Fast Breeder Reactor Research Articles

Measurement of fast neutron induced (n,γ) reaction cross-section of 68Zn, 96Zr, 121Sb and 123Sb in the energy range of 1 to 2 MeV

The (n,γ) reaction cross-section for the elements 68Zn, 96Zr, 121Sb and 123Sb, present in the reactor structural/shielding materials, was measured by neutron activation technique in the neutron energy region of 1–2 MeV as very limited data is available in this energy range. Further, the neutron spectrum peaks in this energy region for the fast breeder reactors and proposed accelerator driven sub-critical systems. The natural strontium (natSr) element was used as a neutron flux monitor by considering effective combined reaction cross-section for 86Sr(n,γ)87Srm and 87Sr(n,n′)87Srm reactions. The pellets of mixture of sample and monitor were irradiated by a quasi-mono energetic fast neutron beam, generated by 7Li(p,n)7Be reaction at FOTIA, Bhabha Atomic Research Centre, Mumbai, India. The activity of activation products was measured by off-line gamma-ray spectrometry using High Purity Germanium Detector (HPGe). The present data with improved uncertainty and covariance analysis enhance the cross-section data base for better constraining the evaluated data and theoretical models. The theoretical (n,γ) reaction cross-sections were calculated using TALYS 1.96, which could reasonably explain the present data with the Fermi gas level density prescription.

Microstructural constituents in compositionally graded NiCrFeSiBC hardfacing alloy/316L SS additive manufactured deposits

Ni-Cr-Fe-Si-B-C hardfacing alloys (D50) provide excellent resistance to abrasive and adhesive wear, high temperature oxidation and corrosion. These alloys are candidate materials for hardface coatings on in-reactor components of fast breeder reactors. Conventionally these deposits are made on steels through the welding route. Due to high fluidity of the molten alloy, generation of residual stress and significant dilution of the deposit, achieving crack free deposits is often a very difficult and time consuming task. To address such issues arising out of conventional fabrication routes, compositionally graded coatings have been tried through laser additive manufacturing. A 3 layered compositionally graded deposit (D50/50D50-50SS/SS) was made on 316 L SS build plate and the phase constituents and possible correlation with micro hardness was analyzed. Uppermost D50 layer consisted of γ-Ni, γ-Ni + Ni3B and γ-Ni + Cr5B3 eutectics, CrB, Cr7C3, Cr23C6, Cr3C2 and Ni3Si precipitates. The middle layer consisted of γ-(Ni,Fe), γ-(Ni,Fe)+Cr5B3, Cr5B3, Cr3C2, Cr7C3,Cr23C6 and Ni3Si. The SS layer closest to the build plate had γ-Fe, Cr3C2 and Cr7C3 precipitates. Corresponding to the variation in microstructure a gradual change in hardness also can be seen starting from the top layer to SS build plate: 580 ± 35 HV0.1, 532 ± 17 HV0.1, 348 ± 21 HV0.1 and 230 ± 10 HV0.1, which is promising.

Enhancing fuel breed-burn performance in a sodium-cooled fast reactor using a novel reactivity control method

This study aims to enhance the fuel cycle and fissile breeding performance of a sodium-cooled fast breeder reactor (FBR) by utilizing minor actinides (MAs) as a means of reactivity control alongside partially-inserted control rods. Choosing the PFBR-500 as the reference design, four core models, designated as Cases A, B, C, and D, utilizing various proportions of minor actinides (MAs) were built and simulated using OpenMC. The MA concentrations were optimized to compensate for the withdrawal of control rods and ensure the same initial reactivity for all cores. Burnup analysis over 365 EFPDs revealed a significant increase in cycle length and burnup for the modified cores along with a modest rise in breeding ratio. Notably, Case C, employing 3.45 wt.% MAs in the 88 inner-core fuel subassemblies achieved an extra 62.25 EFPDs cycle length, a 33.74% rise in single-cycle burnup, and a 3.86% increase in breeding gain compared to the reference. Loading MAs into the inner-core region proved to be more effective in enhancing both fertile-to-fissile conversion (thus extending the cycle length and fuel burnup) and transmutation than utilizing MAs throughout the core due to greater neutron flux at the core center. While Case D utilizing 2.2 wt.% MAs both in the inner and outer core fuel subassemblies had the highest overall MA loading, it demonstrated lower increments in cycle length, burnup, and breeding gain compared to Case C. Case C also exhibited the highest overall destruction rate (approximately 24%/y) for Np and Am isotopes, and successfully transmuted around 24.08 kg 237Np, 9.33 kg 241Am and 3.82 kg 243Am over the course of a single year. The addition of MAs also achieved a slight flattening of the axial and radial flux profile and a decrease in the flux peaking factor. However, it slightly lowered the beta-effective, Doppler constant, and control rod assembly worth, and shifted the coolant void reactivity worth to the positive side.

Influence of thermo-mechanical and solution annealing treatments on the flow behaviour of 14Cr-15Ni stainless steel

ABSTRACT Thermo-mechanical treatment (TMT) involving cold work and annealing is prevalent in enhancing the mechanical properties of precipitation-strengthened stainless steels. This investigation evaluates the effect of TMT and solution annealing treatment (SAT) on the flow properties of a Ti-added 14Cr-15Ni austenitic stainless steel, which is used as cladding material for fast breeder reactors (FBRs). Comparisons drawn with the solution-annealed (SAT) condition showed that TMT enhances the yield stress more significantly compared to the ultimate tensile strength. Analysis of the work-hardening behaviour suggested that though TMT enhanced the initial work-hardening rate, its microstructure was more susceptible to dynamic recovery. Despite exhibiting lower strength, the SAT process enhanced the ductility at all the investigated test temperatures. Adoption of this TMT process on this material has the potential to enhance the performance of the FBR's cladding.

Lessons Learned from Failures and Design Improvements for Large Diameter Seals

Large diameter seals are used in Prototype Fast Breeder Reactor (PFBR) to prevent the escape of radioactive argon cover gas through wide annular gaps. The design of these seals shall address different challenges than normal seals. The large tolerances needed on the width of annulus, achievable tolerances on rubber, clearances in the seal holder, etc., influence the seal performance. The paper discusses analytical and experimental work done to develop robust large diameter seals. The contours of the present study are based on the feedback from the testing and qualification of small size model seals. Various permutations of geometry for seal and its holder are numerically investigated. The study mainly highlighted the influence of seal holder on the seal performance. Various sensitivity studies done on selected seal and seal holder demonstrated robustness of the design solutions.Graphical

Evaluation of Reaction Rate Distribution for Shielding Region in the Prototype Fast Reactor Monju

The reliability and usefulness of the reaction rate distribution data measured in the prototype fast breeder reactor Monju were examined through a comparison with a calculation using JENDL-4.0, mainly focusing on shielding regions around the reactor core. The reaction rates of 238U (n,f) and 58Ni (n,p) sensitive to high-energy neutrons were all judged reliable. The calculation-to-experiment values were slightly worse in the shielding regions; however, those for the 58Ni (n,p) reaction rates were improved by employing JEFF-3.3 instead of JENDL-4.0. A different tendency was observed between the two reactions, probably due to the 238U (n,f) cross section in the energy range of around 700 eV. The reaction rates of 235U (n,f), 239Pu (n,f), 238U (n,γ), and 197Au (n,γ) sensitive to the lower-energy neutrons were mostly judged reliable. The data in the lower axial shielding region are less reliable but may be acceptable for the shielding calculation.

Atmospheric dispersion of radioactive materials for radiological risk assessment in case of hypothetical sodium cooled fast reactor accident

The latest advancements in nuclear power, namely the Generation IV (Gen IV) reactor design, have been successfully chosen in numerous countries. These cutting-edge reactor designs boast superior safety and increased fuel efficiency compared to their predecessors. Among the Gen IV designs is the sodium-cooled fast breeder reactor, and to substantiate the safety assertions, HYSPLIT (Hybrid Single-Particle Lagrangian Integrated Trajectory) dispersion modeling evolved by the National Oceanic and Atmospheric Administration (NOAA) will be employed to scrutinize the potential consequences of an accident. In this context, the analysis focuses on the 1250 MWt/500 MWe sodium-cooled fast breeder reactor (SFR) and its partial fuel melt accident, considering the crucial radioactive reduction factor. Despite existing studies on the consequences of SFR accidents, these studies need to pay more attention to the significance of the radiation reduction factor. This factor is paramount in accurately calculating dose outcomes, as not all radioactive materials can disperse freely within structures, thereby reducing inhalation and cloud shine doses. Furthermore, this investigation underscores the impact of meteorological data and dispersion model input parameters, such as release heights of radioactive materials and measured air concentration heights, on the results of atmospheric dispersion models. Hence, uncertainties in dispersion models arise due to the variability in these factors. The HYSPLIT dispersion model estimates a cumulative dose of 1.65 rem (0.0165 Sv) for a partial fuel melt scenario. The research reveals a maximum dose of approximately 1.8 rem (0.018 Sv) within a 24-hour period, indicating compliance with NRC limitations for Indian fast reactor accident consequences.

Evaluation of Reaction Rate Distribution Measurement in the Prototype Fast Reactor Monju

In the prototype fast breeder reactor Monju, reaction rate distributions of the fission reaction rates of 239Pu, 235U, and 238U and the capture reaction rate of 238U were measured using activation foils during its system startup test. The measurements in the core and radial blanket regions were evaluated in detail, and their reliability and usefulness as the validation data for fast reactor neutronics design methodologies were examined through a comparison with calculations. The reaction rate data measured in Monju were confirmed all reliable and useful as the validation data. The fission reactions of 239Pu, 235U, and 238U can be validated with an accuracy of a few percent in the core and blanket regions. The capture reaction of 238U in the core region also can be validated with a similar accuracy, whereas a precise calculation of the foil cross section is necessary to consider the resonance shielding effects of the surrounding fuel pins and a foil.

Radiochemical and chemical characterization of fuel, salt, and deposit from the electrorefining of irradiated U-6 wt% Zr in hot cells

Abstract Metal fuels are considered as the promising candidates for future fast breeder reactors. Pyro-chemical reprocessing is the ideal method for reprocessing spent metallic fuels due to the inherent process advantages. Electrorefining run was demonstrated in a hot cell facility with irradiated U-6 wt% Zr alloy at 500 °C using LiCl–KCl eutectic melt. In order to understand the behavior of the actinides and various fission products during high-temperature electrolysis, various process streams, viz., irradiated metal alloy fuel, the eutectic salt, and the cathode deposit were analyzed for the uranium, plutonium, and other fission product contents. Various methods employed for characterizing the process streams and the behaviors of some of the fission products during the electrolysis process are highlighted. The major gamma emitting radionuclides present in the irradiated fuel were 106Ru, 125Sb, 134Cs, 137Cs, 144Ce, and 154Eu. During electrorefining, cesium, cerium and europium were oxidized and dissolved in the molten media, whereas ruthenium and antimony remained in the anode basket. A minor contamination of zirconium was found in the cathode deposit.

Understanding the Subsurface Microstructure and Thermal Behavior of Model Oxide Dispersion Strengthened Alloys Through FIB_SEM and TEM

AbstractA potential but an underexplored application of FIB_SEM is its ability to image the subsurface microstructure and capability for an associated chemical analysis. In this article, agglomerated microstructures of two model oxide dispersion strengthened alloy systems, consisting of ZrO2 and Y2O3 dispersions, are evaluated to understand its elevated temperature behaviors. The systems under evaluation are relevant as candidate nuclear structural materials for next-generation fast breeder reactors, and FIB_SEM technique is effectively used along with TEM and EDS for a comprehensive understanding of the material microstructure, and the results are discussed. Distinct microstructures are observed for the two systems with a difference in crystallite size distribution and presence of micron-sized dispersoids in Y2O3. The varied behavior of dispersoids is discussed in terms of its pre-annealing microstructures, and the superiority of ZrO2 over Y2O3 as a dispersoid for ODS alloys is emphasized.

Mechanical and electrical performance of glass fiber reinforced plastic insulation for cryogenic application in fusion magnet irradiated in fast breeder reactor

In a superconducting (SC) magnet fusion reactor, the insulation material used has to withstand significant fluxes of neutrons and neutron-induced gamma radiation. Epoxy-based glass fiber-reinforced plastic (GFRP) composites are the insulation system for fusion magnets and are highly demanding. This insulation system requires excellent mechanical and electrical performance under mixed neutron and gamma radiation at cryogenic temperatures. The insulation material developed consists of boron-free S-glass fiber impregnated with a two-component modified Diglycidyl ether of Bisphenol-A (DGEBA) epoxy resin. An anhydride-cured DGEBA resin system is not suitable for cryogenic temperatures. Therefore, to overcome this thermal stress issue in cryogenic components of SC magnets, a high-toughness, flexible, aromatic amine-cured epoxy resin system was developed for the insulation system. In order to assess the radiation resistance for mechanical and electrical properties, the composite samples were irradiated in the fast breeder test reactor (FBTR) at IGCAR, Kalpakkam, India. The irradiation experiment was carried out for a neutron fluence of 1.0E+21 n/m2 and a gamma dose of 0.4 MGy at a full reactor power of 32 MWt. The tensile strength, interlaminar shear strength, and breakdown strength measurements were investigated before and after irradiation at room temperature and 77 K. The test results indicate that no significant degradation was observed in the properties of the irradiated samples up to the above-mentioned radiation doses.

Evaluation of Specific Wear Coefficient of Sliding Contact Pairs through Tribological Experiments

In fuel handling machine of Prototype Fast Breeder Reactor, colomony-5 is extensively used for hardfacing of rubbing surfaces. To avoid the challenges associated with hardfacing of the components, it is proposed to eliminate hardfacing by using a forged component of a suitable material. 440C steel is a candidate material for this application. Hence, experiments were conducted to evaluate the tribological properties of 440C steel under various conditions seen during the operation. The tests were carried out using pin on disc tribometer by varying the load from 10-20 N and sliding speed from 5-30 mm/s. Friction and specific wear coefficients were evaluated from the experiments. Besides, Global Incremental Wear Model (GIWM) was used for prediction of wear behaviour of the steel.

Performance evaluation and qualification of hard coatings used in absorber rod drive mechanism of Indian fast breeder reactor

Prevention of surface damage at various guiding and bearing locations assumes high importance in ensuring reliable operation of Absorber Rod Drive Mechanisms (ARDM). Contact at an unintended location during any of the operating conditions, including seismic conditions, may cause unacceptable damage, reducing the availability of the ARDMs. Based on international and in-house experiences, the materials and surface coatings for mating surfaces and guiding locations were selected in the ARDMs of Indian Fast Breeder Reactors (FBR).Various mechanical and metallurgical investigations followed by integral assembly tests with full-scale components on hard face coatings were done to evaluate the performance of the hard coatings and to qualify them for reactor operation. In spite of detailed investigations, surface damage-related problems were faced during the initial stages of integral assembly testing, which severely delayed the qualification testing. Key lessons learned from the experimental program are that a holistic approach involving complete kinematics is necessary in the design of the systems to prevent unwanted surface damage-related issues. All these details are discussed in the present paper.From the qualification testing, it can be concluded that the performance of hard coatings for the ARDMs of Indian FBRs is satisfactory. The material and design selections made have formed the basis for the given applications in future FBRs in India.

Investigations of Suction Recirculation in Primary Sodium Pump of Future Indian Fast Breeder Reactors

The design of the next generation of fast breeder reactors has commenced, with the main targets being enhanced safety and improved economy. Nuclear heat generated in the fuel subassembly of fast reactors is removed by circulating sodium through the core using centrifugal pumps. The primary sodium pumps (PSPs) used are large-capacity pumps, and the design of these pumps is different from that of traditional pumps. Though many works have been reported for the performance prediction of centrifugal pumps, most of these works have been carried out in a decoupled way, and only a few works have been reported where the pump is modeled with all the associated geometric structures. Centrifugal pumps are prone to a phenomenon called suction recirculation, which occurs when pumps are operated significantly below the best efficiency point. This suction recirculation has a strong potential to damage the impeller. Correlations given in the literature for the prediction of the onset of recirculation cannot be used for complicated inlet geometries, and three-dimensional computation fluid dynamics (CFD) investigations are most suited for such applications. Many devices have been reported in the literature to reduce the intensity of (or to suppress) suction recirculation. Webs provided in the suction plenum will modify the velocity distribution at the impeller inlet and also can influence suction recirculation. In this work, the centrifugal pump used for primary sodium pumping for fast reactor applications is simulated using CFD techniques in an integrated way. The frozen rotor approach is used to simulate the impeller-diffuser hydraulics. The effect of flow hydraulics in the suction plenum, flow distribution in the standpipe–pump gap, and flow conditions in the pool on the performance characteristics of PSPs are simulated. The flow rate for the onset of suction recirculation is predicted and compared with correlations available in the literature. Simulations are carried out to study the effect of webs on suction recirculation. The effects of the number of webs and the web geometry are also studied.

Thermal conductivity-based hydrogen in argon detector (HAD): Design, development of instrumentation, and performance evaluation.

The Hydrogen in Argon Detectors (HADs) working on the principle of thermal conductivity difference between argon (reference) and argon + H2 (sample) in the cover gas of a sodium-cooled fast breeder reactor using the Wheatstone bridge circuit. The output of HAD is very sensitive to the gas flow rate and to the variation in ambient temperature. The RMSnoise of current was brought within ±1 µA by a constant current source across the bridge. The temperature correction factor fed in the processor brought down the RMSnoise in the HAD signal within ±2.4 mV than the uncorrected one having RMSnoise of ±230 mV. The filtered noise of HAD facilitates enhancing the detection limit of HAD down to 30 ppm (3σ) of hydrogen in argon.

Numerical Simulation of the Effect of Fusion Welding Processes on Residual Stress in Stainless Steel Pipe Weld Joints Used in Fast Reactors

The main structural material used in the building of Fast Breeder Reactors (FBR) is stainless steel of type 316L(N). Finite Element Model (FEM)-based simulations were carried out to understand the effect of various welding processes on the residual stresses induced in 316L(N) stainless steel primary sodium-carrying pipes with an inner diameter of 208 mm and a wall thickness of 5.6 mm. The welding processes considered include multi-pass Tungsten Inert Gas (TIG) welding, A-TIG, laser, hybrid laser-TIG, and Hybrid laser-Metal Inert Gas (MIG). First, single or hybrid heat sources are chosen based on the welding process, and then weld bead profiles obtained during simulation are made to match with that of the actual weld bead profiles obtained during experiments. Then, the optimised heat source is employed for the thermal analysis. The thermal analysis output is then successively linked and provided as input to the mechanical analysis, which uses an isotropic hardening model to predict hoop and axial residual stresses in pipe weld joints. The results show that the A-TIG weld joint exhibited lower hoop stress (165 MPa at inner diameter and -2 MPa at outer diameter) in comparison with that of the other weld joints. The lower residual stress is attributed due to the straight-sided weld bead, lower peak temperature, and slower cooling rate of the A-TIG welding. Therefore, A-TIG welding is recommended among the chosen fusion welding processes for welding of 316L(N) stainless steel primary sodium carrying pipes to minimise the residual stresses and mitigate the premature failure of the pipe weld joints.

A finite volume method based multi-group neutron diffusion model for space-time nuclear reactor kinetics problems in RZ geometry

The transient analysis of Fast Breeder Reactor (FBR) requires the development of high fidelity space-time nuclear reactor kinetics codes for the realistic prediction of spatial and temporal behavior of neutron flux and fission power along with the feedback effects. So, a space-time nuclear reactor kinetics model has been developed for solving the coupled time dependent multi-group neutron diffusion equations and the delayed neutron precursor equations in axi-symmetric cylindrical (r-z) geometry. The numerical model employs an orthogonal mesh finite volume method for spatial discretization and implicit method for time discretization. A computer code named MANDIR-KRZ (Multi-group Analysis of Neutron DIffusion in Reactor for Kinetics problems in RZ geometry) has been developed and employed to reconstruct the keff, transient neutron flux and power results of different benchmark problems. Mathematical formulation of the numerical model and results of the code for Dodds and FBR transient benchmark problems are discussed in this paper.

Numerical investigation on design optimization of core catcher in sodium cooled fast breeder reactor

Core catcher is a passive prevention device that is designed to prevent damage to the lower head of the reactor vessel caused by direct contact with molten corium. In sodium cooled fast breeder reactors (SCFBRs), the structural design of the core catcher directly affects the shape of the debris bed, which in turn impacts the re-criticality and long-term decay heat removal of the debris bed. In this study, the improved discrete element method (DEM) was used to simulate the formation process of the debris bed below the core catcher, and the influencing mechanism of the structure of the core catcher on the shape of the debris bed was explored. Through numerical simulation, the influence on the shape of the debris bed from the projection area of chimney cap, the inclination angle of chimney cap, and the distance between central chimney and surrounding chimneys of the core catcher was examined. It was found that secondary scattering played a significantly positive role in enhancing the uniformity of the debris bed. This study lays a foundation for analysis of the heat transfer performance and evaluation of long-term decay heat removal of debris bed, and can offer engineering guidance for the design and optimization of sodium cooled fast reactor core catcher.

Influence of Anions on the Corrosion Behavior of 9Cr-1Mo Ferritic Steel in Aqueous Media

Modified 9Cr-1Mo steel is used as steam generator material in prototype fast breeder reactors, owing to its good corrosion resistance, creep, and thermal conductivity characteristics. In the present study, the corrosion behavior of modified 9Cr-1Mo steel in aqueous environments containing various corrosive ions at potentiodynamic polarization conditions is reported. In chloride-exposed samples, randomly distributed dense shallow pits were seen, whereas mixed anions (, Cl−, and ) in fresh water resulted in deeper pits uniformly distributed on the entire surface. Laser Raman spectroscopy studies revealed the formation of Fe-oxides/hydroxides and Cr(VI) species, except in alkaline solution. A very thin film of only Fe-oxides (Fe3O4, γ-Fe2O3, and α-Fe2O3) was identified in an OH– ion-dominated solution. X-ray photoelectron spectroscopy results confirm the corroded surfaces comprised of Fe, Cr, and Mo-oxides of varying composition and enrichment of Mo-oxides in alkaline solution. Chloride ions present in the corroded layer influenced the pitting corrosion in neutral chloride and fresh water solutions. Field emission scanning electron microscopy images revealed abundant crystalline lepidocrocite with laminar morphology and doughnut-type magnetite together with pits on the samples exposed in neutral chloride and fresh water media.

Influence of seismic isolation on the earthquake response of internal components in a fast reactor

Nuclear energy is being considered as a frontrunner towards achieving the zero carbon goal. A sustainable use of nuclear energy is possible when a reactor design is deployed across regions with varying seismic hazard. Seismic isolation technology has been proven to reduce earthquake loads in the structures, and is being seen as a key towards the standardized reactor designs. This paper investigates the influence of seismic isolation on the response of a generic fast reactor (GFR) similar to the prototype fast breeder reactor (PFBR) currently under commissioning at Kalpakkam, India. This reactor comprises a head-supported main vessel that houses other reactor components (e.g., core, inner vessel). The major vessels of the reactor are partially filled with liquid sodium. In order to understand the influence of seismic isolation, simplified lumped mass stick models (LMSMs) for the GFR components were developed. These models were based on static analysis results obtained using corresponding finite element (FE) models. The component models were assembled to generate the LMSM and FE model for the GFR. The LMSM was verified through the comparison of dynamic properties obtained using the FE model, and through the comparison of response-history analysis results. These models considered fluid masses lumped at respective shells, thereby ignoring the fluid–structure interaction (FSI) effects. A representative system was analyzed with and without considering FSI effects. These effects did not influence acceleration response significantly. Fixed-base (FB) and isolated base (IB) configurations of the LMSM for the GFR were subjected to a series of response-history analyses. Seismic isolation led to a substantial reduction (e.g., up to 10 times) in the median floor accelerations, particularly for lower sliding friction and longer sliding period. The floor spectral ordinates for the IB configuration of GFR subjected to 1.00 g shaking were less than that for the FB configuration of GFR subjected to 0.25 g shaking at most periods, suggesting that the PFBR can be deployed at locations with seismic hazard much greater than that at Kalpakkam if suitable isolation systems are used.