V. Rajan Babu

Experimental Qualification of Mechanical and Electrical Sub-systems of a Complex Mechanism Against Fatigue Failure

Absorber rod drive mechanisms (ARDM) play an important role in ensuring safety of a reactor by rapid insertion of an absorber rod during abnormal conditions. Various components/sub-systems of ARDMs, both mechanical and electrical, are subjected to different cyclic loadings during service life. Thus, qualifying these systems against fatigue is an important step for gaining confidence in their safe operation for the design life. ASME in Sec. III, Div. 1, Appendices (Para II—1500) provides guidelines for the experimental evaluation of the capability of components to withstand cyclic loading. These rules are developed for static components like pressure vessels. Since no such rules are available for moving components like mechanisms, the same were adopted for the ARDMs, with an understanding that the effect of inertia loads of a moving component are to be accounted in the experiments. In application of these rules to a complex mechanisms such as ARDM, various special cases arise which are not addressed explicitly in the code. The paper describes the intelligent adoption of the fatigue life rules given in ASME to various special cases and their extension to electrical systems. The paper also outlines the experiments carried out for qualifying the ARDM against fatigue.

Design Improvement Studies for Future SFR

Prototype Fast Breeder Reactor (PFBR), a 500 MWe, (U-Pu)O2 fuelled, sodium cooled, pool type fast reactor, is in advanced stage of construction at Kalpakkam, India. Based on the experience gained during the design, manufacture and erection of various reactor components of PFBR, it is planned to construct Sodium cooled Fast Reactors (SFR) by adopting twin unit (2×500 MWe reactors) concept. The future Fast Breeder Reactor (FBR) – 1 & 2 have three main heat transport circuits, namely primary sodium, secondary sodium and steam-water systems. All the reactor internals including core and primary heat transport circuit systems are contained in a single vessel called main vessel and it is closed with top shield. Reactor assembly forms the heart of the Nuclear Steam Supply System. A detailed and exhaustive design / optimisation exercise was initiated towards improving the economic competitiveness and enhancing the safety of future FBRs. It is observed that the overall dimensions of the reactor assembly contribute immensely to the capital cost. In this context, detailed studies were carried out towards optimizing the overall dimensions of the reactor assembly. Further, the reactor assembly design in particular has been engineered to favour manufacture of integrated assembly and erection of the same, as a single unit, in reactor vault to reduce construction time. Various activities undertaken towards technology development of critical components have enhanced the confidence level in the improved design concepts and reducing time for manufacture and erection.In addition to the reactor assembly, specific improvements have been made in decay heat removal systems and sodium purification system. The layout incorporates a twin unit concept in which the ex-vessel fuel handling system and fuel storage building are shared. This paper discusses the basis for undertaking the review exercise and experience gained during construction of PFBR and highlights the design studies and technology development carried out for future SFR.© 2014 ASME

ICONE15-10859 EXPERIMENTAL EVALUATION OF INTEGRITY OF FBR CORE UNDER SEISMIC EVENTS

The core of Prototype Fast Breeder Reactor (PFBR) is designed to produce 1250 MWt at full power. PFBR is under construction at Kalpakkam, India. In PFBR, the core is of free standing type and one of the major safety criteria for the design of core subassemblies is that the integrity of the core subassemblies should not be impaired and they should not be lifted up from the grid plate even during seismic condition. The net downward force acting on the grid plate is less than the weight of the subassembly due to the hydraulic lifting forces acting on it. Experimental analysis has been carried out to ensure that the subassembly does not get lifted off due to vertical seismic excitation. This paper gives the details of the methodology adopted for the experimental seismic analysis carried out on a core subassembly and the upward displacement of the subassembly under the combined effect of upward fluid force and vertical seismic excitations.

Design and Qualification of Control and Safety Rod and Its Drive Mechanism of Fast Breeder Reactor

Prototype Fast Breeder Reactor (PFBR) has two shutdown systems. The absorber rod of the first system is called Control & Safety Rod (CSR). Control & Safety Rod Drive Mechanism (CSRDM) facilitates start-up & controlled shut-down of reactor and control of reactor power by raising and lowering of CSR and shutdown of the reactor on abnormal conditions by rapid insertion of CSR into the core, i.e., by scram action. After the detailed design and analysis of CSR and CSRDM, they were qualified in two stages. In the first stage, the critical assemblies of the mechanism, such as scram release electromagnet, hydraulic dashpot and dynamic seals, were tested individually simulating the operating conditions of the reactor and the design parameters were fine-tuned. Experiments were also carried out on sodium vapour deposition in the annular gaps between the stationary and mobile parts. In the second stage, prototype CSRDM and CSR were manufactured and subjected to various functional tests in air, in hot argon and subsequently in sodium simulating the operating conditions of the reactor. Tests were carried out keeping CSRDM and CSR at aligned condition and with the possible misalignment between them. The performance was checked and recorded maintaining the temperature of sodium starting from 473 K to 823 K. Then the system was subjected to endurance tests. The results show that the performance of CSRDM and CSR is satisfactory and there is no significant change in the performance during endurance testing.© 2006 ASME

Performance Evaluation of Control and Safety Rod and Its Drive Mechanism of Fast Breeder Reactor During Seismic Event

Control and safety rods & their drive mechanisms (CSRDM) and diverse safety rods & their drive mechanisms (DSRDM) are the main constituents of the two independent shutdown systems in Prototype Fast Breeder Reactor (PFBR) of 500 MWe. There are nine CSR and three DSR placed within the hexagonal sheaths, which in turn are located in two radial banks of reactor core. This paper deals with the analysis carried out to predict the performance of CSRDM along with CSR. Analysis is carried out for the static and seismic loading under the fuel handling as well as normal operating conditions with the objective of ensuring structural integrity as well as to estimate the drop time during seismic event. The effects of bowing of sub-assemblies corresponding to the end of life situation have been considered. From the analysis, it is found that the drop time of CSR is 0.82 s, which is less than 1 s, and hence there is no concern of reactor shutdown. Further, it is ensured that there is no mechanical interaction of concern between various parts. The induced stresses are found to be very much less than RCC-MR allowable stress intensity. Thus the performance of CSRDM and CSR is demonstrated to be sound during normal as well as seismic events.Copyright