irradiation effect on mechanical & microstructure behaviour of ...

Sarda, et al., International Journal of Advanced Engineering Research and Studies

E-ISSN2249–8974

Proceedings of BITCON-2015 Innovations For National Development National Conference on : Innovations In Mechanical Engineering For Sustainable Development

Review Article IRRADIATION EFFECT ON MECHANICAL & MICROSTRUCTURE BEHAVIOUR OF BIMETALLIC WELDS Amit Sarda1, S K Moullick2 1

Address for Correspondence

Asso. Professor, Christian College of Engineering, Bhilai, INDIA 2 Professor, BIT Durg (CG) India

ABSTRACT Nuclear reactor pressure vessel is subjected to neutron irradiation in addition to the conventional effects disturbing any conventional pressure vessel during operation. In general, neutron irradiation changes the mechanical properties of the vessel steel, which influence the operational conditions of the pressure vessel, and can bound the, life time of the reactor pressure vessel. To extend the plant life and to enhance the safety at work, integrity and reliability of reactor pressure vessel (RPV) is an influential aspect to consider and study as because the main objective from a nuclear-industry perspective is to maximize and enhance the safety and the reliability of nuclear power plants and to extend their operation lifetime. From this point of view the reactor pressure vessel (RPV) is one of the critical elements, rated as the highest priority category 1 component in all different national safety rankings. As a part of the design of primary heat transport piping of pressurized heavy water reactors, which uses bimetallic weld components, under high neutron flux over considerably long period of time it may become brittle, resulting in cracks which may lead to accidents. Hence, it is essential to assess the structural integrity of the vessel under such scenario. During the operation of the reactors, bimetallic welds become brittle under the effect of irradiation. Such an irradiation embrittlement is manifested by an increase in the ductile-to-brittle transition temperature. Irradiation embrittlement can produce changes in material properties and can cause crack growth. Cleavage fracture and ductile tearing are two competing mechanism in the ductile-to-brittle transition regime of bimetallic welds. Embrittlement results in a raise in the ductile-to-brittle transition temperature, which is normally used as indicator of the degradation status of the material. Information on such fracture behavior of the material at transition region is mandatory to quantify the inherent safety margin available under such undesirable events. A lot of studies have been carried out on the irradiation behaviors of nuclear materials to solve the practical problems associated with a nuclear reactor and to meet the desired mechanical properties to the future design of nuclear reactor. This paper aims to highlight the issue of irradiation in bimetallic welds. KEYWORDS: Bimetallic welds, Irradiation, Embrittlement.

INTRODUCTION This Paper focuses on the effects of neutron irradiation on the mechanical & microstructure behaviour of bimetallic welds used in nuclear reactor pressure vessels. Due to the change in mechanical properties like strength and fracture toughness of the vessel wall materials neutron exposure is a rising anxiety for operating reactors. Such change in properties is known as radiation embrittlement, and when this exposure remains continuous, becomes origin for long-term effects on the structural integrity of reactor pressure vessels. Thus, it is vital to monitor and manage the amount of radiation damage to assure full life time operation of nuclear plants and longterm structural integrity of reactor pressure vessels. In order to control the effects of irradiation, it is necessary to have a good understanding of the mechanisms causing embrittlement in the bimetallic welds that are used in nuclear reactor pressure vessels. Basic understanding of the effects of neutron irradiation damage comes from studies of unirradiated and irradiated materials in power and test reactors. Mechanical test specimens in these reactors have provided a significant amount of data for determining the relevant damage parameters as well as mechanical properties trends. Most of the data currently available are from Charpy V-notch and tension test specimens which are now used to measure the change in properties due to irradiation exposure. These data have been utilized to develop trend curve prediction methods for the embrittlement process which include the effects of variables such as copper, nickel, phosphorus, and neutron fluence in a given temperature range. BIMETALLIC WELDS (BMW’S) For a long time now, bimetallic welds (BMWs) have been a necessity within the pressurized water reactor (PWR) designs which uses enriched (about 3.2% U 235) uranium dioxide as a fuel in zirconium alloy Int. J. Adv. Engg. Res. Studies/IV/II/Jan.-March,2015/167-170

cans. The fuel, which is arranged in arrays of fuel "pins" and interspersed with the movable control rods, is held in a steel vessel through which water at high pressure (at 16 MPa, to suppress boiling) is pumped to reactor vessel via. cold lag pipeline to act as both a coolant and a moderator. After taking heat from core of reactor, the high-pressure water is passed to a steam generator, via. hot lag pipeline which raises steam in the usual way. Both hot lag and cold lag pipe lines are joined by reactor vessel nozzle by a bimetallic weld joint. So it is required to use bimetallic welds, to get the benefit of different properties of two or more different metals and to overcome the problems such as size and other functional requirements like in case of nuclear reactor vessels. These bimetallic welds are used in manufacturing of equipment to satisfy different functional requirements of components. So that the bimetallic weld (BMWs) play a critical and indispensable role in the primary heat transport piping system in the pressurized water reactor. [1,2]. IRRADIATION EFFECTS It is well known that bombardment of metals by energetic neutrons induces considerable changes in their physical and mechanical properties. An increase in yield stress associated with a reduction in ductility was observed in RPV steels used for cladding of RPV. There is considerable difference between the irradiated & unirradiated microstructure of the steel. The variation is depicted in figure 1[3].

Fig 1. Microstructure of steel grains


In the above figure the dark areas are sorbite grains & light areas are ferrite grains. The sorbite grains are very distinct, i.e., the grain boundaries are very clearly defined and always continue in nature. After irradiation, the boundaries are less distinct, becoming smaller, with their grain boundaries broken and unfinished. The embrittlement, neutron irradiation leads to changes of several material properties. These changes can be used as indicators for the state of the degradation, i.e., the decrease of fracture toughness. Neutron embrittlement represents a difficult problem especially in case for PWRs, where the bimetallic welds are attaching the piping system to the various nozzles of the reactor pressure vessel (RPV) and steam generators (SG). Damage Process through Irradiation Exposure The basic mechanisms of embrittlement can be described as follows:  Embrittlement is mostly due to microstructural changes which result in irradiation-induced increases in yield strength which can be quantitatively related to changes in the Charpy V-notch curve, as well as hardness.  Yield strength increase-related changes in fracture properties are due to irradiationinduced fine scale microstructures (about 1 nm) which act as barrier to dislocation motion.  The most likely resultant microstructures are precipitates, micro-voids and interstitial clusters (dislocation loops). Additional possible types of irradiation-induced or enhanced precipitates are phosphides and small carbides.  Other important effects of irradiation are enhanced diffusion rates and defect clustering.  These microstructural changes are kinetic phenomena and are functions of neutron exposure and temperature as well as composition and initial steel microstructure. In general, the changes in the microstructure can be understood from basic principles of alloy thermodynamics and precipitation kinetics, coupled with rate theories of radiation damage. A noticeably greater difference in radiation sensitivity has been observed in weld materials as compared to plates and forgings for approximately the same chemistry and neutron exposure. Microscopic Causes of RPV Embrittlement The microscopic causes of embrittlement lie in the forming of obstacles to dislocation motion, called “hardening centers”, as well as to change the composition and structure of the microscopic interfacial regions along which crystal plane sliding occurs. Both these phenomena are caused by several types of radiation-matter interactions, most of which are, in PWRs, induced by fast neutrons. Also γ-rays, in less frequent circumstances, may give a substantial contribution. From this consideration it follows that fast neutron irradiation is not the only cause of vessel wall embrittlement, even though its contribution is very often the dominant one. KEY FACTORS AFFECTING REACTOR VESSEL EMBRITTLEMENT The studies identify three key factors which affect the radiation damage of reactor pressure vessel steels. These factors include the neutron irradiation and the irradiation temperature. The other key factor is the Int. J. Adv. Engg. Res. Studies/IV/II/Jan.-March,2015/167-170

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material itself which can be different in chemical elements, microstructure, and processing history. The time integrated neutron flux exposure results in continuing loss in toughness and other changes in mechanical properties for most pressure vessel steels until, in some cases, a saturation level is reached. The degree of irradiation damage depends largely on the particular materials involved, especially the microstructure of welds and the amount of neutron irradiation. Neutron Irradiation Neutron irradiation is exposure to penetrating radiation. Irradiation occurs when all or part of the body is exposed to radiation from neutron source. The reactor pressure vessels of commercial power plants are subject to embrittlement mainly due to exposure of high-energy neutrons from the core. Thermal neutrons and γ-rays are also additional sources of damage. For a given exposure, copper, phosphorus and nickel are recognized as the major deleterious element of concern for irradiation degradation of material properties. However, an accurate prediction of the radiation damage cannot account only for the consequences of the occurrence of a postulated collision event, but also for the probability that this event can actually occur. Since boron is often present in RPV steels as impurity, neutron irradiation can yield helium production. Then, the bubbles so formed tend to coalesce at the grain boundaries, initiating the so-called grainboundary cracking, which causes structure embrittlement. Neutron Embrittlement Mechanisms The understanding of irradiation embrittlement of the pressure vessels of nuclear reactors is a key issue for plant lifetime assessment and much effort has been done in the last decades to tackle this complex issue. For the understanding of mechanisms of embrittlement, and the role of elements like Ni, Cu and P, the use of model alloys has proven to be a key methodology in connection with the utilization of material testing reactors or suitable channels of commercial ones. The embrittlement of steels exposed to radiation fields (neutron and γ-radiation mainly) is directly and indirectly caused by displacements of atoms from their original positions due to collisions by energetic particles. After the collision, a so-called displacement cascade is formed containing several vacancies in the middle, surrounded by a cloud of interstitial atoms. Most of the defects recombine and relatively minute defects survive to become freely migrating point defects, which, together with collapsed cascades, take part in the radiation induced damage processes. The radiation damage of RPV materials is proportional to the number of neutrons hitting the material with high magnitude energy to induce atom displacement(s). The embrittlement mechanisms taking place during neutron irradiation of reactor pressure vessel (RPV) steels and welds have been analyzed in several papers [2-4]. The three basic mechanisms contributing to neutron embrittlement for both steels and welds are: direct matrix damage, irradiation induced precipitation and elements segregation. In spite of this fact the models for analysis of radiation embrittlement are mainly based on statistical correlation of large sets of data.


A critical-temperature-shift-based methodology is used for assessment of irradiation embrittlement of pressure vessel steels. Conventionally, the Charpy temperature shift methodology is utilized. Moreover, the attempts are made to use the Master curve technique to solve this problem. Recently, the local approach to fracture (LAF) has been developed, which can be used as a powerful tool for solving this problem. In the present case, the behaviour of fracture mechanism can be understood by using GTN model [5-6] for ductile fracture & Beremin’s models [7-9] for cleavage fracture. Irradiation effects on Mechanical property of bimetallic Welds Embrittlement due to radiation hardening can be quantitatively related to changes in the yield strength following neutron exposure. Microscopically it is possible to describe the role of irradiation produced defects as barriers to dislocation motion which results in hardening and strengthening of the irradiated material. The increase in the yield strength resulting from the presence of defect aggregates has been shown to be temperature independent in irradiated welds. This temperature independent increase in yield stress can in turn be used to explain the shift in transition temperature. This increase then limits the degree of uniform elongation possible thereby creating a "less forgiving" medium under conditions which could be serious to the integrity of a potentially flawed vessel since elastic deformation is reduced. Thus, there is less tolerance for correction in a deformed and flawed vessel if an "overpressure" transient should occur. This projection may be extended to the potential ductile rupture or brittle fracture of a residual (unruptured) section coupled with a rapid separation in the vessel. This scenario must be avoided at all costs in the primary system nuclear pressure vessel. Changes in Transition Temperature (DBT) In bimetallic welds structure the ferritic low alloy steels mostly used and it commonly exhibits the phenomenon of ductile-to brittle transition over a relatively narrow temperature range (about 100°C or less). This range involves a transition from ductile (fracture only under conditions of plastic overloading or tearing) to brittle (fast fracture under elastic loading conditions) as the temperature is reduced. Over this range the micro- and macroscopic nature of the fracture gradually changes from ductile dimpled rupture, with attendant plastic deformation of the adjacent matrix, to cleavage along crystallographic boundaries. This ductile to brittle transition is characterized by a sudden and dramatic drop in the energy absorbed by a metal subjected to impact loading. This transition is practically unknown in FCC metals but is well known in bcc metals. As temperature decreases, a metal's ability to absorb energy of impact decreases. Thus its ductility decreases. At some temperature the ductility may suddenly decrease to almost zero. This transition is often more abrupt than the transition determined by the energy absorbed. This temperature is called the nil-ductility transition temperature (NDT). Micromechanical Modelling Bimetallic welds are used extensively in the construction of nuclear reactor pressure vessels (RPV). Though these welds have a high strength, they are susceptible to a change in fracture mode Int. J. Adv. Engg. Res. Studies/IV/II/Jan.-March,2015/167-170

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from ductile to brittle. This change in fracture behaviour may be due to reductions in temperature, radiation exposure, increased strain rate/stress triaxiality (i.e. constraint) or a combination of these factors. In order to design these welds to exist at cryogenic temperatures, so it is important to understand the micro mechanisms of fracture. The concept of micromechanical modelling as a method for better understanding of fracture mechanism and the characterization of materials is embedded in continuum mechanics. It combines theoretical, experimental and numerical methods enable a less conservative assessment in structural integrity on a micro scale level. It describes the phenomenological, smoothed behaviour of volume elements containing inhomogeneous microstructure by means of uniform stresses and strains. Cleavage fracture and ductile fracture/tearing are competing mechanisms in the ductile/brittle transition regime of bimetallic weld. In this regime, a weld structure can withstand significant amounts of ductile tearing without substantial loss of its load bearing capacity. Experimental data on cleavage fracture tend to be highly scattered. Two reasons have been offered for the large variations in measured cleavage fracture toughness at a given temperature within the transition region. First, cleavage fracture toughness in the transition region is controlled in large part by statistical sampling of a critical cleavage crack nucleus. Second, the level of stress triaxiality ahead of the crack tip is a strong function of the geometry of the crack and the amount of crack growth. Therefore the population of eligible particles for cleavage fracture depends implicitly on crack geometry and this population changes as the crack grows because of alterations in stress triaxiality. In general, the behaviour of fracture mechanism can be understood by using GTN model for ductile fracture & Beremin’s models for cleavage fracture. CONCLUSION This study covers the effects of neutron irradiation on mechanical properties of reactor pressure vessel steels. Major properties affected by high energy neutron exposure are hardness (a physical property) and the associated mechanical properties like strength and toughness. These variation caused by the environmental and materials factors can have severe effects leads to harm to vessel structural integrity and hence the life of a nuclear reactor pressure vessels. The factors of neutron environment (flux, fluence, and energy spectrum) and temperature also have major effects. The most critical factor in the quantitative assessment of embrittlement is steel composition. Further, the role of microstructure is a qualitative factor in the change of mechanical properties with neutron exposure, but more research is required to provide specific and quantitative guidance to these effects and this work tries to enlighten the crucial issue of loss of structural integrity due to neutron embrittlement. Suggestions of fluence rate effects have been made, but examination conditions in previous research shows limited rate effects. In the past investigators believe temperature variations can be so devastating as to cover any flux effect. More research on this subject is required before quantitative assessment of a rate effect can be assigned. Damage or embrittlement can be assessed and assigned in most vessels based upon


knowledge of the weld composition and metallurgical microstructure, neutron irradiation and temperature. From this knowledge, trend curves relating these factors to fracture toughness provide a limiting condition from which operators may judge integrity provided flaw sizes and stresses at critical vessel locations. At the same time with the help of GTN Beremin’s models one can carry out categorization of the mechanical behavior and micro structural assessment of the bimetallic weld at cryogenic temperatures as well as tests can be easily conceded for determination of fracture attributes of the real components at cryogenic temperature through micromechanical modeling approach. It is stated that such operations generally source for two failure modes i.e. ductile & cleavage fractures and working under cryogenic environment the bimetallic welds are very responsive to the failures due to consequences of irradiation & embrittlement behaviour. In order to understand these two failure modes it is call for use of probabilistic models such as GTN & Beremin’s, which are explained in the study work. With the help of this study it may become quite implausible job to apply the experiments on bimetallic joints used in reactor pressure vessels and access the integrity & safety of vessels to improve the life cycle of nuclear reactor. REFERENCE 1.

Faidy C., “BIMET: Structural integrity of Bi-Metallic Components- Program overview”, SMIRT 15. Paper G02-6, Seoul, KOREA August 1999. Nevasmaa 2. Chhibber, R; Arora, N; Gupta, S; Dutta, B. K., “Use of Bimetallic welds in Nuclear reactors: Associated problems and Structural integrity assessment issues”, Journal of Mechanical Engineering Science, Volume 220, Number 8, pp.1121-1133, 2006. 3. O. P. Maksimkin, F. A. Garner “Phase Transformations Observed In EP-450 Ferritic/Martensitic Steel Irradiated at 30 0c to 40.3 Dpa in the Bn-350 Fast Reactor” Institute of Nuclear Physics, National Nuclear Centre, Alma Ata, Kazakhstan and Pacific Northwest National Laboratory 4. Odette G.R. and Lucas G.E., (2001), “Embrittlement of Nuclear Reactor Pressure Vessels”, journal JOM, 53(7) 92001), pp. 18-22. 5. A. L. Gurson (1977) Continuum theory of ductile rupture by void nucleation and growth: Part I—yield criteria and flow rules for porous ductile media. J. Engng Mater. Tech. 99, 2–15. 6. Needleman A, Tvergaard V. An analysis of ductile rupture modes at a crack tip. J Mech and Phys of Solids 1987; 35:151-83. 7. F.M. Beremin, A local criterion for cleavage fracture of a nuclear pressure vessel steel, Metall. Trans. A. vol. 14A (1983) 2277-2287. 8. F. Mudry (1987) A local approach to cleavage fractures. Nuclear Engng Design 105, 65–76. 9. B. K. Dutta, et al. (2007) “Temperature Dependency of Beremin’s Parameters for 20MnMoNi55 Material”, BARC Mumbai. Note: This Paper/Article is scrutinised and reviewed by Scientific Committee, BITCON-2015, BIT, Durg, CG, India

Int. J. Adv. Engg. Res. Studies/IV/II/Jan.-March,2015/167-170

E-ISSN2249–8974