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

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.

Lifetime Management of EDF PWR Pressure Vessels and Pipings

Lifetime management of EDF PWR vessels and pipings are one of the main technical key points of safety and competitivness. This paper describes the EDF global approach in this field, which is applied to the nuclear fleet i.e 58 nuclear power plants, and particularly to the first 34 three loops, as far as lifetime is concerned: • operating procedures and routine maintenance, special maintenance and ten years safety reassessment, • engineering analysis, based on feed back experience, scientific knowledge, degradations mechanisms, causes and consequences management, • operating loadings decrease, • complementary deterministic and cost-benefit analysis, • fit for service justifications, • anticipation strategy to prepare future, based on Non Destructive Testing investigations, ability to repair and/or to replace components, in situ expertises, ... Some examples are given: lifetime management of reactor vessels heads and bottom penetrations of pressure vessels, fit for service of cast stainless steel primary pipings, primary nozzles and auxiliary pipings special maintenance.

Piping Service Life Experience in Commercial Nuclear Power Plants: Progress With the OECD Pipe Failure Data Exchange Project

An extension of a 1994–98 R&D project established in 2002 by certain member countries of the Organisation for Economic Cooperation (OECD), the OECD Pipe Failure Data Exchange (OPDE) Project has produced a major database on the piping service experience applicable to commercial nuclear plants. The 3-year project is operated under the umbrella of the OECD Nuclear Energy Agency (NEA) and organizations producing or regulating more than 80% of nuclear energy generation worldwide are contributing data to the OPDE Project. The Project considers pipe failure data including service-induced wall thinning, part through-wall crack, pinhole leak, leak, and rupture/severance (i.e., events involving large leak rates up to and beyond the make-up capacity of engineered safeguards systems). The part through-wall events include degradation in excess of code allowable for pipe wall thinning or crack depth. OPDE also addresses such degradation that could have generic implications regarding the reliability of in-service inspection. At the end of 2003 the OPDE database included approximately 4,400 records on pipe failure affecting ASME Code Class 1 through 3 and non-Code piping. The database also included an additional 450 records on water hammer events where the structural integrity of piping was challenged but did not fail. This paper summarizes the unique data quality considerations that are associated with piping components. The paper also summarizes the database content.

French R&D Program for RPV Aging Management

Until now French approach for RPV PTS assessment is based on at least 40 years lifetime. This lifetime has been taken into account at each step of the the RPV life: first early in the design, then at each periodic safety demonstration by including the surveillance program, the national and international feedback and R&D results. All of them confirm that all the 3-loop French RPV fulfill the existing criteria for at least 40 years of operation. In order to evaluate their capability to operate for 60 years, an Engineering and Research and Development program has been recently established and engaged by EDF. This large program of activities between all of divisions of EDF is focused on the different fields involved in the risk of fast rupture of the irradiated core vessel. The main purposes of this programme are: • the research of specific data corresponding to a long lifetime of about 60 years; • the studies of new methods to improve the demonstration including several themes such as fluence evaluation, determination of fracture toughness, structural integrity assessment including probabilistic approach, definition of transients; • the evaluation of mitigation methods. This paper contains first a description of what was done at each RPV PTS assessment and an overview of the present program.

Reactor Pressure Vessel of Tricastin Unit 1 Core Zone Inspection: Comparison of Non-Destructive Examination Results Among Two Core Area Inspections Performed in 1999 and in 2003

As part of the unit aging follow-up, a non.destructive examination has been designed to inspect the core zone of the Reactor Pressure Vessel (RPV). This ultrasonic process called “VPM” has been developed and qualified by Intercontroˆle in accordance with EDF specifications. The qualification has been attested by an accredited qualification body. In 1999, the VPM process was used the first time during the in-service inspection of TRICASTIN, unit 1. As a result of the RPV core zone inspection a set of under cladding flaw indications was detected and characterized with the VPM process. These flaws have been analysed as planar manufacturing flaws. In 2003, 4 cycles later, during the TRICASTIN Unit 1 outage inspection, “VPM” was used to characterize again this set of flaw indications in order to verify their dimensions. The changes of the main characteristics (specially height) between both the inspections were compared to the process accuracy. No significant dimension change has been recorded.

Procedures on Fracture Toughness Estimation Safety Assessment and Life Extension for In-Service High Pressure Vessels

Based on the typical in-service high pressure vessels made of PCrNi3MoVA for producing synthetic crystal, a systematic technology of material fracture toughness estimation, structural integrity assessment, and life extension is carried out for the in-service equipment with the following aspects: macroscopically and microscopically analyzing, the tests including KIC, AKV, FATT (50%), the predicting method of fracture, system safety assessment, and the life extension technology. Some practical conclusions can be obtained from the test and analysis as follows: The main failure factors for this kind of high pressure vessels include heat treatment processes, temper brittleness, and stress corrosion; It is found that the value of FATT (50%) increased very significantly; The comparison between the test results and the predicted results of the value of KIC is also performed, and it is shown that the value of KIC of in-service equipment can be estimated by the formula presented by Barsom-Rolfe or in API 579 with the value of AKV, The test temperature is recommended at least to reach 25 C (or room temperature) for the repaired vessels; The life extension technologies are put forward for this kind of in-service super-high pressure vessels.

License Renewal Program for Entergy’s Fleet of Nuclear Power Plants

License renewal of operating nuclear power plants in the United States has become one of the most successful nuclear industry activities in the past few years. Entergy, the second largest nuclear plant operator in the U.S., was one of the pioneers in this new process. In 2000, Entergy submitted a license renewal application to the Nuclear Regulatory Commission (NRC) for Arkansas Nuclear One, Unit 1. By June 2001, less than 17 months later, the NRC issued a renewed license. Due in part to the efficiency and success of this first Entergy license renewal project, a dedicated team of Entergy and Areva (formerly known as Framatome-ANP) personnel was established using virtual office concepts to work on license renewal for the remaining nine Entergy nuclear units over the next decade. Since each license renewal project takes 4 to 5 years and costs $10 to $15 million to complete [1], the dedicated team has focused on improving the schedule and economics of the license renewal process. By early 2004, the dedicated team has worked on five license renewal projects and expects to begin work on at least two additional projects by 2005. The virtual team organization has established standardized processes for managing data and for performing aging management reviews, environmental reviews, and time limited aging analyses evaluations. In addition, the team has worked with the Nuclear Energy Institute (NEI) and the NRC to further improve the efficiency of the license renewal process. This paper discusses the standardized processes established, the virtual team techniques used to manage multiple license renewal projects, and the plans for further process improvements. The ultimate goal of Entergy’s license renewal work is to achieve a highly efficient and effective license renewal process that ensures the safe continued operation of its fleet of nuclear power plants for decades to come.

Life Assessment of Components With Underclad Cracks in PWR

During the in-service inspection in 1999 in Tricastin 1 Nuclear Plant, underclad cracks had been discovered in the core shell. Following the specific demonstration of the acceptability of these defects, other components also sensitive to this kind of defects have been re-examined. The life assessment includes a review of the mechanism of formation of theses underclad flaws by cold cracking and the selection of the components sensitive to theses cracks. This includes various shells of the reactor pressure vessel, the steam generator and the pressurizer. A second part consists in mechanical analysis of theses selected components, with an underclad flaw of very large size. This is compared with the maximum size of a realistic crack, which depends of the size of the heat affected zone during the cladding process. The third part has to do with the In Service Inspection. This concerns only the reactor pressure nozzles — both inlet and outlet — which are examined during the inspection of the core shell of the reactor vessel.

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

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.

Neutron Embrittlement of VVER 1000 and 440/213 RPVs: Learning From EC Projects on RPV Integrity

Since 1991 the European Commission has financed a significant number of Technical Assistance Projects to the Commonwealth of Independent States (TACIS) and EURATOM R&D actions addressing the main safety issues on RPV material embrittlement and integrity assessment. Since the VVER 440 reactors of the reference series 213 are made from recognised neutron embrittlement resistant materials and include comprehensive surveillance programmes, a standard plant life management procedure can be applied to address long-term concerns, mostly aiming at reducing uncertainties in the assessment techniques. Therefore, the open issues (flux effect, upgrading of surveillance results, implementation of toughness measurements and relevant acceptance criteria, behaviour of the cladding) are quite universal. The efficiency of late annealing (at fast [E>0.5 MeV] neutron doses over 1020n/cm2) and the re-embrittlement after annealing remain key issues for any final decision for their operational lifetime. The more recently developed VVER 1000 reactors have some well-known features arising from the original design and manufacturing process (high nickel content in the core weld, location of the surveillance specimens), which have to be carefully considered if appropriate mitigation measures are to be implemented during operation. A precise identification of the issues related to the surveillance programme has been achieved thanks to research on dosimetry evaluation, representativeness (temperature and flux effect) of the specimens and optimisation of the evaluation of their results. Nickle, just as copper and phosphorus, is now recognised as having a detrimental effect on neutron embrittlement in synergy with other elements (e.g. manganese). The analyses of available data for CrNiMn steels do not show a significant effect for fast [E>0.5 MeV] neutron doses below 7.1019n/cm2, but their consistency and relevance might be questionable. A way has already been pioneered which shows how valuable results can be obtained using the existing surveillance programmes specimens. A systematic application on the Russian & Ukrainian plants is now planned in order to get updated figures on design end of life (EOL) integrity assessment. This includes updated dosimetry assessment, multiple specimen testing (reconstitution, impact and static toughness tests) and advanced integrity analyses. An optimised database of representative surveillance results (up to the design end of life) is expected, which should provide a sound basis for further understanding and setting up of relevant prediction tools, considering at the same time any other specific R&D results. The global integrity assessment will also provide for preparing and implementing adequate mitigation measures in due time, if necessary. The paper will report about the knowledge on RPV embrittlement effects, providing evidence of recent contributions to solve shortcomings of the VVER 440/213 and 1000 units. The current state-of-the-art and the remaining open issues have been assessed recently by a group of international experts. The planned R&D activities and the detailed scope of the latest TACIS projects are described.