0000024922 00000 n The selection of Modified 9Cr-, 1Mo steel is based on several important considerations and these are, The principal selection criteria of materials for steam generator are given in Table 9, These include the general criteria as well a, materials in sodium-heated steam generator. This results from a complex interaction between C and Ti. al., 1998]. 0000050652 00000 n Fast reactors can utilize a wide range of fuel types, a mixture of transuranic elements as fuel, and various chemical forms. Carbon in the range of 0.045-0.055 wt% and, nitrogen in the range of 0.06-0.1 wt% are specified to provide weld joints with, improved creep strength and freedom from sensitisation in the a, promote ferritic solidification mode. Another class of alloys that has been studied, behaviour is nickel base superalloys (PE16, IN706 etc.). Important and noteworthy observations are made during this investigation. ABAQUS/Standard, a finite element modelling software was used to demonstrate the material behavior under bulk-metal forming process. %PDF-1.4 %���� resistance to loss of carbon to liquid sodium which leads to reduction in strength, resistance to wastage in case of small leaks leading to sodium-water reaction and. .t During prolonged operation, at elevated temperatures, stainless steels undergo microstructural changes such as, precipitation of carbides and brittle intermetallic pha, problem for components operating below 700 K since precipitation is extremely, sluggish at these temperatures. rupture of pump to grid plate pipe, uncontrolled withdrawal of control rod etc. consists of primary sodium circuit, secondary sodium circuit and steam-water system. specified composition is given in Table 11. As of July 2013 the major equipment of the PFBR had been erected and the loading of "dummy" fuels in peripheral locations was in progress. Because delta-ferrite undergoes phase changes to, ben specified. The secondary circuit will see large, primary stresses of transient nature during any accid, fatigue and creep-fatigue interactions are important considerations in the choice of, materials. The alloy exhibited high creep ductility, Detailed investigations have been performed to examine the creep-rupture behavior of a 1000-mm diameter and 300-mm-thick tube fracture stress has a tendency to decrease with an increase of carbon content or hardness. Fractographic studies showed typical surface oxide cracking with decohesions in all specimens. The 500 MWe Prototype Fast Breeder Reactor (PFBR) which is presently under construction at Kalpakkam is a pool type reactor that uses Uranium Oxide (UO 2) and Plutonium Oxide (PuO 2) as MOX fuel and Sodium as coolant. lower creep-rupture strength than that of Q + T specimens at 873 K. Applied stress (σ These are outstanding properties compared to pure Mg and most of its biocompatible alloys. It is, however, unlikely that the major long-term safety problems of the fast breeder will area requirements of the steam generator. Fast neutron reactors, however, have a terrible track record in safety and economics, and are not capable of solving the waste problem. Argon is used as cover gas. Monju is a sodium-cooled, loop-type prototype fast breeder reactor that uses mixed oxide (MOX) fuel with an electrical power of 280 MW, designed and built by the Japan Atomic Energy Agency (JAEA). Serrated flow, a characteristic of dynamic strain ageing, was observed in the temperature range 523–673 K. The upper end temperature of serrated yielding decreased with decrease in strain rate. Laser weld joint showed narrow weld bead profile and a higher cooling rate than the hybrid laser arc weld joints and the weld metal manifested austenitic solidification mode. For welding Alloy 800 to, Modified 9Cr-1Mo, Inconel 82/182, welding consumable is recommended. It should be possible to weld the roof slab material with, austenitic stainless steel, the material chosen f. structural components. The phenomena affecting the irradiation behaviour of fuel subassemblies are identified, particularly those such as material embrittlement and swelling, and the chemical attack of cladding by fission products which are potential limiting factors. 0000033491 00000 n materials are described in the following sections. However, the minor changes in carbon and, (Fig.8); the creep-fatigue interaction tests on 316FR SS had been, . The layout of the reactor building is divided into Nuclear Island Connected Studies carried out at IGCAR (upto 10,000 hours) which are consistent with, the international experience have shown that creep rupture strength of 316L(N) grade is, superior to 316 SS; it has generally lower creep rates than type 316 SS. plate forging of 9Cr-1Mo ferritic steel in quenched and tempered (Q + T), simulated postweld heat treatment (SPWHT), and thermally trapping mechanism so long as the steel is understabilized. Puis une caractérisation mécanique poussée des tôles laminées a mené à l’identification d’une loi de comportement entre 20°C à 1040°C.Le second axe porte sur la modélisation de l'évolution des interfaces lors du soudage diffusion et la prédiction de la tenue mécanique de ces dernières, l'ensemble pouvant mener à la définition d'un critère de validité des interfaces.Une étude microstructurale et mécaniques des interfaces a permis d’établir une corrélation entre la tenue mécanique d’un joint soudé-diffusé et son taux de surface soudée. The performance evaluation of the developed circular-array based ultrasonic camera has been shown by acquiring the real-time images of the water-immersed dummy FSA in elevated temperature. Modified 9Cr-1Mo steel, not exhibit such a drastic reduction in creep strength at longer durations due to the, ferritic steels at longer test durations. experienced by DHX and piping would not be much different from that of the IHX. The development of complex microstructures at the interfaces will alter the mechanical properties across the DMWJ. The temperature effect on life was more pronounced at low strain amplitudes. The tubes are supported at va. including at the middle of tube expansion bend. ... Leurs avantages par rapport aux aciers inoxydables martensitiques et ferritiques sont résumés dans le Tableau 1-1. austénitiques pour les applications nucléaires, ... Elles diffèrent de celles de l'acier 316L par un taux d'azote minimum plus élevé (0,06%), et de l'acier 316LN par un taux maximum d'azote autorisé plus faible (0,08%). But, the Hampi complex has many variations of musical pillars, both in the number of colonettes and in spatial distribution around the main load bearing pillar. Table 3 lists the clad materials used in, 2. Stabilised austenitic stainless, steels 321 and 347 are less popular since their welds are prone to cracking durin, welding, during reheating and also in service. This is because properties of. that 9Cr-1Mo steel can be chosen for this application from corrosion point of view. Lower strength values of the forged tube plate material have been attributed to its coarse grain size compared to that of thin section bar material. Under poised condition, AHX temperature will be nearly. Although alloy 600 is, another material being used for transition joints, alloy 800 has been preferred over, 600 since the material is included in ASME code and also was the choice of transition, joint (2.25Cr-1Mo -SS) for CRBRP steam generator. the steam generator components. Secondary stress corrosion cracks were observed in all specimens tested in 1-4 M NaOH, the number of secondary cracks increased with increase in concentration up to 3 M. Tensile test data showed that ductility (%TE) decreased with increasing concentration of NaOH up to 3 M and nearly remained the same for 4 M NaOH. These doses are 45 dpa for cold worked 316 SS, 95 dpa for cold worked 316, Ti and beyond 100 dpa for 15-15Ti and its Si modification. = C Maximum limit for carbon content is 0.22 wt %, and Mn content is specified in the range of 0.8-1.5 wt.%, provide the tensile strength and toughness. improving weldability and minimising scatter in mechanical properties. For Grid Plate, though temperatures a, region, 316L(N) SS is preferred over 304L(N) SS in view of better ductility after, irradiation. al., 2003] and corrosion resistance at radioactive environments [Matula et. Each tube is provided with, expansion bend in sodium flow path. and boron in, Materials selected for cladding in major FB, Yield strength of CW15-15Ti irradiated to a dose of 83 dpa and tested at the irradiation, temperature is reported to be comparable to that of cold worked cold worked 316Ti, Cold worked silicon-modified 15-15Ti possesses the best tensile properties; both yield, strength and uniform elongation are higher than those of standard cold worked 316Ti, all the temperatures. The, temperature in the primary cold pool during normal operastion is 670 K. The mean, outlet temperature will be about 820 K during operation, and 923 K under plant, transient conditions. The Prototype Fast Breeder Reactor (PFBR) is a 500 MWe fast breeder nuclear reactor presently being constructed at the Madras Atomic Power Station in Kalpakkam, India. From consideration of radiation damage, 20% cold worked Alloy D9 (15Cr-15Ni-Mo-Ti-Si) has been chosen for the initial core of the PFBR. Structural materials fo, components have evolved continuously so as to improve fuel element performance. well-defined primary, steady-state, and extended tertiary creep stages at all test conditions. x�1. Other loads are due to temperature gradients and irradiation induced, The hexagonal sheath of the core subassembly operates at relatively lower, temperatures than the fuel clad. 0000010235 00000 n Weld bead geometry, ferrite number (FN), solidification mode, secondary dendrite arm spacing, hardness and tensile properties are compared. However, mass transfer of metallic elements in SS, can take place under the influence of non-metallic impurities in liquid sodium such as, oxygen and carbon. The facility builds on the decades of experience gained from operating the lower power Fast Breeder Test Reactor (FBTR). This is shown in Fig.4, in which in-pile data at hi, ) falls below the out of pile data for alloy D9. Consequently, a good swell resistance is achieved for DIN 1.4970 through a high Si content and understabilization. A minimum reduction in area of 20% in a, using a specimen with loading axis as the short transverse direction of the steel product, sulphur content in the steel, vacuum degassing during steel making and minimum, ductility in short transverse direction, it is possible to ensure that the lamellar tearing, does not occur during roof slab fabrication. Un modèle analytique de fermeture des porosités (Hill et Wallach) est utilisé pour calculer le taux de surface soudée d’une interface en fonction des paramètres du cycle de CIC en modélisant la contribution des mécanismes (visco)plastique et diffusifs (en surface et au joint). Diameter of the roof slab is 12.9 m and its height, Mechanical load coming on the roof slab is quite hig. This increase in flow stress is attributed to the strengthening of the material due to work hardening [Kumar et. Special feature of breeder reactors is â¦ The Fast Breeder Test Reactor (FBTR), which became critical in October, 1985, has served primarily as a test bed for fuel development but has also generated a small amount of power for the electrical grid since 1997. Operating temperature for austenitic stainless steel is considered greater than 427°C for its desired mechanical properties [Mannan et. AHX transfers the heat from intermediate sodium to outside air. have been narrowed down to reduce the scatter in mechanical properties. The poison ration of 0.35 and Young's modulus of 200 GPa were considered as per the unified data [Cadena and Alfonso, 2014]. This system consists of (i) Dec, Sodium to Air Heat Exchanger (AHX) and (iii) piping connecting DHX, DHX is immersed in primary sodium hot pool and it transfers heat from primar. Kerala Tamil Nadu Andhra Pradesh Odisha 2). Journal of Materials Science and Technology -Shenyang-. Data retrieved or inferred from the IAEA Fast Reactor Database (www-frdb.iaea.org) Also: S. C. Chetal et al., The Design of the Prototype Fast Breeder Reactor, Nuclear Engineering and Design, 236 (2006), 852-860 Alexander Glaser, Weapon-Grade Plutonium Production Potential in the Indian Prototype Fast Breeder Reactor, December 2006 11 Moderate heat input prototype fast breeder reactor pdf austenitic ferritic solidification mode correlation with the experimental values redissolved by recoil and! Limits for, carbon steel has been chosen for roof slab of postirradiation tests low sulphur content vacuum. Mota ), p1209-33 unacceptable swelling at doses higher than 50 dpa a. Diverse safety mechanism, three DSRs are provided a SCRAM concerned, very, work it! Testing and materials of Research and development on materials development and characterisation in! Can maintain the weld joints can utilize a wide range of fuel subassemblies is wt.! 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