Webserver Date: 24-May-2019

Fast Breeder Reactor Technology


Fast Breeder Reactor Technology


S. B. Bhoje
Director, Reactor Group, IGCAR




As a part of development of FBR in India, a 40 MWt Fast Breeder Test Reactor (FBTR) was commissioned in Oct 1985. Though the design of FBTR was partly obtained from France, the construction was essentially an indigenous effort - the fuel, the sodium coolant and the components. Experience with unique new carbide fuel and sodium systems including steam generator has been very good. Design of 500 MWe Prototype Fast Breeder Reactor (PFBR) has been undertaken at IGCAR as next logical step towards commercial deployment of FBR. PFBR is the forerunner of a series of reactors that are to follow. In the following paragraphs, a brief description of PFBR and the associated R&D program with its evolution are presented.


Main Options


Sodium coolant and pool type concept are chosen for the primary circuit of PFBR. Power is fixed as 500 MWe. The well proven mixed oxide fuel is chosen for PFBR. The core inlet/outlet temperatures are chosen as 670 K/820 K, based on detailed thermal hydraulic and structural mechanic studies. The materials of construction are 20% CW D9 for fuel clad and wrapper (low irradiation swelling and high creep rupture strength), SS 316 LN for sodium circuit components (corrosion resistance and high temperature strength), 9 Cr - 1 Mo for steam generator (high strength and freedom from stress corrosion cracking) and A 48 P2 for top shield (high impact strength). The systematic design methodology adopted (that is consistent with concurrent worldwide FBR technology) has resulted in compact, simplified layout with 2 primary sodium pumps and 2 secondary loops. The reactor site is Kalpakkam and is designed for a life of 45 years.


Reactor Assembly


The reactor assembly consists of core, grid plate, Core Support Structure (CSS), main vessel, safety vessel, inner vessel, top shields and absorber rod drive mechanisms. The core is homogeneous with two enrichment zones having radial and axial blankets. The target burn-up is 100 GWd/t with a maximum linear power of 450 W/cm. Adequate diversity and redundancy for reactor shut down are provided in the form of two independent, fast acting, diverse shutdown systems. The Sub Assemblies (SA) are supported on a Grid Plate (GP). The GP forms the inlet plenum for distributing coolant flow from the pumps to the core. It is supported on core support structure (CSS). The CSS is a box type orthogonally stiffened structure designed for effective use of structural material to withstand various loads. It also supports the in-vessel transfer post through which SA are handled. The CSS is supported on the main vessel, which is closed at the top by top shields and serves as boundary against release of radioactivity under operating and accident conditions. It holds 1150 t of primary sodium, blanketed by argon cover gas. Its shape is designed to enhance the buckling resistance and is suspended by a cylindrical shell supported on the reactor vault. To minimize thermal aging and creep, it is cooled by sodium. A safety vessel surrounds the main vessel with a nominal gap of 300 mm. This gap permits in-service inspection of the vessels and ensures that the sodium level in the hot pool does not fall below Intermediate Heat Exchanger (IHX) inlet windows, in the unlikely event of main vessel leak. The SS thermal insulation fixed on its outer surface reduces heat flux to the reactor vault. The safety vessel is supported on reactor vault. Inner vessel separates the sodium in the hot and cold pools and is supported on the grid plate. The shape of the inner vessel is arrived at based on thermal- hydraulic and structural considerations. Top shield consists of roof slab, Large Rotatable Plug (LRP), Small Rotatable Plug (SRP) and Control Plug (CP). It provides biological and thermal shielding in the upper axial direction of the reactor. The roof slab supports the LRP, PSP, IHX and heat exchangers of decay heat removal system. The roof slab, LRP and SRP are of box type structures made of 30 mm thick carbon steel plates. Concrete of density 3,000 kg/m3 is used as the shielding material. Air is used for cooling and inflatable seals are used for sealing. CP provides support for the Absorber Rod Drive Mechanisms (ARDM), core outlet temperature monitoring thermocouple tubes (210 nos.) and failed fuel location modules.


Heat Transport System


The heat transport system consists of primary sodium circuit, secondary sodium circuit and steam-water system. From the considerations of reduced capital cost, construction schedule and outage time due to failure of components, a 2 loop arrangement is adopted for the secondary sodium circuit together with 2 Primary Sodium Pumps (PSP) for the primary sodium circuit. There are 2 Intermediate Heat Exchangers (IHXs), 1 secondary sodium pump and 4 Steam Generators (SG) per loop. The 2 IHX/PSP is selected based on the existing pool type reactor.


Component Handling System


The core subassemblies (fuel, blanket, absorber and shielding SA) are handled with reactor in shutdown condition (473 K). Refueling is done after 185 efpd. The handling system has been divided into two parts i.e. in-vessel handling and ex- vessel handling. The spent fuel SA are stored inside the main vessel for 8 months and then shifted to spent fuel storage bay, which is a water pool. Handling of fresh SA consists of receipt from transport flask, flow test for any gross blockage and storage in fresh SA transfer chamber. They are then transferred to the reactor using cell transfer machine and inclined fuel transfer machine. Handling of PSP, IHX, ARDM are done by leak-tight shielded flasks. Storage pits, decontamination facility and dismantling facility are provided for these radioactive components in the Reactor Containment Building (RCB).


Balance of Plant


Steam-water system adopted is very similar to that of conventional thermal power stations of same capacity. A transmission voltage of 220 kV with indoor switchyard and conventional switch gear equipment is selected. The plant will be connected to the southern regional grid to export power generated and to provide off-site power supply to the station auxiliaries. Emergency power supply is provided by 4 DG sets, UPS (Class II) and 48 V DC (Class I) supplies. Auxiliary systems such as raw water system, demineralized water plant, service water system, fire protection system, A/C and ventilation system, nitrogen supply system, argon supply system and compressed air system are provided to suit the requirements of the reactor.


Plant Layout


The nuclear island has a common base raft for structurally inter connected buildings. Condenser is cooled by sea water and pump house is located off shore with pipes on jetty. The plant layout provides back up control room, physical separation of 2 SG buildings and 4 safety grade decay heat removal (SGDHR) circuits and location of turbine building with respect to safety related items.


Various Failure Modes considered in the design of PFBR


Instrumentation and Control


Neutron flux is monitored by fission chambers located in hot sodium above the core. SA outlet sodium temperatures are monitored to detect SA fault events (e.g. under cooling). Failed fuel detection is done by monitoring cover gas fission product activity and delayed neutrons in the primary coolant. Provision is made for continuous monitoring of SG tube integrity by detection of hydrogen in sodium, hydrogen in argon of the surge tank, argon pressure of surge tank etc. For detection of sodium leaks, wire type leak detectors, spark plug leak detectors, sodium ionization detectors and mutual inductance type level probes are provided. Reactor is tripped by dropping the absorber rods once safety parameters such as neutron flux, temperatures, flows cross their threshold values. The power is regulated manually.




FBR systems are designed with the defence-in-depth approach having redundancy, diversity and independence. The safety measures provided are two diverse reactor shutdown systems, two decay heat removal systems, a core catcher and RCB. Control and safety rod system is used for reactivity compensation, power control and shutdown, while diverse safety rod system is used only for shutdown. Generally, the decay heat is removed by normal heat transport path through SG. In case of loss of off-site power or non-availability of secondary or steam-water circuit, the decay heat is removed via a class I safety grade passive direct reactor cooling system. It consists of four independent circuits of 8 MWt capacity each having a sodium to sodium heat exchanger dipped in reactor hot pool and a sodium to air heat exchanger. A core catcher is provided to collect the molten fuel, suitably disperse it and ensure long term cooling in case of melt down of seven SA. With a similar defense in depth approach, an RCB is provided for Anticipated Transients Without Scram (ATWS) which are of event frequency < 10-6 per reactor year.


Detailed Analysis and Design Studies


For design of PFBR components, the French RCC-MR code is followed, backed by special procedures not covered by the code. Apart from the codal limits, certain functional limits are to be respected on deflection and slopes for core SA, GP, CSS and CP from the considerations of reactor scramability and reactivity addition. The high thermal stresses in conjunction with high temperature are responsible for failure modes such as creep, fatigue and ratcheting. The design codes call for detailed inelastic analysis which requires constitutive models having ability to simulate creep, plasticity, creep-plasticity interaction, cyclic hardening, strain memory and thermal aging. The ‘23-parameter Chaboche model’ and ‘13-parameter reduced Chaboche model’ are utilized. To have a better thermomechanical behavior and also from economic considerations, the component wall thicknesses are kept as low as possible, since the basic thickness for the pressure loading is generally small. The high R/t ratio of components causes concerns over other failure modes, viz. buckling, seismic loading and fluid structure interaction effects. Large sodium mass in association with sodium free surfaces within thin closely spaced concentric vessels amplifies the severity of seismic events. Large heat transfer coefficient of sodium leads to convective resistance of the fluid and conductive resistance of the structures of comparable magnitudes, despite lower thickness of the structures. High level of temperature coupled with large temperature differences leads to significant buoyancy effects and radiative interaction among structures. These call for heat transfer analysis in coupled modes involving convection, conduction and radiation simultaneously. Various failure modes considered in the design of PFBR are summarized in Fig. 1. These complexities call for specialized analysis methods both in thermal-hydraulics and structural mechanics. For the purpose of theoretical analyses, many thermal-hydraulic and structural mechanic computer codes have been developed at IGCAR. A large experimental R & D programme has been launched to validate the computer codes and the design. At present, 40 experiments are in progress at various places in IGCAR, BARC and outside institutes.


The performance testing of various subsystems of absorber rod drive mechanisms viz electromagnets, seals, dashpot, gripper assembly etc. have been carried out. For the reliability testing of prototype control and safety rod drive mechanism, a sodium test rig has been built. Development of sodium resistant concrete is in progress at the centre.


FBR requires highly sophisticated and extensive analytical capabilities in core physics, radiation shielding, thermal-hydraulics, structural mechanics and safety. Such capabilities have been developed at IGCAR during the last decade.


Development of Manufacturing Technology


Several indigenous materials development programmes have been initiated. For example, development of D9 material has been completed and sample clad tubes have been manufactured successfully. The developmental activities for 20%CW D9 material and fuel SA fabrication are in progress. Development regarding manufacture of 8m long 316 LN tubes for IHX has been completed and 23m long SG tubes of 9Cr-1Mo have been manufactured. Manufacturing technology development program for key components was initiated with the participation of Indian industries. The prototype control and safety rod drive mechanism, diverse safety rod drive mechanism, transfer arm and inclined fuel transfer machine, full sized sectors of components like main vessel, inner vessel, roof slab have been manufactured. Providing infrastructure requirements for the site such as site assembly shop, construction office, levelling, roads are being started.


Present Status of PFBR


The detailed design of NSSS is nearing completion and is being reviewed by the Atomic Energy Regulatory Board. Appointment of consultants for balance of plant design and preparation of environmental impact assessment report are under progress. It is planned to start construction of PFBR in April 2001 and the project is expected to be completed in 8 years. After successful commissioning of PFBR, it is proposed to construct 4 x 500 MWe FBR at a suitable site in a phased manner.