Liquid Metal Cooled Fast Breeder Reactors Research Articles

Advanced Methods for the Fabrication of Mixed Uranium Plutonium Oxide, Monocarbide and Mononitride Fuels for Fast Breeder Reactors

The reference fuel for operating and forthcoming prototype and commercial liquid metal cooled fast breeder reactors (LMFBR) all over the world, including India, is mixed uranium plutonium oxide (MOX). Mixed uranium plutonium monocarbide (MC) and mononitride (MN) are considered as advanced LMFBR fuels. “Powder-pellet (POP)” is the conventional method for production of MOX, MC and MN fuel pellets starting from UO2 and PuOa powders. The POP route involves generation and handling of plutonium bearing fine powder or dust particles which has the problem of radiotoxic dust hazard. Further, fine powders have poor flowability which makes automation and remote fabrication difficult. The combination of ammonium uranyl plutonyl carbonate (AUPuC) process and low temperature oxidative sintering (LTS) minimises the radiotoxic dust hazard and the fabrication cost of MOX fuel and is considered as advanced POP route.Vibro-sol (or sphere-pac) and sol-gel microsphere pelletisation (SGMP) are advanced methods for fabrication of MOX, MC and MN fuels. Here, dust-free and free-flowing gel microspheres of the oxide or oxide-carbon are prepared from heavy metal nitrate feed solutions by “internal” or “external” gelation processes. The gel microspheres are subjected to controlled calcination for MOX or carbothermic reduction for MC and MN. Thereafter, the microspheres are either vibropacked in fuel pins or directly pelletised and sintered. These processes minimise radiotoxic dust hazard, facilitate automation and remote processing and ensure excellent microhomogeneity.The present paper summarises the state-of-art of the POP, vibrosol and SGMP processes for the fabrication of MOX, MC and MN fuels, highlighting the author's experience in SGMP process.

The thorium cycle for fast breeder reactors

The role that could be played by liquid metal-cooled fast breeder reactors (LMFBRs) in the utilization of India’s considerable thorium resources is reviewed in this article. Distinct advantages of thorium-based fuels over plutonium-uranium fuels in LMFBRs pertain to a more favourable coolant voiding reactivity coefficient and better fuel element irradiation stability. The poorer breeding capability of thorium-fuelled fast reactors can in principle be overcome by improved core design and development of advanced fuel concepts. The technical feasibility of such advanced thorium fuels and core designs must be established by sustained research and development. It is also necessary to efficiently close the thorium fuel cycle of fast breeder reactors by appropriate development of the fuel reprocessing and refabrication stages. The Fast Breeder Test Reactor (FBTR) at Kalpakkam is expected to be an important tool for development of thorium fuel and fuel cycle technology. A quick look at the economics of the thorium cycle for fast reactors, vis-a-vis the more conventional uranium cycle indicates only a small and acceptable cost disadvantage on account of the need for remote fabrication of recycled thorium fuel.

Plutonium metallurgy in India

The embryo of plutonium metallurgy in India, as well as in Asia, was formed nearly 25 years back in Trombay. Thereafter, in the intervening 25 years, significant amount of research and development work have been carried out atbarc on this man-made fissile element and indigenisation has been achieved in the fabrication, characterisation and property evaluation of plutonium metal, delta-stabilised alloys, Al-Pu fuels, Pu-Be neutron sources, (UPu)O2 fuels for thermal and fast reactors and (UPu)C and (UPu)N advanced fuels for liquid metal-cooled fast breeder reactors (lmfbr).This paper summarises the development of plutonium metallurgy in India highlighting the safety aspects, our achievements and potential of plutonium for generation of nuclear electricity in the coming decades.

Numerical Simulation of Thermal-Hydraulic Transients in Lmfbr Piping Systems

Abstract The transient performance characteristics of liquid metal-cooled fast breeder reactor (LMFBR) systems are strongly influenced by the energy transport in the piping interconnecting the system components. A one-dimensional coolant-wall model is used to simulate the thermal transport processes, with energy equations formulated by using nodal heat balances. An implicit scheme is applied to the coolant equation with an explicit formulation of the wall heat flux, allowing the solution to march in the flow direction. The impact of uncertainties in Nusselt number predictions based on generally used correlations is examined. For coolant dynamics, each pipe run between components is treated as a lumped control volume. From the transient cases studied, the following results are obtained: (1) The degree of axial nodalization required to obtain a converged temperature solution is bounded. (2) Time step control for numerical integration is most efficiently achieved by using the characteristic time approach. (3...

Heat and Radiation Resistant Cable for the Use on LMFBR

The heat and radiation resistant cable for the use on the liquid metal cooled fast breeder reactor (LMFBR) has been newly developed. Stainless steel sheathed mineral insulated type cable has been usually used in the vinicity of the primary cooling circuit of LMFBR because of the harsh environmental condition in this area. Sometimes, this type of cable causes problems for its inflexibility in terms of installation. This newly developed cable (named M. Flex cable) consists of inorganic materials such as mica and glass, and the heat and radiation resistant organic materials such as polyimide for the insulation and stainless steel wire braid for the protection layer. Such tests as gamma irradiation test up to total dose of 1.4 × 1010R, flexibility test, and other tests, have been performed and results show that M. Flex cable has been satisfactory for the LMFBR applicaltion. And also, these results implied the applicability of this cable to the other fields than nuclear plants.

The Analysis and Design of Improved Control Systems for Recirculation Boilers

The importance of providing effective boiler control system performance in order to maintain high power plant availability is explained. Methods of predicting likely boiler perturbations are developed and the limitations of conventional boiler control systems investigated. Multivariable frequency response analysis methods are applied to the design and integrity analysis of an improved control system for an LMFBR (Liquid metal cooled fast breeder reactor) power plant boiler. This improved control scheme provides large stability margins and low interaction between loops. Its improved response to disturbances and resilience to system failures will contribute to greater plant availability.

Liquid metal natural convection from plane surfaces: A review including recent sodium measurements

The design of various components of Liquid Metal Cooled Fast Breeder Reactors (LMFBR) requires knowledge of the natural convection heat transfer conditions. A review has therefore been made of natural convection heat transfer for low Prandtl number fluids: this includes both theoretical and experimental studies. The importance of the surface orientation is emphasized with reference to vertical, inclined, and horizontal surfaces. The most recent measurements made by the authors are compared with the analytical predictions and measurements made in other low Prandtl number fluids.

Experimental Parameter Investigations on the Tribological Behavior of Stellite 6 in Liquid Sodium

The friction and wear behavior of various material combinations was studied in a liquid-sodium loop up to fluid temperatures of 900 K to guarantee self-alignment and safe operation of liquid-metal-cooled fast breeder reactor (LMFBR) components, such as subassemblies, control and shutdown units, instruments, and centering pads. Pin-on-disk, plate-on-plate, and tilting plane test section systems were used. Numerous material combinations exhibiting good corrosion resistances in liquid sodium were investigated under identical operating conditions. Those revealing the lowest wear rates and the lowest friction coefficients were subject to parameter tests. The influences on friction and wear behavior of friction velocity, contact force, sodium temperature, equipment criteria, and tribochemistry were studied. The experimental results are shown for a Stellite 6-Stellite 6 combination. They revealed friction and wear coefficients to be clearly below the set limits. However, this material combination has a high cobalt content and is therefore no optimum choice with respect to contamination problems of the primary system of an LMFBR. Other experiments are necessary to develop materials with zero contamination effects.

Transient flow and heat transfer of liquid sodium coolant in the outlet plenum of a fast nuclear reactor

A calculation procedure for axisymmetric elliptic flows is applied to predict the transient velocity and temperature fields of a heavy fluid jet issuing vertically into a volume of relatively light fluid. This situation arises in the outlet plenum of a Liquid-Metal-Cooled Fast Breeder Reactor (LMFBR) during reactor transients. The time averaged conservation equations for momenta and heat transfer were solved on a CDC 6600 digital computer for various plenum inlet transients, along with a two-equation model of turbulence and proper modelling of the buoyancy terms. Predictions are presented of flow and heat transfer in the form of velocity vector plots and temperature contours. Predictions are in qualitative agreement with expectations, invariably establishing that the flow by-passes the outlet plenum.

Experimental study of whole core thermal-hydraulic phenomena in EBR-II

Abstract In the analysis of core-wide temperature distributions in a liquid-metal-cooled fast breeder reactor (LMFBR) under both normal and abnormal operating conditions, it is commonly assumed that there is little, if any, thermal interaction between adjacent subassemblies. Since intersubassembly heat transfer tends to reduce the transverse temperature gradients in a reactor, thereby ameliorating the effects of local overheating, this assumption is conservative. In order to assess the importance of this effect as well as of flow redistribution in a reactor core, an experimental study was conducted in EBR-II covering a wide range of operating power and flow conditions, including both forced and natural convection. The results of this study indicate that radial heat transfer and flow redistribution are important mechanisms in the thermal-hydraulics of LMFBR cores, especially at low flow rates.

Energy

An energy supply and demand scheme proposed by researchers at the Swiss Federal Institute for Reactor Research is described. An underground nuclear energy park using high-temperature gas cooled reactors and liquid metal-cooled or gas-cooled fast breeder reactors is proposed. The proposed site is in the Giuv-Syenite ridge located in the Eastern Aare-Gorrhard granite formation. The excavated (425 million tons) granite will provide an estimated 34,000 tons of uranium and thorium, enough to provide nuclear fuel for twelve centuries of operation at the stabilized energy-consumption level for the year 2050.

Structural response of fast breeder reactors to hypothetical core disruptive accidents

Liquid metal-cooled fast breeder reactors (LMFBRs) so far have been analyzed for the consequences on the plant and the environment for hypothetical core disruptive accidents (HCDAs). To provide the appropriate analytical tools for this effort, analysis and codes are currently under development in several countries. They combine the hydrodynamics and solid mechanics (and more recently the bubble dynamics) phenomena to gage stresses, strains, and deformations of the important components of the system, and the overall adequacy of the primary and secondary containments. The effort is partitioned into the structural analysis of (a) the core components, and (b) the primary system components beyond the core. The core mechanics effort covers the structural response of fuel pins, hexcans, fuel elements, and fuel element clusters to transient pressures and thermal loads. Two- and three-dimensional finite element codes are under development for these core components. The results of these analyses would permit evaluation of the adequacy of the heat removal process to continue following severe core component deformations. Also, these analyses are currently being combined with neutronics, for the core transition phase, to allow for the mass movements for realistic neutronic calculations. The primary system and containment program treats the structural response of the components beyond the core, starting with the core barrel. Combined hydrodynamics-solid mechanics codes provide transient stresses and strains and final deformations for components such as the reactor vessel, reactor cover, cover holddown bolts, as well as the pulses for which the primary piping system is to be analyzed. Both, Lagrangian and Eulerian two-dimensional codes are under development, which in their combined form provide greater accuracy and longer durations for the treatment of HCDAs. More recently the codes are being augmented with bubble migration capability pertaining to the latter stages of the HCDA, after slug impact. This step will permit treatment of the instabilities following slug impact, the ultimate reversal of the sodium slug with the rising bubble, the bubble break-up, and the calculations of sodium splillage and radioactive gases, if any, in the secondary containment. The extent of sodium spillage and sodium fires should be known for evaluation of the secondary containment. The mechanics of bubble migration are needed for radiological study of post-accident phenomena. More recently dynamic fracture mechanics considerations are being incorporated to remove arbitrary failure criteria imposed on components such as the core barrel and vessel. Most recent developments involve the adaptation of the 2-D Eulerian primary system code to the 2-D elastic-plastic treatment of the primary piping. The pulses are provided at the vessel primary piping interfaces of the inlet and outlet nozzles. The calculation includes the elbows and pressure drops along the components of the primary piping system. Pressures larger than the ones used as input at the inlet and outlet nozzles were observed. As expected, they occur far from the nozzles, in the pipe, where the pulses meet. Recent improvements to the primary containment codes include introduction of bending strength in materials, Lagrangian mesh regularization techniques, and treatment of energy absorbing materials for the slug impact. A further development involves the combination of a 2-D finite element code for the reactor cover with the 2-D finite-difference hydrodynamic code for continuous monitoring of stresses, strains, and deformations in the cover, as well as pressure changes in the hydrodynamic code. Substantial experimental effort is in progress in various countries on the response to energy releases of vessels and internals, piping systems, subassemblies, and subassembly clusters. These experimental results are being utilized for the verification or modification of the analyses and codes under development.

Wear resistant coatings for reactor components in liquid sodium environments

Rubbing surfaces of critical components for the coming generation of the nation's liquid–metal-cooled fast-breeder reactors must withstand severe environments of high temperature, liquid–metal corrosion, and nuclear irradiation. For some component requirements, the use of the better known bearing materials and hardfacings has been shown to be inadequate or impractical. For several applications, bulk (> 25-μ thick) coatings have provided the only successful solution to friction problems in high-temperature sodium. Coating processes that have been or are being evaluated include plasma spray, detonation gun, spark-discharge, sputtering, and diffusion coating processes. Testing included friction and wear tests in up to 650 °C sodium, sodium corrosion tests for up to one year, thermal cycling, bond strength tests, and irradiation tests in a fast reactor. For one particularly severe application, a specially modified commercial coating of chromium carbide in a 15 vol% nichrome binder, applied by a commercial detonation gun process, has so far been found to be most acceptable in meeting the friction, corrosion, and irradiation damage criteria. Much of the test data reported is on this coating and its antecedent modifications.

Prediction of the Incipient Boiling Conditions Following a Blocked Lmfbr Subassembly Accident

The pressure-temperature history model of incipient boiling of liquid alkali metals is applied to a single subassembly blockage accident of a liquid-metal-cooled fast breeder reactor (LMFBR) where ...

Heavy Ion Bombardment of Stainless Steel

A problem of current concern in the design of liquid-metal cooled, fast breeder reactors (LMFBR) for commercial power generation is the volumetric expansion of structural components due to the agglomeration of vacancies produced during irradiation. Present plans call for LMFBR fuel cladding and major structural components to be fabricated of an austenitic stainless steel; with operating temperatures and fuel cycles now anticipated, projected values of the volumetric expansion in fuel clad at the end of a fuel cycle are as high as 10-12%. The attainment of such fluences in present test reactors is costly and time-consuming, requiring an irradiation time of some 4-5 years.