Indian Prototype Fast Breeder Reactor Research Articles

CFD investigation of helical wire-wrapped 7-pin fuel bundle and the challenges in modeling full scale 217 pin bundle

Flow and temperature distributions of sodium in a heat generating fuel pin bundle with helically wound spacer wire have been predicted from basic principles by solving the three-dimensional conservation equations of mass, momentum and energy, for a wide range of Reynolds number. Turbulence has been modeled using the k– ɛ turbulence model. The geometry details of the bundle and heat flux from the fuel pin are similar to that of the Indian Prototype Fast Breeder Reactor (PFBR) that is currently under construction. The focus of the study is to assess the effect of transverse flow in promoting flow and temperature uniformity. It is seen that the ratio of maximum transverse velocity to the maximum axial velocity is nearly equal to the tangent of the rolling up angle of the spacer wire. Due to the wire wrap, the difference in bulk sodium temperature between the peripheral and central sub-channels is reduced to by a factor of 4 when compared to that without spacer wire. The film drop at the junction between wire and the pin is found to be only ∼70 °C. The predicted results are found to be in close agreement with that of the experimental results reported in literature. The present study considers a 7-pin bundle assembly of one helical pitch. The computational time and memory required for a 217 pin with 15 pitches assembly is ascertained to be 500 times that required for the current study. Hence, research activities have been directed towards developing a parallel CFD code and structural mesh generation software.

06/02130 Reliability analysis of safety grade decay heat removal system of Indian prototype fast breeder reactor: Arul, A. J. et al. Annals of Nuclear Energy, 2006, 33, (2), 180–188.

06/02130 Reliability analysis of safety grade decay heat removal system of Indian prototype fast breeder reactor: Arul, A. J. et al. Annals of Nuclear Energy, 2006, 33, (2), 180–188.

06/00122 Selection of hardfacing material for components of the Indian prototype fast breeder reactor: Bhaduri, A. K. et al. Journal of Nuclear Materials, 2004, 334, (2–3),109–114

06/00122 Selection of hardfacing material for components of the Indian prototype fast breeder reactor: Bhaduri, A. K. et al. Journal of Nuclear Materials, 2004, 334, (2–3),109–114

Reliability analysis of safety grade decay heat removal system of Indian prototype fast breeder reactor

The 500 MW Indian pool type Prototype Fast Breeder Reactor (PFBR), is provided with two independent and diverse Decay Heat Removal (DHR) systems viz., Operating Grade Decay Heat Removal System (OGDHRS) and Safety Grade Decay Heat Removal System (SGDHRS). OGDHRS utilizes the secondary sodium loops and Steam–Water System with special decay heat removal condensers for DHR function. The unreliability of this system is of the order of 0.1–0.01. The safety requirements of the present generation of fast reactors are very high, and specifically for DHR function the failure frequency should be less than ∼1E-7/ry. Therefore, a passive SGDHR system using four completely independent thermo-siphon loops in natural convection mode is provided to ensure adequate core cooling for all Design Basis Events. The very high reliability requirement for DHR function is achieved mainly with the help of SGDHRS. This paper presents the reliability analysis of SGDHR system. Analysis is performed by Fault Tree method using ‘CRAFT’ software developed at Indira Gandhi Centre for Atomic Research. This software has special features for compact representation and CCF analysis of high redundancy safety systems encountered in nuclear reactors. Common Cause Failures (CCF) are evaluated by β factor method. The reliability target for SGDHRS arrived from DHR reliability requirement and the ultimate number of demands per year (7/y) on SGDHRS is that the failure frequency should be ⩽1.4E-8/de. Since it is found from the analysis that the unreliability of SGDHRS with identical loops is 5.2E-6/de and dominated by leak rates of components like AHX, DHX and sodium dump and isolation valves, options with diversity measures in important components were studied. The failure probability of SGDHRS for a design consisting of 2 types of diverse loops (Diverse AHX, DHX and sodium dump and isolation valves) is 2.1E-8/de, which practically meets the reliability requirement.

Reliability analysis of shutdown system

This paper presents the results of reliability analysis of Shutdown System (SDS) of Indian Prototype Fast Breeder Reactor. Reliability analysis carried out using Fault Tree Analysis predicts a value of 3.5 × 10 −8/de for failure of shutdown function in case of global faults and 4.4 × 10 −8/de for local faults. Based on 20 de/y, the frequency of shutdown function failure is 0.7 × 10 −6/ry, which meets the reliability target, set by the Indian Atomic Energy Regulatory Board. The reliability is limited by Common Cause Failure (CCF) of actuation part of SDS and to a lesser extent CCF of electronic components. The failure frequency of individual systems is <1 × 10 −3/ry, which also meets the safety criteria. Uncertainty analysis indicates a maximum error factor of 5 for the top event unavailability.

Selection of hardfacing material for components of the Indian Prototype Fast Breeder Reactor

Nickel-base hardfacing alloys have been chosen to replace cobalt-base alloys as hardfacing material for components of the Indian Prototype Fast Breeder Reactor, for minimising the dose rate to personnel during maintenance and decommissioning, and to reduce the shielding thickness required for component handling. Induced activity, dose rate and shielding computations showed that replacing cobalt-base alloys with nickel-base alloys for hardfacing of components would result in a marked reduction in both the dose rate from the components and the thickness of lead handling flasks. Long-term ageing studies on the nickel-base hardface deposits on austenitic stainless steel showed that the hardface deposit would retain adequate hardness at the end of the components' design service-life of 40 years of exposure at 823 K.

Microstructural characterisation of a dissimilar weld of alloy D9 and AISI 316M stainless steel produced using a 2.5 kW CW CO 2 laser

The dissimilar weld, a fuel clad tube (D9 alloy) and an end plug made of AISI 316M forms a part of the fuel pin in the 500 MW Indian Prototype Fast Breeder Reactor (PFBR). The present study aimed at the development of a crack resistant microstructure of the dissimilar weld by optimisation of 2.5 kW CO 2 laser welding parameters. The D9 alloy, because of its low Cr eq /Ni eq ratio (close to 1), solidified first as primary austenite which makes its welds particularly susceptible to solidification cracking. In this study, the primary mode of solidification in the fusion zone was modulated by controlling the relative degree of fusion of the fuel clad tube and the end plug. A defect free weld with predominantly a primary ferrite mode of solidification was obtained by optimising laser parameters and displacing the focused laser beam towards the end plug side to achieve a greater contribution of end plug material to the fusion zone. The formation of a crater and associated defects at the termination site of laser welding were eliminated successfully by ramping the laser power during welding.

Inherent safety concepts in nuclear power reactors

Different inherent safety concepts being considered in fast and thermal reactors are presented after outlining the basic goals of nuclear reactor safety, the ‘defence in depth’ philosophy to achieve these goals and the characteristics affecting the safety of liquid metal fast breeder and light water reactors. The inherent safety potential of fast reactors with respect to different sizes and types of fuel is also discussed. Finally, the approach proposed for the Indian Prototype Fast Breeder Reactor (Pfbr), which is in the detailed design stage, is also presented.