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«Report EUR 26964 EN European Commission Joint Research Centre Institute for Energy and Transport Contact information Miguel Peinador Veira Address: ...»

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European Clearinghouse:

Report on Leaks and Cracks of the

Reactor Coolant Pressure Boundary

Summary Report of an

European Clearinghouse

Topical Study

Radian Marius Sanda

Miguel Peinador Veira

Report EUR 26964 EN

European Commission

Joint Research Centre

Institute for Energy and Transport

Contact information

Miguel Peinador Veira

Address: Postbus 2, 1755 ZG Petten, the Netherlands

E-mail: Miguel.Peinador-Veira@ec.europa.eu

Tel.: +31 22 456 5176

Fax: +31 22 456 5637 https://ec.europa.eu/jrc Legal Notice This publication is a Science and Policy Report by the Joint Research Centre, the European Commission’s in-house science service. It aims to provide evidence-based scientific support to the European policy-making process. The scientific output expressed does not imply a policy position of the European Commission. Neither the European Commission nor any person acting on behalf of the Commission is responsible for the use which might be made of this publication.

All images © European Union 2014, except cover image JRC 92718 EUR 26964 EN ISBN 978-92-79-44422-7 (pdf) ISBN 978-92-79-44423-4 (print) ISSN 1831-9424 (online) ISSN 1018-5593 (print) doi: 10.2790/316405 Luxembourg: Publications Office of the European Union, 2014 © European Union, 2014 Reproduction is authorized provided the source is acknowledged.

Abstract Leaks and cracks of the Reactor Coolant Pressure Boundary are a clear indication of degraded conditions inside the nuclear power plants, challenging the barriers against radioactive releases into the environment. Thus it is important to investigate the existing operational experience in this kind of events. The objective is to determine the adequacy of protection of nuclear power plants against leaks and cracks and the effectiveness of the corrective actions implemented, as well as to provide recommendations on how to prevent or mitigate the impact of such events on NPP operation.

Leaks and cracks in the Reactor Coolant Pressure Boundary cannot be avoided. Because of this, even the Technical Specifications (or the Plant Normal Operation Procedures) allow a limited leak rate, or a specific time of unavailability, for different systems needed for normal operation of the plant. However, any leak and crack of the Reactor Coolant Pressure Boundary which remains undetected long enough will raise particular challenges for the Systems, Structures and Components within the affected part of the Reactor Coolant Pressure Boundary. Even if the leak is timely detected, there is usually a significant impact on the overall radioprotection of the plant. In some cases, the total reactor coolant loss is important and prolonged outage periods are required for recovery. So, even in the case of successful detection of such events, to analyze them is necessary in order to fully understand the mechanism which provoked such events.

This Summary Report presents the results of a comprehensive study performed by the European Clearinghouse on Operating Experience Feedback of NPP with the support of IRSN and GRS.

TABLE OF CONTENTS

–  –  –

LIST OF ACRONYMS

GRS Gesellschaft für Anlagen- und Reaktorsicherheit mbH HAZ Heat Affected Zone IAEA International Atomic Energy Agency IRS International Reporting System for Operating Experience jointly operated by IAEA and OECD/NEA IRSN Institut de Radioprotection et de Sûreté Nucléaire IGSCC Inter-Granular Stress Corrosion Cracking LER Licensee Event Report NPP Nuclear Power Plant OECD/NEA Nuclear Energy Agency OEF Operating Experience Feedback QA Quality Assurance R&D Research and Development RCP Reactor Coolant Pump RCPB Reactor Cooling Pressure Boundary RCS Reactor Cooling System RPV Reactor Pressure Vessel RVH Reactor Vessel Head SCC Stress Corrosion Cracking SICC Strain Induced Corrosion Cracking SG Steam Generator SGTR Steam Generator Tube Rupture SRV Safety Relief Valve SSC Systems, Structures and Components TGSCC Trans-Granular Stress Corrosion Cracking US NRC United States Nuclear Regulatory Commission

1 INTRODUCTION

Leaks and cracks on the primary circuit boundary are evidence of degradation of a barrier against radioactive releases into the environment.

Any leak and crack of the Reactor Coolant Pressure Boundary (RCPB) that remains undetected long enough will raise particular challenges for the systems, structures and components (SSCs) within the affected part of the RCPB. Even if the leak is timely detected, there is usually a significant impact on the overall radioprotection of the plant. In some cases, the total reactor coolant loss is important and prolonged outage periods are required for recovery. Thus it is important to investigate the existing operating experience in this kind of events.

The objective of this study is to determine trends and lessons learned, in order to draw recommendations and conclusions to prevent re-occurrence of such events. The main goal is to identify and share good practices among the nuclear community.

This Summary Report presents the results of a comprehensive study [1] performed by the European Clearinghouse on Operating Experience Feedback (OEF) of NPP with the support of IRSN (Institut de Sûreté Nucléaire et de Radioprotection) and GRS (Gesellschaft für Anlagen und Reaktorsicherheit mbH). The study [1] was performed by analysing the events contained in four different databases, for a period covering 20 years.





2 METHODOLOGY

Four different event report databases have been searched in order to analyse the operating experience of events related to leaks and cracks of the reactor coolant pressure

boundary, namely:

 The International Reporting System for Operating Experience (IRS), operated jointly by the International Atomic Energy Agency (IAEA) and the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA).

The fundamental objective of the IRS is to contribute to improving the safety of commercial nuclear power plants (NPPs) operated worldwide. The IRS is a worldwide system containing events which are reported on a voluntary basis, and the reporting criteria vary with countries.

 United States Nuclear Regulatory Commission Licensee’s Event Reports (US NRC LER) national database which consists of reports from the licensees for those types of reactor events and problems that are believed to be significant and useful to the NRC in its effort to identify and resolve threats to public safety.

 The French database developed and operated by the IRSN, in order to classify and to analyse the OEF. It contains events reported by French nuclear power plants under French regulations.

 The German database, operated by GRS in Germany. This database contains specific information on the behaviour of pressurized components in German NPPs.

The events are reported by German plants under German reporting criteria for systematic evaluation of operating experience.

All databases were searched using keywords and other specific searching tools, yielding a list of potentially relevant events. The reports of these events were reviewed individually to determine the pertinence of these events for the study, thus obtaining a screened list of relevant events.

Screening the databases' search results helped identify the events that are relevant for the topic of this report. Event screening results are largely dependent on the criteria used to report significant nuclear safety and/or radiation protection events. The criteria used to report cracks and leaks events in one country may be (and usually are) different from those used in another country, even for similar types of events and plants.

After screening, the following families were created for a short quantitative analysis:

 Event causes.

 Part of component affected.

The families created for the analysis of lessons learned are the following:

 Reactor pressure vessel and pressurizer (RPV).

 Steam generators (SGs).

 Reactor cooling pumps (RCPs).

 Piping.

 Safety relief valves (SRVs).

 Valves, other than safety relief valves.

 Flanges.

For every family of events, the report summarizes the lessons learned as well as the recommendations which can be drawn from the operating experience.

3 MAIN FINDINGS

This section presents the main results of the quantitative analysis performed on the events identified.

When performing the analysis for the databases selected as sources of information, it is imperative to take into account that these databases are not built in respect to the same reporting criteria, thus no common integrated analysis can be carried out. For the same reason, comparison should be made in a very careful manner.

The search for incidents in the IRS revealed 144 events involving leaks and cracks of the RCPB. As referring to US NRC LER database, screening resulted in 74 applicable events.

In the case of SAPIDE and KomPass databases, the search resulted in 129 and 61 applicable events respectively. Given to the different nature of the databases, the study concentrated on statistical analyses on the data to gain insights of the events distribution.

3.1 Components affected by events Although there was no intention to account the number of events occurred at a certain type of reactor, the analyses were made taking into account the differences between plant designs.

Figures 1 through 4 show the distribution of the events according to the types described in the previous section, for every database.

Figure 1 – Distribution of leaks and cracks reported to the IRS Figure 2 – Distribution of leaks and cracks in United States of America Figure 3 – Distribution of leaks and cracks in France Figure 4 - Distribution of leaks and cracks in Germany Referring to all events, the analysis revealed that the most affected component by leaks and cracks are the pipes (116 events). The other most affected components by these types of events are SGs (88 events, most of them involving tubes and installation of nozzle dams), RPV and pressurizer (66 events) and valves (65 events). The less affected component affected by leaks and cracks proven to be the safety relief valves (17 events).

3.2 Event causes For all the 408 events, Table 1 presents the distribution of events regarding their causes.

–  –  –

Corrosion is the main root cause of the events analysed. Sometimes, this was just the 'apparent' cause. Indeed, in about 50 % of the cases corrosion was triggered by the chemical reactions between the materials being in contact one with each other, in the accidental presence of a catalyst (e.g. Stress Corrosion Cracking (SCC) phenomenon, for alloy 600, in the presence of a chloride medium and oxygen, or Trans-Granular Stress Corrosion Cracking (TGSCC) of stainless steel — from exterior to interior — in the presence of chlorides). In about one third of the cases, corrosion was triggered by manufacturing defects that were the precursors for Inter-Granular Stress Corrosion Cracking (IGSCC). The other major contributors for triggering corrosion are maintenance defects and faulty operations (e.g. strain-induced corrosion cracking (SICC) as a result of (1) Fourteen events were identified as having double root causes. This led to a total of 75 events for the German database.

a groove at a weld seam that acted as a starting point for the degradation, or inhibited movement of components, which induced additional strain).

Manufacturing defects are the second largest root cause of the events. These events are nearly all related to welding faults, though in some cases inappropriate Quality Assurance (QA) measures at the manufacturer can be considered as precursors.

However, no unique trend can be identified for this root cause.

Fatigue (both mechanical and thermal) is also an important contributor to the root causes family.

Mechanical fatigue appeared, in certain cases, in combination with another cause, like manufacturing or maintenance defects, especially on welds. Mechanical fatigue appeared especially on those SSCs exposed repeatedly or periodically to various aggressive conditions (i.e. instrumentation line during outage operations, particularly during vessel head removal and set-back operations, or while operating/testing valves and pumps).

Thermal fatigue usually appears mainly in conjunction with the Farley–Tihange phenomenon2, but also could be triggered by high temperature differences from one side to the other of a tube, for example — but in this case thermal fatigue is only the initiating mechanism, the propagating mechanism being mechanical fatigue.

Maintenance defects and faulty operations, together, share 20 % of the causes of the events. Most of these were triggered by human factors, either by non-appropriate establishment of the 'Foreign Materials Exclusion Zone', or by incomplete manoeuvring or inappropriate positioning especially of hand-operated (isolation) valves. Installation errors, especially regarding SG inlet or outlet nozzle dams and drain plugs, were the source of recurring leaks. However, not all the events within these two categories can be explicitly attributed to human errors.

3.3 Part of component affected Defects, cracks and leaks are often located in specific areas with geometric or metallurgical (weld) discontinuities. Table 2 presents the most affected part of components by these events.

–  –  –

The base metal is the part of the component most frequently involved in events related to leaks and cracks. As base material, alloy 600 is widely used in NPPs, and not only for the SG tubes (i.e. vessel head penetrations). Another base material widely used is the austenitic stainless steel, which can be titanium or niobium stabilised (i.e. piping, pump This phenomenon can be described as thermal fatigue in the terminal section of the safety injection lines, which are connected to the reactor coolant loops, due to the simultaneous mixing of cold water from the charging system and hot water from reactor coolant loops.

housings and valve bodies).



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