A neutronics investigation simulating fast reactor environments in the thermal-spectrum advanced test reactor

The second stage of Indian nuclear programme envisages the deployment of fast reactors on a large scale for the effective use of India’s limited uranium reserves. The Fast Breeder Test Reactor (FBTR) at Kalpakkam is a loop type, sodium cooled fast reactor, meant as a test bed for the fuels and structural materials for the Indian fast reactor programme. The reactor was made critical with a unique high plutonium MK-I carbide fuel (70% PuC+30%UC). Being a unique untested fuel of its kind, it was decided to test it as a driver fuel, with conservative limits on Linear Heat Rating and burn-up, based on out-of-pile studies. FBTR went critical in Oct 1985 with a small core of 23 MK-I fuel subassemblies. The Linear Heat Rating and burn-up limits for the fuel were conservatively set at 250 W/cm & 25 GWd/t respectively. Based on out-of-pile simulation in 1994, it was possible to raise the LHR to 320 W/cm. It was decided that when the fuel reaches the target burn-up of 25 GWd/t, the MK-I core would be progressively replaced with a larger core of MK-II carbide fuel (55% PuC+45%UC). Induction of MK-II subassemblies was started in 1996. However, based on the Post-Irradiation Examination (PIE) of the MK-I fuel at 25, 50 & 100 GWd/t, it became possible to enhance the burn-up of the MK-I fuel to 155 GWd/t. More than 900 fuel pins of MK-I composition have reached 155 GWd/t without even a single failure and have been discharged. One subassembly (61 pins) was taken to 165 GWd/t on trial basis, without any clad failure. The core has been progressively enlarged, adding MK-I subassemblies to compensate for the burn-up loss of reactivity and replacement of discharged subassemblies. The induction of MK-II fuel was stopped in 2003. One test subassembly simulating the composition of the MOX fuel (29% PuO2) to be used in the 500 MWe Prototype Fast Breeder Reactor was loaded in 2003. It is undergoing irradiation at 450 W/cm, and has successfully seen a burn-up of 92.5 GWd/t. In 2006, it was proposed to test high Pu MOX fuel (44% PuO2), in order to validate the fabrication and fuel cycle processes developed for the power reactor MOX fuel. Eight MOX subassemblies were loaded in FBTR core in 2007. The current core has 27 MK-I, 13 MK-II, eight high Pu MOX and one power reactor MOX fuel subassemblies. The reactor power has been progressively increased from 10.5 MWt to 18.6 MWt, due to the progressive enlargement of the core. This paper presents the evolution of the core based on the progressive enhancement of the burn-up limit of the unique high Pu carbide fuel.

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Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

10.2172/1048404 ◽

2012 ◽

Author(s):

Timothy A. Hyde

Keyword(s):

Rare Earth ◽

Fast Reactor ◽

Reactor Fuel ◽

Minor Actinide ◽

Fast Reactor Fuel ◽

Test Reactor ◽

A Minor

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Design and Development of Diverse Safety Rod and Its Drive Mechanism for PFBR

Volume 3: Thermal Hydraulics; Current Advanced Reactors: Plant Design, Construction, Workforce and Public Acceptance ◽

10.1115/icone17-75851 ◽

2009 ◽

Author(s):

R. Vijayashree ◽

P. ChellaPandi ◽

K. Natesan ◽

S. Jalaldeen ◽

S. C. Chetal ◽

...

Keyword(s):

Fast Reactor ◽

Design Stage ◽

Design And Development ◽

Flow Induced Vibration ◽

Design Specifications ◽

Active Core ◽

Final Design ◽

Endurance Testing ◽

Test Reactor ◽

Fast Acting

Prototype Fast Breeder Reactor (PFBR) is U-PuO2 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.

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Operating Experience of Fast Breeder Test Reactor

Volume 1: Plant Operations, Maintenance, Engineering, Modifications and Life Cycle; Component Reliability and Materials Issues; Next Generation Systems ◽

10.1115/icone17-75592 ◽

2009 ◽

Author(s):

K. V. Suresh Kumar ◽

V. Ramanathan ◽

G. Srinivasan

Keyword(s):

Steam Generator ◽

Fast Reactor ◽

Residual Life ◽

Detection System ◽

Water System ◽

Operating Experience ◽

The Core ◽

Grid Plate ◽

Test Reactor ◽

Carbide Fuel

Sodium cooled fast breeder reactors constitute the second stage of India’s three-stage nuclear energy programme, for effective utilization of the country’s limited reserves of natural uranium and exploitation of its large reserves of thorium. The Fast Breeder Test Reactor (FBTR) at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, is a sodium cooled, loop type fast reactor. Its main aim is to provide experience in fast reactor operation and large scale sodium handling and to serve as a test bed for irradiation of fast reactor fuels and materials. FBTR was built on the lines of the French Rapsodie-Fortissimo reactor, with modifications to make it a generating plant. FBTR heat transport system consists of two primary sodium loops, two secondary sodium loops and one common tertiary steam and water circuit. The steam water system mainly consists of a once-through steam generator, which produces super heated steam at a pressure of 125 bars and temperature of 480° C, feed water system and condensate system. The steam produced is supplied to a condensing turbine coupled to an alternator. The reactor achieved first criticality in Oct 85 with a small core of 22 fuel subassemblies (SA) having a unique carbide fuel rich in Pu. This fuel (called MK-I) was developed and made in India and has a composition of 70% PuC-30% UC. The steam generator was put in service in Jan.1993 and turbine generator was synchronized to the grid in July 1997. In the light of the excellent performance of the carbide fuel, which has endured a burn-up of 155 GWd/t without any clad failure, the core has been gradually expanded by the addition of mark II (55% PuC-45% UC) and MOX (44% PuO2−56% UO2) fuel SA to compensate for the burn-up reactivity loss, and the reactor power has been progressively raised from 10.6 MWt to a maximum of 17.4 MWt. Over the years, several safety related experiments have been conducted. These include natural convection tests and experiments to validate the Failed Fuel Detection System. The challenges faced include a major fuel handling incident, primary sodium leak and reactivity transients. Two major modifications were carried out — one on the Steam Generator Leak Detection System and the other in the steam-water circuit. These helped in improving the campaign availability from less than 50% to more than 90%. The main component limiting the life of reactor is the grid plate supporting the core. The fast flux at the grid plate was measured using Np foils. The residual life of the grid plate has been estimated as 11 effective full power years. The paper presents operating experience of the reactor, performance of the carbide fuel, safety related experiments done in the reactor, various challenges faced, various modifications carried out to improve system reliability and availability and residual life assessment of the reactor.

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