PSI - Issue 60

Gopal Sanyal et al. / Procedia Structural Integrity 60 (2024) 311–323 Author name / StructuralIntegrity Procedia 00 (2019) 000–000

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1. Introduction Ferritic steels are well-known materials of construction of various structural components in nuclear reactors (Gupta, 2001; Murty et. al. 2001). Reactor pressure vessels (RPVs) and reactor support structures in the pressurized water reactor (PWR) and boiling water reactor, heat transport piping, and end fitting in the pressurized heavy water reactor are some of the important examples of applications where ferritic steels were employed as component materials (Gupta, 2001). In the application of materials for structural applications as critical components of nuclear reactor systems, there exists a great concern for the influence of embrittlement phenomena on properties such as strength, toughness, and ductility that are essential for sustenance of designed functions of component throughout its lifetime. Thus, the research to prevent premature failures of components due to enervated technological properties under the influence of thermal, irradiation, andenvironmental embrittlement have resulted in the development of a number of alloys with better resistance to operating embrittlement phenomena (Briant and Banerji 1978). Low alloy steels are used as materials for nuclear reactor pressure vessel components. Two classes of low alloy steels are predominately used for the pressure vessel application. These are the Ni-Mn-Mo, ‘‘Western’’ vessel steels and the Cr-Mo-V, ‘‘Eastern’’ vessel steels (Davies, 1999). The modern methods of nuclear reactor pressure vessel production rely on the use of sections of forged rings that are welded together (Davies, 1999). Hydrogen uptake in steels is considered to be a cause of concern due to degradation of the load bearing ability caused by its presence in elevated levels. The possibilities of hydrogen uptake in components of steel can occur during fabrication (eg.during melting or in liquid weld pool) (Loidl et. al., 2011), in processes such as electroplating/pickling (Hillier, 2004), or due to service environment such as found in gas pipelines, battery etc. (Gabetta et. al., 2018). Generally, the absorbtion of hydrogen in steels may occur directly in a single step or via an adsorbtion step (Bockris et. al., 1965, Zheng et. al., 1995). In the latter, hydrogen that adsorbs on the steel surface by the Volmer reaction is ingressed in the matrix limited by the acidic or alkaline medium of environment, the overpotential or fugacity (Bockris et. al., 1965) and the extent of Tafel recombination reaction. Due to weak Fe – H bond (as compared with Fe – O bond) (Gabrielli et. al., 2006) on the steel surface at room temperature, generally high fugacities are required to efficiently enable the ingress of the adsorbed hydrogen in the matrix. Once inside the steel, hydrogen can interact with carbides to form methane and lead to damage called as hydrogen attack (Louthan, 2008). It can accumulate at inclusions or laminations to cause blistering or can precipitate within micropores as hydrogen gas (Zapffe and Sims, 1941)to cause cracking or can localize plastic flow by enhancing dislocation mobility and planar slip (Brinbaum and Sofronis, 1994; Robertson et. al., 2015) or reduce the cohesive strength (Oriani, 1987). With such diverse consequences of hydrogen led damage in steels particular attention has been given to develop the grade of low alloy steels for RPV applications with minimum hydrogen ingress during manufacture and fabrication of the vessel. Until recently the presence of hydrogen led damage during service in PWR has not been considered to be of serious issue. However, the detection of thousands of cracks, classified as hydrogen flaking, within a region of 30 – 120 mm from the water side of the belt line core ring section of Belgian nuclear power reactors Doel 3 and Tihange 2 after 30 years of service (Bogaerts, 2017), has rekindled the debate on implication of H-ingress from corrosion reactions and radiolysis of coolant water. Although these cracks were evaluated to be benign for the structural integrity of the vessels (Materials Reliability Program MRP-367,2013), the lack of inconvertible proof that the austenitic clad can altogether prevent the hydrogen generated by the coolant waterto reach and be ingressed in the underlying ferritic vessel as a result of long-term use under radiation, provides strong reasons to study the basic manifestation of hydrogen embrittlement in low alloy steels for RPV applications. The low alloy Cr-Mo-V steel, in the present study, was selected for a new generation reactor pressure vessel material. The reactor pressure vessel material, being one of the critical components of the PWR nuclear reactor, by virtue of being the housing for thein-core structural components, has naturally attracted intensive investigations concerning temper and irradiation embrittlement and dynamic strain ageing (DSA) (Gupta, 2010). Information on

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