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Significance of material technology for plant life management in nuclear power plants
Nuclear Engineering and Design 193 (1999) 337 – 347 www.elsevier.com/locate/nucengdes
Significance of material technology for plant life management in nuclear power plants Manfred Erve * Power Generation Group, KWU NTI, Siemens AG, Postfach 3220, D-61050 Erlangen, Germany Accepted 21 July 1999
Abstract The life-limiting mechanisms for components and systems are physical aging and wear. Both of them are related to changes of microstructure in the bulk material or at the phase boundaries medium/material and material/material. They are triggered during operation by factors such as temperature, mechanical load, and environment. Thus, to achieve an utmost effective aging management it is necessary, to understand the underlying aging and wear mechanisms such as neutron irradiation, fatigue, corrosion, fretting, etc. Definition and qualification of suitable corrective and preventive actions against accelerated aging, requires precise knowledge of the aging processes and life-limiting situations and thresholds. It is obvious, then, that materials engineering plays a large part in effective and economical plant life management. Within this paper, the role of materials science and technology in plant aging management during the various stages within a whole life cycle of a power plant is described: (1) the correct choice of materials as part of a well-based materials concept in the design stage is very important for later plant operation. As an example steam generator materials are presented. (2) The parameters of the individual manufacturing processes during erection of components and systems must be optimally selected in order to guarantee long-term operation. As an example the reasons for core shroud cracking in a BWR NPP are discussed. (3) Aging mechanisms must be accounted for in operation of components and systems, and their effects have to be counteracted in order to prevent service-life limiting situations. Details are described with respect of corrosion and neutron irradiation. Demanding future tasks for materials science and technology are presented, which are necessary to continue to contribute to an optimized plant life management and to cost-effective operation of nuclear power plants at high safety levels. © 1999 Elsevier Science S.A. All rights reserved.
1. Introduction The degradation mechanisms limiting the design life of a part or a component are aging and wear. Both are the result of microstructural pro* Tel.: +49-9131-18-2457; fax: +49-9131-18-2911. E-mail address: [email protected] (M. Erve)
cesses in the bulk material or at material–fluid or material–material phase boundaries, triggered during operation by such environmental influences as temperature, constant or changing mechanical loads, neutron irradiation, corrosion or frictional forces. The significance of materials science and technology for effective and economic plant life management is thus clear: the mecha-
0029-5493/99/$ - see front matter © 1999 Elsevier Science S.A. All rights reserved. PII: S 0 0 2 9 - 5 4 9 3 ( 9 9 ) 0 0 1 8 9 - 2
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M. Er6e / Nuclear Engineering and Design 193 (1999) 337–347
nisms underlying aging and wear processes must be understood in order to define and qualify suitable corrective and preventative measures. However, forward-looking ageing and plant life management do not begin with plant operation. Significant decisions taken already in the planning and construction phases before start-up can substantially influence not only the overall service life of a component or a plant but also its reliability and availability. Material technology contributes significantly both in the scientific background of
fundamental research as well as in everyday engineering work during all phases of the life cycle of nuclear power plants (Fig. 1). Examples are given below.
2. Material technology in the planning phase determines the quality and design life of components Design, material, manufacturing and inspection determine the quality of a component in the
Fig. 1. Material technology within the life cycle of a nuclear power plant.
Fig. 2. Contribution of material technology to basic safety.
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Fig. 3. Integrity of SG-tubes is governing the life time of steam-generators.
Fig. 4. IGSCC in the core shroud of KWW.
‘begin of life’ stage (Kußmaul, 1984), according to the German basic safety concept. A balanced consideration of these basic elements of assured safety prevents component failure due to manufacturing deficiencies. Consistent application of this principle has led to continuous further development of materials manufacturing and processing for components related to safety and reliability in nuclear power plants, affecting such aspects as ingot weights, forging, forming and welding (Erve et al., 1988 Erve and Schmidt, 1991). This has enabled a reduction in the number of welds or the re-location of welds to low-stress
areas of the components, use of more complex forgings instead of castings, and reduced fluence in the core area of the RPV by increasing the water gap between the core and RPV wall, permitted by the use of large forgings (Fig. 2). As a result, not only were maintenance costs reduced (e.g. by decreasing the duration of in-service inspections), but also the best conditions were created for trouble-free operation and optimum plant life management. The correct choice of materials as part of a well-based materials concept is very important for later plant operation. The material properties are
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determined and described in laboratory tests which may be complex and long-term for sufficiently realistic simulation of the various types of loads. Component tests are used to verify the materials rules determined in the laboratory tests and to check the long-term behavior of the materials under conservative loads. This provides the design engineers with the basic tools for dimensioning components and systems for the design service life. The significance of the selection of a suitable
material for the design life of a component and hence the cost-effectiveness of a plant is exemplified in the selection of the steam generator tube material in western pressurised water reactors. As the steam generator tubes ( : 4000 Ushaped tubes with an overall length of : 85 km and a surface area of : 5400 m2) serve as barrier between the radioactive reactor coolant and the nonradioactive secondary coolant, they are subject to especially stringent requirements regarding leak-tightness and hence corrosion resistance in
Fig. 5. Evaluation of the degree of sensitization using the ‘S-value-approach’ (according to Cihal, 1984 and Rocha, 1962).
Fig. 6. Potential aging mechanisms and resulting effects on components.
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Fig. 7. Component integrity overview on life-limiting mechanisms for LWR-components.
particular. Two materials concepts were initially developed in the western world (Fig. 3) (Tenckhoff and Ro¨sler, 1986). In the USA and other countries, the SG tubes were manufactured from the Ni alloy Inconel 600. Although its high Ni content of approx. 70% makes this material completely resistant to transgranular stress corrosion cracking, it lies in the range of increased susceptibility to intergranular stress corrosion cracking. Based on this knowledge, an alloy with a moderate Ni content, Incoloy 800 containing : 30% Ni and 20% Cr, was qualified quite early in Germany. This material is insensitive to either of these types of corrosion under the given reducing water chemistry conditions. The success of this materials decision can be seen in the evaluation of worldwide availability and repair statistics for steam generators: In the early 1990s, :8000 – 10 000 Inconel 600 tubes were sealed off with plugs each year, with the associated losses in output. By the year 1995, the number of tubes sleeved in corroded areas was : 95 000 (IAEA 1997). Meanwhile, many steam generators with Inconel 600 tubes were replaced for reasons of cost effectiveness, or such replacement projects are already planned for the near future. In contrast, PWR steam generators with Incoloy 800 tubes have since completed operating times up to :25 years, during which only one case of stress corrosion cracking has been incurred
(secondary side, in the area of the ‘sludge-pile’ above the tube sheet) (Stieding et al., 1990). It is impossible to write a more impressive success story of effective and preventive plant life management of a component, which, although clearly based on an overall consideration of materials, design and water chemistry, was significantly influenced by material technology in the suitable choice of materials, verifying laboratory investigations and intervention in design details and water chemistry procedures.
3. Material technology in the construction phase and its significance for design life of components The manufacturing steps of casting, forging, forming, welding and heat treatment are performed in logical sequence and with innumerable frequency in series along with many other aspects of fabrication and construction to erect the plant component by component and system by system. The parameters of the individual manufacturing processes must be optimally selected in order to guarantee not only the specified material properties such as strength and toughness, but also other application-related properties such as creep and corrosion resistance for long-term plant operation. The following case, considered in retrospect from our current level of understanding of materi-
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als, illustrates the serious consequences of incorrect selection of one of these parameters on the design life of a component. Crack formation in reactor pressure vessel internals in non-German plants (Erve et al., 1997c) triggered visual inspection of the welds in the core shrouds in German boiling water reactors as well. In one of these inspections performed in September and October of 1994 at Wu¨rgassen nuclear power station, indications were detected in the lower and upper support rings, in the top guide reinforcement ring, as well as in some areas of the core plate ring (Fig. 4) (Wachter et al., 1995). Ultrasonic examination to determine crack depth was performed at some selected points on the upper and lower flange of the core shroud. Maxi-
mum crack depths of 25–30 mm were determined in the upper support ring. Laboratory tests on material samples from the crack areas indicated that the cause of damage was sensitization of the base material (No. 1.4550) by formation of chromium carbide and resultant chromium removal from the grain boundaries during furnace annealing (stress-relief heat treatment of weld segments in the rings at 600°C for up to 6.5 h, Fig. 5) and welding, in connection with a low stabilizing ratio (Nb/C : 8) and high carbon content (0.067%). The given water chemistry conditions and the increased corrosion potential due to radiolysis led to intergranular stress corrosion cracking in operation as a result of this unfavorable combination of materials manufac-
Fig. 8. Component integrity integrated concept of RPV safety analysis.
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Fig. 9. NPP Kozloduy Unit 1 reactor pressure vessel, assessment of rest-life-time.
Fig. 10. Remedies and preventive measures to avoid IGSCC.
turing parameters (heat treatment temperature and time as well as chemical composition) (Wachter et al., 1995, Erve et al., 1997c). Fluid dynamic and structural mechanical analyses have demonstrated that only slight loads act on the individual components even under accident condi-
tions, enabling achievement of the protection goals of ‘shutdown ability’ and ‘emergency cooling ability’ or ‘residual heat removal ability’. Corresponding considerations of limiting cases were also performed for the individual plants in the USA and evaluated positively by the NRC, for which
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reason the cracks detected to date in 25 non-German units did not lead to shutdowns. However, the detection of cracking under the given boundary conditions did result in a premature end to the service life of the affected component and ultimately for the entire plant at Wu¨rgassen.
4. Operational plant life management based on fundamentals of material technology A significant task of the nuclear power plant operators is to maintain the standards of safety and availability prevailing at the time of start-up even after many years of plant operation, or even
to improve these based on new findings and developments. Among other things, this task requires exact knowledge of the aging and wear mechanisms as well as the situations limiting service life. Fig. 6 gives an overview of the significant aging and wear mechanisms which must be accounted for in components of nuclear power plants, as well as their possible effects which must be counteracted in order to prevent service-life limiting situations such as decrease of wall thickness below allowable minimum limits, reaching the stress usage factor regarding fatigue, exceeding the allowable brittle fracture transition temperature shift due to neutron irradiation, or loss of functional capability (Fig. 7). Of particular impor-
Fig. 11. Economical plant operation in the interest of the customer.
Fig. 12. Component integrity monitoring of aging and measures to prevent occurrence of failures.
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tance are the effects of neutron irradiation near the core in the reactor pressure vessel and the many different types of corrosion. Examples are given below.
4.1. Example: irradiation beha6ior Accelerated irradiation surveillance programs are used to monitor the properties of the base materials and weld metals used in the reactor core area under the influence of neutron irradiation up to the end of the service life of the components in the RPVs of plants in western countries and in most plants in Central and Eastern European countries. The goal is to always ensure sufficient toughness to rule out the initiation of brittle failure for all conceivable loads under loss-ofcoolant accident conditions or other pressurized thermal shock (PTS) transients. In some reactors of the first generation of the Russian VVER 440/ 230 series, irradiation samples of this type were not planned, but rather the influence of irradiation was calculated using theoretical rules based primarily on the chemical composition of the materials. Bulgaria’s Kozloduy Unit 1 plant received special attention from the experts when an especially high phosphorous content lying outside the prior validity of the calculation rules (measured values B0.052%) was detected in samples from the weld subject to the highest neutron irradiation. As conservative engineering evaluations could not completely rule out brittle failure in the event of a PTS event, only a testing of the material condition could provide a reliable conclusion with regard to the remaining service life, or at least answer the question regarding the necessity of recovery annealing of the material with a view of temporarily improving the situation. The plant operator therefore decided in mid-1996 to provide sample material for suitable material tests (Fig. 8) by sample removal from the RPV weld, as had already been performed successfully at Greifswald 1 and 2, Kozloduy 2 and Novovoronesh 3 and 4. The chemical composition of the boat samples taken by electric-discharge machining was determined, the notched bar impact energy was ascertained for small samples with a test cross-section of 5 ×5
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mm, and the fluence was calculated. The values obtained in this way enabled sufficiently precise characterization of the actual material condition. It was shown that restart could proceed without prior annealing and that reliable operation of the component is possible at least until after :2 years accelerated irradiation at the Rovno plant of the boat samples currently removed enables a reliable conclusion, following corresponding material testing, regarding material behavior over the next 10–15 years (Fig. 9) (Erve et al., 1997b). Materials science has thus made an essential contribution to the safety of the plant and has supplied a reliable basis for determining the remaining service life of the RPV and hence for life management of the overall plant.
4.2. Example: corrosion Intergranular stress corrosion cracking in unstabilized austenitic stainless steels had already become the challenge for plant life management of safety-related piping in non-German BWR plants by the early 1970s. This problem was the reason for such actions as the decommissioning of Gundremmingen A, the only plant constructed under GE license in Germany. When the first cracks appeared in Ti-stabilized austenitic piping in the Brunsbu¨ttel BWR plant in 1993 and finally clearly traced to intergranular stress corrosion cracking, materials science initially appeared to have failed to counteract this type of crack initiation in boiling water reactors with the implementation of stabilized austenitic steels. Today, a well-grounded material technology damage analysis enables more specific determination of the cause of damage (Erve et al., 1997d). Local sensitization of welds in the area of an overheated joint generated in welding (with chromium carbide deposits at the grain boundaries) led to crack initiation and crack propagation by intercrystalline stress corrosion cracking in operation under the influence of the boiling water reactor medium (high temperature water containing O2 and H2O2). Extensive materials laboratory investigations enabled the definition and qualification of corrective measures (Fig. 10) (Erve et al., 1997a), such as:
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Exclusive use of niobium-stabilized X6 CrNiNb 18 10 steel (material No. 1.4550) with optimized chemical composition. Optimized welding methods capable of reproducibly establishing low internal tensile or compressive stresses in the weld root. Prevention of root defects and extensive root shrinkage in the course of welding. Weld post-treatment methods. Targeted grinding of the weld root as required. Material technology has thus provided the operator with reliable measures for replacing defective piping and for operating the replaced system sections without additional in-service inspection requirements and without the risk of systematic appearance of findings in the future. Although the current materials concept must therefore be updated and adapted to the new state of the art in science and technology (as already implemented in the changes to KTA 3201.1), the basic advantages of stabilized austenitic steels and hence the correctness of the materials decisions taken at the time were not called into question by this research: The concentration of free carbon in the matrix and thus the risk of possible formation of chromium carbide is minimized by binding the free carbon in a special carbide. Sensitization of stabilized steels presupposes prior damage of the matrix in the course of manufacturing, such as due to overheating during welding (dissolution of special carbides). Stabilized steels have better corrosion resistance under consideration of additional boundary conditions such as cold working.
5. Future challenges to materials science and technology with regard to plant life management in nuclear power plants The current pressures of competition and hence costs in the power generation industry also require implementation of optimized plant life management methods for cost-effective operation of nuclear power plants at high safety levels. Materials science and technology will continue to contribute to this goal in the future. Demanding tasks can be seen to which solutions have not yet been found
(Fig. 11): The microstructural processes which proceed in the RPV material during irradiation are insufficiently understood. The use of high-resolution microscopy tools such as the transmission electron microscope and small-angle neutron scattering should enable better description of the relationships between the macroscopic parameters such as fluence, chemical composition and material toughness, permitting reliable conclusions regarding remaining service life for vessels even without an irradiation program and obviating measures such as complex sampling of clad vessels in VVER 440 plants. This is particularly the case for material behavior during re-irradiation following prior recovery annealing for which the theoretical predictions have since proven too conservative in comparison with the measured values, as well as for targeted plant life extension beyond the fluence range covered by irradiation surveillance programs. Quantification of the safety margin against brittle failure of the RPV with advanced micromechanistical methods and concepts (e.g. local approach and master curve concepts) which will enable the elimination of unnecessary conservatism opens possibilities for service life extension and extended or alternative procedures for determining the safety margin between material load limits and maximum loads in special cases. Localizing of areas damaged by fatigue in the vicinity of cracks caused by unforeseen stress cycles using nondestructive examination methods and, in general, the early detection of fatigue by such methods can be significant in plants which have reached or exceeded their design service life. The assignment of microstructural changes such as the configuration of dislocations, martensite formation and microcrack formation to physically measurable parameters such as ultrasound absorption, positron life and Barkhausen noise, etc. requires future research in this area. The synergistic influence of irradiation and corrosion on the components of RPV internals and the question of possible crack initiation by irradiation-assisted stress corrosion cracking (IASCC) under the loading and ambient condi-
M. Er6e / Nuclear Engineering and Design 193 (1999) 337–347
tions of the structural parts of these components both in BWR as well as in PWR plants can become a task of forward-looking plant life management with increasing neutron fluence of these components. The basis of this task must be created through materials science with both engineering considerations as well as further fundamental investigations such as in the international EPRI-CIR program in which the VGB and Siemens’ KWU Group are involved on the German side. Work is already in progress on the preparation of on-line and on-site measuring techniques for checking and monitoring the risk of corrosion with EPR (electrochemical potentiokinetic reactivation) measurements and electrochemical noise. The use of neural networks in the analysis of damage and the increased development and implementation of expert systems for materials engineering open new directions in plant life management of systems and plants. Advanced manufacturing and repair methods must continue to be qualified with materials testing to determine the results from plant operation in order to implement cost-effective, long-term, safe solutions. This is also the case for new alternative water chemistry methods such as are currently under investigation for reducing the risk of intercrystalline crack initiation in austenitic piping in BWR plants. The list of topics could be expanded with many innovative ideas, the goal of whose implementation is the reduction of operating and inspection costs, the maintenance of plant safety with increasing age, or the reduction of the personnel dose rate with effective and forward-looking plant life management oriented to the actual condition of the components. Common to all parts of the plant life management ‘feedback loop’ (Fig. 12) is that materials science and technology make a significant and recognized contribution, the significance of which is more likely to increase with increasing age of the plants based on the relationships between aging, wear and materials described at the beginning of this report, as well as on material structures and properties.
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References Cihal, V., 1984. Intergranular corrosion of steels and alloys. In: Material Science Monographs. Elsevier, Oxford –New York – Tokyo, p. 227. Erve, M., Bru¨mmer, G., Kleen, U., Maier, V., Ilg, U., Ba¨umler, H.J., Seibold, A., Blind, D., 1997a. Planned and implemented remedial measures to prevent IGSCC and to ensure safe plant operation with BWR piping systems made of stabilized austenitic steels. Nucl. Eng. Des. 171, 125 – 136. Erve, M., Dragunov, Y., Gledachev, I., Monette, Ph., Rieg, C.Y. Upgrading of PTS assessment for VVER 440/230 reactor pressure vessels — the case of Kozloduy 1. SMIRT 14, Lyon, August 17 – 22, 1997b. Erve, M., Maier, V., Bouecke, R., Leibold, F., Marschke, D., Senski, G. Measures to assess and to assure the integrity of RPV-internals at Isar 1 nuclear power station. In: Seventh German – Japanese Joint Seminar, September 29– 30, 1997c. Erve, M., Wesseling, U., Kilian, R., Hardt, R., Bru¨mmer, G., Maier, V., Ilg, U., 1997d. Cracking in stabilized austenitic stainless steel piping of German boiling water reactors — characteristic features and root cause. Nucl. Eng. Des. 171, 113 – 123. Erve, M., Papouschek, F., Fischer, K., Maidorn, Ch., 1988. State of the art in the manufacture of heavy forgings for reactor components in the Federal Republic of Germany. Nucl. Eng. Des. 108, 487 – 495. Erve, M., Schmidt, J. Modern manufacturing techniques in fabrication of components for advanced water-cooled reactors. IAEA technical committee meeting on ‘Materials for Advanced Water-Cooled Reactors’. Pilsen, CSFR, May 14 – 17, 1991. IAEA-TECDOC-981. Assessment and Management of Aging of Major NPP Components Important to Safety: Steam Generator, November 1997. Kußmaul, K. German basis safety concept rules out possibility of catastrophic failure. Nucl. Eng. Int., Dec. 1984, pp. 41 – 46. Rocha, H.I. Die Sensibilisierung stabilisierter und nicht stabilisierter austenitischer Sta¨hle durch Chromcarbid. Zeitschrift fu¨r Schweißtechnik 3, 1962, pp. 98 – 106. Stieding, L., Stellwag, B., Wieling, N., Rieß, R. Steam generator tubing and associated operating performance. Kerntechnik 55 No. 6, December 1990, pp. 340 – 345. Tenckhoff, E., Ro¨sler U. Werkstoffe fu¨r leichtwasserreaktoren — anforderungen und betriebsbewa¨hrung. Atomwirtschaft, March 1986. Wachter, O., Wesseling, U., Bruns, J., Kilian, R. Interkristalline Spannungsrißkorrosion an Nb-stabilisierten austenitischen Stahl in Kerneinbauten eines Siedewasserreaktors. In: VGB Conference, Korrosion und Korrosionsschutz in der Kraftwerkstechnik, Essen, November 29 – 30, 1995.
Significance of material technology for plant life management in nuclear power plants Manfred Erve * Power Generation Group, KWU NTI, Siemens AG, Postfach 3220, D-61050 Erlangen, Germany Accepted 21 July 1999
Abstract The life-limiting mechanisms for components and systems are physical aging and wear. Both of them are related to changes of microstructure in the bulk material or at the phase boundaries medium/material and material/material. They are triggered during operation by factors such as temperature, mechanical load, and environment. Thus, to achieve an utmost effective aging management it is necessary, to understand the underlying aging and wear mechanisms such as neutron irradiation, fatigue, corrosion, fretting, etc. Definition and qualification of suitable corrective and preventive actions against accelerated aging, requires precise knowledge of the aging processes and life-limiting situations and thresholds. It is obvious, then, that materials engineering plays a large part in effective and economical plant life management. Within this paper, the role of materials science and technology in plant aging management during the various stages within a whole life cycle of a power plant is described: (1) the correct choice of materials as part of a well-based materials concept in the design stage is very important for later plant operation. As an example steam generator materials are presented. (2) The parameters of the individual manufacturing processes during erection of components and systems must be optimally selected in order to guarantee long-term operation. As an example the reasons for core shroud cracking in a BWR NPP are discussed. (3) Aging mechanisms must be accounted for in operation of components and systems, and their effects have to be counteracted in order to prevent service-life limiting situations. Details are described with respect of corrosion and neutron irradiation. Demanding future tasks for materials science and technology are presented, which are necessary to continue to contribute to an optimized plant life management and to cost-effective operation of nuclear power plants at high safety levels. © 1999 Elsevier Science S.A. All rights reserved.
1. Introduction The degradation mechanisms limiting the design life of a part or a component are aging and wear. Both are the result of microstructural pro* Tel.: +49-9131-18-2457; fax: +49-9131-18-2911. E-mail address: [email protected] (M. Erve)
cesses in the bulk material or at material–fluid or material–material phase boundaries, triggered during operation by such environmental influences as temperature, constant or changing mechanical loads, neutron irradiation, corrosion or frictional forces. The significance of materials science and technology for effective and economic plant life management is thus clear: the mecha-
0029-5493/99/$ - see front matter © 1999 Elsevier Science S.A. All rights reserved. PII: S 0 0 2 9 - 5 4 9 3 ( 9 9 ) 0 0 1 8 9 - 2
338
M. Er6e / Nuclear Engineering and Design 193 (1999) 337–347
nisms underlying aging and wear processes must be understood in order to define and qualify suitable corrective and preventative measures. However, forward-looking ageing and plant life management do not begin with plant operation. Significant decisions taken already in the planning and construction phases before start-up can substantially influence not only the overall service life of a component or a plant but also its reliability and availability. Material technology contributes significantly both in the scientific background of
fundamental research as well as in everyday engineering work during all phases of the life cycle of nuclear power plants (Fig. 1). Examples are given below.
2. Material technology in the planning phase determines the quality and design life of components Design, material, manufacturing and inspection determine the quality of a component in the
Fig. 1. Material technology within the life cycle of a nuclear power plant.
Fig. 2. Contribution of material technology to basic safety.
M. Er6e / Nuclear Engineering and Design 193 (1999) 337–347
339
Fig. 3. Integrity of SG-tubes is governing the life time of steam-generators.
Fig. 4. IGSCC in the core shroud of KWW.
‘begin of life’ stage (Kußmaul, 1984), according to the German basic safety concept. A balanced consideration of these basic elements of assured safety prevents component failure due to manufacturing deficiencies. Consistent application of this principle has led to continuous further development of materials manufacturing and processing for components related to safety and reliability in nuclear power plants, affecting such aspects as ingot weights, forging, forming and welding (Erve et al., 1988 Erve and Schmidt, 1991). This has enabled a reduction in the number of welds or the re-location of welds to low-stress
areas of the components, use of more complex forgings instead of castings, and reduced fluence in the core area of the RPV by increasing the water gap between the core and RPV wall, permitted by the use of large forgings (Fig. 2). As a result, not only were maintenance costs reduced (e.g. by decreasing the duration of in-service inspections), but also the best conditions were created for trouble-free operation and optimum plant life management. The correct choice of materials as part of a well-based materials concept is very important for later plant operation. The material properties are
340
M. Er6e / Nuclear Engineering and Design 193 (1999) 337–347
determined and described in laboratory tests which may be complex and long-term for sufficiently realistic simulation of the various types of loads. Component tests are used to verify the materials rules determined in the laboratory tests and to check the long-term behavior of the materials under conservative loads. This provides the design engineers with the basic tools for dimensioning components and systems for the design service life. The significance of the selection of a suitable
material for the design life of a component and hence the cost-effectiveness of a plant is exemplified in the selection of the steam generator tube material in western pressurised water reactors. As the steam generator tubes ( : 4000 Ushaped tubes with an overall length of : 85 km and a surface area of : 5400 m2) serve as barrier between the radioactive reactor coolant and the nonradioactive secondary coolant, they are subject to especially stringent requirements regarding leak-tightness and hence corrosion resistance in
Fig. 5. Evaluation of the degree of sensitization using the ‘S-value-approach’ (according to Cihal, 1984 and Rocha, 1962).
Fig. 6. Potential aging mechanisms and resulting effects on components.
M. Er6e / Nuclear Engineering and Design 193 (1999) 337–347
341
Fig. 7. Component integrity overview on life-limiting mechanisms for LWR-components.
particular. Two materials concepts were initially developed in the western world (Fig. 3) (Tenckhoff and Ro¨sler, 1986). In the USA and other countries, the SG tubes were manufactured from the Ni alloy Inconel 600. Although its high Ni content of approx. 70% makes this material completely resistant to transgranular stress corrosion cracking, it lies in the range of increased susceptibility to intergranular stress corrosion cracking. Based on this knowledge, an alloy with a moderate Ni content, Incoloy 800 containing : 30% Ni and 20% Cr, was qualified quite early in Germany. This material is insensitive to either of these types of corrosion under the given reducing water chemistry conditions. The success of this materials decision can be seen in the evaluation of worldwide availability and repair statistics for steam generators: In the early 1990s, :8000 – 10 000 Inconel 600 tubes were sealed off with plugs each year, with the associated losses in output. By the year 1995, the number of tubes sleeved in corroded areas was : 95 000 (IAEA 1997). Meanwhile, many steam generators with Inconel 600 tubes were replaced for reasons of cost effectiveness, or such replacement projects are already planned for the near future. In contrast, PWR steam generators with Incoloy 800 tubes have since completed operating times up to :25 years, during which only one case of stress corrosion cracking has been incurred
(secondary side, in the area of the ‘sludge-pile’ above the tube sheet) (Stieding et al., 1990). It is impossible to write a more impressive success story of effective and preventive plant life management of a component, which, although clearly based on an overall consideration of materials, design and water chemistry, was significantly influenced by material technology in the suitable choice of materials, verifying laboratory investigations and intervention in design details and water chemistry procedures.
3. Material technology in the construction phase and its significance for design life of components The manufacturing steps of casting, forging, forming, welding and heat treatment are performed in logical sequence and with innumerable frequency in series along with many other aspects of fabrication and construction to erect the plant component by component and system by system. The parameters of the individual manufacturing processes must be optimally selected in order to guarantee not only the specified material properties such as strength and toughness, but also other application-related properties such as creep and corrosion resistance for long-term plant operation. The following case, considered in retrospect from our current level of understanding of materi-
342
M. Er6e / Nuclear Engineering and Design 193 (1999) 337–347
als, illustrates the serious consequences of incorrect selection of one of these parameters on the design life of a component. Crack formation in reactor pressure vessel internals in non-German plants (Erve et al., 1997c) triggered visual inspection of the welds in the core shrouds in German boiling water reactors as well. In one of these inspections performed in September and October of 1994 at Wu¨rgassen nuclear power station, indications were detected in the lower and upper support rings, in the top guide reinforcement ring, as well as in some areas of the core plate ring (Fig. 4) (Wachter et al., 1995). Ultrasonic examination to determine crack depth was performed at some selected points on the upper and lower flange of the core shroud. Maxi-
mum crack depths of 25–30 mm were determined in the upper support ring. Laboratory tests on material samples from the crack areas indicated that the cause of damage was sensitization of the base material (No. 1.4550) by formation of chromium carbide and resultant chromium removal from the grain boundaries during furnace annealing (stress-relief heat treatment of weld segments in the rings at 600°C for up to 6.5 h, Fig. 5) and welding, in connection with a low stabilizing ratio (Nb/C : 8) and high carbon content (0.067%). The given water chemistry conditions and the increased corrosion potential due to radiolysis led to intergranular stress corrosion cracking in operation as a result of this unfavorable combination of materials manufac-
Fig. 8. Component integrity integrated concept of RPV safety analysis.
M. Er6e / Nuclear Engineering and Design 193 (1999) 337–347
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Fig. 9. NPP Kozloduy Unit 1 reactor pressure vessel, assessment of rest-life-time.
Fig. 10. Remedies and preventive measures to avoid IGSCC.
turing parameters (heat treatment temperature and time as well as chemical composition) (Wachter et al., 1995, Erve et al., 1997c). Fluid dynamic and structural mechanical analyses have demonstrated that only slight loads act on the individual components even under accident condi-
tions, enabling achievement of the protection goals of ‘shutdown ability’ and ‘emergency cooling ability’ or ‘residual heat removal ability’. Corresponding considerations of limiting cases were also performed for the individual plants in the USA and evaluated positively by the NRC, for which
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reason the cracks detected to date in 25 non-German units did not lead to shutdowns. However, the detection of cracking under the given boundary conditions did result in a premature end to the service life of the affected component and ultimately for the entire plant at Wu¨rgassen.
4. Operational plant life management based on fundamentals of material technology A significant task of the nuclear power plant operators is to maintain the standards of safety and availability prevailing at the time of start-up even after many years of plant operation, or even
to improve these based on new findings and developments. Among other things, this task requires exact knowledge of the aging and wear mechanisms as well as the situations limiting service life. Fig. 6 gives an overview of the significant aging and wear mechanisms which must be accounted for in components of nuclear power plants, as well as their possible effects which must be counteracted in order to prevent service-life limiting situations such as decrease of wall thickness below allowable minimum limits, reaching the stress usage factor regarding fatigue, exceeding the allowable brittle fracture transition temperature shift due to neutron irradiation, or loss of functional capability (Fig. 7). Of particular impor-
Fig. 11. Economical plant operation in the interest of the customer.
Fig. 12. Component integrity monitoring of aging and measures to prevent occurrence of failures.
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tance are the effects of neutron irradiation near the core in the reactor pressure vessel and the many different types of corrosion. Examples are given below.
4.1. Example: irradiation beha6ior Accelerated irradiation surveillance programs are used to monitor the properties of the base materials and weld metals used in the reactor core area under the influence of neutron irradiation up to the end of the service life of the components in the RPVs of plants in western countries and in most plants in Central and Eastern European countries. The goal is to always ensure sufficient toughness to rule out the initiation of brittle failure for all conceivable loads under loss-ofcoolant accident conditions or other pressurized thermal shock (PTS) transients. In some reactors of the first generation of the Russian VVER 440/ 230 series, irradiation samples of this type were not planned, but rather the influence of irradiation was calculated using theoretical rules based primarily on the chemical composition of the materials. Bulgaria’s Kozloduy Unit 1 plant received special attention from the experts when an especially high phosphorous content lying outside the prior validity of the calculation rules (measured values B0.052%) was detected in samples from the weld subject to the highest neutron irradiation. As conservative engineering evaluations could not completely rule out brittle failure in the event of a PTS event, only a testing of the material condition could provide a reliable conclusion with regard to the remaining service life, or at least answer the question regarding the necessity of recovery annealing of the material with a view of temporarily improving the situation. The plant operator therefore decided in mid-1996 to provide sample material for suitable material tests (Fig. 8) by sample removal from the RPV weld, as had already been performed successfully at Greifswald 1 and 2, Kozloduy 2 and Novovoronesh 3 and 4. The chemical composition of the boat samples taken by electric-discharge machining was determined, the notched bar impact energy was ascertained for small samples with a test cross-section of 5 ×5
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mm, and the fluence was calculated. The values obtained in this way enabled sufficiently precise characterization of the actual material condition. It was shown that restart could proceed without prior annealing and that reliable operation of the component is possible at least until after :2 years accelerated irradiation at the Rovno plant of the boat samples currently removed enables a reliable conclusion, following corresponding material testing, regarding material behavior over the next 10–15 years (Fig. 9) (Erve et al., 1997b). Materials science has thus made an essential contribution to the safety of the plant and has supplied a reliable basis for determining the remaining service life of the RPV and hence for life management of the overall plant.
4.2. Example: corrosion Intergranular stress corrosion cracking in unstabilized austenitic stainless steels had already become the challenge for plant life management of safety-related piping in non-German BWR plants by the early 1970s. This problem was the reason for such actions as the decommissioning of Gundremmingen A, the only plant constructed under GE license in Germany. When the first cracks appeared in Ti-stabilized austenitic piping in the Brunsbu¨ttel BWR plant in 1993 and finally clearly traced to intergranular stress corrosion cracking, materials science initially appeared to have failed to counteract this type of crack initiation in boiling water reactors with the implementation of stabilized austenitic steels. Today, a well-grounded material technology damage analysis enables more specific determination of the cause of damage (Erve et al., 1997d). Local sensitization of welds in the area of an overheated joint generated in welding (with chromium carbide deposits at the grain boundaries) led to crack initiation and crack propagation by intercrystalline stress corrosion cracking in operation under the influence of the boiling water reactor medium (high temperature water containing O2 and H2O2). Extensive materials laboratory investigations enabled the definition and qualification of corrective measures (Fig. 10) (Erve et al., 1997a), such as:
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Exclusive use of niobium-stabilized X6 CrNiNb 18 10 steel (material No. 1.4550) with optimized chemical composition. Optimized welding methods capable of reproducibly establishing low internal tensile or compressive stresses in the weld root. Prevention of root defects and extensive root shrinkage in the course of welding. Weld post-treatment methods. Targeted grinding of the weld root as required. Material technology has thus provided the operator with reliable measures for replacing defective piping and for operating the replaced system sections without additional in-service inspection requirements and without the risk of systematic appearance of findings in the future. Although the current materials concept must therefore be updated and adapted to the new state of the art in science and technology (as already implemented in the changes to KTA 3201.1), the basic advantages of stabilized austenitic steels and hence the correctness of the materials decisions taken at the time were not called into question by this research: The concentration of free carbon in the matrix and thus the risk of possible formation of chromium carbide is minimized by binding the free carbon in a special carbide. Sensitization of stabilized steels presupposes prior damage of the matrix in the course of manufacturing, such as due to overheating during welding (dissolution of special carbides). Stabilized steels have better corrosion resistance under consideration of additional boundary conditions such as cold working.
5. Future challenges to materials science and technology with regard to plant life management in nuclear power plants The current pressures of competition and hence costs in the power generation industry also require implementation of optimized plant life management methods for cost-effective operation of nuclear power plants at high safety levels. Materials science and technology will continue to contribute to this goal in the future. Demanding tasks can be seen to which solutions have not yet been found
(Fig. 11): The microstructural processes which proceed in the RPV material during irradiation are insufficiently understood. The use of high-resolution microscopy tools such as the transmission electron microscope and small-angle neutron scattering should enable better description of the relationships between the macroscopic parameters such as fluence, chemical composition and material toughness, permitting reliable conclusions regarding remaining service life for vessels even without an irradiation program and obviating measures such as complex sampling of clad vessels in VVER 440 plants. This is particularly the case for material behavior during re-irradiation following prior recovery annealing for which the theoretical predictions have since proven too conservative in comparison with the measured values, as well as for targeted plant life extension beyond the fluence range covered by irradiation surveillance programs. Quantification of the safety margin against brittle failure of the RPV with advanced micromechanistical methods and concepts (e.g. local approach and master curve concepts) which will enable the elimination of unnecessary conservatism opens possibilities for service life extension and extended or alternative procedures for determining the safety margin between material load limits and maximum loads in special cases. Localizing of areas damaged by fatigue in the vicinity of cracks caused by unforeseen stress cycles using nondestructive examination methods and, in general, the early detection of fatigue by such methods can be significant in plants which have reached or exceeded their design service life. The assignment of microstructural changes such as the configuration of dislocations, martensite formation and microcrack formation to physically measurable parameters such as ultrasound absorption, positron life and Barkhausen noise, etc. requires future research in this area. The synergistic influence of irradiation and corrosion on the components of RPV internals and the question of possible crack initiation by irradiation-assisted stress corrosion cracking (IASCC) under the loading and ambient condi-
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tions of the structural parts of these components both in BWR as well as in PWR plants can become a task of forward-looking plant life management with increasing neutron fluence of these components. The basis of this task must be created through materials science with both engineering considerations as well as further fundamental investigations such as in the international EPRI-CIR program in which the VGB and Siemens’ KWU Group are involved on the German side. Work is already in progress on the preparation of on-line and on-site measuring techniques for checking and monitoring the risk of corrosion with EPR (electrochemical potentiokinetic reactivation) measurements and electrochemical noise. The use of neural networks in the analysis of damage and the increased development and implementation of expert systems for materials engineering open new directions in plant life management of systems and plants. Advanced manufacturing and repair methods must continue to be qualified with materials testing to determine the results from plant operation in order to implement cost-effective, long-term, safe solutions. This is also the case for new alternative water chemistry methods such as are currently under investigation for reducing the risk of intercrystalline crack initiation in austenitic piping in BWR plants. The list of topics could be expanded with many innovative ideas, the goal of whose implementation is the reduction of operating and inspection costs, the maintenance of plant safety with increasing age, or the reduction of the personnel dose rate with effective and forward-looking plant life management oriented to the actual condition of the components. Common to all parts of the plant life management ‘feedback loop’ (Fig. 12) is that materials science and technology make a significant and recognized contribution, the significance of which is more likely to increase with increasing age of the plants based on the relationships between aging, wear and materials described at the beginning of this report, as well as on material structures and properties.
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