Inner surface of most of primary circuit components in PWR/BWR/WWER type reactors is covered by austenitic cladding that serves primarily as anticorrosion protection. This is also supported by the requirements for stress analysis of the vessel by most of the codes — austenitic cladding is not taken into account in the calculation of vessel wall thickness and on allowance of stress intensities for operating conditions. Its effect is taken, in some codes like for WWER components, in fatigue calculation and also for evaluation of vessel resistance against fast fracture during pressurized thermal shock (PTS) events.

Reactor internals are components that are no typical pressure boundary but they are nevertheless very important as they hold fuel elements and all reactor control system elements and thus must ensure their safe and reliable operation during the whole reactor life under all operating and even beyond bases regimes. In principle, reactor internals can be replaced but their weight, quantity of very high activated material and cost such possibility practically excluded. Thus, evaluation of the reactor internals condition and prediction of their behavior during the whole or even extended lifetime is of high importance. Reactor internals are subjected to very high neutron irradiation that could initiated not only stress corrosion (irradiation assisted) cracking but also large embrittlement and changes in dimensions (swelling and creep). VERLIFE – “Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation” was initiated and co-ordinated by the Czech and was developed within the 5th Framework Program of the European Union in 2003 and later upgraded within the 6th Framework Program “COVERS – Safety of WWER NPPs” of the European Union in 2008. This Procedure had to fill the gap in original Soviet/Russian Codes and Rules for Nuclear Power Plants (NPPs) with WWER (Water-Water-Energetic-Reactor = PWR type) type reactors, as those codes were developed only for design and manufacture and were not changed since their second edition in 1989. VERLIFE Procedure is based on these Russian codes but incorporates also new developments in research, mainly in fracture mechanics, and also some principal approaches used in PWR codes. Within the last upgrading and principal extending of this VERLIFE Procedure was developed within the 3-years IAEA project (in close co-operation with another project of the 6th Framework Program of the European Union “NULIFE – Plant Life Management of NPPs”) that started in 2009 with final approval and editing in 2013”) a part dealing with the evaluation of reactor internals lifetime was elaborated.. This IAEA VERLIFE procedure for internals has been implemented into the existing Normative Technical Documentation (NTD) ASI (Czech Association of Mechanical Engineers), Section IV – Evaluation of Residual Lifetime of Components and Piping in WWER type NPPs. Main damaging mechanisms that should be taken into account in reactor internals and the procedure are described in detail with necessary formulae for materials of internals: - Radiation hardening - Radiation embrittlement - Radiation swelling - Radiation creep - Swelling under stress effect - Swelling inducing embrittlement - Irradiated assisted stress corrosion cracking - Transformation austenite-ferrite and also the method for evaluation of the resistance against non-ductile failure of postulated defect. The paper will describe these main principles and also more detailed information on the procedure for evaluation of reactor internals will be given.

Small punch test specimens are widely used for a long time as they are simple to produce and requires only a small volume of material. This fact is advantageous especially for high activity materials but also for assessment of operational damage in components materials when component integrity and strength may not be affected. In the same time, no test standard exists and several different specimen types and test procedures have been developed in different place. Thus, to unify this activity, considerable attention has been paid since 2012 to the standardization of small punch test technique within the American Society of Testing and Materials (ASTM). In 2016 a large InterLaboratory Study has been launched within the ASTM subcommittee E10.02 - Behavior and Use of Nuclear Structural Materials, involving 12 laboratories and 6 evaluated structural materials from the nuclear and non-nuclear power plant components. Paper describes the current status of ASTM standardization, results of the InterLaboratory Study, first analysis of the results with respect to some important test parameters, lessons learned and open questions remaining to be solved for the successful completion of the standardization process.

VERLIFE – “Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation” was initiated and co-ordinated by the Czech and was developed within the 5th Framework Program of the European Union in 2003 and later upgraded within the 6th Framework Program “COVERS – Safety of WWER NPPs” of the European Union in 2008. This Procedure had to fill the gap in original Soviet/Russian Codes and Rules for Nuclear Power Plants (NPPs) with WWER (Water-Water-Energetic-Reactor = PWR type) type reactors, as those codes were developed only for design and manufacture and were not changed since their second edition in 1989. VERLIFE Procedure is based on these Russian codes but incorporates also new developments in research, mainly in fracture mechanics, and also some principal approaches used in PWR codes. Last upgrading and principal extending of this VERLIFE Procedure was realized within the 3-years IAEA project (in close co-operation with another project of the 6th Framework Program of the European Union “NULIFE – Plant Life Management of NPPs”) that started in 2009 with final approval and editing in 2013. As all versions of the VERLIFE procedure were coordinated by the Czech and first version was based on the Czech version of the NTD ASI, there have been simultaneously incorporated into the Czech NTD ASI (Normative Technical Documentation of the Czech Association of Mechanical Engineers) guidelines that are accepted by the Czech State Office for Nuclear Safety for the use in evaluation of Czech NPPs. This document has several parts: Section IV – “Evaluation of Residual Lifetime of Components and Piping in WWER type NPPs” deals with the evaluation during NPP operation. Main part of the document is divided into four main parts: - Evaluation of resistance of components and piping against non-ductile failure - Evaluation of resistance of component and piping against fatigue damage, - Evaluation of resistance of components and piping against corrosion-mechanical damage - Evaluation of residual lifetime of components and piping with defects found during in-service inspections Additionally, several appendices are included for detailed description of individual parts of evaluation, e.g. - Determination of neutron fluences in reactor pressure vessel and internals - Determination of degradation of materials during operation - Requirements for evaluation of pressurized thermal shock regimes - Evaluation of corrosion-errosion effects in piping - Environmental fatigue evaluation - Evaluation of reactor pressure vessel failure probability Finally, the following appendices dealing with components integrity have been included: - Lifetime of reactor pressure vessel internals - Leak-before-break concept for WWER piping - No-break-zone for WWER piping The paper will describe structure and main principles of this Section IV.

Reactor pressure vessel integrity and lifetime evaluation is based on the use of fracture mechanics apparatus but most of the material vessel material data and their degradation during operation are based on results from Charpy V-notch impact tests. Then, empirical correlations between transition shift of temperature dependence of notch toughness and fracture toughness are applied. Elaboration of „Master Curve“ approach for fracture toughness experimental data analysis allows to use fracture toughness data directly to the reactor pressure vessel integrity evaluation. Wider use of this approach is limited by the lack of appropriate database from surveillance specimen test data, as mostly only Charpy impact specimen are included into the Surveillance specimen programs. Fortunately, all WWER Surveillance programs contain also fracture toughness specimens, either pre-cracked Charpy size or CT-0.5. Thus, database of fracture toughness data from Surveillance programs of WWER-440/V-213C type reactor pressure vessels, operated in the Czech Republic, Slovakia and Hungary and manufactured only by one manufacturer - SKODA JS, was collated and analyzed. These vessels were manufactured from 15Kh2MFAA type steel and appropriate weld metal, both of Cr-Mo-V type with low content of detrimental impurities — P and Cu. Analysis of the data in fluence interval up to 6×10 24 m −2 (with neutron energies E n larger than 0.5 MeV) show that transition temperature shifts in fracture toughness temperature dependence are higher than for Charpy impact tests. Several formulae have been applied for fitting these shift dependencies with chemical composition of materials and finally new Embrittlement Trend Curves for Charpy shifts have been corrected. Additionally, new Embrittlement Trend Curves for fracture toughness shifts based on “Master Curve” approach have been also proposed. Both trends are using simple power law on fluence with exponents around 0.6 and depend on phosphorus and copper contents even though effect of other elements has been also checked.

The AGE60+ project aimed towards the application of ageing related data bases and methodologies to ensuring the safe operation of Light Water Reactors beyond 60 years, and the longer-term safe storage of waste. It acted by encouraging European researchers to share data in order to maximize its utility. As an initial step, the AGE60+ consortium considered the development and utilization of a number of databases set up to collate data and derive information relevant to the management of nuclear power plant. This review was used to identify the factors which lead to a database successfully making a useful contribution. The AGE60+ project has been a very fruitful collaboration. It has: • Raised the concept of sharing data and setting up databases related to the ageing of concrete in waste disposal, and received a generally positive response to this first step. • Identified an area in which a new database would provide the opportunity to advance predictive modeling of the degradation of internals. • Collated and reviewed data on the thermal ageing of low alloy steels. • Identified information gaps in the area of internals degradation and the thermal ageing of low alloy steel components. • Produced an up-to-date database on WWER-440 RPV steel embrittlement. • Produced a new embrittlement trend curve (ETC), relating the fracture toughness transition temperature to fluence and composition for WWER-440 weld and base metal. • Expanded and utilized a database on MnMoNi steel embrittlement. This AGE 60+ project was realized as a part of the NUGENIA+ Call and lasted 18 months with participating organizations: UJV Rez a.s., Czech Republic, AREVA GmbH, Germany, CIEMAT, Spain, MTA EK, Hungary and NNL, UK as a leading organization.

Reactor pressure vessels of PWR/BWR/WWER type reactors are covered by austenitic cladding made by welding on their inner wall. Austenitic materials usually have no transition temperature behavior as they have fcc crystallographic structure. But, austenitic cladding made by welding contain usually up to 8 % of delta-ferrite that results in some transition behavior of fracture properties. This transition can be observed in temperature region below room temperature. Surprisingly, this transition behavior in static fracture toughness of both cladding layers can be well described by “Master curve” approach. Results from testing austenitic cladding for WWER type reactors will be shown and discussed, ether in unirradiated as well as irradiated conditions — only small changes in fracture toughness properties in this transition region are observed as a result of irradiation.

Evaluation of integrity and lifetime of reactor pressure vessels is usually based on fracture mechanics approach using empirical correlation between transition temperatures from impact tests and static fracture toughness test results in the form of “design curve”. Moreover, material degradation during operation is also usually monitored by impact surveillance specimen testing under the assumption that shifts in temperature dependencies if impact toughness and static fracture toughness are the same. To verify this assumption, study of the correlation between these two shifts has been performed on WWER steels — 15Kh2MFA (Cr-Mo-V) and 15Kh2NMFA (Ni-Cr-Mo-V) types. Several sources of results have been used : (a) reconstitution of tested remains of Charpy V-notch impact test specimens from irradiated programs was performed to obtain pre-cracked Charpy size specimens for three point bending type fracture toughness testing, (b) comparison of tests results from surveillance programs irradiated by similar fluences, (c) experimental irradiation programs with accelerated irradiation in research reactor. Additionally, some results from the recent study of irradiation embrittlement of high nickel weld are included — its behavior shows to some extraordinary tendency. Thanks to the use of reconstitution, both series of specimen were irradiated under the same conditions — temperature and neutron fluence and comparison is reliable. Results show that transition temperatures from fracture toughness testing are larger than those from Charpy impact tests. Similar results have been obtained also for other two groups of results.

Special weld of 15Kh2MFA (Cr-Mo-V) type steel was investigated in several laboratories within the IAEA Coordinated round robin exercise. Specimens were irradiated by several fluences and impact tests of standard Charpy size and 5×5 and 3×4 mm specimens were then tested to determine appropriate transition temperatures and their shifts. When applied normally used criteria (impact energy, lateral expansion, ductile fracture appearance), shifts in standard specimens and subsize specimens correlate well but not in 1:1 ratio — shifts in subsize specimens were observed to be smaller than in standard ones — the smaller size of specimens, the smaller transition temperature shifts have been obtained. The paper tries to find appropriate criteria for better correlation between shifts in different sizes of impact specimens that can be supported by proper analysis.

Nuclear reactor materials are affected by many stressors during their operation that result either from the nuclear reactions in the reactor core or by operation conditions (temperature, pressure) and water environment. Generally, several different methods of monitoring damage in reactor materials are applied, depending on their type, design and conditions: - Destructive, usually represent by surveillance specimens, - Semi-destructive, usually performed by cutting small specimens from component surface, e.g. for small punch tests, - Non-destructive, representing by non-destructive inspection methods applied during in-service inspections, like traditional ultrasonics, optical, dye-penetrant, eddy current etc. and also some new like automated ball indentation, thermoelectric power measurements etc. The paper summarizes the possibility of application of these methods on main reactor components and shows some typical results and problems.

Integrity and lifetime of reactor pressure vessels are practically determined by their material resistance against fast/non-ductile failure and consequently by their radiation damage resulting in irradiation embrittlement and hardening. Mechanism of irradiation embrittlement of RPV materials depends on selected materials and operation conditions but their values depend in a large extent on reactor design, i.e. on neutron flux/fluence depending on RPV wall. Generally, new RPV design allows smaller neutron fluences but absolute value of irradiation embrittlement still depends on a choice of RPV material. Even though radiation damage (especially irradiation embrittlement) is important for RPV behavior, integrity and lifetime depends, in principle, on final value of applied fracture mechanics parameter — transition temperature. Thus, its initial value as well as its shift due to irradiation embrittlement is of interest but only the embrittlement can be affected during operation. The paper discusses absolute and relative effect of both these parameters for integrity and lifetime assessment for different type of reactors — of old and new as well as the newest design. Then, role of surveillance specimen programs for the evaluation of radiation embrittlement of RPV materials is discussed, first of all from point of view of existing codes and standards. Additionally, some recommendations are given for decreasing the effect of irradiation embrittlement on final integrity and lifetime evaluation.

Integrity of reactor pressure vessels (RPV) are of the most importance for safety of the whole NPP. From all potential regimes, Pressurized Thermal Shock (PTS) regimes during emergency cooling conditions are the most severe and most important. Several nuclear codes are based in similar approaches but their procedures differ and are based on national knowledge and approach to fracture mechanics as well as non-destructive methods of reactor pressure vessel testing. The paper will compare requirements and procedures for PTS evaluation in accordance with RCC-M code in France [2], KTA in Germany [3], Russian original code PNAEG from 1989 [5] and new procedure from 2004 for WWER vessels [4], as well as VERLIFE procedure and IAEA-NULIFE VERLIFE [6] procedure for WWER RPVs and finally ASME Code requirements [1] including US NRC RG approach. Detailed comparison of individual parameters in calculations are compared — material properties, degradation of materials, calculated defects size and form, fracture mechanics approach, warm pre-stressing possibility etc.

Integrity of reactor pressure vessels (RPV) are of the most importance for safety of the whole NPP. From all potential regimes, Pressurized Thermal Shock (PTS) regimes during emergency cooling conditions are the most severe and most important. Several nuclear codes are based in similar approaches but their procedures differ and are based on national knowledge and approach to fracture mechanics as well as non-destructive methods of reactor pressure vessel testing. The paper will compare requirements and procedures for PTS evaluation in accordance with RCC-M code in France [2], KTA in Germany [3], Russian original code PNAEG from 1989 [5] and new procedure from 2004 for WWER vessels [4], as well as VERLIFE procedure and IAEA-NULIFE VERLIFE [6] procedure for WWER RPVs and finally ASME Code requirements [1] including US NRC RG approach. Detailed comparison of individual parameters in calculations are compared — material properties, degradation of materials, calculated defects size and form, fracture mechanics approach, warm pre-stressing possibility etc.

Plant life extension (as well as Long Term Operation) to 60 or 80 years of operation also requires a reliable information about the potential irradiation embrittlement (and also thermal ageing) of reactor pressure vessel materials. Such information is usually obtained from testing specimens within the surveillance specimen program that is designed for the design reactor pressure vessel (RPV) life, regularly for 40 years only. Life extension requires modification of such program (if there is still time to perform it) or a design of a new – extended one. Such program should have to contain RPV archive materials that are not in every case available. Thus, combination of archive materials and possible surrogate materials must be taken into account for this program. Some complication can be expected with thermal ageing data as some laboratory tests at higher temperatures must be realized. The paper describes such program for Nuclear Power Plant (NPP) Dukovany, Czech Republic with WWER-440 type reactors.

VERLIFE — “Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation” was developed within the 5th Framework Program of the European Union in 2003 and later upgraded within the 6th Framework Program “COVERS – Safety of WWER NPPs” of the European Union in 2008. This Procedure had to fill the gap in original Soviet/Russian Codes and Rules for WWER type NPPs, as these codes were developed only for design and manufacturing and were not changed since their second edition in 1989. VERLIFE Procedure is based on these Russian codes but incorporates also new developments in research, mainly in fracture mechanics, and also some principal approaches used in PWR codes. To assure that VERLIFE Procedure will remain a living document, new 3-years IAEA project (in close co-operation with the another project 6th Framework Program of the European Union “NULIFE – Plant Life Management of NPPs”) has started in 2009. Within this project, upgrading/updating of the VERLIFE procedure is prepared together with the extension by (at least) following procedures: - Leak-before-break concept for WWER NPPs; - Reduction of Probability of Break procedure for evaluation of integrity of high-energy piping in NPPs of WWER-440 and WWER-1000 types; - Lifetime of reactor pressure vessel internals; - Risk informed In-service inspection implementation process and organization; - Methodology for Qualification of In-Service Inspection Systems for WWER Nuclear Power Plants; - Component and piping supports; - Monitoring and evaluation of erosion-corrosion damage in piping materials. Final document, after its approval by expert groups of the IAEA and NULIFE, will be issued as “IAEA/NULIFE Guidelines for Integrity and Lifetime Assessment of Components and Piping in WWER NPPs”. The paper will describe these main principles and also future plans.

Long Term Operation (LTO) to 60 or 80 years of operation also requires a reliable information about the potential irradiation embrittlement (and also thermal ageing) of reactor pressure vessel materials. Such information is usually obtained from testing specimens within the surveillance specimen program that is designed for the design RPV life, regularly for 40 years only. Life extension requires modification of such program (if there is still time to perform it) or a design of a new – extended one. Such program should have to contain RPV archive materials that are not in every case available. Thus, combination of archive materials and possible surrogate materials must be taken into account for this program. Some complication can be expected with thermal ageing data as some laboratory tests at higher temperatures must be realized. The paper describes such program for NPP Dukovany, Czech Republic with WWER-440 type reactors that are now more than 20 years of operation.

VERLIFE – “Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation” was developed within the 5th Framework Programme of the European Union in 2003 and later upgraded within the 6th Framework Programme “COVERS – Safety of WWER NPPs” of the European Union in 2008. This Procedure has to fill the gap in original Soviet/Russian Codes and Rules for WWER type NPPs, as these codes were developed only for design and manufacture and were not changed since their second edition in 1989. VERLIFE Procedure is based on these Russian codes but incorporates also new developments in research, mainly in fracture mechanics, and also some principal approaches used in PWR codes. To assure that VERLIFE Procedure will remain a living document, new 3-year IAEA project (in close cooperation with the another project 6th Framework Programme of the European Union “NULIFE – Plant Life Management of NPPs”) has started in 2009. Within this project, upgrading/updating of the VERLIFE procedure is prepared together with the extension by (at least) following procedures: - Leak-before-break concept for WWER NPPs; - No-Break-Area Assessment Procedure for WWER Piping; - Lifetime of reactor pressure vessel internals; - Probabilistic approach for failure and lifetime assessment of WWER reactor pressure vessels and primary piping (including Risk-informed ISI); - Qualification of NDT procedures; - Supports of components and piping; - Monitoring and evaluation of erosion-corrosion damage in piping. Final document, after its approval by expert groups of the IAEA and NULIFE, will be issued as “IAEA/NULIFE Guidelines for Integrity and Lifetime Assessment of Components and Piping in WWER NPPs”. The paper will describe these main principles and also future plans.

The through-wall attenuation of neutron fluence of reactor pressure vessel (RPV) steels is often expressed using an exponential decay function based on some estimate of displacements per atom (dpa). In order to verify this function, an irradiation project was performed in which 18 layers of Charpy specimens and one central temperatue control layer were stacked in a block to simulate a 190 mm thick RPV wall. Three western-type RPV steels (medium and low copper plates and a high copper Linde 80 flux weld) were irradiated in this project. Mechanical property tests of these materials have been performed under a consortium of EPRI, CRIEPI, NRI-Rez and ATI Consulting to fully characterize the mechanical properties in terms of Charpy transition temperature and upper-shelf energy, as well as reference fracture toughness using the Master Curve. Some results have been reported at previous PVP conferences. In this paper, we report the results of microstructural characterization using three-dimensional atom probe tomography (APT) of the medium copper plate and the high copper weld metal. The microstructures obtained by APT reasonably explain the changes in mechanical properties of these materials, and the difference in the response of these materials to irradiation was also identified. The mixed effect of fluence/flux/spectrum is discussed from the microstructural point of view.

VERLIFE – “Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation” was developed within the 5th Framework Program of the European Union in 2003 and later upgraded within the 6th Framework Program project “COVERS – Safety of WWER NPPs” of the European Union in 2008. This Procedure has to fill the gap in original Soviet/Russian Codes and Rules for WWER type NPPs, as these codes were developed only for design and manufacture and were not changed since their second edition in 1989. VERLIFE Procedure is based on these Russian codes but incorporates also new developments in research, mainly in fracture mechanics, and also some principal approaches used in PWR codes. To assure that VERLIFE Procedure will remain a living document, new 3-year IAEA project (in close co-operation with another project 6th Framework Program of the European Union “NULIFE – Plant Life Management of NPPs”) has started in 2009. Within this project, upgrading/updating of the VERLIFE procedure is prepared together with the extension by following procedures: - Lifetime of reactor pressure vessel internals. - Leak-before-break concept for WWER NPPs. - No-break-zone for WWER NPPs. - Probabilistic approach for failure and lifetime assessment of WWER reactor pressure vessels and primary piping (including Risk-informed ISI). - Thermal fatigue evaluation. Final document, after its approval by expert groups of the IAEA and NULIFE, will be issued as “IAEA Guidelines for Integrity and Lifetime Assessment of Components and Piping in WWER NPPs”. The paper will describe these main principles and also future plans.