with Design Codes… These Codes are extremely important for In-service inspection programs in particular and essential tools to justify long term operation of Nuclear Power Plants.

The Next EDF Tracking System

2009 ◽  
Author(s):  
Emmanuelle Julli ◽  
Bertrand Lantes
Keyword(s):  
Radioactive Waste ◽  
Nuclear Power ◽  
Power Plants ◽  
Tracking System ◽  
Treatment Plant ◽  
Computer Application ◽  
Pressurized Water ◽  
Waste Package ◽  
Additional Processing ◽  
Water Reactors

EDF’s network of nuclear power plants (NPP) comprises 58 pressurized water reactors. Solid waste arising during plant operation (mainly VLLW, LLW and ILW) are conditioned and sent either to interim storage, an off site treatment plant for additional processing (e.g. the Centraco incinerator or the melting facilities of SOCODEI) or directly to one of the two final repositories operated by ANDRA, the French national radioactive waste management agency. The tracking system allows: - the checking of waste package characteristics against acceptance criteria of the final disposal facilities or off site treatment facilities; and - the transmission of the waste package data to ANDRA and SOCODEI. Since 1992, the EDF computer application DRA has been run on networked computers at EDF and ANDRA, and more recently at SOCODEI. DRA is also a key element in the management of radioactive waste. It allows a large range of inter comparisons to be made between the NPPs in operation and is thus the principal tool used optimize technical and economic performance. After 15 years of use, DRA was technically obsolete and could no longer be successfully developed to meet evolving regulatory requirements. It was therefore decided to completely replace the DRA system and in so doing to introduce new functionality.

Sign in / Sign up

Export Citation Format

Share Document