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The accident of Fukushima nuclear power plant (NPP) underscores the necessity of conducting probabilistic safety assessment (PSA) of an NPP against the hazards caused by external events of natural origin. One of the important tasks of an external event PSA (EE-PSA) of an NPP is determination of component fragility. The fragility of a component is the probability of its reaching a limit state condition on a particular level of demand caused by the hazard. The seismic fragility of component is expressed by means of a double lognormal model with three parameters – the median ground acceleration capacity A m , random variability β R and uncertainty β U in the median capacity. There are direct and indirect methods to calculate fragility parameters A m , β R and β U . Direct method involves either analysis or testing. Indirect method is experience based method that maximizes the use of past experience in conjunction with plant walk down. The parameters for wind fragility V m , β R and β U are estimated for different SSCs using design data (e.g. wall thickness, reinforcement, and anchorage) and variability associated with the wind response and capacity. The parameters for flood fragility H m , β R and β U are also estimated for different structures, systems and components (SSCs) using design data and variability associated with the flood response such as inundation depth and capacity such as installation height. However, it generally shows more cliff-edge shape without any special barriers (e.g. protective walls, water-tight doors and sealing countermeasures). It means that flood fragility is highly path-dependent and thus flood hazard is influenced by barriers’ fragility. The paper reviews the available open literature and summarizes the development of methods for deriving the fragility parameters of NPP components for use in PSA against seismic ground motion, wind and tsunami induced flood hazard. The paper also briefly discusses the component fragility for external event PSA of multiple NPP units at a site for natural hazards like earthquake, wind and flood. The state-of-the-art of seismic fragility analysis for NPP component is in matured level, while that of flood hazard in early stage of development and wind hazard in nascent stage. [ABSTRACT FROM AUTHOR]

Abstract: Seismic fragility of structure, system or component (SSC) is the probability of its reaching a limit state, for a given seismic demand. It is evaluated in terms of ground motion parameters, which is generally peak ground acceleration. Seismic fragility of a nuclear power plant (NPP) is derived from the fragility of its SSC. Seismic qualification, prerequisite for determination of seismic fragility of the SSC of a NPP, is conducted by either direct method involving analysis and testing; or indirect one involving experience based method. The paper surveys and summarizes the methods available to derive the seismic fragility of SSC(s) of an NPP, which are qualified by direct as well as indirect methods. [ABSTRACT FROM AUTHOR]

Abstract: A number of shake table tests had been conducted on the scaled down model of a concrete wall as part of CAMUS experiment. The experiments were conducted between 1996 and 1998 in the CEA facilities in Saclay, France. Benchmarking of CAMUS experiments was undertaken as a part of the coordinated research program on ‘Safety Significance of Near-Field Earthquakes’ organised by International Atomic Energy Agency (IAEA). Technique of deflection-based method was adopted for benchmarking exercise. Non-linear static procedure of deflection-based method has two basic steps: pushover analysis, and determination of target displacement or performance point. Pushover analysis is an analytical procedure to assess the capacity to withstand seismic loading effect that a structural system can offer considering the redundancies and inelastic deformation. Outcome of a pushover analysis is the plot of force–displacement (base shear–top/roof displacement) curve of the structure. This is obtained by step-by-step non-linear static analysis of the structure with increasing value of load. The second step is to determine target displacement, which is also known as performance point. The target displacement is the likely maximum displacement of the structure due to a specified seismic input motion. Established procedures, FEMA-273 and ATC-40, are available to determine this maximum deflection. The responses of CAMUS test specimen are determined by deflection-based method and analytically calculated values compare well with the test results. [Copyright &y& Elsevier]

Delamination of the inner (primary) containment dome of the Kaiga Atomic Power Project, Unit-1 (Kaiga-1) occurred during construction. Subsequent to the investigation of the delamination phenomenon, rehabilitation of the containment dome was initiated. A methodology for safety evaluation of rehabilitation of the containment dome was suitably devised based on the outcome of the investigation work. Emphasis was laid, during safety evaluation, on the safety of the existing ring beam, development of review basis for the re-engineered dome, design review and quality assurance in design, quality assurance of materials used, examination of constructibility using the new design, and safety review of construction activities of the re-engineered dome. The inner containment structure was accepted after the successful pre-commissioning proof (structural integrity) and leakage rate tests. [Copyright &y& Elsevier]

• The site evaluation is a multi-task process consisting of site characterization, derivation of design basis of external natural and human induced events and examination of implementation of emergency management planning. • The subject is challenging as site evaluation is not confined to technical activities only; it needs administrative and social input. • All these lead to different countries having different practices in defining site and its external zones, and for site evaluation activities at different stages of life cycle, management system, and regulatory interfaces. • Effort is made in this paper to outline the state-of-the-practice of site evaluation for nuclear power plants. • The site evaluation processes of three countries USA, France and India are reviewed and summarized. • The summary is then integrated with the requirements and guidance on this subject given in IAEA Safety Standard and insights gained from the NEA/CNRA document. Selection of appropriate site and robust site evaluation have significant impact on the design, construction and operation of nuclear power plant (NPP) in meeting the nuclear safety requirements following the concept of Defense-in-Depth (DID). The site evaluation is a continuing process over the life cycle activities of an NPP including its site selection. The objective of site evaluation is that the site-plant interaction does not cause radiation exposure to the public and environment beyond a threshold value during both normal operation and accident situations. The site evaluation is a multi-task process consisting of site characterization, derivation of design basis of external natural and human induced events and examination of implementation of emergency management planning. Radiological impact assessment and environmental (non-radiological) impact assessment are part of investigations of the suitability of site. The subject is challenging as site evaluation is not confined to technical activities only; it needs administrative and social input. All these lead to different countries having different practices in defining site and its external zones, and for site evaluation activities at different stages of life cycle, management system, and regulatory interfaces. Effort is made in this paper to outline the state-of-the-practice of site evaluation for nuclear power plants. The site evaluation processes of three countries USA, France and India are reviewed and summarized. The summary is then integrated with the requirements and guidance on this subject given in IAEA safety standard (IAEA, 2016) and insights gained from the NEA/CNRA document (NEA/CNRA, 2010). Experience gained from recent accidents/incidents of nuclear power plant is also studied and suitably incorporated in this review work. [ABSTRACT FROM AUTHOR]

Abstract: An experimental study has been conducted to know the structural behaviour of a representative test specimen for a typical Prestressed Concrete Inner Containment dome with due regards to the local behaviour at the steel plate–concrete interfaces at large steam generator openings, under initial condition at prestress transfer. The primary thrust of the work has been in the objective of predicting possibilities of separation at the steel–concrete interface zones adjacent to the embedded plates of the steam generator openings of the inner containment dome. Important observations in line with the desired objectives have been made based on the results of the experiment. [Copyright &y& Elsevier]

Abstract: A finite element (FEM) study has been made to know the overall structural behaviour of a PSC inner containment (IC) dome having large steam generator (SG) openings with emphasis on the local behaviour of the steel–concrete interfaces near SG openings, under initial prestress transfer. The primary thrust of the work has been in the objective of predicting the possibilities of separation at the steel–concrete interface zones adjacent to the embedded plates of the SG openings. For the FEM analysis, the interface zone has been modeled using gap elements, the properties of which are derived from the results of the past experiments conducted on steel plate–concrete interface specimens. Important observations have been made regarding dome deformation and the stresses with special emphasis on the local behaviour of steel–concrete interfaces at and around SG openings. [Copyright &y& Elsevier]