Feasibility of Safe Operation of WWER-440-Type Nuclear Power Plants for Up to 60–70 Years

: Most WWER-440 reactors are operated over the planned operational lifetime of 30 years. Now, the owners/operators are preparing for the second phase of extending the operation. The paper presents an overview of the most important aspects of ageing of the main components of the WWER-440-type reactors based on the operational and ageing management experiences and primary research efforts supporting the operation. The paper aims to demonstrate that the expectations for the operability of these reactors for up to 60–70 years are realistic. The scope of the review is limited to the ageing of the reactor pressure vessel, vessel internals


Introduction
Nuclear electricity generation, mainly due to geopolitical changes that dramatically affect supply chains of energy sources, is an indispensable element of the electricity system of several national economies in the European Union.In addition to developing new capacities and deploying new types, such as small and modular reactors, continuing the safe operation of existing nuclear power plants (NPPs) and extending their authorized operating time is a reasonable approach for nuclear operating countries.At many NPPs worldwide, the second period of operational time extension is already prepared.
The rationale for operating beyond the planned operating lifetime is two-fold.First, from the safety point of view, nuclear power plants shall comply with the requirements even at the very last minute of operation.This indicates that the entire functioning of plant safety systems should be ensured despite the ageing, i.e., from a safety point of view, the operation plants are working as new ones.Second, the operating plants have already reimbursed capital costs, and specific replacements, well-organized management of ageing, and maintenance require certain financial expenses to prolong the operation.
Safety enhancement programs of existing plants were rational responses to accidents at the Three Mile Island NPP (1978) and the Chernobyl NPP (1986).Since the reactors worldwide would have reached the end of their authorized operating time, preparing for the long-term operation of existing reactors was the only solution for keeping the nuclear option alive.Pioneering work has been carried out in the USA, where the NRC launched a comprehensive nuclear plant ageing research program in 1985 to identify and resolve technical safety issues related to the ageing of nuclear power plant systems, structures, and components (SSCs).In 1995, the 10 CFR Part 54 License Renewal Act was issued, which allows US nuclear power plants to extend their service life license for 40 years and receive an operating license extension for up to 20 years [1].The technical bases for the renewal of operating licenses are summarized in [2].In the US, the technical and scientific basis and legal framework for the second 20-year service life extension were also developed in the USA [3,4].Similar to the US practice, many countries have established systematic ageing management programs at nuclear power plants and extended the operational lifetime over the initially defined term.At the international level, the International Atomic Energy Agency (IAEA) launched the program for Safety Aspects of Long-Term Operation (SALTO) [5] and published its first guidelines for ageing management [6].The IAEA also established technical services for the Member States [7].Furthermore, starting in 2010, the IAEA has been coordinating international cooperation in the frame of the International Generic Ageing Lessons Learned (IGALL) program, guiding the ageing management programs and long-term operation (LTO) activities of the Member States [8,9].
Recently, the countries in the European Union have been operating their NPPs over the originally established lifetime and preparing repeated extensions.Specific interest is devoted to the WWER-440 reactors.The WWER reactors are light-water-moderated and water-cooled pressurized water reactors.The name comes from the Russian "вoдo-вoдянoй энергетический реaктoр", which transliterates as Vodo-Vodyanoi Energetichesky Reaktor (Water-Water Energetic Reactor, WWER, but the Russian-type acronym, VVER is also often used).The 440 indicates the nominal rated electrical power of 440 MWe.There are several subtypes of the WWER-440 reactor design.The WWER-440/213-type operates in Slovakia, Hungary, Czech Republic, Finland, Ukraine, and Russia, the WWER-440/270 in Armenia, the WWER-440/230 and WWER-440/179 in Russia [10].The reliable power supply and the stability of the economy of these countries largely depend on the operation of the WWER-440 plants.Since the operational lifetime was initially set to 30 years, the extension for the operating time was necessary since these reactors were placed into operation between 1971 and 2000 (except for the Unit 3 of the Mochovce NPP in Slovakia, connected to the grid in 2023).
The technical and moral basis of long-term operation is the regular safety evaluation and upgrading of these plants.Practically, the safety of these reactors can no longer be judged based on their original design.After the Chernobyl accident, the safety of all WWER-440 plants has been evaluated and assessed by national and international review programs.Safety deficiencies have been identified, and extensive safety improvement programs have been implemented [11].Moreover, the rated power has been increased above 500 MWe at several plants.These safety improvements form the technical and moral basis for the first and repeated service life extensions (e.g., in [12]).Furthermore, the post-Fukushima stress test and improvements enhanced the external hazard safety and preparedness for severe accidents.
The long-term operability of WWER-440-type reactors has been intensively investigated by the operating countries and by international organizations, such as the IAEA, as mentioned above.In the European Union In the 5th and 6th Framework Programs, the subsequent research projects called "NULIFE" and "VERLIFE" have made an essential effort regarding irradiation embrittlement of the WWER-440 reactor pressure vessel (RPV) and other components, e.g., [13].In 2003, an "Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation" was developed and upgraded in 2008 [14].The final procedure, the "Guidelines for Integrity and Lifetime Assessment of Components and Piping in WWER NPPs during Operation", was developed as a joint product of the European Union and AEA project in 2011.Recently, the Electric Power Research Institute (EPRI) US coordinates the evaluation of operational experiences regarding ageing and ageing management experiences and the related research activity in the frame of the Pressurized Water Reactor Materials Reliability Program (MRP) [15].The regulatory approval for LTO depends on national nuclear safety regulations.LTO can be approved via formal operation license renewal for the fixed term, or the prolongation of operation can be justified by periodic safety reviews every 10 years.For example, in Hungary, the regulatory authority approves and controls the long-term operation via a combination of license renewal and periodic safety reviews.In all cases, the owners of plants have some target total operational lifetime in mind.Table 1 shows the status of the operational time extension at all 35 operating WWER-440 plants.The four oldest plants, Novovoronezh-4, and Kola -1 and -2 received the first operating time extension permit for 15 years of operation from 2001 to 2004.Units 1-4 of the Paks NPP received a 20-year service extension permit between 2012 and 2017.Some WWER-440 Units (Kola-1, Kola-2, Novovoronezh-4, Armenian-2) have already been granted an operating license for ~60 years thanks to the subsequent extension of the operating time.The Finish Regulator approved the operation of the Loviisa WWER-440/213 reactors for 70 years, up to 2057.For Bohunyice 3-4 Units, it is also planned to license the 60-year operating time.At Paks NPP, Hungary, preparation for the subsequent license renewal (up to 70 years) is launched.The paper presents an overview of the lifetime-limiting ageing processes of the essential reactor components based on the operational and ageing management experiences of the WWER-440-type reactors to justify the operability of these reactors for the targeted period of 60-70 years.
Recently, the lifetime-limiting ageing processes identified for WWER-440 reactors are related to the material reliability of reactor pressure vessels (RPVs), especially the irradiation-assisted stress corrosion cracking (IASCC) of the core baffle-former bolts and understanding of the void swelling and its effect on material properties of WWER-440 reactor pressure vessel internals (RVIs).Understanding the susceptibility of stainless steels used in WWER-440 to IASCC and the environmental effects on the fatigue for pressure boundary components of WWERs for the targeted LTO period is essential.According to the operational experiences, at nominal water chemistry conditions, the late-life stress corrosion cracking (SCC) is not a significant limitation for the targeted term of operation of WWER primary components.At the WWER-440 plants, titanium-stabilized stainless steel is used for the steam generator heat-exchanging tubes.
Although the material has an outstanding performance record, especially at WWER-440 steam generators, the outer diameter stress corrosion cracking (ODSCC) and corrosion susceptibility in the crevice environment are interesting.The ageing phenomena of the WWERs have been identified and investigated in the EPRI MRP "MRP-471-WWER Issue Management Tables: Identifies material research gaps for WWER light water reactors" [15].
In this paper, the scope of the review of the ageing processes of WWER-440 reactors is limited to the material features and ageing of the reactor pressure vessel, vessel internals, and steam generator.Although the reactor pressure vessel (RPV) is not considered limiting for the target time of 60 to 70 years, the reliability of the RPV remains the fundamental question for the justification of the long-term operation.The analyses to justify the first operational lifetime extensions are briefly presented and compared with some recent findings regarding RPV irradiation embrittlement.Based on the operational experience of the core baffle-former bolts, the ageing management plans for 50 to 70 years of operation are also presented.The questions related to ensuring the outstanding performance of the WWER-440 steam generators are also discussed.Most important corrective actions, improvements of in-service inspections, maintenance, and ageing management are also mentioned in the paper.It is shown, despite the unavoidable material degradation processes, there are no unresolvable issues that would limit the operation of the WWER-440 reactors for up to 70 years under the established operational environment and ageing management practice.

Reactor Pressure Vessel of WWER-440/213-Type NPPs
Reactor pressure vessel (RPV) is vital for pressurized water reactors (PWRs) and WWER reactors.Their resistance against potential brittle/non-ductile failure practically determines the lifetime of PWR and WWER reactor pressure vessels.The damaging effect of neutron radiation governs this resistance.
The WWER-440 RPVs operated in Armenia, Russia, Ukraine, and Finland have been made by Soviet manufacturers, and the other RPVs were manufactured by Skoda in Czechoslovakia.Depending on the manufacturer, there are specific differences in the manufacturing procedures and chemical compositions of materials.
WWER-440 RPVs have some significant features that are different from PWR designs.First, it was assumed that they must be transportable by rails.This requirement has some crucial consequences on vessel design, such as a smaller pressure vessel diameter, which results in a smaller water gap thickness, and thus a ~5 to 10 times higher fast neutron flux (E > 0.5 MeV) on the RPV is relatively high, about 10 15 m −2 s −1 .Therefore, this requires materials with high resistance against radiation embrittlement.The WWER-440 RPVs are made from low alloy steel 15Cr2MVA (12Cr2MFA at Soviet-made RPVs) without longitudinal welds, and the circumferential submerged arc welding was made using Sv-10CrMoVTi wire.The RPV is covered internally by a welded clad of two stainless steel layers.The inner layer is non-stabilized stainless steel (Sv-07Cr25Ni13) and the outer layer is niobium-stabilized stainless steel Sv-08Cr19Ni10Mn2Nb, but Sv-07Cr19Ni10Nb is in the case of the Soviet-made RPV.The phosphorus and copper contents in the welds of WWER-440/213 range from 0.010-0.028%for P and 0.03-0.18%for Cu.For primary data of material composition of WWER-440 RPVs see, e.g., in [16][17][18].
The early experiences of WWER operating countries and the results and international research programs on material embrittlement for WWER-440 RPV have been summarized, e.g., in [18,19].Before 2000, several ageing management measures were implemented for the reduction in irradiation embrittlement and its effects on the safe operation of the RPVs, as follows: • reduction in the neutron flux on the RPV by low leakage core design and dummy assemblies; • lessening the thermal shock by heating the water for the emergency core cooling; • introduction of advanced surveillance programs; • annealing of the critical circumferential RPV welds, where it was needed.
Thanks to the abovementioned measures and development of the evaluation methodology and fluence calculations, the operability of RPVs of all reactors could be justified for the first 10 to 20 years of extension above the 30 years of the original operational lifetime based on pressurized thermal shock (PTS) calculations.The critical welds of the RPVs of Loviisa-1, Rovno -1 and -2, and Kola NPPs have been annealed only.
For example, in the case of Paks NPP, the PTS calculations based on analysis of postulated embedded flaws endorsed the possibility of 50 years of operation for all four reactors.The approach applied for PTS analysis of Paks RPVs complied with the Hungarian Regulatory Guidance.According to this, the structural integrity against brittle fracture of the RPV is ensured if the actual ductile-brittle transition temperature (DBTT), T k of its critical components is less than the maximum allowable component-specific DBTT, T allow k .The analysis compares the material's static fracture toughness K Ic and stress intensity factor K I calculated from the given loading situation.The significant steps of the analysis were as follows (see, e.g., in [20][21][22][23][24]): 1.
Identification of the critical components of the RPV.
Figure 1 shows the critical RPV locations subjects of fracture mechanics analysis.
In the case of WWER-440 RPV, significant are the base metal, circumferential weld No. 5/6, weld heat affected zone, cladding in the belt line region, and other circumferential welds, including those in the nozzle region.

2.
Selection of the PTS initiating events with higher frequency than 10 −5 /year based on the probabilistic safety analysis (PSA).
Neutron fluence calculations.The neutron transport calculation methodology has been validated by comparison with calculation benchmarks and measurements.The first validation test has been performed in the frame of the REDOS Project [22,23].
The experimental data stem from measurements performed on a mock-up simulating WWER-440 core and vessel wall installed in the LR-0 zero-reactor in the Nuclear Research Institute, Řež, Czech Republic.The second validation test of the calculation methodology has been performed with plant-specific data.Here, the calculated and measured reaction rates on the activation detectors besides the surveillance specimens were used to validate the calculation model.In the case of Paks NPP, the measurements performed for the 8-11 and 9-11 fuel cycles of unit 2, and the measurements performed for the 7-11 cycles of unit 4 were used.In both cases, the results of flux uncertainty estimates were less than 10% [21].
Based on the refueling history and future core configurations, the end-of-life fluences (for 50 and 60 operating years) are calculated for the RPV wall and the surveillance position.
In the case of Paks NPP, the end-of-life fluence for the base metal of the RPVs at Paks NPP for 60 years varies between 3.21 × 10 20 cm −2 and 3.36 × 10 20 cm −2 and for the critical No. 5/6 welding varies between 2.33 × 10 20 cm −2 and 2.36 × 10 20 cm −2 .
In the case of Paks NPP, the end-of-life fluence for the base metal of the RPVs at Paks NPP for 60 years varies between 3.21 × 10 20 cm −2 and 3.36 × 10 20 cm −2 and for the critical No. 5/6 welding varies between 2.33 × 10 20 cm −2 and 2.36 × 10 20 cm −2 . 5. Temperature and stress field calculations for RPV wall.Fracture mechanics calculations assume an under-cladding crack with a depth equal to 0.1 times RPV wall thickness and with an aspect ratio of 1/3, oriented in the base metal normal to the principal stress.In the circumferential weld, it is oriented circumferentially.

5.
Temperature and stress field calculations for RPV wall.Fracture mechanics calculations assume an under-cladding crack with a depth equal to 0.1 times RPV wall thickness and with an aspect ratio of 1/3, oriented in the base metal normal to the principal stress.In the circumferential weld, it is oriented circumferentially.
The results of qualified in-service inspections justified the assumption of the embedded postulated crack.Two types of inspection were applied for the entire cladding area: (1) Ultrasonic inspection from the inner surface and (2) an additional Eddy current inspection, overlapping the first 5-mm thickness of the RPV inner wall area.
The transients with screening criteria 10 −5 /year were analyzed using the linear elastic fracture mechanics.
The most significant transients have been analyzed by applying nonlinear fracture mechanics.
The stress intensity factor, K I , is calculated at the crack tip and the boundary between the cladding and base metal or weld.The material's static fracture toughness, K Ic , is calculated via the following reference curve: where T k is the nil-ductile temperature.
In evaluating the shift of the critical temperature (∆T k ), the effects of irradiation, thermal ageing, and fatigue were considered.
T k and ∆T k were determined by evaluating the Charpy impact test results of surveillance specimens for every RPV.
The VERLIFE proposed form for the dependence of T k versus fast neutron fluence, F n was applied, i.e., ∆T k = A F (F n /F 0 ) n + δT M , where A F and n are empirical constants, δT M is equal to 10 and for the weld as The condition of postulated defect stability is K I ≤ K Ic from what the allowable temperature, T allow k could be derived.

6.
Qualified in-service inspections and nondestructive testing verify the integrity of the cladding [21].
Based on the analysis above, the extension for 20 years above the original operational lifetime of 30 years could be justified for the RPVs of Paks NPP.
A similar procedure as above has been applied to the other WWER-440-type reactors listed in Table 1.Among national regulations and guidelines, the references [17], and [25][26][27][28][29] have been followed for evaluating end-of-life RPV conditions and justification of lifetime extensions for WWER-440 RPV.
Moreover, with more than a thousand reactor years of operational experience, an extensive database of RPV material surveillances, and results of excessive research, a factual statement can be drawn on the RPV material embrittlement of the WWER-440 reactors.
Notably, statistical analysis of the data obtained during impact testing of surveillance specimens of 15 WWER-440 RPVs operating in Russia, Ukraine, Armenia, Hungary, the Czech Republic, and Slovakia was performed [30].The raw data used for the analysis were obtained from the IAEA International Database of RPV Materials.
As a result, the empirical dependences of the shift of the nil-ductile transition temperature ∆T k on the neutron fluence have been obtained from surveillance specimens irradiated up to a fluence of 5 × 10 20 cm −2 .The tested specimens were grouped as per copper, nickel, and phosphorus content as "clean", "almost clean", "dirty", and "highly dirty".For example, the phosphorus copper content in "clean" base metal surveillance specimens was less than 0.012% and 0.07%, respectively.
Based on the results of the irradiation tests, the influence of P and Cu to ∆T k was evaluated at fluences up to 5 × 10 20 cm −2 (E > 0.5 MeV), more than double the design end-of-life WWER-440 fluence value that was 2.4 × 10 20 cm −2 for the base metal and 1.8 × 10 20 cm −2 for the weld.
Using the empirical correlation for T k versus fast neutron fluence, F n proposed in [28] and the surveillance data for RPV of reactor No. 1 at Paks NPP and extrapolating for the of-life WWER-440 fluence value that was 2.4 × 10 20 cm −2 for the base metal and 1.8 × 10 20 cm −2 for the weld.
Using the empirical correlation for   versus fast neutron fluence,   proposed in [28] and the surveillance data for RPV of reactor No. 1 at Paks NPP and extrapolating for the 70 years end-of-life fluence that is less than 5 × 10 20 cm −2 , 70 years of safe operation can be predicted as shown in Figure 2.An amendment of the regulatory limits has been proposed [30] and seems to be empirically justified.The application range of normative dependences ΔTk(Fn) on the reactor pressure vessel can be increased from 3 × 10 20 cm −2 to 5 × 10 20 cm −2 .
Increasing the maximum fluence to 5 × 10 20 cm −2 will allow the WWER-440 RPV with relatively low impurity contents (P 0.017% and Cu 0.14%) to be operated for 60-80 years without annealing the base metal and, in some cases, without annealing the irradiated welds.

Reactor Pressure Vessel Internals
The primary function of the reactor vessel internals (RVI) is to support the core and the control rod assemblies.The RVI has the additional function of directing the flow of the reactor coolant and providing shielding for the reactor pressure vessel.The RVI is subjected to neutron irradiation and exposure to the primary coolant.The core basket or barrel is part of the RVI structure consisting of the core baffle-former built from horizontal forming plates that follow the shape of the core baffles and the shape of the peripheral fuel assemblies.In the case of WWERs, the former follows the shape of the hexagonal fuel assemblies.The vertical plates/baffles surrounding the outer face peripheral fuel assemblies.The vertical plates are bolted to the horizontal former plates that are bolted to the An amendment of the regulatory limits has been proposed [30] and seems to be empirically justified.The application range of normative dependences ∆T k (F n ) on the reactor pressure vessel can be increased from 3 × 10 20 cm −2 to 5 × 10 20 cm −2 .
Increasing the maximum fluence to 5 × 10 20 cm −2 will allow the WWER-440 RPV with relatively low impurity contents (P 0.017% and Cu 0.14%) to be operated for 60-80 years without annealing the base metal and, in some cases, without annealing the irradiated welds.

Reactor Pressure Vessel Internals
The primary function of the reactor vessel internals (RVI) is to support the core and the control rod assemblies.The RVI has the additional function of directing the flow of the reactor coolant and providing shielding for the reactor pressure vessel.The RVI is subjected to neutron irradiation and exposure to the primary coolant.The core basket or barrel is part of the RVI structure consisting of the core baffle-former built from horizontal forming plates that follow the shape of the core baffles and the shape of the peripheral fuel assemblies.In the case of WWERs, the former follows the shape of the hexagonal fuel assemblies.The vertical plates/baffles surrounding the outer face peripheral fuel assemblies.The vertical plates are bolted to the horizontal former plates that are bolted to the core basket.These Energies 2023, 16, 4170 9 of 17 core baffle-former bolts hold together a structure inside the reactor vessel.This structure and the bolts are subjected to significant mechanical stress and high neutron flux.
Core baffle-former bolt degradation was first noted in the late 1980s in French PWRs.In the US, in some pressurized water reactors, the operation experiences indicate susceptibility to IASCC degradation of the baffle-former bolts [31].Adequate in-service inspections and replacement of the damaged bolts manage the issue [32].
Figure 3 shows the core baffle-former's shape and the two bolts used for fixing the former of the WWER-440/213 reactors at Paks NPP in Hungary.In the WWER reactors, solution-annealed titanium-stabilized stainless steels (0X18H10T corresponding to Type AISI 321) have been used for RVI due to their corrosion resistance, toughness, ductility, strength, and fatigue characteristics.
Energies 2023, 16, x FOR PEER REVIEW 9 of 17 core basket.These core baffle-former bolts hold together a structure inside the reactor vessel.This structure and the bolts are subjected to significant mechanical stress and high neutron flux.Core baffle-former bolt degradation was first noted in the late 1980s in French PWRs.In the US, in some pressurized water reactors, the operation experiences indicate susceptibility to IASCC degradation of the baffle-former bolts [31].Adequate in-service inspections and replacement of the damaged bolts manage the issue [32].
Figure 3 shows the core baffle-former's shape and the two bolts used for fixing the former of the WWER-440/213 reactors at Paks NPP in Hungary.In the WWER reactors, solution-annealed titanium-stabilized stainless steels (0X18H10T corresponding to Type AISI 321) have been used for RVI due to their corrosion resistance, toughness, ductility, strength, and fatigue characteristics.There is only one case of known cracked baffle bolts in WWER-440 internals [33] found by volumetric ultrasonic examination.The damaged bolts could have cracked due to improper alignment and restricted thermal expansion.One bolt has been damaged due to irradiation-assisted intergranular stress corrosion cracking.The case has been thoroughly investigated, and the bolt has been replaced [33][34][35][36].According to analysis [37] in There is only one case of known cracked baffle bolts in WWER-440 internals [33] found by volumetric ultrasonic examination.The damaged bolts could have cracked due to improper alignment and restricted thermal expansion.One bolt has been damaged due to irradiation-assisted intergranular stress corrosion cracking.The case has been thoroughly investigated, and the bolt has been replaced [33][34][35][36].According to analysis [37] in the Czech Republic, the baffle-former bolts are the most susceptible subcomponent of reactor vessel internals.They are loaded by fatigue and IASCC.Both degradation mechanisms are influenced and accelerated by swelling development.Since the very limited cases of the baffle bolts were damaged, the baffle bolt IASCC has not been a high-priority issue for the first phase of the lifetime extension.Regarding swelling and baffle-former bolts ageing, the operability of the RVI structures has been justified by in-service inspections and ultrasonic testing.The performance of WWER-440 RVI materials has also been investigated and demonstrated using specimens harvested after 45 years of operation with a damaging dose of 7.9-43.0dpa (displacement per atom) and irradiation temperature of 280-315 • C that showed negligible swelling [38].
Preparing for the extended operation above 40-50 years, the WWER operators developed in-service inspections and repair methodologies and performed ageing and stress analyses.For example, in the case of Paks NPP, Hungary, the applicable inspection and repair techniques have been reviewed.The acceptance criteria for the crack depth and cross-section for the bolts have been analyzed for different periods of testing based on the standard ASME BPVC Section III, Division 1, Subsection NG and Nonmandatory Appendix F. Despite the good operational experiences, a research effort is needed for a better understanding of the conditions contributing to void swelling and its effect on the material properties of WWER RPV internals, and for the definition of empirically justified allowable neutron doses (or dpa).
From the nuclear safety point of view, the degradation of baffle-former bolts does not directly endanger the safe operation.Since the bolts are practically replaceable, their damage does not limit the reactor's lifetime.However, the deformation of the geometry and loss of integrity of the core former structure due to the swelling and degradation of bolts should be avoided.Therefore, as part of the licensee's activity during the extended operation, especially during operations above 50 years, the utility establishes periodical inspection of the reactor vessel internals, including the volumetric inspections of the baffleformer bolts, that assure safety.

Steam Generator
The steam generators (SGs) of the type PGV-440 at WWER-440/213 are lifetimelimiting components since they are not replaceable within reasonable expenses.
The heat-exchanging tubes and the steam generator tube headers (collectors) are manufactured from titanium-stabilized stainless steel (equivalent to AISI 321).The experience regarding the ageing of WWER steam generators is summarized in the TECDOC-1577 of the International Atomic Energy Agency [39].Based on the operational experiences, the WWER-440 steam generators demonstrated outstanding performance as it is shown in the Table 2 for the Paks NPP.Since the WWER-440 reactors are six-loop systems, there are experiences of 35 × 6 × 40 = 8400 steam generator operational years available for assessing the capability to operate for 60 to 70 years (40 years is assumed as the approximate average operational time).Throughout this operational period, the steam generator's material failures could be detected at the proper time and could be repaired.A relatively early defect of the feed-water distributor inside the SG arose due to accelerated erosion.These elements were replaced at all WWER-440 plants.The heatexchanging surface of the SG is overdesigned by approximately 15% for the reactors with a design power output of 440 MWe and 10% for the reactors operating at an uprated power level of ∼ =500 MWe.The acidic crevice pH was a concern regarding the integrity of the heat exchange tubes since the titanium-stabilized stainless steel is susceptible to ODSCC in an acidic environment.However, the plugging statistics were acceptable, and the trends do not show serious operability limitations up to 50 years [40][41][42].
At several WWER-440 plants, the plugging trends were improved by removing copper alloys from the secondary system and introducing the high-pH secondary water chemistry that essentially slows down the ODSCC.The effect of the power uprate on the heat-exchanging tube integrity is negligible.Nevertheless, the evaluation of operational experiences is needed for establishing an even more effective mitigation technology of titanium-stabilized stainless steel steam generator tubing, mostly in crevices, while at the same time minimizing corrosion product transport and deposition in the steam generators.In accordance with some research results, these efforts are essential for extending over 50 years since this steel might be susceptible to degradation in highly alkaline pH [43,44].
In the future, alternative chemical mitigation materials should be found to replace hydrazine if its use will be restricted due to its health and environmental effects [45].
There has been an isolated case of the cracking of WWER-440 SG primary collector threaded holes exposed to primary water [46,47].The threaded holes should not be in direct contact with the primary medium, but it was expected that the water had entered during assembly/disassembly. Failure analysis indicated that the primary cause of the cracking was the lubricant, which contained high concentrations of Sulphur and Molybdenum, high stresses, and material impurities.The collector has been repaired by cutting away and replacing the cracked part.
Critical parts of the WWER-440 steam generators are the dissimilar material welds (DMW) connecting the austenitic stainless-steel primary pipe to the K22 ferritic steel body of the steam generator shown in Figure 4.
The WWER-440-related experiences are summarized in several documents and studies, e.g., [48][49][50][51][52][53][54][55].During the long-term operation of the SG, corrosion damages have been found that were related to intercrystallite corrosion of metal 10Ch16N25AM6, built-up by electrodes EA 395/9 [48].According to [48], the degradation mechanisms for the flaws in the dissimilar material welds at WWERs are usually a combination of corrosion on the carbon steel side and stress-corrosion-cracking interdendritic character.The factors determining the cracking, i.e., the stresses, the material, and the environment have been thoroughly investigated, e.g., [48][49][50][51][52][53][54][55][56][57][58].Cases of DMW degradations have been found and repaired at Dukovany NPP and Kola NPP [48,57,58].The approach of ASME, 2013 Section XI was applied to evaluate allowable flaw sizes in the DMW connecting the austenitic stainless-steel primary pipe to the K22 ferritic steel body of the steam generator [52,53].A thorough nondestructive examination indicates the performance of these newly repaired DMWs [58].From the point of view of SG operability for up to 60-70 years, the DMW issue is manageable since verified methods are available for monitoring the degradation and repair.The practical questions are the insurance of the effectiveness of ageing monitoring of DMW that is shown in [49,53].Moreover, improving the draining, sludge removal is essential, if possible.
The DMW issue is not limited to the steam generator.Since there are several critical locations where DMWs are applied including WWER-440 RPV safe end welds, hydroaccumulator pipe welds, and welds under the pressurizer, the DMW issues are of high priority.In the future, efforts should be made to collect, evaluate, and analyze the operators' operational data and in-service inspection results [44].The WWER-440-related experiences are summarized in several documents and studies, e.g., [48][49][50][51][52][53][54][55].During the long-term operation of the SG, corrosion damages have been found that were related to intercrystallite corrosion of metal 10Ch16N25АM6, built-up by electrodes EA 395/9 [48].According to [48], the degradation mechanisms for the flaws in the dissimilar material welds at WWERs are usually a combination of corrosion on the carbon steel side and stress-corrosion-cracking interdendritic character.The factors determining the cracking, i.e., the stresses, the material, and the environment have been thoroughly investigated, e.g., [48][49][50][51][52][53][54][55][56][57][58].Cases of DMW degradations have been found and repaired at Dukovany NPP and Kola NPP [48,57,58].The approach of ASME, 2013 Section XI was applied to evaluate allowable flaw sizes in the DMW connecting the austenitic stainless-steel primary pipe to the K22 ferritic steel body of the steam generator [52,53].A thorough nondestructive examination indicates the performance of these newly repaired DMWs [58].From the point of view of SG operability for up to 60-70 years, the DMW issue is manageable since verified methods are available for monitoring the degradation and repair.The practical questions are the insurance of the effectiveness of ageing monitoring of DMW that is shown in [49,53].Moreover, improving the draining, sludge removal is essential, if possible.
The DMW issue is not limited to the steam generator.Since there are several critical locations where DMWs are applied including WWER-440 RPV safe end welds, hydroaccumulator pipe welds, and welds under the pressurizer, the DMW issues are of high priority.In the future, efforts should be made to collect, evaluate, and analyze the operators' operational data and in-service inspection results [44].

Generic Conditions for Long-Term Operation of WWER-440 Plants
In addition to the above-discussed technical aspects, there are several non-technical aspects related to controlling material degradation and resolving issues arising during operation in the long term.These are as follows: 1.
The operating countries established comprehensive regulations on controlling operator ageing management activities and approving long-term operations based on license renewal or the periodic safety review.From the technical point of view, in both cases, the Regulatory Authorities control the information on the ageing of the critical structures and components and the effectiveness of the ageing management programs.For example, in the case of Hungary, the regulation includes the basic elements of the US NRC 10 CFR Part 54 [1].The control of compliance with the current licensing basis is maintained via the annual updating of the Final Safety Analysis Report and its Periodic Safety Review every 10 years.The license renewal itself is a two-step process.
First, the LTO Program should be developed and submitted to the regulator at least 4 years before the design life expires, but not before 20 years of operation.Second, the formal license renewal application should be submitted 1 year before the design lifetime expires.Four years of experience implementing the LTO Program should demonstrate that the licensee's Program is effective; it ensures long-term operation and the licensee's assessments regarding the safe lifetime are appropriate.Obtaining the environmental license for an extended term of operation is a precondition to applying for the new operating license.
International cooperation exists for gathering and evaluating operational experiences, e.g., the Pressurized Water Reactor Materials Reliability Program organized by EPRI US [14] or by the IAEA [8].Guidances and methodologies for the operators are the main products of these activities, e.g., [6,25,26,62].
Although the national regulatory frames and the applied standards are different, a generalized scientific-technical basis exists for the evaluation and management of the ageing of the critical components thanks to the intensive exchange of the data and operational experiences in the frame of international projects, coordinated by the IAEA, European Union, EPRI.The international benchmarking and comparison of the testing and surveillance methods and qualification and evaluation of the tests are very important.These form the common basis for safe long-term operation.

5.
The IAEA SALTO program and services support the operator's practical ageing management activities, e.g., [62][63][64][65][66]. Independent of the national regulation and differences in operator practices, the baseline is defined by the safety requirements, guidelines of the IAEA, and internationally accepted standards and best practices.
The IAEA review missions enforce this generic baseline.

Discussion and Conclusions
The paper aimed to demonstrate that the expectations for the operability of the WWER-440-type reactors for up to 60-70 years are realistic considering the material aspects of ageing of the main components: The reactor pressure vessel, the reactor pressure vessel internals, and the steam generator.Reviewing the operational experiences and research results related to ageing issues of the WWER-440-type reactors, three main aspects for the justification of long-term operability have been considered:

•
Whether the ageing processes and the stressors have been already identified?

•
Whether the methods for monitoring and evaluating the material conditions and analysis of time limits operability ensure confidence in the long-term operability of this reactor type?

•
Are the already implemented ageing management programs, methods, and mitigative or corrective actions effective?
Regarding the WWER-440 reactor pressure vessels (base metal and welds), irradiation embrittlement is the lifetime-limiting aging effect.The enhanced surveillance and in-service inspection methods are qualified, and the measures for reducing fast neutron fluences and temperature stresses are effective in accordance with experience.Methodology for the analysis of time limits of safe operation could be qualified as proven.As an ultima ratio, the annealing of the RPV welds can be implemented.Decisions on the necessity of annealing can be made based on the evaluation of the material condition.
Regarding the ageing processes of the reactor pressure vessel internals, the operating experience and the related research provided sufficient data and knowledge to prepare for monitoring programs and repair technologies for the ageing-related damages of core baffle-former bolts.The ageing degradation of the reactor pressure vessel internals is receiving more importance during the operation over 40-50 years.The operator's efforts and the supporting research activity should ensure confidence in long-term operability.
The operational performance of the WWER-440 steam generators demonstrates the adequacy of the material selection and the design of steam generators.The critical locations and material degradation processes have been identified, and the operators implemented effective ageing management programs and corrective measures.In-service inspection of heat-exchanging tubes, their plugging technique, also the repair technologies for the dissimilar material welds are proven by experience.
Despite the generic preparedness of the operators for long-term operation and positive experiences regarding the adequacy of material selection and design of the main components of the WWER-440-type reactors, internationally coordinated efforts are needed for gathering and evaluation of operational experiences and data and for ensuring proper scientific basis of long-term operation of the nuclear power plants considered.The ageing phenomena subjects of further investigations have been identified in the paper.The international cooperation will back up the practical efforts of the WWER-440 operators and national regulatory authorities in approving and supervising the long-term operation.

Figure 1 .
Figure 1.RPV of WWER-440/213 at Paks NPP and the locations selected for fracture mechanics analysis.

Figure 1 .
Figure 1.RPV of WWER-440/213 at Paks NPP and the locations selected for fracture mechanics analysis.
end-of-life fluence that is less than 5 × 10 20 cm −2 , 70 years of safe operation can be predicted as shown in Figure2.

Figure 2 .
Figure 2. Nil-ductile temperature for the RPV of reactor No. 1 at Paks NPP versus fast neutron fluences compared to the prediction via the empirical formula proposed in [30].

Figure 2 .
Figure 2. Nil-ductile temperature for the RPV of reactor No. 1 at Paks NPP versus fast neutron fluences compared to the prediction via the empirical formula proposed in [30].

Figure 3 .
Figure 3. Core baffle-former shape (A), the connection of the plate to the former (B), and the drawing and pictures (C) and (D) show the two types of bolts (case Paks NPP).

Figure 3 .
Figure 3. Core baffle-former shape (A), the connection of the plate to the former (B), and the drawing and pictures (C,D) show the two types of bolts (case Paks NPP).

Figure 4 .
Figure 4. Cross-section of the WWER-440/213 steam generator showing the welds at the collector area.The Dissimilar Material Weld is indicated as 1.6.1.

Figure 4 .
Figure 4. Cross-section of the WWER-440/213 steam generator showing the welds at the collector area.The Dissimilar Material Weld is indicated as 1.6.1.

Table 1 .
The WWER-440 nuclear power plants and status of operational lifetime extension.

Plant/Type Connected to the Grid Years of Extension/Valid to Approved Operation 1st 2nd Time
• C for base metal and 16 • C for the weld.For example, for reactor No. 1 at Paks NPP, the dependence of T k versus fast neutron fluence, F n was found for the base material as T k = −34.2+ 65.747 F n /10 20 0.511 + 10 , [°C],

Table 2 .
Steam generator plugging statistics for Paks NPP, Hungary.