G. Vaidyanathan

Design, Development, Testing and Qualification of Diverse Safety Rod and Its Drive Mechanism for a Prototype Fast Breeder Reactor

Prototype fast breeder reactor is U-PuO 2 fueled sodium cooled pool type fast reactor and it is currently under construction at Kalpakkam, India. Prototype fast breeder reactor is equipped with two independent fast acting and diverse shutdown systems. A shutdown system comprises of sensors, logic circuits, drive mechanisms, and neutron absorbing rods. The two shutdown systems of prototype fast breeder reactor are capable of bringing down the reactor to cold shutdown state independent of the other. The absorber rods of the second shutdown system of prototype fast breeder reactor are called as diverse safety rods (DSRs) and their drive mechanisms are called as diverse safety rod drive mechanisms (DSRDMs). DSRs are normally parked above active core by DSRDMs. On receiving scram signal, the electromagnet of DSRDM is de-energized and it facilitates fast shutdown of the reactor by dropping the DSR into the active core. For the development of prototypes of DSR and DSRDM, three phases of testing, namely, individual component testing, integrated functional testing in room temperature, and endurance testing at high temperature sodium, were done. The electromagnet of DSRDM has been separately tested at room temperature, in furnace, and in sodium. Specimens simulating the contact conditions between electromagnet and armature of DSR have been tested to rule out self-welding possibility. The prototype of DSR has been tested in flowing water to determine the pressure drop and drop time. The functional testing of the integrated prototype DSRDM and DSR in aligned and misaligned conditions in air/water has been completed. The performance testing of the integrated system in sodium has been done in three campaigns. During the third campaign of sodium testing, the performance of the system has been verified with 30 mm misalignment at various temperatures. The third campaign has qualified the system for 10 years of operation in reactor. This paper presents design, development, testing, and qualification of the prototype DSR and DSRDM. Salient design specifications for both DSRDM and DSR are listed initially. The conceptual and detailed design features are explained with the help of figures. Details on material of construction are given at appropriate places. Test plans and criteria for endurance testing in sodium for qualification of DSRDM and DSR for operation in reactor are briefed. Brief explanation of test setups and typical test results are also given.

Application of Acoustic Technique for Surveillance and Anomaly Detection in LMFBRs

Acoustic techniques find wide application in Liquid Metal Fast Breeder Reactors (LMFBRs) for ensuring its high reliability, safety and plant availability. Various surveillance methods based on acoustic technique can be employed in these reactors to detect deviations from normal operating conditions. This could be used for the measurement of drop time of Diverse Safety Rods (DSRs) in the core, detection of in-sodium water leaks in Steam Generators, cavitation detection in sodium pumps and reactor core components. An active R&D program is being pursued in these areas at Indira Gandhi Centre for Atomic Research. Acoustic measurement technique has been used to determine the drop time of Diverse Safety Rods in sodium. 3 nos of Diverse Safety Rods (DSRs) are provided in Prototype Fast Breeder Reactor (PFBR) for its safe shut down in case of a SCRAM. An online drop time measurement system using acoustic technique is planned to detect the proper insertion of DSRs into their corresponding Subassemblies. Experiments were conducted during the performance testing of DSRs in sodium using accelerometer instrumented wave-guide system and free fall time and braking time of DSR have been measured. For detection of in-sodium water leaks in Steam Generators, acoustic method serves as a supplementary monitoring technique with an intermediate sensitivity and instantaneous response. To develop an acoustic leak detection system for Steam Generators of Prototype Fast Breeder Reactor, preliminary studies on the behavior of micro leak and its propagation has been carried out in Sodium Water Reaction Test Rig, injecting steam into sodium. Acoustic technique was employed to characterize the onset of leak. Cavitation in LMFBRs can occur in fuel subassemblies, pressure drop devices, pumps etc. It is important to minimize cavitation to reduce the risk of damage from erosion. Acoustic technique was extensively used in qualifying Prototype Fast Breeder Reactor components against cavitation phenomenon. This paper discusses the various experiments carried out towards the development of the acoustic surveillance methods for FBRs, instrumentation involved, results obtained from experiments and brief details of the future programme.© 2008 ASME

Assessment of Dynamic Instability in Once Through Steam Generator

Hydrodynamic flow instability in Once Through Steam Generators (OTSG) is one of the important problems in the design and operation of Liquid Metal Fast Breeder Reactors (LMFBRs). Under certain operating conditions, water flow in OTSG is susceptible to instability due to the close coupling between the thermal and hydraulic processes. Sustained flow oscillations due to instability are undesirable since they result in flow mal-distribution among the tubes resulting in thermal stress, mechanical vibrations and system control problems. It is therefore, necessary to assess the operating conditions, under which instability occurs so that the system may be designed to operate always under stable conditions. The cause of the main type of instability, important for the design of SGs is the propagation of density waves. This type of low frequency instability is referred to in literature as parallel-channel, density wave, time delay or mass flow-void feedback instability. Dynamic instability (density wave oscillation DWO) occurs because of the phase mismatch between the primary perturbation (water flow) and the response to this perturbation (pressure drop). As many tubes are operating under essentially constant pressure heads, this mismatch can lead to sustained/diverging oscillations. Water flow oscillation in tubes manifests as oscillations in the steam temperature at the tube outlet/pressure fluctuations. However it is difficult to instrument individual tubes in SG for such measurement in an operating plant. If the flow oscillation in the tube manifests itself in the overall module flow, then fluctuation in the overall flow/flow noise could be utilized for on-line stability measurements. Towards this, experiments were conducted in the sodium heated once through steam generator in an OTSG model. To confirm the extent of oscillation in the steam temperature and in inlet water flow, 3 tubes out of 19, were monitored besides overall module flow. Main objective of the present study was to assess the occurrence of dynamic instability in SG through module inlet flow perturbations, measured by ΔP measurements across the orifice at entry to the tubes and steam temperature fluctuation measurement at the outlet of tube by bare thermocouples. This paper discusses the experiments carried out in the Steam Generator model of Prototype Fast Breeder Reactor to investigate the instability phenomenon, the instrumentation details, the results and its discussion.Copyright