Fast Breeder Reactor Research Articles

Palladium telluride within nuclear waste containing borosilicate glass

As a part of its clean nuclear energy program, India is developing Fast Breeder nuclear reactor. High level nuclear wastes, which are likely to be produced upon reprocessing of its spent fuel will contain platinum group of elements in concentrations higher than their respective solubility limits within borosilicate melts. Towards this, interaction study of palladium telluride (PdTe) with sodium barium borosilicate base glass and waste glass were carried out. Mostly, spherical PdTe got settled down at the bottom of ceramic crucibles without any compositional exchanges with base glass. However, significant compositional variations took place in case of waste glass. Interaction studies between Alloy 690 and base glasses show development of wide reaction zones (including intergranular attacks within Alloy 690 coupons, formation of Cr2O3 layer and crystallization of needle shaped phases within adjacent glass domains) at the interfaces. In case of waste glass, tiny PdTe beads crystallized at the reaction zone/borosilicate glass interfaces.

A micromechanical analysis of swelling-induced embrittlement in neutron-irradiated austenitic stainless steels

Swelling is commonly observed in austenitic stainless steels irradiated at high doses and high temperatures, such as fuel claddings in Fast-Breeder Reactors (FBR), as a consequence of the formation of irradiation-induced nanovoids. Macroscopic embrittlement has been reported above a critical swelling level. Several physical explanations have been proposed such as void-related fracture mechanisms and phase transformation associated with local changes of microchemistry. In addition, structural effects can affect the critical swelling level, tentatively explained by local corrosion, swelling gradient and bending effects. In order to address these phenomena, an extended experimental database is first presented, based on fuel pins irradiated in the Phénix fast reactor, showing the evolution of conventional tensile properties with swelling. In the present database, 15-15Ti specimens were irradiated from 30 to 120 dpa, with irradiation temperatures ranging from 400 °C to 625 °C and a maximum swelling of 11%. Other data published in scientific literature were also included in the analysis to extend the swelling range up to 30%. SEM and TEM analysis are also summarized to highlight specific embrittlement mechanisms related to the presence of voids. Theoretical analysis and finite element simulations are then performed based on porous materials constitutive equations in order to rationalize the experimental observations. Analytical and numerical results show that swelling-induced embrittlement can be understood from a mechanical perspective as a transition from void growth to void coalescence deformation mode. The effect of the spatial heterogeneity of void distribution is quantified and shown to be a key parameter. Structural effects coming from tests performed on ring and axial tensile specimens and from the presence of a swelling gradient in cladding thickness are also quantified. Numerical results are compared to experimental data, and a good agreement is observed.

Impact properties of modified 9Cr 1Mo steel welds: Comparison between cold wire and hot wire gas tungsten arc welding processes

Modified 9Cr 1Mo steel is used for the manufacture of Steam Generators (SG) for Prototype Fast Breeder Reactor at Kalpakkam, India. The hot wire Gas Tungsten Arc Welding (GTAW) process was selected for the manufacture of SGs wherever feasible in addition to the conventional cold wire GTAW process due to various advantages. Drop weight tests on modified 9Cr 1Mo steel welds produced by hot wire and cold wire GTAW processes were carried out in accordance with ASTM E208. It is found that the Nil Ductility Transition (NDT) behavior of the hot wire and cold wire GTA welds were similar during drop weight tests performed up to −10 °C. Subsequently, Charpy V notch impact tests were carried out at +18 °C and 0 °C in accordance with ASME Boiler and Pressure Vessel Code. The impact strength at 0 °C of welds produced by cold wire GTAW process was found to be superior compared to the welds of hot wire GTAW process. Subsequently, the analysis of impact strength was carried out using a two-sided students t-test. With a 95% confidence level, the results of the t-test revealed that there is a genuine difference in impact strength at 0 °C between the welds of hot wire and cold wire GTAW processes.

Irradiation behaviour of alloy D9 clad and wrapper in FBTR

Titanium modified austenitic stainless steel 316 (Alloy D9) is the material of clad and wrapper for the 500 MWe Prototype Fast Breeder Reactor (PFBR) at Kalpakkam, India. Towards evaluating the irradiation performance of the mixed oxide (MOX) fuel design and Alloy D9 structural material, a test fuel subassembly (FSA) made of Alloy D9 wrapper with 37 MOX fuel pins of Alloy D9 cladding was irradiated in Fast Breeder Test Reactor (FBTR) to a peak burnup of 112 GWd/t. The dimensional changes, swelling and mechanical properties of alloy D9 clad and wrapper at displacement damage levels of 40-62 dpa were evaluated through post-irradiation examinations (PIE) and compared with that of AISI 316 clad/wrapper of FBTR. This study has revealed superior performance of alloy D9 compared to AISI 316 counterparts and suggests that MOX fuel subassembly of PFBR with Alloy D9 cladding and wrapper can safely attain the initial target burnup of 100 GWd/t and corresponding displacement damage of 85 dpa.

Investigation of swelling behaviors of U-10Zr metallic fuel in the low temperature regime via a cavitational void swelling model

The uranium zirconium (U-Zr) metallic nuclear fuel has long been considered as a promising candidate fuel type for fast breeder reactors. One of the key performance issues of the U-10Zr metallic fuel is its rather strong swelling, which would lead to significant fuel cladding mechanical interaction (FCMI) at relatively high fuel burn-ups. It is therefore imperative to understand the fuel swelling behaviors and the underlying physics mechanisms. Recently published experimental results of U-10Zr fuel swelling behaviors in a low temperature regime (400-600K) showed perceivable swelling of the in-pile irradiated fuel. However, the underlying physics mechanism behind this swelling is not clear. In this work, we performed in-depth analyses on the microstructural features of the irradiated fuel with metallography and conducted a finite-element based simulation on a computational platform, COMSOL, to determine the hydrostatic stress in the irradiated fuel. The spherical objects in the metallography result have been identified as cavities with relatively low fission gas pressure instead of equilibrium gas bubbles. A rate theory code based on the cavitational void swelling model has been developed and the fuel swelling behaviors of U-10Zr fuel in the low temperature regime have been assessed. The simulation results suggest a much lower migration energy for vacancy diffusion in the metallic fuel compared to what is currently used, and a new set of key parameters for the rate theory model were determined.

Asymptotic Crack Size of a Prototype Sized Pipe Bend and Comparison With A16 Master Curve

AbstractIn the design of a fast breeder reactor (FBR) piping system, asymptotic crack size (2Cs) is traditionally estimated using the leak-before-break (LBB) master curve given in RCC MRx A16. The current approach is seen not to be sufficiently efficient in the LBB demonstration of FBR piping systems. Numerical crack growth studies have received much attention in the past decade. Deployment of fracture mechanics based numerical crack growth studies on prototype-sized pipe geometries generates considerable interest in estimating asymptotic crack size. This paper outlines the advantages of numerical finite element based crack growth studies to estimate the Cs on a prototype-sized pipe bend. The numerical procedure has resulted in an accurate and more realistic estimate of Cs than the traditional LBB Master curve as per A16. Considering the economic advantages of a relatively lower value of Cs for LBB demonstration, it is recommended to adopt the numerical method to estimate Cs for the FBR piping system.

Study on startup characteristics of prototype once-through steam generator for China fast reactor

Once-through steam generator (OTSG) is a vital heat transfer equipment of China Fast Reactor-600 (CFR-600), and its startup characteristics are of great importance to the safe operation of nuclear power plant. In this paper, first of all, the Thermal-hydraulic analysis Code Of Sodium heated once-through Steam generator (TCOSS) has been further optimized, and the code was introduced in the text. Secondly, the Steam Generator Test Facility (SGTF) and India Commercial Fast Breeder Reactor (CFBR) were selected for steady-state verification, and the feedwater flow rate increase condition of the prototype OTSG experiment was selected for transient verification. All the verifications were in good compliance, which proved the reliability of the TCOSS code for subsequent calculation of startup conditions. Finally, taking the prototype OTSG of CFR-600 as the research object, and the transient characteristics of cold startup and hot startup have been predicted and analyzed. The results show that both processes of cold startup and hot startup for prototype OTSG can be realized stably. The simulation results of startup processes showed a good agreement with expectations, which can provide support for the development of CFR-600 steam generator.

Effect of Material Behaviour on Flow Uniformity over Different Spillway Weir Profiles

The spillway weir system finds an important application to achieve uniform flow in the open canal stream passages, for irrigation and in many industrial processes, majorly including cooling of reactor components with liquid sodium in prototype fast breeder reactors, process plants, and power industries. The optimal design of weir profile and the velocity of flow over the weir are the critical factors that need to be optimized to minimize flow separation in order to avoid entrainment of gas in flowing liquid, and also, the nonuniform cooling resulted in uneven temperature distributions in many thermohydraulic structures. In this research, the spillway weir profile made from SS-316, SS-304, concrete, and teak wood is compared for optimization of flow uniformity by means of hydraulic experiments conducted on the water-air model and also to investigate the behavior of flow past over the rectangular weir of three different radii 6 mm, 6.47 mm, and 7 mm, respectively, and the minimum of 1.2% flow separation was observed with 6.74 mm profile, and also, the computational fluid dynamics (CFD) based numerical simulation was performed in the fluid domain to validate the experimental data to attain the flow uniformity when flow past over the spillway weir profile and the results are highlighted. The experimental results on air entrapment, water film thickness, and velocity of water over the weir profile on the perspective of uniformity of flow are also discussed in detail.

Concept of a small fast breeder reactor system capable of transmuting MA and LLFP

The issues of nuclear power systems are the effective use of uranium resources and the reduction of radioactive wastes. Important radioactive wastes are minor actinides (MA : 237Np, 241Am, 243Am, etc.) and long-lived fission products (LLFP : 129I, 99Tc, 79Se, etc.). The purpose of this study was to construct a system that simultaneously achieves the breeding of fissile materials and the transmutation of MA and LLFP using small fast reactors. Transmutation was achieved by loading the innovative Duplex type MA-containing fuel in the core region and the LLFP-containing moderator in the first layer of the radial blanket. Breeding of the entire small fast reactor system was achieved by combining two small fast reactors: one for transmutation and the other for breeding. As a result, it was clarified that in this small fast reactor system, a breeding ratio of approximately 1.03 was obtained, and support ratios (SRs) of 2 or more were achieved for MAs and LLFPs.

Challenges in the Long-Term Behaviour of Highly Radioactive Materials

Highly radioactive materials are at the core in many useful applications ranging from operating nuclear reactors (including fast breeder reactors) to vitrified high-level radioactive waste, which is currently stored and awaiting final disposal into dedicated facilities within deep geological formations [...]

Runge–Kutta type-2 method for solving reactor point kinetics equations and its validation by analysing thermal and fast reactor benchmarks

Runge – Kutta type-2 method is used for solving point kinetic equations. Time step size chosen ranging from 10−3-10−4 sec. RK-2 method is applied to estimate power transients of thermal reactor and fast breeder reactors with step, ramp and sinusoidal neutron reactivity perturbations. Results are compared with the standard problem results at different time steps with different numerical methods ie. Modified ETD and Cohen’s methods. This comparison of transient power values shows that RK-2 method is predicting accurately for fast and thermal reactors for longer duration of time.

A case study on elimination of premature failure source from manufacture of fluoroelastomer inflatable seals for Sodium-cooled Fast Reactor towards sustainability

An innovative model of engineering failure analysis leading to a prospective model of sustainability and decarbonisation, supported by design-material-failure threads, emerge from two defective test inflatable seals (Seals; ∼2 m dia) subjected to laboratory investigations i.e. visual, dimensional, physico-mechanical, material fingerprinting and analysis of variation. The Seals of 500 MW(e) Prototype Fast Breeder Reactor (PFBR) design were industrially manufactured (year: 2007) by cold feed extrusion and continuous cure (CFECC) of a peroxide cured, Viton GBL 200S:600S blend compound, APA-1, 2007. Primary aim was to implement life maximization (global design thread) potential (≥30 y) of advanced polymer architecture (APA) fluoroelastomer (FKM) in reactor inflatable seals (∼6.3 m/∼4.2 m dia) with repeatability and reproducibility by eliminating premature failure source from improper manufacture.Unfeasibility of or unlikely value addition from validation-modelling-simulation-microstructure study necessitated innovative treatment i.e. extracting maximum from the minimalist investigation by maximizing interpretation-data mining-cross verification and utilizing past (validated) FEA results of standard FKM PFBR inflatable-backup seals (life: 10 y) through design-material threads. The treatment was reinforced by reorientation of perspective. Distribution of laboratory data along Seal circumference was interpreted as phenomenological manifestation of thermomechanical-electrochemical interactions in the APA-1, 2007 macromolecular realm, brought about by thermomechanical loads from the progressing CFECC with time. The data was thus utilised to reconstruct the CFECC (2007) in terms of processing parameter variance with respect to a benchmark CFECC study on identical seal design. Incorrect cure index (of APA-1, 2007; global failure thread) was ascertained as the source of premature failure by tracing the origin of departures e.g. tracking the straying of APA-1, 2007 tensile stress-strain behaviour (structure property correlation) and cracking failure limits (global material thread) from the desired (i.e. ≥30 y) along the abstracted echelons (i.e. reconstructing manufacture, molecular-phenomenological interplay) of the CFECC in terms of unacceptable macromolecular transitions. The novel approach and methodology predict new-seal-failure-situation in reactor (by residual stress) and/or continuing operation-safety uncertainty of inflatable seal and reactor (by undercure) from the incorrect cure index. The identified source and template for permanent rectification are crucial for FBR-Gen IV cover gas elastomeric sealing. The applicability extends to other nuclear and inaccessible-hazardous, critical/very critical elastomeric usage (marine, air, space, oilfield, gas etc.) as well.The research findings provide complete description of (APA-FKM) residual stress and tensile stress-strain behaviour by the first principles and an analogy of mechanical-entropy spring, accompanied by a new commentary on structure-property-performance correlation of APA-FKM. The findings are expected to enrich understanding of rubber failure. The failure analysis of a very critical nuclear component grows vastly in connotation using global threads to suggest a prospective conceptual benchmark template for sustainability-decarbonisation-augmentaion i.e. better realization of life maximization potentials in applications by synthesizing improved specifying-manufacturing-quality monitoring with the global threads.This study could provide basis for new research on defect modeling-simulation-validation and encourage formulation-modeling-validation of new constitutive relations, considering incorrect cure index and its effects. The concepts and routes of maximum from minimum, working by first principal, reconstruction, threads and molecular-phenomenological interplay are expected to find wider scope beyond this research.

Early History of the Nuclear Fuel Cycle in France -The Honeymoon Era of Radioactive Waste Reprocessing and Fast Breeder Reactors

Early History of the Nuclear Fuel Cycle in France -The Honeymoon Era of Radioactive Waste Reprocessing and Fast Breeder Reactors

A highly precise micro-analytical XRF method for compositional characterization of fast breeder reactor fuels

Accurate and precise determination of plutonium and uranium in Fast Breeder Reactor (FBR) fuels is one of the most important steps in the chemical quality control of these fuels.

Quantification of sodium contaminant on steel surfaces using pulse CO2 laser-induced breakdown spectroscopy

Corrosion is one of the main reasons for in-core accidents in liquid sodium-cooled fast reactors, especially accidents due to fuel cladding pipe damage. It is urgently required to investigate what kind of compound is produced as a corrosion product after the interaction between stainless steel and sodium in fast breeder reactors (FBR). In this work, the identification and quantification of sodium contaminant on steel surfaces has been conducted using laser-induced breakdown spectroscopy utilizing the specific characteristics of a pulse transversely excited atmospheric CO 2 laser. Experimentally, a pulse TEA CO 2 laser (Shibuya, 10.64 μm, 200 ns) was directed and bombarded onto the sodium contaminant deposited on the surface of stainless steel. An excellent emission spectrum of sodium from the contaminant was obtained without any disturbance from analytical lines from the steel itself. A quantification of sodium contaminant on the steel surface has been successfully made by a linear calibration curve obtained from steel containing various concentrations of sodium. The limit of detection of sodium on the metal surface was estimated to be 0.5 mg/kg. Also, a comparative sodium analysis study was qualitatively made by using LIBS utilizing a pulse Nd:YAG laser. The results demonstrate that the present technique of TEA CO 2 LIBS is far superior to the case of Nd:YAG LIBS, as proven by an excellent emission spectrum of sodium with optimum intensity, and low noise and background emission.

Analysis of Fuel Burnup and Transmutations at High Burnup of Sodium Fast Breeder Reactor

In this paper, the Monte Carlo N-Particle extended computer code (MCNP) were used to design a model of the European Sodium-cooled Fast Reactor. The multiplication factor, conversion factor, delayed neutrons fraction, doppler constant, control rod worth, sodium void worth, masses for major heavy nuclei, radial and axial power distribution at high burnup are studied. The results show that the reactor breeds fissile isotopes with a conversion ratio of 0.994 at fuel burnup 70 (GWd/T), and minor actinides are buildup inside the reactor core. The study aims to check the efficiency of the model on the calculation of the neutronic parameters of the core at high burnup.

In Situ Creep Rupture Life Prediction Methods for 316FR Austenitic Stainless-Steel Weldment

The present paper proposes new creep life prediction methods that can be used in situ as an in-service inspection tool for the condition-based monitoring of high-temperature structure components of 316FR stainless steel that has been developed in Japan for fast breeder reactors. The methods were based on two types of the modified theta methods and the combination of the modified theta and modified omega methods and were applied to creep deformation data during creep tests made at different applied stress levels at 823K and 923 K in air. Creep lives were predicted by the proposed three methods at arbitrary time in the continuous measurement of deformation during the creep tests. The results show that the combination of the modified and modified omega methods gives the most accurate prediction of creep rupture time at the two levels of test temperature for different applied stress levels.

Design and development of probe release mechanism for cable pusher module of PFBR SG inspection system

Inspection of SG tubes in Prototype Fast Breeder Reactor (PFBR) requires pushing and retrieving of the probe along with the cable. A design of cable pusher module (CPM) has been carried out in this research work to deliver adequate axial force for the pushing and retrieval of the probe and cable. The change of the probe calls for the removal of the inspection system from the SG manhole causing extended inspection down-time. Considering all these uncertainties, a probe release mechanism (PRM) is designed, and tested for use. This is a major technological advancement to the PFBR SG inspection system. The probe release mechanism assists with moving the pulleys away to allow cable and probe replacement during maintenance. This is a major technological advancement to the PFBR SG inspection system. This paper describes the exact design of the pre-compressed springs that provide radial force on the inspection cable through the pulleys for smooth insertion and retrieval. Futhermore, the needed actuator force and the wedge configuration for the separation of the pulley modules during probe release are covered. An actuator is designed using a maximum cable weight and tube friction of 2.5 N and 2.3 N, respectively.

Life prediction of steam generator shell nozzle junction for Fast Breeder Reactor

In the field of nuclear power plants, structural components working at high temperature under cyclic conditions are the locations of many complex phenomena. The severe requirements in design procedures require the development of lifetime prediction techniques which take into account a crack initiation under creep and fatigue conditions. The prediction of macroscopic crack initiation is generally made through more or less simple parametric equations, relating lifetime to some stabilized cyclic quantities such as elastic or inelastic strain ranges. In steam generator, the shell nozzle junction is subjected to cyclic loads with subsequent plastic strains that cause propagation of fatigue cracks. It is possible to predict the life of this mechanical component by utilizing the life strain curve ε-Nf. In this paper life predictions are performed for shell nozzle junction and these results are compared with experimental results.

Experimental investigation on sodium leak behaviour through a pinhole

Sodium is most preferred coolant in Fast Breeder Reactors (FBR) due to its favourable neutronic and excellent heat transfer properties. However, when exposed to atmosphere, sodium reacts with oxygen and moisture resulting in sodium fire leading to rise in ambient temperature and release of harmful sodium oxide aerosols. Owing to this, prevention and early detection of sodium leak is of highest priority in the design of sodium systems. In all sodium systems, wire type leak detectors are provided for early detection of sodium leak. However, sodium leak from small defects such as hairline crack and pinholes may lead to oxidation of sodium and deposition of reaction products plugging the leak path. Such leaks may go undetected, cause local corrosion of structural material and may lead to sudden failure. Hence, investigation on sodium leaks from small openings is vital for sodium safety. Series of experiments were carried out to understand such sodium leak scenario through a small aperture simulating a pinhole defect. Sodium plugging phenomenon was studied at various oxygen concentrations from 0.5% to 3% in the leak zone, as a function of driving pressure. Conditions for unplugging were also established by varying sodium temperature and driving pressure, for plugged cases.