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Abstract: The 500MWe Prototype Fast Breeder Reactor (PFBR) is under construction at Kalpakkam, India. The main vessel of this pool type reactor acts as the primary containment in the reactor assembly. In order to keep the main vessel temperature below creep range and to reduce high temperature embrittlement and also to ensure its healthiness for 40 years of reactor life, a small fraction of core flow (0.5m3/s) is sent through an annular space formed between the main vessel and a cylindrical baffle (primary thermal baffle) to cool the vessel. The sodium after cooling the main vessel overflows the primary baffle (weir shell) and falls into another concentric pool of sodium separated from the cold pool by the secondary thermal baffle and then returned to cold pool. The weir shell, where the overflow of liquid sodium takes place, is a thin shell prone to flow induced vibrations due to instability caused by sloshing and fluid–structure interaction. A similar vibration phenomenon was first observed during the commissioning of Super-Phenix reactor. In order to understand the phenomenon and provide necessary experimental back up to validate the analytical models, weir instability experiments were conducted in a 1:4 scale stainless steel (SS) model installed in a water loop. The experiments were conducted with flow rate and fall height as the varying parameters. The experimental results showed that the instability of the weir shell was caused due to fluid structure interaction. This paper discusses the details of the model, the modeling laws, similitude criteria adopted, analytical prediction, the experimental results and conclusion. [Copyright &y& Elsevier]

Abstract: The construction of Prototype Fast Breeder Reactor (PFBR), a 500MWe liquid sodium cooled reactor, is in progress at Kalpakkam in India. Control plug (CP) is located right above the core subassemblies in the hot pool. Control plug is an important component as many of the critical reactor parameters are sensed and controlled by the components housed in the control plug assembly. In PFBR primary circuit, components are basically thin walled, slender shells with diameter to thickness ratio ranging from 100 to 650. These components are prone to flow induced vibrations. The existence of free liquid (sodium) surfaces, which is the source of sloshing phenomenon and the operation of primary sodium pump in the primary pool are other potential sources of vibration of reactor components. Control plug is a hollow cylindrical shell structure and provides passages and support for 12 absorber rod drive mechanisms (ARDM) which consists of 9 control and safety rods and 3 diverse safety rods, 210 thermo wells to measure the sodium temperature at the exit of various fuel subassemblies, three failed fuel localization modules (FFLM) and acoustic detectors. It consists of a core cover plate (CCP), which forms the bottom end, two intermediate supports plate, i.e. lower stay plate (LSP) and upper stay plate (USP) and an outer shell. The CCP is located at a distance of 1.3m from the core top. With such a gap, there will be long free hanging length of the thermocouple sleeves, Delayed neutron detector (DND) sampling tubes and ARDM shroud tubes and hence they are susceptible to flow induced vibrations. Since control plug is partially immersed in hot sodium, the reactor transients are felt by the components, hence it is very much essential to understand the vibration response of the control plug components. The main vibration excitation mechanisms are vortex shedding, turbulence buffeting, fluid-elastic instability, etc. In order to assess the susceptibility of CP internals against flow induced vibrations (FIVs), to measure structural response and to validate the analytical codes developed for FIV analysis of CP along with the assumptions of supports for various CP internals therein, a flow induced vibration testing program was formulated in 1:4 scale model of PFBR CP. As the first phase of this program, experimental modal analysis of CP internals was carried out in air to estimate the modal parameters. Subsequently, flow induced vibration studies were conducted in the 1:4 scale model of CP in water. In this model, size of the tubes, shell and plates are reduced to ¼ size of the PFBR CP. The stiffness of CP parts is reduced by 4 times and mass decreased by 64 times which results in scaling up of modal frequencies by 4 times. The CP internals in the model were instrumented with accelerometers and strain gages. The studies were conducted in water with flows derived on the basis of velocity similitude. The output signals from the sensors were acquired and analyzed to obtain frequency spectra, overall vibration amplitude and strain values at various locations inside CP. The study carried out confirms the absence of resonance due to flow induced vibration mechanisms for the entire operating range. This paper elucidates the modeling details, similitude criteria, instrumentation employed and experimental results obtained with discussion on results. [Copyright &y& Elsevier]