PERFORMANCE EVALUATION OF CONTROL & SAFETY ROD AND ITS DRIVE MECHANISM OF FAST BREEDER REACTOR DURING SEISMIC EVENT

ABSTRACT 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.

ICONE17-75851 DESIGN AND DEVELOPMENT OF DIVERSE SAFETY ROD AND ITS DRIVE MECHANISM FOR PFBR

ABSTRACT Prototype Fast Breeder Reactor (PFBR) is U-PuO 2 fuelled sodium cooled Pool type Fast Reactor and it is currently under advanced stage of construction at Kalpakkam, India. The Fast Breeder Test Reactor (FBTR) which is the only fast reactor currently operational in India is having only one shutdown system. However the IAEA and Atomic Energy Regulatory Board (AERB) Guide Lines call for two independent fast acting diverse shutdown systems for the present generation reactors. Hence PFBR 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 PFBR are capable of bringing down the reactor to cold shutdown state independent of the other. The absorber rods of the second shutdown system of PFBR are called as Diverse Safety rods (DSR) and their drive mechanisms are called as Diverse Safety Rod Drive Mechanisms (DSRDM). DSR are normally parked above active core by DSRDM. On receiving scram signal, Electromagnet of DSRDM is de-energised and it facilitates fast shutdown of the reactor by dropping the DSR in to the active core. This paper presents chronological design and development of the prototype DSR and DSRDM starting from the design specifications. Salient design specifications for both DSRDM and DSR are listed initially. The conceptual & detailed design features are explained with the help of figures. Various important design options considered in the initial design stage, choice of final design along with brief explanation for the particular choice are also given for some of the important components. Details on material of construction are given at appropriate places. Details on various analysis such as large displacement analysis for buckling, bending analysis for determining reactive forces and friction in the mechanism, thermal stress analysis of electromagnet during scram, flow induced vibration analysis of DSRDM and DSR and hydraulic analysis for estimating the pressure drop and drop time of DSR are also given. Test plans for design verification, manufacturing and shop testing experience of prototype systems, and criteria for endurance testing in sodium for qualification of DSRDM and DSR for operation in reactor are also briefed

Residual stress measurement round robin on an electron beam welded joint between austenitic stainless steel 316L(N) and ferritic steel P91

This paper is a research output of DMW-Creep project which is part of a national UK programme through the RCUK Energy programme and India's Department of Atomic Energy. The research is focussed on understanding the characteristics of welded joints between austenitic stainless steel and ferritic steel that are widely used in many nuclear power generating plants and petrochemical industries as well as conventional coal and gas-fired power systems. The members of the DMW-Creep project have under- taken parallel round robin activities measuring the residual stresses generated by a dissimilar metal weld (DMW) between AISI 316L(N) austenitic stainless steel and P91 ferritic-martensitic steel. Electron beam (EB) welding was employed to produce a single bead weld on a plate specimen and an additional smoothing pass (known cosmetic pass) was then introduced using a defocused beam. The welding re- sidual stresses have been measured by five experimental methods including (I) neutron diffraction (ND), (II) X-Ray diffraction (XRD), (III) contour method (CM), (IV) incremental deep hole drilling (iDHD) and (V) incremental centre hole drilling (iCHD). The round robin measurements of weld residual stresses are compared in order to characterise surface and sub-surface residual stresses comprehensively

Effect of Dopants on Zirconia Stabilization—An X-ray Absorption Study: I, Trivalent Dopants

Peer Reviewedhttp://deepblue.lib.umich.edu/bitstream/2027.42/65562/1/j.1151-2916.1994.tb06964.x.pd

Design basis and materials requirements for sodium heated steam generators for fast reactors

An attempt has been made to categorise the various materials, which have been considered for a sodium heated steam generator. Choice of steam generator, selection of material, problems of sodium heated generators, R&D efforts for indigenous materials for steam generators etc. have been discussed. It is concluded that stabilised ferritic steel will be a good selection as a sodium heated steam generator material under Indian Conditions. (Sarvashri S.C. Chetal & M.C. Sabherwal, Reactor Research Centre, Deptt. Of Atomic Energy, Kalpakkam

Application of advanced mechanics for the structural design of sodium cooled fast reactor components

Some of the safety critical components of sodium-cooled fast reactors (SFR) are large dimensioned thin-walled shell structures. These components hold a large sodium mass and are subjected to high temperatures (820 K), high temperature gradients and cyclic thermal loads. The components are designed for long design life (>40 y). These features introduce complicated cyclic thermo-plastic deformations, the need for applying the principles of fracture mechanics under high cycle thermal fatigue, structural instabilities, fluid-structure interactions including elastic instabilities and high strain rate deformations. Hence, the structural design of such components is performed by addressing the relevant concepts in mechanics. In this paper, results of application of such advanced concepts for the structural design of components of the 500 MWe Prototype Fast Breeder Reactor (PFBR), which is under construction at Kalpakkam, are highlighted. New rules are recommended for inclusion in design codes like RCC-MR

Effects of nuclear island connected buildings on seismic behaviour of reactor internals in a pool type fast breeder reactor

The seismic analysis of reactor assembly housing the primary circuit of a typical 500 MWe capacity pool type fast breeder reactor (PFBR) is reported. The reactor assembly is supported on the reactor vault within the nuclear island connected buildings (NICB). The seismic responses, viz. critical displacements, sloshing heights, stresses and strain energy values in the vessels are determined for the reactor assembly by detailed finite element analysis including the fluid-structure interaction and sloshing effects. Analysis is carried out to quantify the effects of inter-connection of the reactor vault with the adjacent buildings under the assumptions that the reactor vault along with reactor assembly is: (1) an isolated structural system from the adjacent buildings within reactor containment building (RCB) and (2) connected with the adjacent civil structures through floor slabs. Analysis indicates that, by inter-connecting the vault with the NICB, there are overall increases of all the governing parameters which decide the seismic design criteria. The significant effects are increases of: (1) radial and axial displacements of core top and absorber rods and vertical accelerations of core subassemblies which are of concern to reactor safety, (2) primary membrane stress intensities for the inner vessel and (3) strain energies developed at the critical portions which can enhance the buckling risks of main vessel, inner vessel and thermal baffles. Hence, it is preferable to isolate the reactor vault, directly constructing from the base raft without inter-connecting it with the NICB, from the seismic loading considerations

Investigation of fluid-elastic instability of weir shell in a pool type fast breeder reactor

This article does not have an abstract

Investigation on buckling of FBR vessels under seismic loadings with fluid structure interactions

The critical thin walled shell structures in the reactor assembly of a pool type fast breeder reactor (FBR) are the main vessel, inner vessel and thermal baffles. On these structures, the seismic events impose major forces by developing high dynamic pressures, thereby causing a concern on structural integrity due to buckling. An integrated analysis for determining realistic forces and critical buckling loads at any instant during the seismic event has been carried out for the vessels of a typical 500 MWe pool type fast breeder reactor. The dynamic forces including pressure distributions generated on the vessel surfaces are extracted from the seismic analysis carried out for the reactor assembly. The seismic forces thus computed from axisymmetric analysis are transmitted appropriately to 3D shell geometries for the buckling analysis. In view of high computational time needed for carrying out buckling analysis at every time increment, the elastoplastic analysis is carried out only at a few critical time steps which are identified based on strain energies that are associated with the shear and compressive stresses developed at the portions of the vessels prone to buckle. The shear buckling of main vessel straight portion and buckling of toroidal portion of inner vessel and thermal baffles are found to be important. The possible randomness of support excitation time histories is accounted for by compressing as well as expanding the nominal time histories by 10%. Buckling strength reduction factors due to the initial geometrical imperfections are adopted from the literature. The inner vessel is found to be the most critical component which buckles under seismic forces induced by a safe shutdown earthquake with a load multiplier of 1.52, which is higher than the minimum factor of safety of 1.3 required as per the design code RCC-MR [RCC-MR: edition, 2002. Design and construction rules for mechanical components for FBR nuclear islands, vol. 1, section I, subsection B. AFCEN, Paris, in press] for service level D conditions