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Procedía CIRP 38 (2015) 8- 13
The Fourth International Conference on Through-life Engineering Services
Life Cycle Management of Structural Components of Indian Nuclear Reactors and Reprocessing Plants
Baldev Raja*, P. Chellapandib and U. Kamachi Mudalic
aNational Institute of Advanced Studies, Bengaluru 560 012, India bBharatiya Nabhikiya Vidyut Nigam Limited, Kalpakkam 603 102, India cIndira Gandhi Centre for Atomic Research, Kalpakkam 603 102, India
Corresponding author. Tel.: +91-80-2360-1969; fax: +91-80-2218-7076. E-mail address: baldev_dr@gmail.com
Abstract
The increasing demands on reliability, safety and economics of nuclear systems translate to challenges in realization of high-performance components for operation at steady state, transient and severe accident conditions. The authors, based on their four decades of research, development and deployment experiences, present a review of findings relating to life cycle management of critical structural components in Indian thermal, fast reactors and reprocessing facilities. The challenges relating to specific structural components are described with highlights of materials to improve life for prolonged service with safety and economics.
© 2015TheAuthors.PublishedbyElsevierB.V. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/).
Peer-review under responsibility of the Programme Chair of the Fourth International Conference on Through-life Engineering Services. Keywords: Nuclear System; Life Cycle; Nuclear Material; Sodium; Spent Fuel; Energy Efficiency
1. Introduction
The life cycle management of key structural components in any nuclear system assumes significant importance and requires a high degree of systematic, professional and specialized expertise to provide, reliability, safety and economics for operation at steady state, transient and severe accident conditions [1-3]. In order to maximize the efficiency of new systems and enhance the performance of ageing components, it is essential to understand degradation in a qualitative and quantitative manner. Also to achieve high level of reliability and excellent performance, several issues related to design, materials selection, fabrication, quality control, transport, storage, condition monitoring, failure analysis, etc. have to be adequately addressed and implemented, for structural components used in nuclear systems [1-6]. In addition to meeting the demands of extended service life, with cost competitiveness along with enhanced safety requirements; it is essential to demonstrate materials integrity and reliability so as to enable nuclear power plants to operate, beyond their initial design life. At the forefront of any energy challenge in the life cycle management of structural components, it is a necessity that plants operate for
as long as possible in a safe, reliable and cost-effective manner.
Fig 1. Indian nuclear power reactor portfolio [9, 10].
Comprehensive approaches desire trans-disciplinary teams capable of testing, evaluation, analysis and implementation towards achieving the robust design, materials, manufacturing, modeling, simulation, monitoring and
2212-8271 © 2015 The Authors. Published by Elsevier B.V. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/).
Peer-review under responsibility of the Programme Chair of the Fourth International Conference on Through-life Engineering Services. doi:10.1016/j.procir.2015.08.023
effective management. It has been demonstrated that with holistic approaches, the life cycle can be managed with safety and cost-effectiveness. The authors explain their approaches with a few substantial examples including aging and degradation of zircaloy-2 pressure tubes of Pressurised Heavy Water Reactor (PHWR) that has been extensively investigated and understood to decide en-mass replacement of coolant channels in earlier PHWRs of Rajasthan atomic power station and Madras atomic power station [7, 8]. During service, pressure tubes are susceptible to hydrogen pick up, delayed hydrogen cracking, creep, and formation of blisters. The life of pressure tubes could be prolonged by replacing zircaloy-2 with Zr-2.5Nb material. Similarly, the Monel-400 alloy tubes used in steam generator of PHWRs where tube leaking due to under deposit attack and pitting corrosion was encountered have been replaced with Inconel 600 alloy tubes [7, 8]. In Sodium Cooled Fast Reactor (SFR) with closed fuel cycle programme of India, comprehensive studies to understand behaviour of materials in liquid sodium by exposing the representative samples and components in liquid sodium loops operating at reactor conditions for long durations and evaluating their performances with extensive validated modelling have enabled right choices of materials, monitoring Uingstrategies and manufacturing [1-4]. Steam generators for SFRs are key to realize safe operation with high availability of the nuclear system.
The paper addresses life cycle management of this important component as an illustrative example of our comprehensive approach. In the spent nuclear fuel reprocessing and waste management systems, a systematic approach has been pursued to study degradation causes and mechanisms during service with life cycle analysis approach, by taking into consideration the material, fabrication, comprehensive quality management, service, operating conditions etc. The study has identified susceptible locations, where high degradation due to corrosion is expected in nitric acid medium leading to adaptation of methodologies to eliminate such occurrences during service. In this paper, the authors describe experiences of life cycle management of critical structural components in Indian nuclear systems, based on their four decades of research, development and deployment experiences.
2. Materials and challenges in safe and reliable operation of the nuclear components
Design, materials, manufacturing health monitoring techniques and methodologies enable component to perform the intended function with higher, efficiency, improved availability, increased safety and lower maintenance. Even though the design life of a nuclear power plant (NPP) is typically 30 or 40 years, NPP including reprocessing plant materials do suffer various failures during the design life, which are expensive and involve repair, inspection and replacements [1, 5, 7, 8]. In addition to satisfying the materials design criteria based on tensile properties, thermal creep, cyclic fatigue, creep-fatigue, fracture toughness, etc. structural materials for current and proposed future nuclear energy systems must provide adequate resistance to
environmental degradation phenomena and radiation damage, chemical compatibility and damage tolerance to fracture [2, 9]. Hence, materials aging degradation are significant concern and thus pursuit for nuclear and reprocessing plants materials. The different Indian nuclear power reactor configurations are shown in Fig 1 [9, 10].
The global installed nuclear power capacity has remained nearly dormant for the past few decades and the world wide developments scenario has changed after the Fukushima nuclear accidents in some countries. The present nuclear power share in India is about 3.5 %. In order to improve the average world per capita consumption of energy, India has envisaged a "Three Stage Nuclear Power Programme" for achieving energy independency by effective utilization of its limited natural uranium reserves and exploitation of its large thorium deposits. India has high emphasis on sodium cooled fast breeder reactor with closed fuel cycle. This approach is central to nuclear energy sustainability of India.
Indian reactor power plants have witnessed over 300 reactors year of accident-free safe operation with high overall capacity factor of 80% and availability factor of 90% during the last operation [1, 7]. This could only be possible due to sound design of components and systems, appropriate material selection, maintaining quality assurance at all stages and strict chemistry control supported by extensive in-service examinations using the state of art and chemistry based technologies. The sound design of equipment has played physics a vital role in corrosion control, and reliability materials performance [7, 8]. A variety of materials are required for the construction of BWR and PHWR. For example, materials like Ni alloys, are used for steam generator (SG) tubes (Monel-400, Inconel 600, 690, Incoloy 800, etc.). The majority of the components holding radioactive water or gas are made of 300 series of SS. Zr 2.5Nb and zircaloy-4 are used as cladding and pressure tube. Admiralty brass, aluminium brass, cupronickel, titanium, etc. are mainly used in condenser tubes and balance of plant heat exchangers [6, 7]. Most of these materials and component and system technology have been fully developed indigenously and the technology has been translated into production plants, which are successfully operating as various units of Department of Atomic Energy (DAE), India [7, 8]. Modeling of hydrogen pick up in pressure tube and cladding tube in PHWR, experimental back up, prediction models of life estimation, inservice inspection technologies and replacements of these components are some of the key achievements leading to success of PHWRs life cycle management.
3. Structural components management in sodium cooled fast reactor
In SFR with closed fuel cycle programme of India, comprehensive studies to understand the behaviour of materials in liquid sodium by exposing the representative samples and components in liquid sodium loops operating at reactor conditions for long durations and evaluating their performances has enabled right choices of materials, monitoring strategies and manufacturing. Developments of
advance materials are central to a safe, reliable, economic operation and effective life cycle management of NPP. In SFR; Prototype Fast Breeder Reactor (PFBR), the hostile environment of high neutron flux, liquid sodium and elevated operating temperature in poses demanding operating conditions on structural materials. To achieve high degree of reliability and at the same time meet the imposing challenges of the stringent material specification and acceptance criteria a comprehensive transdisciplinary approach has been pursued [2, 3]. The schematic illustration shows major components of PFBR with various forms of material degradation by corrosion (Fig 2). Though austenitic stainless steels are compatible with sodium in SFR, they undergo corrosion in liquid sodium. The different corrosion process by liquid sodium are : (i) leaching of elements (i.e., Ni, Cr, Mn, etc.) and consequent wall reduction of tubes and pipes, (ii) dissolution of metal elements by uniform or general corrosion in hot section, (iii) formation of carburized/decarburized and precipitation of carbide, and (iv) impurities in sodium like oxygen, hydrogen and nitrogen induce changes in microstructure that reduces mechanical properties. Waterside corrosion problems in steam generator materials include high temperature oxidation, accelerated corrosion in two phase region, corrosion in sodium-water reaction product, impingement wastage, corrosion in crevices, stress corrosion cracking etc [2, 5].
Fig 2. The schematic of PFBR showing the major components along with various forms of possible corrosion initiation sites.
The SFR components can broadly be classified components in three categories: (i) core structural, including clad and wrapper (ii) main structural vessel and (iii) steam generator [12] and mentions prominent corrosion mechanisms. Major challenges for life cycle management are the effect of neutrons on embrittlement, creep, erosion, corrosion resistance, radiation-induced growth, swelling, hydrogen embrittlement, robust nuclear design, fabrication, learning from international experience, economics, etc [1-3]. In addition; the structural materials in a primary sodium circuit require sodium compatibility whereas the material to be used in the secondary circuit of the steam generators requires the compatibility with both sodium, water and steam media.
In a typical SFR, like PFBR of India, stainless steel (SS) of type 316LN has been chosen as the major structural material for core components and primary and secondary loops and 90% of the structural components in the reactor assembly [2, 12]. Nitrogen alloying to type 316L SS enhances design life of reactor components. However, the welds are found to be the major life limiting factors as these have caused minor sodium leaks [3, 9]. It is worth to mention that SFRs are designed for longer design life (40-60 years) under high temperature environment (773-873 K). The out-of-pile components are mainly made of austenitic stainless steels, specifically 316 LN SS for operations at high temperatures (> 723 K) and 304 LN SS for lower temperature applications. The operating pressures are relatively lower, however, thermo-mechanical loads are very important. Accordingly, the design should address several short term and long term degradation mechanisms originating from materials, mechanics and manufacturing [3, 12]. The effective and successful operation of SFRs is largely dependent on the performance of core structural materials, i.e. clad and wrapper materials of the fuel subassembly, which are subjected to intense neutron flux of 4 x 1015 n/cm2s-1 about two orders of magnitude higher than thermal reactors. This leads to unique materials problems from consideration of radiation damage, resistance to void swelling, irradiation creep, and irradiation embrittlement [2, 3].
Structural materials for SFR core components have evolved progressively with capability to withstand an increase of neutron dose so as to achieve higher burn-up for improving fuel element performance. The evolution trajectory using from 316 SS with 20% cold worked (CW) to CW 316LTi to D9 (15Cr15Ni2.2Mo-Ti). Austenitic stainless steel (Alloy D9) is chosen as the in-core material for cladding tube and hexagonal wrapper for PFBR. However, for achieving higher fuel burn up for future commercial SFRs, clad materials with improved materials are being developed, which include Alloy D9I with higher phosphorus and silicon contents. In addition, the optimized D9 alloy with Ti/C = 6 with 0.75 wt. % Si, and 0.054 wt. % P, designated as Indian Fast Reactor advanced Clad-1 (IFAC-1), is developed for fuel pin cladding and wrapper applications to allow safe operation up to 120,000 Mwd/t. Austenitic SS and its related variant 316LN SS are preferred candidate materials for SFRs due to excellent high temperature mechanical properties, compatibility with sodium, excellent formability, ease of fabrication, weldability and availability [1,2,13]. Limitation to achieve high burn-up comes from the current generation core structural materials (i.e., austenitic SS) owing to its excessive high void swelling, creep deformation and degradation at higher neutron influence rather than from the fuel, and the austenitic SS are not sustainable for next level performance of fuel [2,12,13]. Well conceived roadmap for the development of structural materials and test facilities to further increase the target burn up levels up to 200,000 MWd/t are being explored. Oxide dispersion strengthened (ODS) steel also referred to as nano-strengthened alloys or nano-structured ferritic alloys are being studied extensively as main candidate materials for fuel pin application worldwide [6, 12, 14, 15]. ODS steels are
endowed with adequate high temperature strength, adequate creep resistance and resistance to oxidation and irradiation [14, 15]. Fig 3 shows the benefits of a ferritic/martensitic (FM) matrix with respect to high swelling resistance of ODS steel as compared to austenitic SS [15]. ODS ferritic-martensitic alloys exhibit adequate mechanical properties at high temperature. It is attributed to finely dispersed nano-oxides that serve as a obstacles for mobile dislocations thus improving high-temperature strength, and as a sink of point defects produced by neutron displacement. These mechanisms are responsible for maintaining superior neutron damage resistance [14, 15].
60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 dose (dpa)
Fig 3. Deformation of irradiated austenitic and FM steels in Phénix versus irradiation dose [15].
Steam generator materials should possess good resistance to creep, low cycle fatigue, creep-fatigue interaction, sodium corrosion and stress corrosion cracking [2,3,12]. Modified 9Cr-1Mo steel is the choice for steam generator [2, 12, 14]. Modified 9Cr-1Mo, normalized and the tempered condition is used. Additions of V, Nb and N ensure the formation of highly stable V, Nb-carbonitrides (MX) particles to confer high creep strength [12, 13]. Stringent chemical composition specifications, weldability, formability, oxidation and corrosion resistance and other mechanical properties have been achieved for PFBR materials with resultant improvement of reliability of components [2, 16]. Materials choice, controlled specifications, right manufacturing, careful choice of welding consumables, robust design processes, and quality assurance combined with comprehensive model based robotic remote field eddy current inspection and ppm level hydrogen pick up sensor (hydrogen is produced due to steam-water-sodium reaction) are the factors responsible for life cycle confidence of SG in SFRs for 40 years. In addition, learning from failures of SGs worldwide has been vital to design and manufacturing of high quality SGs for SFRs.
4. Life cycle management issues related to SFR
Fast Spectrum Reactor particularly employing sodium coolant have accumulated more than 400 reactor years of operation [1, 9, 10]. The overall operating experience with sodium and structural materials is found to be sufficient mature for launching commercialization. Specifically, the science of SFRs involves understanding of unique fuel and structural materials under high temperature, sodium, severe irradiation
environments over the long reactor life, sodium chemistry, aerosol behavior, sodium fire and sodium water reactions, special sensors for sodium applications (detection of micro water leaks in steam generator, sodium leaks, purity measurements, level detectors, etc) [10, 16]. Thermal hydraulics and structural mechanics aspects should involve thorough understanding of turbulences, instabilities, gas entrainments, thermal striping, stratifications, ratcheting, etc. A large number of failures, in the past, are attributed to stresses arising out of fast heat transfer of sodium in inadequately designed and manufactured components [9, 13]. The design of components, operating at high temperature calls for material damage interactions among creep-low cycle and high cycle fatigue behavior under monotonic and cyclic thermo-mechanics loading conditions, very high and very low strain rate effects and fluid structural interactions under normal, transient and seismic loading conditions [1, 2, 16]. The design envisages elaborate testing of mechanisms operating in sodium and argon cover gas space. These apart, design should recommend provisions for handling any minor and major sodium leaks and sodium-water reaction effects. Realization of technology demands the use of mature codes and standards, (which are available), manufacture of large dimensioned thin walled shell structures made of austenitic stainless steel petals with close tolerances, machining of large dimensioned and tall slender components with stringent tolerances (grid plate, absorber rod drives and component handling systems), fabrication of large size box structures with controlled distortions, hard facing technology with nickel base materials such as colmonoy (need to be used to avoid the induced radioactivity problems from Co60 isotopes during decommissioning stage) [10]. The opaqueness of sodium provides high impetus for the developments of innovative sensors for carrying out reliable inspects for the cracks in the load bearing structures (ISI) as well as to accomplish under sodium repair. The out of pile structural components are generally designed as per international codes such as ASME or French code RCC-MR. The current versions of these codes do not provide design criteria for considering the effects of sodium,low level radiation damage, thermal ageing especially at moderate temperatures, very high strain rates possible during postulated severe accident scenarios. In case of PFBR, certain analysis and experience based design rules have been developed to address these specific issues. For example, to address the effects of sodium: material data generated in air could be safely applied for the design of components with wall thickness larger than 2 mm. The influence of corrosion, carburization, decarburization, formation of ferritic substructure layers and sensitization are considered for thin walled components (t < 2 mm) [9, 10]. The effect of neutron dose level less than 1 dpa is negligible for the cold pool components. Accordingly, it is ensured that for such components, viz. bottom portions of main and inner vessels and grid plate, the dose is less than 1 dpa. The thermal neutron irradiation produces helium, mainly by the reaction with boron present in the steel at high temperature. Helium affects creep rupture strength and ductility of austenitic stainless in the creep regime. In this respect, the design data recommended for the European Fast Reactor (EFR) has been
adopted for PFBR. As regards to high strain rates, reduced ductility is recommended at high strain rates, which is derived from uniform elongation after taking into account effects of welds, irradiation, thermal ageing, accumulated creep and fatigue damage and strain rate on the ductility. Based on the material tests, the acceptable strain limit for 316 LN SS is recommended as 15 %. Subsequently, further factors of safety are applied. The allowable value is found to be as low as 3 % for the main vessel, the primary containment in PFBR [2, 9]. It can be said that the life limiting factors of SFR components and systems have been understood comprehensively and addressed adequately in the design with strong focused R&D, especially in the domain of materials, sensor developments, thermal hydraulics and structural mechanics aspects. The confidence has been enhanced through application of robust testing, evaluation, and manufacturing and in service inspection technologies.
The life-cycle benefit in terms of extra years in service ensuring economic competitiveness and enhanced safety is vital for commercial deployment of future SFR. Based on the feedback from design, research & development experiences of FBTR, and 500 MWe Prototype Fast Breeder Reactor (PFBR) robust & economical SERs and reprocessing plants are being realized. Beyond PFBR, cost effective design has been evolved by decreasing the number of components and introducing major features of twin unit concept, for sharing non-safety related systems and services. The capital cost reduction for the reactor assembly by net material saving, by way of enhancing the plant design life from 40 years to 60 years, new concepts of grid plate, primary pipes, top shield and fuel handling system, optimizing the main vessel diameter and bottom dished head shape through these improved concepts is estimated to be about 25% of capital cost [17, 18]. Furthermore, reduction in number of steam generators from 8 to 6, integrated primary sodium purification, compact plant layout and enhanced burn-up to 200 GWd/t in future SFR have the possibilities of significant capital cost and reduction in per unit cost of electricity [1, 18]. Similarly, continued research and development activities in the field of In-Service Inspection, would play key rule in forthcoming years to realize longer design life (40 to 60 and 60 to 100 years) with high confidence, acceptable to designers, utilities regulators of future advanced SFRs being conceived nationally as well as internationally.
5. Materials challenges for aqueous reprocessing plant
The success of the three-stage power generation programme evolved by DAE, India depends on the early introduction of SFRs in an expeditious manner and closing the fuel cycle. Closing the fuel cycle is an important strategy to ensure and sustain the desired growth of nuclear energy [1-3]. In the spent nuclear fuel reprocessing and waste management plants, a systematic approach has been pursued to study degradation causes and mechanisms during service with life cycle analysis approach by taking into consideration the material, fabrication, service, operating conditions, etc. Our studies have identified susceptible locations where high degradation
due to corrosion is envisaged in nitric acid medium leading to solutions for eliminating such occurrences during service [3, 4]. Similarly, in order to achieve and sustain an efficient operation of any nuclear reprocessing plant, materials performance is critical issue and is dependent on understanding and mitigating specific environmental degradation processes (e.g., corrosion, mechanical, and radiation effects). The reprocessing by the PUREX (plutonium uranium extraction) process of high burn-up SFRs spent nuclear fuel involves several stages that employ headend treatment involving chemical de-cladding followed by dissolution of fuel in nitric acid (boiling 11.5M), feed clarification, room temperature solvent extraction (1-4 M) to intermediate (warm, raffinate waste storage tanks), and evaporation (boiling 4-6 M) [3,4,19,20]. The materials degradation in such a harsh high radioactive and corrosive environment can lead to reduced performance, or in severe cases, catastrophic failure [3, 4, 16, 19]. Materials challenges must be successfully met.
The development of various advanced materials improves the reprocessing plants performance by increasing safety margins and design flexibility. The most commonly used materials of constructional in reprocessing plants are austenitic grades of stainless steel along with smaller quantities of titanium, zirconium and their alloys [16, 19, 20]. The corrosion problems are more severe during dissolution of fast reactor fuel of high burn-up due to large yields of fission products and transuranium elements in the spent fuel, and as well as complex oxidizing species in the fuels [19, 20]. Corrosion failures in reprocessing equipment made of stainless steel due to general, end grain, intergranular corrosion attack etc. attributed to transpassive corrosion in highly oxidizing nitric acid are reported [16, 20]. To avoid such material failure problems, a programme has been initiated to develop special high nitrogen steel (HNS) nitric acid grade (NAG) type 304LN SS with 0.13, 0.19 and 0.41% N along with filler wire composition (for welding) containing 0.36% N. Based on extensive studies carried out; AISI type 304LN SS with approximately 0.13%N has been considered as the choice for components operating at lower temperatures like continuous dissolver, nuclear waste storage tanks, etc. for nitric acid service in Fast Reactor Fuel Cycle Facility under construction by the DAE at Kalpakkam, India [21]. In addition, several NAG SS having compositions similar to AISI types 304L, 310L, and several new proprietary alloys with very low carbon content have been developed worldwide [19, 21]. Ti, Zr, Hf, Nb, Ta and alloys such as Ti-5%Ta, Ti-5%Ta-1.8%Nb etc. are explored alternatives to austenitic stainless steel [3, 21]. Titanium and its specific alloys developed for the purpose exhibit enhanced corrosion resistance on recirculating nitric acid process streams [3, 20]. However, significant corrosion in hot and pure nitric acid solutions and vapor condensates are observed in nitric acid for Ti alloys [3, 20, 21].
Zirconium has been utilized in reprocessing plant because of its superior corrosion resistance. Zr and its alloys are not influenced by vapor and condensates in boiling nitric acid [3,
21]. The primary concern with the use of zirconium and alloys is susceptibility to corrosion fatigue and stress corrosion cracking in highly concentrated and hot nitric acid conditions which are overcome by controlling texture and microstructures during fabrication of the components.
6. Summary
NPP and associated closed fuel cycle facilities operate under extreme conditions, i.e., boiling water, high pressure, liquid sodium, steam water, sea water, oxidizing nitric acid conditions, etc. In addition, factors that influence the service life of the structural components include: component design, selecting appropriate material, corrosion resistance, mechanical properties, fabrication, quality control, availability, cost, etc. The efficient life cycle management of critical structural components of nuclear system is a fundamental step to get reliable, safe performance and prolonged service. This requirement demands careful choice and implementation approaches to materials, advanced manufacturing technologies and total quality management. Life of PHWRs are extended from 25 to 40 years by robust life predictions and management through measurements, modelling and replacement technologies. In addition, based on the feedback from design and construction experiences of PFBR, cost effective design has been evolved by decreasing the number of components and introducing major features of twin unit concept for achieving improved economy and enhanced safety. Similarly, the challenges and the approaches adopted towards the efficient life cycle management of some critical structural components in Indian nuclear systems are highlighted. Development of right materials namely, steels, nickel based alloys, stainless steels, titanium and zirconium and their alloys are discussed. The approaches to achieve the desired results are the key elements of the paper.
Acknowledgments
The authors thank Dr. S. Ningshen, IGCAR, Kalpakkam, for valuable help in the preparation of manuscript. The authors are thankful to many of their colleagues and collaborators who have richly contributed to the successes of the programme.
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