integrated (FAMOSi), necessary tools to monitor local fatigue and to provide realistic

Transkript

1 Content in brief German Experiences in Local Fatigue Monitoring (Page 284) E. Abib, St. Bergholz and J. Rudolph The ageing management of nuclear power plants (NPP) has gained an increasing importance in the last years. The reasons are mainly due to the international context of extending period of plants operation. The fatigue damage process takes a central position in ageing mechanisms of components. In the design phase of NPP, fatigue analyses are still based on theoretical considerations and empirical values, which are summarized in the design transient catalogue, necessary for licensing. These analyses will also provide the fatigue-relevant positions in the NPP and give a basis for future design improvements and optimization of operating modes. The design transients are in practice conservatively correlated with the real transients occurring during operation. Uncertainties reveal very conservative assumptions. During operation of the plant, it has to be recurrently proved, that the plant is being operated under designed boundary conditions. Moreover, operating signals are constantly acquired to enable a fatigue evaluation. Experience shows that some significant differences occur between the design transients and the real occurred transients during plant operation. Nowadays a direct derivation of the complete stress tensor at the fatigue-relevant locations is enabled thanks to the recorded local loads and combination with finite element (FE) analyses. So, additionally to the recorded temperature curves, a representation of the time evolution of the 6 stress components for each monitored component is possible. This allows the application of the simplified elastoplastic fatigue check according to design codes. The fatigue level can be realistically analyzed with a suitable cycle-counting method. Areva offers the Areva fatigue concept (AFC) and the new fatigue monitoring system integrated (FAMOSi), necessary tools to monitor local fatigue and to provide realistic assessment. Fatigue Evaluation Including Environmental Effects for Primary Circuit Components in Nuclear Power Plants (Page 289) J. Seichter, S.H. Reese and D. Klucke The influence of LWR coolant environment to the lifetime of materials in Nuclear Power Plants is in discussion internationally. Environmental phenomena were investigated in laboratory tests and published in recent years. The discussion is mainly focused both on the transition from laboratory to real plant components and on numerical calculation procedures. Since publishing of the NUREG/CR-6909 report in 2007, formulae for calculating the F en factors have been modified several times. Various calculation procedures like the so called Strain-integrated Method and Simplified Approach have been published while each approach yields to different results. The recent revision of the calculation procedure, proposed by ANL in 2012, is presented and discussed with regard to possible variations in the results depending on the assumptions made. In German KTA Rules the effect of environmentally assisted fatigue (EAF) is taken into account by means of so called attention thresholds. If the threshold value is exceeded, further measures like NDT, inservice inspections including fracture mechanical evaluations or detailed assessment procedures have to be performed. One way to handle those measures is to apply sophisticated procedures and to show that the calculated CUF is below the defined attention thresholds. On the basis of a practical example, methods and approaches will be discussed and recommendations in terms of avoiding over-conservatism and misinterpretation will be presented. CARINA A Program for Experimental Investigation of the Irradiation Behaviour of German Reactor Pressure Vessel Materials (Page 297) H. Hein, E. Keim, E. Bechler, P. Efsing, J. Ganswind, R. Knobel, G. König, P. Barreiro, M. Widera and A. de Jong The proof of a sufficient safety margin against brittle fracture of the reactor pressure vessel (RPV) is an important part of the operational safety of nuclear power plants. The RPV safety assessment procedure applicable in Germany is described in KTA of the Nuclear Safety Standard Commission (KTA). This deterministic assessment concept is based on the comparison of load curves with the material resistance curve in terms of fracture toughness. The fracture toughness curve can be determined either indirectly according to the RT- NDT concept based on Charpy tests or directly according to the more appropriate RT T0 approach based on Master Curve analysis of fracture toughness tests, respectively. In the recently completed research project CARI- NA the data base for pre-irradiated original RPV steels of German PWR construction lines was extended by comprehensive fracture toughness testing. The data obtained up to neutron fluences of n/cm 2 (E > 1 MeV) are analysed and discussed particularly in terms of Master Curve applications. The experimental results show that optimized RPV manufacturing specifications and reactor designs are advantageous for a long-term plant operation in comparison to less optimized materials with lower toughness and to reactor designs with substantial higher neutron irradiation. With the obtained data, experiences and insights an essential contribution was also made to the integration of the Master Curve concept in German safety standards. Isar-2 Nuclear Power Station Twenty-five Years (Page 304) E. Fischer and M. Luginger The Isar-2 nuclear power station (KKI 2) began commercial power operation on April 9, 280

2 Content in brief In these past 25 years the plant generated a total of approx. 285 billion kwh of electricity. The annual electricity production of KKI 2 of approx. 12 billion kwh corresponds to a share of approx. 15 % in the cumulated Bavarian electricity production. This amount of electricity, theoretically, could supply some 3 million three person households, or meet two thirds of the electricity requirement of the Bavarian industry, for one year. In its 25 years of power operation the Isar-2 nuclear power plant has recorded the highest annual gross electricity production of all nuclear power plants in the world nine times so far. A plant performance as impressive as this necessitates a plant availability far above the average. This, in turn, is based on short revision times and faultfree plant operation. However, high plant safety and availability must not be taken for granted, but are the result of responsible, safety-minded plant operation combined with continuous plant optimization and permanent execution of comprehensive checks, inspections, and maintenance measures. Besides plant technology also organization and administration were permanently advanced and adapted to changing requirements so as to safeguard reliable, safe, and non-polluting plant operation. Using Management of Aging in German Nuclear Power Plants Aspects of KTA 1403 Pertaining to Residual Power Operation, Postshutdown Operation and Residual Operation (Page 308) P. Barreiro, Th. Bever, G. Brast, B. Elsche, P. Grossmann, F. Hüttner, Th. Linnemann, S.H. Reese, S.-O. Smit, M. Widera and R.-M. Zander Management of aging in nuclear power plants originated in the United States of America and became a topic of debate in Germany from the late 1990s onward. On the basis of the existing plant-specific measures practiced comprehensively, KTA 1403, Management of Aging in Nuclear Power Plants, was drafted and finalized in This publication first presents the context of AM with regard to German nuclear power plants, including references to national and international historical developments. Against this backdrop, the difference between management of aging and lifetime management is discussed next. This is followed by a description of the status of the AM process in nuclear power plants currently in operation, especially organizational plant-specific implementation. As a consequence of the decision by the German federal government to discontinue the peaceful use of nuclear power in Germany and the associated 13 th amendment to the Atomic Energy Act of July 31, 2011, a considerable part of the German nuclear power plant park already lost its right of power operation. In this situation, aspects of AM are discussed for plants in the no-power, post-operation and residual operation phases. Finally, experience accumulated in plantspecific execution of the AM process on the basis of KTA 1403 is considered and summarized. Reactor Safety Research in Times of Change (Page 316) R. Zipper, Since the early 1970ies reactor safety research sponsored by the German Ministry of Economics an Technology and its predecessors and pursued independently from interests of industry or industrial associations as well as from current licensing issues significantly contributed to the extension of knowledge regarding risks and possible threats associated with the operation of nuclear power plants. The results of these research activities triggered several measures taken by industry and utilities to further enhance the internationally recognized high safety standards of nuclear power plants in Germany. Furthermore, by including especially universities in the distinguished research activities a large number of young scientists were given the opportunity to qualify in the field of nuclear reactor technology and safety thus contributing to the preservation of competence during the demographic change. The nuclear phase out in Germany affects also issues of reactor safety research in Germany. While Germany will progressively decrease and terminate the use of nuclear energy for public power supply other countries in Europe and in other parts of the world are continuing, expanding and even starting the use of nuclear power. As generally recognized, nuclear safety is an international issue and in the wake of the Fukushima disaster there are several initiatives to launch a system of internationally binding safety rules and guide lines. The German Competence Alliance therefore has elaborated a framework of areas were future reactor safety research will still be needed to support German efforts based on own and independent expertise to continuously develop and establish highest safety standards for the use of nuclear power supply domestic and abroad. German Nuclear Engineering is High Tech (Page 322) J. Reinartz Despite the political decision of Germany to opt out of electricity production from nuclear power, know-how in nuclear technology will continue to be needed not only for demolition of nuclear plants and final storage of radioactive waste, but especially in the field of safety technology. For decades after the last nuclear power plant in Germany has been shut down we will continue to need know-how in nuclear technology, innovative solutions, and well-trained personnel to do the job, said Dr. Ralf Güldner, President of DAtF. Nuclear technology in this country is more than just electricity generation, Güldner continued. At the Energy in a Dialog event organized in Berlin on March 12, 2013, these were topics of debate. Hans-Christoph Pape of the German Federal Ministry of Economics and Technology and Stefan vom Scheidt, Technical Manager and Spokesman of the Board of Management of Areva GmbH, offered comments and explained the status and perspectives. atw Vol. 58 (2013) No. 5»atomwirtschaft-atomtechnik«is published monthly by INFORUM Verlags- und Verwaltungsgesellschaft mbh Robert-Koch-Platz 4, Berlin, Germany phone fax Publisher: atw@atomwirtschaft.de Editorial: editorial@atomwirtschaft.com 281

3 Content in brief (German) Deutsche Erfahrungen mit der lokalen Ermüdungsüberwachung (Page 284) E. Abib, St. Bergholz und J. Rudolph Das Alterungsmanagement bei Kernkraftwerken (KKW) ist in den letzten Jahren immer wichtiger geworden. Die Gründe dafür liegen hauptsächlich im internationalen Kontext der Laufzeitverlängerung. Schäden durch Ermüdung spielen bei Alterungsmechanismen von Bauteilen eine zentrale Rolle. In der Konstruktionsphase von KKW sind Ermüdungsanalysen noch auf theoretischen Überlegungen und empirischen Werten aufgebaut, die für das Genehmigungsverfahren im Auslegungstransientenkatalog zusammengefasst sind. Diese Analysen liefern dann auch die ermüdungsrelevanten Stellen im KKW und stellen damit eine Grundlage für künftige Konstruktionsverbesserungen und die Optimierung der Betriebsweise dar. In der Praxis korrelieren die Auslegungstransienten konservativ mit den wirklichen Transienten im Betrieb. Unsicherheiten zeigen sehr konservative Annahmen auf. Im Anlagenbetrieb muss immer wieder nachgewiesen werden, dass das KKW innerhalb der Auslegungs-Grenzbedingungen läuft. Außerdem werden ständig Betriebssignale gewonnen, mit denen eine Ermüdungsbewertung durchgeführt werden kann. Die Erfahrung zeigt, dass zwischen den Auslegungstransienten und den im Betrieb tatsächlich auftretenden Transienten erhebliche Unterschiede vorkommen. Heute ist dank der aufgezeichneten lokalen Lasten in Verbindung mit einer Analyse finiter Elemente eine direkte Ableitung des vollständigen Belastungstensors an ermüdungsrelevanten Stellen möglich. Neben den aufgezeichneten Temperaturverläufen kann auch die zeitliche Entwicklung der 6 Lastkomponenten für jedes überwachte Bauteil dargestellt werden. Damit ist die vereinfachte Überprüfung der elastisch-plastischen Ermüdung nach den Auslegungsvorschriften möglich. Das Maß der Ermüdung lässt sich mit einer geeigneten Methode zur Zählung der Lastwechselspiele realistisch analysieren. Areva bietet das Areva-Ermüdungskonzept (AFC) und das neue integrierte Ermüdungsüberwachungssystem (FAMOSi) an notwendige Hilfsmittel zur Überwachung lokaler Ermüdung und zur realistischen Beurteilung. Bewertung der Ermüdungsbeanspruchung von Primärkreiskomponenten in Kernkraftwerken unter Berücksichtigung des Mediumseinfluss (Page 289) J. Seichter, S.H. Reese und D. Klucke Der Einfluss des Primärkreismediums auf die Ermüdungslebensdauer von Werkstoffen, die in Kernkraftwerken für druckführende Komponenten eingesetzt werden, wird international diskutiert. Dazu wurden Ergebnisse diverser Laboruntersuchungen veröffentlicht. Hauptdiskussionspunkte sind die Übertragbarkeit der Laborergebnisse auf KKW- Komponenten sowie die vorgeschlagenen Berechnungsmethoden. Seit Veröffentlichung des NUREG/CR-6909 Berichtes im Jahr 2007 wurde z.b. der Formelsatz für die F en -Faktoren etliche Male revidiert. Es wurden verschiedene Berechnungsmethoden wie die Strain-Integrated Method oder die Simplified Method veröffentlicht, wobei jede zu anderen Ergebnissen führt. Speziell wird anhand der letzten Revision der Berechnungsmethode durch ANL aus dem Jahr 2012 die Abhängigkeit der Ergebnisse von den getroffenen Annahmen diskutiert. Die KTA-Regel berücksichtigt den Mediumseinfluss durch sog. Aufmerksamkeitsschwellen. Bei Überschreiten dieser Werte sind zusätzliche Maßnahmen wie z.b. wiederkehrende Prüfungen einschließlich bruchmechanischer Bewertungen oder eine rechnerische Quantifizierung des Mediumseinflusses nötig. Eine Möglichkeit für diese rechnerische Betrachtung ist die Anwendung qualifizierter Berechnungsverfahren, um zu zeigen, dass die rechnerischen Erschöpfungsgrade die Schwellenwerte nicht überschreiten. An einem Beispiel werden verfügbare Berechnungsverfahren diskutiert und Empfehlungen gegeben, wie überkonservative Ergebnisse vermieden werden können. CARINA Ein Programm zur experimentellen Untersuchung des Bestrahlungsverhaltens von deutschen Reaktordruckbehälter-Werkstoffen (Page 297) H. Hein, E. Keim, E. Bechler, P. Efsing, J. Ganswind, R. Knobel, G. König, P. Barreiro, M. Widera und A. de Jong Der Nachweis eines ausreichenden Sicherheitsabstandes gegen Sprödbruch des Reaktordruckbehälters (RDB) ist ein wichtiger Bestandteil der Betriebssicherheit von Kernkraftwerken. Zur sicherheitstechnischen Bewertung des RDB wird in Deutschland die KTA des Regelwerks des Kerntechnischen Ausschusses (KTA) zugrunde gelegt. Dieses deterministische Bewertungskonzept beruht auf dem Vergleich der Belastungskurven mit dem Materialwiderstand hinsichtlich Bruchzähigkeit. Die Bruchzähigkeitskurve kann entweder auf indirektem Weg nach dem auf Kerbschlagbiegeversuchen basierten RT NDT -Konzept oder direkt nach dem besser geeigneten RT T0 -Konzept, welches auf einer Master- Curve Analyse von bruchmechanischen Versuchen beruht, bestimmt werden. Mit dem Forschungsvorhaben CARINA (Characteristics of Irradiated German RPV Materials) wurde die experimentelle Datenbasis für bestrahlte deutsche RDB-Werkstoffe durch umfangreiche Bruchmechanikversuche erweitert. Die bis zu Neutronenfluenzen von 7, n/cm 2 (E > 1 MeV) erhaltenen Versuchsdaten werden insbesondere mit Bezug auf Master-Curve-Anwendungen ausgewertet und diskutiert. Die experimentellen Ergebnisse zeigen, dass optimierte RDB-Herstellungsspezifikationen und Reaktorauslegungen von Vorteil für einen langfristigen Kraftwerksbetrieb sind, im Vergleich zu weniger optimierten Werkstoffen mit niedrigerer Zähigkeit und zu Reaktorbaulinien mit wesentlich höherer Neutronenbestrahlung. Mit den gewonnenen Daten, Erfahrungen und Erkenntnissen konnte auch ein wesentlicher Beitrag zur Einarbeitung des Master-Curve- Konzepts in das deutsche Regelwerk geliefert werden. 25 Jahre Kernkraftwerk Isar 2 (Page 304) E. Fischer und M. Luginger Das Kernkraftwerk Isar 2 (KKI 2) nahm am 9. April 1988 seinen kommerziellen Leistungsbetrieb auf. Seither sind 25 Jahre vergangen, in denen das Kraftwerk insgesamt ca. 285 Mrd. kwh Strom erzeugt hat. Die jährliche Stromerzeugung des KKI 2 von etwa 12 Mrd. kwh entspricht einem Anteil von ca. 15 % an der gesamten bayerischen Stromerzeugung. Mit dieser Strommenge könnten rechnerisch ein Jahr lang rund 3 Millionen 3-Personen-Haushalte versorgt bzw. 2/3 des Strombedarfs der bayerischen Industrie gedeckt werden. Während 282

4 Content in brief (German) des 25-jährigen Leistungsbetriebs erzielte das Kernkraftwerk Isar 2 bisher insgesamt 9-mal die höchste Jahresbruttostromproduktion aller Kernkraftwerke weltweit. Zur Erreichung einer derart beeindruckenden Leistungsbilanz bedarf es einer überdurchschnittlich hohen Verfügbarkeit der Anlage. Diese wiederum basiert auf kurzen Revisionszeiten und einem störungsfreien Betrieb der Anlage. Hohe Sicherheit und Verfügbarkeit der Anlage sind jedoch keine Selbstverständlichkeit, sondern das Ergebnis des verantwortungsvollen und sicherheitsgerichteten Anlagenbetriebs in Verbindung mit kontinuierlicher Anlagenoptimierung sowie permanenter Durchführung umfassender Prüfungen, Inspektionen und Instandhaltungsmaßnahmen. Neben der Anlagentechnik wurden auch Organisation und Administration permanent weiterentwickelt und sich ändernden Anforderungen angepasst, um einen zuverlässigen, sicheren und umweltfreundlichen Betrieb der Anlage sicherzustellen. Nutzen des Alterungsmanagements in deutschen Kernkraftwerken Aspekte der KTA 1403 für den restlichen Leistungsbetrieb, Nachbetrieb und Restbetrieb (Page 308) P. Barreiro, Th. Bever, G. Brast, B. Elsche, P. Grossmann, F. Hüttner, Th. Linnemann, S.H. Reese, S.-O. Smit, M. Widera und R.-M. Zander Das Alterungsmanagement (AM) in Kernkraftwerken hat seinen Ursprung in den USA und wurde in Deutschland ab Ende der 1990er-Jahre thematisiert. Auf der Grundlage der bereits anlagenspezifisch vorhandenen, umfassend praktizierten Maßnahmen wurde die KTA 1403 Alterungsmanagement in Kernkraftwerken erarbeitet und im Jahr 2010 finalisiert. In der vorliegenden Veröffentlichung wird zuerst der Kontext des AM in Bezug auf die deutschen Kernkraftwerke geschildert und sowohl auf nationale als auch auf internationale historische Entwicklungen eingegangen. Vor diesem Hintergrund wird anschließend die Abgrenzung des Alterungsmanagements vom Lebensdauermanagement diskutiert. Danach wird der Status des AM-Prozesses in den laufenden Kernkraftwerken beschrieben, insbesondere die organisatorische anlagenspezifische Umsetzung. Durch den Beschluss der Bundesregierung, die friedliche Nutzung der Kernenergie in Deutschland zu beenden, und die damit einhergehende 13. Novelle des Atomgesetzes vom (13. AtG- Novelle) hat ein beträchtlicher Teil des deutschen Kernkraftwerksparks die Berechtigung zum Leistungsbetrieb bereits verloren. Vor diesem Hintergrund werden Aspekte des AM für Anlagen in der Nichtleistungs-, Nachbetriebs- und Restbetriebsphase erörtert. Abschließend werden die Erfahrungen mit der anlagenspezifischen Umsetzung des AM-Prozesses unter Anwendung der KTA 1403 reflektiert und zusammengefasst. Reaktorsicherheitsforschung in Zeiten des Wandels (Page 316) R. Zipper, Seit Anfang der 1970er-Jahre hat die vom Bundesministerium für Wirtschaft und Technologie und dessen Vorgängern geförderte und unabhängig von Interessen der Industrie oder von Industrieverbänden ebenso wie von aktuellen Genehmigungsfragen durchgeführte Reaktorsicherheitsforschung erheblich zur Erweiterung der Kenntnisse über Risiken und mögliche Bedrohungen im Zusammenhang mit dem Betrieb von Kernkraftwerken beigetragen. Die Ergebnisse dieser Forschungsaktivitäten haben bei Industrie und Energieversorgungsunternehmen einige Maßnahmen zur weiteren Verbesserung der international anerkannten hohen Sicherheitsstandards der Kernkraftwerke in Deutschland ausgelöst. Außerdem haben durch die Einbeziehung vor allem von Hochschulen in die hervorragenden Forschungsaktivitäten zahlreiche Nachwuchswissenschaftler die Möglichkeit bekommen, sich auf den Gebieten Reaktortechnik und Sicherheit zu qualifizieren und damit zur Kompetenzerhaltung während des demographischen Wandels beizutragen. Der Ausstieg Deutschlands aus der Nutzung der Kernenergie wirkt sich auch auf Fragen der Reaktorsicherheitsforschung in Deutschland aus. Während Deutschland die Nutzung der Kernenergie zur öffentlichen Energieversorgung immer weiter zurückfährt und schließlich einstellt, setzen andere Länder in Europa und anderswo auf der Welt die Expansion oder sogar den Neubeginn der Kernenergienutzung fort. Wie allgemein anerkannt, ist nukleare Sicherheit eine internationale Angelegenheit. Im Gefolge der Katastrophe von Fukushima gibt es einige Initiativen zur Einführung eines international verbindlichen Systems von Sicherheitsvorschriften und Richtlinien. Der Deutsche Kompetenzverbund hat dazu einen Rahmen von Bereichen ausgearbeitet, in denen auch in Zukunft Reaktorsicherheitsforschung erforderlich ist, um die deutschen Bemühungen zu unterstützen, auf Grund eigenen und unabhängigen Fachwissens die höchsten Sicherheitsstandards zur Nutzung der Kernenergie für die Energieversorgung im In- und Ausland stetig weiter zu entwickeln und einzuführen. Deutsche Kerntechnik ist Hightech (Page 322) J. Reinartz Trotz des politisch beschlossenen Ausstiegs Deutschlands aus der Stromerzeugung mit Kernenergie wird kerntechnisches Knowhow auch weiterhin benötigt. Und dies nicht nur für den Rückbau der kerntechnischen Anlagen und die Endlagerung radioaktiver Abfälle, sondern insbesondere auch im Bereich der Sicherheitstechnik. Wir brauchen über Jahrzehnte nach dem Abschalten des letzten Kraftwerks in Deutschland noch kerntechnisches Knowhow und weiterhin innovative Lösungen und das dafür gut ausgebildete Personal, sagte Dr. Ralf Güldner, Präsident des DAtF. Kerntechnik hierzulande ist mehr, als nur die reine Stromgewinnung, so Güldner weiter. Auf der Veranstaltung Energie im Dialog am 12. März 2013 in Berlin standen diese Themen zur Debatte. Hans-Christoph Pape vom Bundesministerium für Wirtschaft und Technologie und Stefan vom Scheidt, Technischer Geschäftsführer und Sprecher der Geschäftsführung der Areva GmbH, nahmen Stellung und erläuterten Status und Perspektiven. atw Vol. 58 (2013) No. 5»atomwirtschaft-atomtechnik«is published monthly by INFORUM Verlags- und Verwaltungsgesellschaft mbh Robert-Koch-Platz 4, Berlin, Germany phone fax Publisher: atw@atomwirtschaft.de Editorial: editorial@atomwirtschaft.com www. atomwirtschaft.de 283

5 NPP Ageing Management German experiences in local fatigue monitoring Elodie Abib, Steffen Bergholz and Jürgen Rudolph, Erlangen/Germany I. Introduction During the early operation of NPPs in the 1970s and 1980s local loads occurred at different locations causing damaging fatigue cracks. These were either due to new loading conditions which were not considered in the design phase (e.g. temperature stratification) or insufficient manufacturing quality (e.g. welded joints). These problems constituted the starting signal for the development of fatigue monitoring systems. In Germany, the fatigue monitoring system (FAMOS) was for instance developed by then Siemens/KWU at the end of the 1980s and installed in German NPPs. This way, the compliance with authority demands was assured. This background explains that in Germany, fatigue evaluation is based on decades of experience and regulatory requirements. The rule KTA establishes the rules for qualified fatigue monitoring. Recently, a new rule DIN on fatigue monitoring systems is being compiled. In practice, the systematic measurements of local loading effects at fatigue relevant locations in NPPs were enabled. These experiences gave rise to a better understanding of the ongoing loading phenomena. In recent years, the ageing management of NPPs has gained an increasing importance. The reasons are mainly due to the international context of extending periods of new plants operation (60 years) and lifetime extension projects. Moreover, applicable fatigue design rules are tightened particularly with respect of considering the environmentally assisted fatigue (EAF) effect e.g. by means of fatigue environmental factors (F en ). Fatigue monitoring during plant operation usually shows significant differences between design transients (based on design data before plant operation) and the Addresses of the authors: Elodie Abib, Steffen Bergholz, Jürgen Rudolph AREVA NP GmbH Henri-Dunant-Strasse Erlangen, Germany real transients respectively idealized model transients (based on measured data during plant operation). It is a straightforward strategy to compare design transients and model transients in order to show the validity of design data in the sense of conservative covering. Experience has shown significant differences between design and model transients. These differences are mainly revealed by means of continuous local measurements at fatigue relevant locations, e.g. a local fatigue monitoring strategy. Global fatigue monitoring strategies have to rely on operational signals and some transfer functions. The operational instrumentation was not originally intended for fatigue monitoring. Usually, it is neither located near the fatigue relevant positions nor it is able to resolve highly transient phenomena due to its own measuring dynamics characteristics. Hence, while addressing fatigue monitoring requirements, a clear distinction has to be made between global and local approaches. II. Local fatigue phenomena observed in NPPs Cyclic loading conditions in NPPs are predominantly of thermal cyclic character with operational thermal transients of different temperature change rates at maximum temperatures of about 350 C. The predominance of thermal cyclic loadings due to the operation of NPPs as well as relevant stress/strain amplitudes in the low cycle fatigue (LCF) regime is a strong argument in favor of the application of an appropriate monitoring strategy. In contrast, fatigue is a highly local phenomenon requiring knowledge both of the local geometry of relevant components and the local loading conditions. Local fatigue monitoring is located at fatigue relevant locations at the outer surface of pipes and is based on additional temperature measurement by means of thermocouples. The complex fluid flow events occurring during the operation of NPPs are influenced by the automatic operational control processes. Nevertheless, as a consequence of the manifold manual intervention opportunities equal technological processes may induce different local loading sequences for the components. In other words, an assessment of components exclusively based on operational measuring instrumentation in connection with transfer functions does not guarantee the detection of local load phenomena. Thus, local data acquisition and monitoring of local loads at the fatigue relevant components is the appropriate solution. Operating experience shows that the fatigue ageing mechanism is normally due to cold and hot feed operations. These operations have to be monitored and recorded in order to provide for a qualified fatigue check based on realistic load data. Local effects such as the swapping flow after feeding interruption can only be recorded in the load data set this way. It is to be pointed out that the safety check against cyclic loads of the components has to be a permanent operation accompanying procedure. There are examples of the German fatigue monitoring experience where the fatigue assessment induced the necessity of retrofitting of components or the modification of the operating modes due to the results of local fatigue monitoring. For instance, the feedwater sparger of the steam generator was subsequently designed in a way that the stresses of cyclically occurring stratification transients were minimized. This raises the question how complex stratification flows in NPP components can be detected and properly recorded for subsequent fatigue assessment. Note that even the local instrumentation for local fatigue monitoring may not give all required information for a substantiated fatigue assessment in case of highly complex loading situations. Complex operational surge line loadings are an according example. In these cases, the load determination process should be assisted by detailed computational fluid dynamics (CFD) calculations. Even in these rare cases, the local load measurements of the fatigue monitoring program constitute valuable input data for realistic boundary conditions of the CFD analyses. As an example, the components of the spray line of a pressurized water reactor are well-known fatigue monitoring candidates and are usually subject of detailed code-based fatigue analyses. They may be subject of thermal shock like transients with high cold-hot temperature differences and stratification flows may occur. The load input for detailed fatigue calculation (DFC) should cover the temperature transients measured during operation based on local fatigue monitoring. Figure 1 shows an example of a model transient, represented by the green area in Figure 2, characterized by the temperature difference as a function of time, the temperature change rate, the 284

6 NPP Ageing Management 3 components were generated and integrated into an overall shell type model of the spray line. In the course of the DFC the transient temperature fields have to be analyzed for all relevant model transients (as for example Figure 1). These transient temperature fields are the input data for the subsequent transient linear or nonlinear structural mechanical analyses yielding the local stress and strain ranges for the fatigue assessment, according to the requirements of the design codes. An example of the temperature distribution in the T-joint is shown in Figure 3. It represents one point of time of one model transient during a cold water injection process. III. German local fatigue monitoring tools Fig. 1. Example of a model temperature transient. The rule of the German nuclear code KTA is recommending the implementation of monitoring solutions in NPPs to check fatigue phenomena. In this frame, the mechanical analyses department of Areva NP developed over the years a global management concept for NPPs, the socalled AFC. This multidisciplinary fatigue approach was further expanded and 2 modules of AFC evolved: FAMOSi and the FFE. Later on, these two recently developed modules will be described in details. III.A. AREVA Fatigue Concept (AFC) Fig. 2. Modules of the AREVA fatigue concept (AFC). hold time and a possible offset of the 12 o clock position due to stratification. Note that this transient specified for the fluid (bulk temperature) in connection with an appropriate heat transfer coefficient conservatively covers the measured loads with respect of the mentioned parameters and the frequency of occurrence. Although it might be similar to a specified design transient, represented by the blue area in Figure 2, the crucial differences are the underlying measured operational loads. The idealization of measured loads in specified model transients and an additional conservative grouping are characteristic for DFC with regard of calculation efforts, particularly in the case of the application of general elasto-plastic analysis. The fast fatigue evaluation (FFE) process offers an option of direct load data processing without derivation of model transients. In that case, the underlying fatigue analysis is simplified elasto-plastic. All in all it is obvious that the realistic load specification takes a decisive influence on the fatigue usage calculation. In the given example of a spray line, the nozzle, the spool, and the T-section were identified as fatigue-relevant components within the spray line system. In the context of the DFC, detailed brick type FE models of these The AFC is a global tool developed in Germany gathering several kinds of modules, in order to propose plant operators a solution to cope with fatigue problems and to optimize the availability of the plant. The different modules of AFC are represented in Figure 2. According to the international nuclear codes, such as ASME, RCC-M and KTA, requirements on fatigue analyses are given and fatigue studies shall be performed in the design and operation phases of plants. To cope with this requirement, several methods are given within the AFC. These methods rely on some real measurements performed at the heart of the Fig. 3. Exemplary temperature distribution in the T-joint. 285

7 NPP Ageing Management zones prone to fatigue. Indeed, some sensors are implemented on the selection of locations, which are judged to be sensitive by plant experts. These locations depend on the type of plant and also on the needs of the operators. As an example, on the EPRTM, the monitoring selection is focused especially on the following locations: Primary loops Surge line Spray lines Chemical and volume control system Safety injection system Feedwater system Emergency feedwater system Up to now, the fatigue monitoring system was integrated in the Teleperm XP (TXP) technology of the plant instrumentation and control (I&C). The TXP system is used in order to pilot the I&C environment inside the plant. In this way, plant measurements are retrieved and broadcast on the process information and control system (PICS). For fatigue monitoring, plant operators required a stand-alone solution being able to retrieve measured data and analyze them online, so as to quickly get fatigue status. Therefore, the new FAMOSi solution was designed with brand-new properties. FAMOSi consists now of an economic solution offering online functions to match operators expectations in terms of thermal fatigue compliance, time and money saving. III.B. Fatigue Monitoring System integrated (FAMOSi) The brand new AFC fatigue module, FAMOSi, offers a revolutionary monitoring solution, optimized via a fieldbus. This high performance system is collecting measurement via some sensors. The sensors of the monitored locations are implemented all around the pipe sections and are led through extension cables to junction boxes. The junction boxes are gathering the extension cables of several measurement sections together in trunk cables. The trunk cables are led to the information modules (IM), located inside the containment. The IM are led most of the time via fiber optic to the processing unit (PU). For this stage, the implementation of a fieldbus all around the inside of the containment enables to connect several IM to a single PU. The PU transfers the acquired digitized data outside the containment to the server unit (SU), via fiber optic cable. Finally, the SU is saving and archiving the database. In this way, it is possible to directly display the measurements. Moreover, FAMOSi offers a new user interface via the FAMOSi software. This one enables plant operators to have the results of the data posttreatment very quickly, and to get online a fatigue status of the selected components. 286 Figure 4 and Figure 5 give an overview of FAMOSi location in the reactor building (RB). The IM and PU cabinets are optimized to use a limited place in the containment. Their reduced dimensions make them very easy to install and to locate inside the containment, as shown in Figure 6. The system layout embraces the whole measurement chain from the sensors to the SU. The new technology of the system takes a real advantage of having its 3 types of dedicated cabinets IM, PU and SU: IM are getting the signals from the junction boxes, digitizing the data and transferring them to the PU via the fieldbus. The number of IM depends on the number of sections to be monitored; thus several IM can be implemented in the RB. PU is broadcasting live data to all connected network stations and providing power supply to the IM; one single PU is needed in the RB. SU is gathering the data on the SQL database in the control room. The different cabinets are connected via a fieldbus composed of fiber optic cables. The fiber optic cables are chosen for their lightness and high efficiency in transmitting data. This is an economic solution for cable material. Moreover, it enables to reduce the amount of cables and of cable trays within the containment, to reduce the amount of wall penetrations and to save erection and maintenance time. With this new layout, shown in Figure 7, one single wall penetration is needed to make the fiber optic junction between PU and SU. Once acquired, the data are post processed via the FAMOSi software. This user interface offers a large range of possibilities to post process the plant measurements. Miscellaneous modules are implemented to elaborate different types of data evaluations, such as: Plausibility check Matrix classification of ranges and frequencies Fig. 4. FAMOSi locations of measurement sections. Fig. 5. FAMOSi location in the RB.

8 NPP Ageing Management The measurement sections of FAMOSi are installed at the outer surface of the pipe. FFE calculates the corresponding inner surface temperature as indicated in Figure 9. The thermal load cycles are well known after that step and the stress time history are calculated with the Green s function approach via unit transients of +/-100 K, in order to scan the original temperature time history at each time step. By means of unit transients, stresses are calculated for all monitored locations. After the calculation of the equivalent stresses, the mechanical load cycles can be classified by the use of the Rainflow algorithm. Then, comparisons with the fatigue curve results in fatigue levels are performed for all relevant locations and the method provides the cumulative usage factors (CUF). The steps are summarized in Figure 10. The great advantage of the FFE method is its high-speed calculation time to assess CUF. This enables to display via the FAMOSi software a real time fatigue evaluation and to calculate an enveloping fatigue level. Finally, this method takes into account the environmental effects described in the report NUREG/CR-6909 ANL-06/08 Effects of LWR Coolant Environments on the Fatigue Life of Reactor Material via EAF and by application of Fen. Fig. 6. Size of FAMOSi cabinets. IV. Benefits of local monitoring for long term operation In Germany, there is no discussion about local versus global fatigue monitoring, since the German nuclear rules require a local monitoring system, see following extract from KTA : Fig. 7. FAMOSi layout. Calculation of inner wall thermal field Calculation of equivalent stress field Calculation of cumulative usage factors Figure 8 gives an overview of the software possibilities. The currently developed software can evolve and add some new functions depending on the plant operators expectations. Its usage can also be extended to post-process some other plant data and monitoring solutions. An important module offered by the software consists of the revolutionary method FFE, described in details in the paragraph below. 9.2 Monitoring of loadings side surface of a pipe and can evaluate the fatigue level of the component for different thermal loads (plug flow, stratification) Monitoring of quasi-static mechanical and thermal loadings (1) It shall be ensured that temporal and local temperature changes relevant to fatigue III.C. Fast Fatigue Evaluation (FFE) For highly loaded components, the detailed method FFE, can be used to calculate the load factors in a realistic way. This method uses FAMOSi measured data from the out- Fig. 8. FAMOSi software architecture. 287

9 NPP Ageing Management measurements is still open or without any evidence. The above described local fatigue approach will give the answers. Additionally, a lot of other advantages, e.g. functionality observation of valves, can be archived. However, the main point is that the most realistic knowledge of local loads leads to the most realistic stress history and consequently, to the most realistic actual fatigue usage. Nowadays, the new NPPs are designed for 60 years of service operation, followed by potential discussions of further operation, after 60 years. It means that three generations of employees shall operate these power plants. Therefore, getting knowledge about local loads is very important and shall be monitored by the best standard monitoring solution possible to erect in NPPs: FAMOSi. The current changes in calculation rules taking into account EAF effects show also that rules are constantly evolving and tightened, especially over such a long operating period. The local knowledge of loads and fatigue usage also leads to a strongly improved maintenance philosophy, since comparable results will exist on different points of interest (POI). Non destructive tests (NDT) can be performed at the most relevant positions, correlated to actual usage factors. A real time-benefit between 2 NDT campaigns can be gained from a low-load service operation. The follow up of fatigue usage at relevant POI gives at an early stage real possibilities to improve and optimize operating modes, aiming at reducing local loads. Moreover, the value of investment will increase and the risk of unplanned outages is significantly reduced. Finally, the fact that, German NPPs are the ones with the highest specific electricity production in the world is also due to this useful sustainable local monitoring concept. Fig. 9. Thermal field transfer from outer wall to inner wall. V. Conclusions Fig. 10. FFE calculation steps. are monitored by a sufficiently dense net of measuring points of the standard instrumentation. When selecting the measuring points the effects of the mode of operation (little mass flows, indifferent pressure conditions, switching operations, temperature differentials) and the design (pipeline installation, isolating function of valves) shall be taken into account (2) Where thermal stratification is expected to occur, the temperature measuring points shall be located such that all relevant loading variables across the pipe cross-section and axially to the pipe run can be measured. Indeed, the reasons for such a strong requirement are described in the chapter quoted above and extracted from the German nuclear code. Recent developments of calculation rules for fatigue (e.g. EAF rules to determine the factor F en ) show that the knowledge of local loads is necessary to prove the fatigue state at these positions. If only some global measurements are implemented, based on in-service installation, the answer to the question of the correctness of local loading assumptions and global The strategy of employing local fatigue monitoring is a straightforward solution enabling the direct measurement of loads on the zones sensitive to fatigue. Therefore, in the context of power plants lifetime extension, and to overcome the potential differences existing between the design transients and the real occurred transients during plant operation, Areva developed 2 new modules in the frame of AFC. FAMOSi is a revolutionary stand alone system which can easily be installed and embedded, if wished, in the I&C frame of operating plants. With its new layout and new user interface, the measurements can be displayed and the corresponding fatigue status of sensitive components can be calculated online. Indeed, the FFE method allows online an extremely fast and precise calculation of CUF. In the end, a local fatigue monitoring approach, widely supported by measured data, is the basis for decisions of optimized operating modes. Thus, it influences the fatigue usage factors, helps to comply with nuclear codes and to operate plants in an economic and efficient way. Acknowledgments The authors wish to express special thanks to all contributors to the AFC within Areva. Nomenclature AFC Areva Fatigue Concept ASME American Society of Mechanical Engineers CFD Computational Fluid Dynamics CUF Cumulative Usage Factor DFC Detailed Fatigue Check DIN Deutsches Institut für Normung EAF Environmentally Assisted Fatigue FAMOS Fatigue Monitoring System FAMOSi Fatigue Monitoring System integrated 288

10 NPP Ageing Management FE F en FFE KTA LCF NDT Finite Element Fatigue environmental factors Fast Fatigue Evaluation Kerntechnischer Ausschuss Low Cycle Fatigue Non Destructive Testing NPP PICS POI PWR I&C Nuclear Power Plant Process Information and Control System Point Of Interest Pressurized Water Reactor Instrumentation and Control IM Information Module PU Processing Unit RB Reactor Building SQL Structured Query Language SU Server Unit TXP Teleperm XP Fatigue evaluation including environmental effects for primary circuit components in nuclear power plants Johannes Seichter, Dresden /Germany Sven H. Reese and Dietmar Klucke, Hannover/Germany Introduction For several years environmentally assisted fatigue (EAF) has been an object of investigation in the context of designing and evaluating NPP (nuclear power plant) components lifetime. The publication of the US- NRC Regulatory Guide [1] in 2007, defining a guideline for evaluating fatigue analyses for new builds incorporating the life reduction of metallic components due Addresses of the authors: Johannes Seichter Head of Analyses Siempelkamp Prüf- und Gutachter-Gesellschaft mbh Am Lagerplatz 6a Dresden/Germany Dr. Sven H. Reese and Dietmar Klucke Component Technology Global Unit Generation E.ON Kernkraft GmbH Tresckowstraße Hannover/Germany to effects of the LWR (light water reactor) coolant environment, brought the topic into the international discussion. In this guideline the report NUREG/CR-6909 by [2] based on the investigations at Argonne National Laboratory (ANL) in the USA was referenced, containing a new and comprehensive set of laboratory fatigue data both in air and under LWR conditions for some relevant ferritic and austenitic stainless steel grades as well as Ni-Cr-Fe alloys. These studies were based on materials relevant to ASME in general, while general transferability of the effect itself cannot be excluded for piping material relevant to German NPPs. There are 2 main results of these investigations: the first results are updated fatigue mean-data-curves in air and following updated fatigue-designcurves for the same environment as well. In this article, this topic is not further addressed. But the second result is a calculation procedure based on these data and presented in [2] defining an empirical numerical correction factor, the so called F en, being a main topic of this paper. For the sake of completeness, there was a couple of preceding publications, dealing with this lifetime influencing effect, [3] to [5] Ageing Management for instance. The calculation procedure according to [2] has been added to ASME code, Section III, Division 1 as Code Case N-792 [6]. The fact that this Code Case has not been officially endorsed by the US NRC ([10]) underlines that discussions have not been terminated. But not only ASME, other regulatory codes take the EAF phenomena into account, too, e.g. JSME [7] or KTA [8]. Despite the fact that [2] focuses on new builds only, an intensive discussion in terms of application to currently operating reactors in the framework of the KTA was started in Germany after the publication of the report. The calculation procedure as presented in [2] has been updated several times in the meantime ([9 to 11]), but the general approach, quantifying the phenomena of EAF by the F en correction factor, is unquestioned. While the consideration of EAF in the framework of the US-NRC Regulatory Guide [1] is limited to new reactor builds only, in Germany the application to nuclear reactors being currently in operation is state of the art and practice. In the German KTA rules (e.g. Standard No [8]), the environmentally-assisted fatigue (EAF) is under consideration in terms of the introduction of so-called attention thresholds for the cumulative fatigue usage factors (CUF). It is defined that, if the CUF, calculated without consideration of EAF, exceeds the threshold, further measures like for example non-destructive in-service inspections including fracture mechanical analysis or detailed assessment procedures (e.g. as described within [2, 9, 10 or 11]) have to be taken into account. The application of sophisticated and more detailed procedures with respect to fatigue calculation is a feasible approach to take these topics of the KTA into account by reducing the CUF below the attention thresholds. Based on the conservatisms in fatigue calculations, a couple of effective methods to numerically reduce the degree of conservatism are presented. Additionally examples will be given in terms of a practical application to NPP components taking into account measured operational data. 289