Liquid Metal Cooled Fast Breeder Reactors Research Articles

Hydraulic design optimization of spherical header for future fast reactors of India

The liquid sodium as coolant in a typical pool type Liquid Metal cooled Fast Breeder Reactor (LMFBR), flows through primary sodium pump, spherical header and finally enters the grid plate, which acts as an inlet plenum to heat-generating sub-assemblies. The reduction of pressure drop in primary circuit would reduce the operating cost of the system. Spherical header, a vital component in primary circuit suffers major flow non-uniformities resulting in higher pressure drop. Reduction of the pressure drop by uniformization of sodium flow in spherical header is considered. To this end, a 3D computational fluid dynamics (CFD) analysis is carried out to investigate the hydrodynamics of flow, inside the spherical header. Towards this a relationship between pressure loss coefficient and geometrical dimensions, viz., diameter and height of the cone, inside the header are established in order to minimize the pressure drop. The variation of pressure loss coefficient is estimated for header without and with the internals. The static pressure loss coefficient inside the spherical header without an internal cone is found to be 5.59. With a suitable and optimized geometrical dimension of the cone as internal flow guiding device, the loss coefficient is reduced to 0.53. The loss coefficient predicted for the spherical header is validated against the experimental estimates. Consequently, the simplified design of the flow guiding device is achieved by removing baffles.

COUPLED MODELLING OF THE EBR-II SHRT-45R INCLUDING PHOTON HEAT DEPOSITION

The Fast REactor NEutronics/Thermal-hydraulICs (FRENETIC) code has been developed during the last years at Politecnico di Torino, implementing a full-core coupled neutronic/thermal-hydraulics model for steady-state and transient analysis of liquid-metal cooled fast breeder reactor (LMFBR). In the framework of the validation activities for the code, an analysis of the sodium-cooled reactor EBR-II, previously carried out in the frame of a IAEA Coordinated Research Project, is performed with FRENETIC including the most recent physics models. In particular, photon transport and heat deposition are taken into account, a feature which has been proved in previous studies to be relevant to the correct study of the EBR-II core. To this purpose, a set of nuclear data for photons has been generated by means of the Monte Carlo code Serpent-2, and it is demonstrated that the code is able to take into account the photon heat deposition in the EBR-II.

Implications of large scale sodium water reactions in an LMFBR

Abstract Large leak sodium-water reaction in the steam generator is one of the important events to be considered for the Liquid Metal cooled Fast Breeder Reactor (LMFBR). For the design of secondary sodium circuit components a design basis water/steam leak has been defined. For PFBR (Prototype Fast Breeder Reactor), under commissioning at Kalpakkam, India, a computer code SWEPT developed for the pressure transient analysis is used for estimating the maximum design pressure of the components. In order to study the possibility of any cliff edge effect, large scale sodium-water reaction analyses beyond the design basis leak have been carried out using SWEPT and CAST3M computer codes and the results are presented in this paper. Based on the studies it is found that simultaneous double ended guillotine failure of 15 tubes can be managed with the available design margin. Possibility of core sub assembly vibration is also studied using the CAST3M computer code.

Vibration of core subassemblies due to large sodium–water reaction in the steam generator of a Liquid Metal Fast Breeder Reactor

In a Liquid Metal cooled Fast Breeder Reactor (LMFBR), liquid sodium is used as the reactor coolant to extract the heat energy from the core of the reactor. Liquid sodium has a chemical characteristic that makes it to violently react with water or steam. When large sodium-water reaction occurs due to defect, leakage or breakage in the steam generator tubes, high pressure waves will be generated and get transmitted throughout the secondary sodium circuit and interact with the components like sodium headers on top/bottom of steam generator, surge tank, intermediate heat exchanger etc., which also affects the safety of primary sodium circuit and core subassemblies. In LMFBR, sodium water reactions are included in the Design Basis Events. This dynamic event needs to be analyzed from the perspective of integrity of core subassemblies considering the phenomenon of fluid structure interaction. To evaluate the integrity of core subassemblies, the behavior of high pressure waves at intermediate heat exchanger is considered and the response of the subassemblies to the dynamic pressure load has been studied. The dynamic analysis of the reactor model is carried out using a three-dimensional finite element analysis in CAST3M computer code. In the code, the fluid elements are represented by volume elements and the coupling between fluid and structure by surface elements. Fluid damping is considered in the simulation. Validation of the code has been carried out using an experimental setup. Analysis results shows the maximum displacement of the primary circuit components viz., IHX, Inner vessel and subassemblies are 5.0 mm, 2.0 mm and 0.07 mm respectively for a design basis sodium water reaction event. Due to this displacement there is negligible change in the reactivity of the core.

Tensile Properties Variation Across the Dissimilar Metal Weld Joint Between Modified 9Cr–1Mo Ferritic Steel and 316LN Stainless Steel at RT and 550 °C

In liquid metal cooled fast breeder reactors (LMFBR), modified 9Cr–1Mo ferritic steel (P91 or Grade 91) is a preferred material for constructing steam generators due to its creep strength and stress corrosion cracking resistance. The austenitic stainless steels (SS 316LN and SS 304LN) are widely used for primary and secondary piping systems because of its oxidation resistance and excellent creep strength. So, the dissimilar metal weld joint (DMWJ) between P91 and SS 316LN is inevitable. Nickel-based consumables (Alloy 82 and Alloy 182) are preferred to join these materials. The DMWJ will experience the temperature up to 550 °C. For accurate integrity assessment, the mechanical properties of individual regions are to be evaluated at room temperature (RT) and 550 °C. Hence, the present investigation is focused on evaluating the mechanical properties of various regions of DMWJ at RT and 550 °C. From this investigation, it is understood that the tensile properties are heterogeneous across the DMWJ at RT and 550 °C. The high-temperature (550 °C) tensile properties are significantly lower with respect to the RT properties. The development of complex microstructures at the interfaces will alter the mechanical properties across the DMWJ.

Modification in the Microstructure of Mod. 9Cr-1Mo Ferritic Martensitic Steel Exposed to Sodium

Mod. 9Cr-1Mo is used as the structural material in the steam generator circuit of liquid metal-cooled fast breeder reactors. Microstructural modifications on the surface of this steel are investigated after exposing to flowing sodium at a temperature of 798 K (525 °C) for 16000 hours. Sodium exposure results in the carburization of the ferritic steel up to a depth of ~218 µm from the surface. Electron microprobe analysis revealed the existence of two separate zones with appreciable difference in microchemistry within the carburized layer. Differences in the type, morphology, volume fraction, and microchemistry of the carbides present in the two zones are investigated using analytical transmission electron microscopy. Formation of separate zones within the carburized layer is understood as a combined effect of leaching, diffusion of the alloying elements, and thermal aging. Chromium concentration on the surface in the α-phase suggested possible degradation in the corrosion resistance of the steel. Further, concentration-dependent diffusivities for carbon are determined in the base material and carburized zones using Hall’s and den Broeder’s methods, respectively. These are given as inputs for simulating the concentration profiles for carbon using numerical computation technique based on finite difference method. Predicted thickness of the carburized zone agrees reasonably well with that of experiment.

A computational study of flow mal-distribution on the thermal hydraulic performance of an intermediate heat exchanger in LMFBR

The flow and thermal non-uniformities occurring in the intermediate heat exchanger (IHX) of a liquid metal-cooled fast breeder reactor have been characterized through numerical simulations. For modeling the primary and secondary sodium flow through the IHX, an equivalent anisotropic porous medium approach has been used. The pressure drop in the equivalent porous medium is accounted through the inclusion of additional pressure drop terms in the Navier–Stokes equations, with the help of standard correlations for cross flow or parallel flow over tubes. For secondary sodium flow, the effects of a flow distributor device with orifices and baffles at the inlet have also been included, in addition to axial flow through the tubes. The heat exchange between primary and secondary streams is incorporated in the form of a volumetric heat source or sink term, which is corrected iteratively. The resulting flow distributions are in reasonable agreement with available experimental results. The study shows that the temperature of the secondary sodium flow at the exit can be made more uniform by exchanging less heat near the inner wall of IHX, as compared to the region close to the outer wall, using suitable flow distribution devices.

Sodium removal from sodium wetted under sodium ultra sonic scanner

Sodium is used as a coolant in primary and secondary heat transport systems of Liquid Metal cooled Fast Breeder Reactors (LMFBR). The Prototype Fast Breeder Reactor (PFBR) is in advanced stage of construction at Kalpakkam, India. Under Sodium Ultra Sonic Scanner (USUSS) is used to image the components dipped in sodium and more specifically the gap between Sub Assembly (SA) heads and the control plug. This is essential to confirm that no SA is protruding from its position. USUSS has to be qualified in sodium before its use in the reactor. After sodium testing, sodium adhering to the USUSS needs to be removed to prevent reaction with oxygen and moisture present in air when it is taken out. The sodium removal process is based on the use of moist carbon-di-oxide (CO2) at a temperature of about 353K. The reaction rate is controlled by varying the ratio of CO2 to moisture. Hydrogen released during the reaction of sodium with moisture is continuously monitored to ensure safe limits of hydrogen concentration. This paper describes the details of the experimental setup used in a cleaning pit for the removal of sodium adhering on USUSS and the technical details of the cleaning process.

Dynamic simulation of accidental closure of intermediate heat exchanger isolation valve in a pool type LMFBR

In a pool type liquid metal cooled fast breeder reactor (LMFBR), core and other internals such as pumps, heat exchangers are immersed in a pool of sodium. Heat exchange from primary sodium circuit (pool) to secondary sodium circuit (loop) is through four intermediate heat exchangers (IHX) immersed in primary sodium pool. Each IHX is provided with a sleeve valve at its primary sodium inlet window for the purpose of isolating the shell side of IHX from the sodium pool. With such a provision, an inadvertent partial or complete closure of a sleeve valve of one of the IHX during normal operation of the reactor has been considered as a design basis event for the reactor. One dimensional transient thermal hydraulic models of the primary and secondary sodium circuits have been developed to study the thermal hydraulic consequences of such an event. The main areas of concern in the plant and the availability of safety parameters for the detection of the event have been evaluated.

Turbulent flow simulation in a wire-wrap rod bundle of an LMFBR

The pressure drop and heat transfer characteristics of wire-wrapped 19-pin rod bundles in a nuclear reactor subassembly of liquid metal cooled fast breeder reactor (LMFBR) have been investigated through three-dimensional turbulent flow simulations. The predicted results of eddy viscosity based turbulence models ( k– ɛ, k– ω) and the Reynolds stress model are compared with those of experimental correlations for friction factor and Nusselt number. The Re is varied between 50,000 and 150,000 and the ratio of helical pitch of wire wrap to the rod diameter is varied from 15 to 45. All the three turbulence models considered yield similar results. The friction factor increases with reduction in the wire-wrap pitch while the heat transfer coefficient remains almost unaltered. However, reduction in the wire-wrap pitch also enhances the transverse flow velocity in the cross-sectional plane as well as the local turbulence intensity, thereby improving the thermal mixing of coolant. Consequently, the presence of wire wrap reduces temperature variation within each section of the subassembly. The associated reduction in differential thermal expansion of rods is expected to improve the structural integrity of the fuel subassembly.

Reliability Assessment of Advanced Materials and Structures

This paper discusses the reliability assessment of advanced structures such as micro/nano structures, electronics packaging for modern electronic devices, high temperature structures such as nuclear power plants, rocket engines for space application. Characterization of advanced materials includes free-standing electrodeposited nano-crystalline nickel thin film, polysilicon, and piezoelectric thin film for MEMS/nano structures. Lead-free solders under pseudo-power cyclic thermal loading, impact loading and mechanical bending loading were evaluated for reliability assessment of the advanced electronic packaging. Reliability assessment of Cu/SnAg double bump structures was performed for the 150µm pitch flip chip electronic packaging application. The 316L austenitic stainless steel for high-temperature components application such as in the primary side of liquid metal cooled fast breeder reactor (LMFBR) was tested under various temperatures where dynamic strain ageing occurs. The effects of dynamic strain ageing on the tensile behavior and fatigue behavior of 316L stainless steel have been identified.

Effect of sodium environment on the low cycle fatigue properties of modified 9Cr–1Mo ferritic martensitic steel

Modified 9Cr–1Mo ferritic steel is the material of current interest for the steam generator components of liquid metal cooled fast breeder reactors (LMFBRs). The steam generator has been designed to operate for 30–40 years. It is important to accurately determine the life of the components in the actual environment in order to consider the extension of life beyond the design life. With this objective in view, a programme has been initiated at our laboratory to evaluate the effects of flowing sodium on the LCF behaviour of modified 9Cr–1Mo steel. LCF tests conducted in flowing sodium environment at 823 K and 873 K exhibited cyclic softening behaviour both in air and sodium environments. The fatigue lives are significantly improved in sodium environment when compared to the data obtained in air environment under identical testing conditions. The lack of oxidation in sodium environment is considered to be responsible for the delayed crack initiation and consequent increase in fatigue life. Comparison of experimental lifetimes with RCC-MR design code predictions indicated that the design curve based on air tests is too conservative.

Evaluation of high plutonia (44% PuO 2) MOX as a fuel for fast breeder test reactor

Uranium plutonium mixed oxide (MOX) containing up to 30% plutonia is the conventional fuel for liquid metal cooled fast breeder reactor (LMFBR). Use of high plutonia (>30%) MOX fuel in LMFBR had been of interest but not pursued. Of late, it has regained importance for faster disposition of plutonium and also for making compact fast reactors. Some of the issues of high plutonia MOX fuels which are of concern are its chemical compatibility with liquid sodium coolant, dimensional stability and low thermal conductivity. Available literature information for MOX fuel is limited to a plutonium content of 30%. Thermodynamic assessment of mixed oxide fuels indicate that with increasing plutonia oxygen potential of the fuel increases and the fuel become more prone to chemical attack by liquid sodium coolant in case of a clad breach. In the present investigation, some of these issues of MOX fuel have been studied to evaluate this fuel for its use in fast reactor. Extensive work on the out-of-pile thermo-physical properties and fuel-coolant chemical compatibility under different simulated reactor conditions has been carried out. Results of these studies were compared with the available literature information on low plutonia MOX fuel and critically analyzed to predict in reactor behaviour of this fuel containing 44% PuO 2. The results of these out-of-pile studies have been very encouraging and helped in arriving at a suitable and achievable fuel specification for utilization of this fuel in fast breeder test reactor (FBTR). As a first step of test pin irradiation programme in FBTR, eight subassemblies of the MOX fuel are undergoing irradiation in FBTR.

A road map for the realization of global-scale thorium breeding fuel cycle by single molten-fluoride flow

For global survival, we need to launch a rapid regeneration of the nuclear power industry. The replacement of the present fossil fuel industry requires a doubling time for alternative energy sources of 5–7 years and only nuclear energy has the capability to achieve this. The liquid metal cooled fast breeder reactors (LMFBR) have the best breeding criteria but the doubling time exceeds 20 years. Further, the use of plutonium in these systems has the potential of nuclear proliferation. The Thorium Molten-Salt Nuclear Energy Synergetic System [THORIMS-NES], described here is a symbiotic system, based on the thorium–uranium-233 cycle. The production of trans-uranium elements is essentially absent in Th–U system, which simplifies the issue of nuclear waste management. The use of 233U contaminated with 232U as fissile material, instead of plutonium/ 235U makes this system nuclear proliferation resistant. The energy is produced in molten-salt reactors (FUJI) and fissile 233U is produced by spallation in Accelerator Molten-Salt Breeders (AMSB). This system uses the multi-functional “single-phase molten-fluoride” circulation system for all operations. There are no difficulties relating to “radiation-damage”, “heat-removal” and “chemical processing” owing to the simple “idealistic ionic liquid” character of the fuel. FUJI is size-flexible, and can use all kinds of fissile material achieving a nearly fuel self-sustaining condition without continuous chemical processing of fuel salt and without core-graphite replacement for the life of the reactor. The AMSB is based on a single-fluid molten-salt target/blanket concept. Several AMSBs can be accommodated in regional centers for the production of fissile 233U, with batch chemical processing including radio-waste management. FUJI reactor and the AMSB can also be used for the transmutation of long-lived radioactive elements in the wastes and has a high potential for producing hydrogen-fuel in molten-salt reactors. The development and launching of THORIMS-NES requires the following three programs during the next three decades: (A) pilot-plant: miniFUJI (7–10 MWe), (B) small power reactor: FUJI-Pu (100–300 MWe), (C) fissile producer: AMSB for globally deploying THORIMS-NES.

Thermochemical degradation of limestone aggregate concrete on exposure to sodium fire

Limestone aggregate concrete blocks were subjected to sodium fire conforming to a realistic scenario in order to qualify them as protective sacrificial layers over structural concrete flooring in liquid metal-cooled fast breeder reactors. Mid infrared absorption measurements were carried out on these sodium fire-exposed samples as a function of depth from the affected surface. Definite signatures of thermochemical degradation indicating dehydration and structural modification of the limestone concrete have been obtained. Control runs were carried out to delineate the thermal effects of sodium fires from that of the chemical interaction effects. Measurements on limestone aggregate samples treated with fused NaOH provided direct evidence of the exact mechanism of the sodium attack on concrete. The observed degradation effects were correlated to the mechanical strength of the concrete blocks and to the intensity of the sodium fire experienced.

Strategy for combining fast growth in electricity generation capacity with the early introduction of the thorium cycle in India

The purpose of this paper is to describe a strategy for combining fast growth in electricity generation capacity, with the early introduction of the thorium cycle in India. Three types of Liquid Metal Cooled Fast Breeder Reactors (LMFBRs) are considered: LMFBR 1, with plutonium/depleted uranium driver zones and depleted uranium blankets; LMFBR 2, with self-sustaining plutonium/depleted uranium driver zones and depleted uranium axial blankets, while the radial thorium blanket produces excess U 233; LMFBR 3, with U233-Th driver zones and Th blankets. The strategy suggested is a combination of 50% LMFBR 1 and 50% of a symbiotic mix of LMFBR 2 and LMFBR 3 all the three using metal fuel. With this strategy, the delay in reaching an installed electrical capacity of 350 GWe is likely to be only about 15 years beyond the approximately 50 years envisaged with only metal fuelled LMFBR 1.

Atmospheric Dispersion of Sodium Aerosol due to a Sodium Leak in a Fast Breeder Reactor Complex

Liquid sodium at high temperatures (470 K to 825 K) is used as the primary and secondary coolant in Liquid Metal cooled Fast Breeder Reactors (LMFBR). In the event of a postulated sodium leak in the Steam Generator Building (SGB) of a LMFBR, sodium readily combusts in the ambient air, especially at temperatures above 523 K. Intense sodium fire results and sodium oxide fumes are released as sodium aerosols. Sodium oxides are readily converted to sodium hydroxide in air due to the presence of moisture in it. Hence, sodium aerosols are invariably in the form of particulate sodium hydroxide. These aerosols damage not only the equipment and instruments due to their corrosive nature but also pose health hazard to humans. Hence, it is essential to estimate the concentration of sodium aerosols within the plant boundary for a sodium leak event. The Gaussian Plume Dispersion Model can obtain the atmospheric dispersion of sodium aerosols in an open terrain. However, this model does not give accurate results for dispersion in spaces close to the point of release and with buildings in between. The velocity field due to the wind is altered to a large extent by the intervening buildings and structures. Therefore, a detailed 3-D estimation of the velocity field and concentration has to be obtained through rigorous computational fluid dynamics (CFD) approach. PHOENICS code has been employed to determine concentration of sodium aerosols at various distances from the point of release. The dispersion studies have been carried out for the release of sodium aerosols at different elevations from the ground and for different wind directions.

Analytical study on elimination of severe recriticalities in large scale LMFBRS with enhancement of fuel discharge

The possibility of severe recriticality could be excluded if the molten core materials are discharged from reactor core in the early stage of core disruptive accident (CDA). Based on this idea, several design measures for future commercial liquid metal-cooled fast breeder reactors (LMFBRs) have been proposed to enhance the molten fuel discharge from core in order to prevent formation of the core-wide molten pool with high mobility. One promising concept in these design candidates is modified-FAIDUS (Fuel subassembly with Inner DUct Structure). The event progression in unprotected loss of flow (ULOF) accident in a sodium-cooled large scale FBR with modified-FAIDUS was analyzed to assess the effectual performance of modified-FAIDUS in preventing severe recriticality using the SAS4A and SIMMER-III codes. Two parametric cases were performed covering the uncertainty of duct wall failure mechanism, one with stable fuel crust and another with unstable crust condition. The calculation showed that the final amount of discharged fuel from core in both cases was more than 20% of initial core inventory. The degraded core after fuel discharge is composed of the mixture of solidified fuel, swollen fuel chunks and molten steel, of which low mobility prevents massive fuel motion. The reactor power lowered to decay heat level and the reactivity lowered around −20 $, thus, the possibility of severe recriticality was eliminated.

Study on creep-fatigue life prediction methods for low-carbon nitrogen-controlled 316 stainless steel (316FR)

Low-carbon, nitrogen-controlled 316 stainless steel called 316FR was developed and is regarded as a principal candidate for a main structural material of liquid metal-cooled fast breeder reactor plants in Japan. To develop a creep-fatigue evaluation method suitable for this steel, a number of uniaxial creep-fatigue tests have been conducted for three products of this steel. Long-term data up to about 35,000 h were obtained and applicability of failure life prediction methods was studied based upon their results. Cruciform shaped specimens were also tested under biaxial loading conditions to examine the effect of stress multiaxiality on failure life under creep-fatigue condition.

Simulation of chemical kinetics in sodium-concrete interactions

Abstract Sodium-concrete interaction is a key safety-related issue in safety analysis of liquid metal cooled fast breeder reactors (LMFBRs). The chemical kinetics model is a key component of the sodium-concrete interaction model. Conservation equations integrated in sodium-concrete interaction model cannot be solved without a set of relationships that couple the equations together, and this may be done by the chemical kinetics model. Simultaneously, simulation of chemical kinetics is difficult due to complexity of the mechanism of chemical reactions between sodium and concrete. This paper describes the chemical kinetics simulation under some hypotheses. The chemical kinetics model was integrated with the conservation equations to form a computer code. Penetration depth, penetration rate, hydrogen flux, reaction heat, etc. can be provided by this code. Theoretical models and computational procedure were recounted in detail. Good agreements of an overall transient behavior were obtained in a series of sodium-concrete interaction experiment analysis. Comparison between analytical and experimental results showed that the chemical kinetics model presented in this paper was creditable and reasonable for simulating the sodium-concrete interactions.