Sodium-cooled Fast Breeder Reactor Research Articles

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.

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.

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 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.

Coolant leak rate through fatigue cracked pipe bend: An experimental and design standard based investigation

Leak-Before-Break (LBB) requirement plays a vital role in demonstrating the structural integrity of sodium-cooled Fast Breeder Reactor (FBR) piping system. Crack size at penetration (2CS), crack size that corresponds to detectable coolant leak rate (2CL), and unstable crack size (2CG) are the primary crack parameters associated with the LBB design of FBR piping system. The detectable leak rate through a piping system is directly related to the differential pressure (ΔP) and crack size. The design pressure of FBR piping systems is relatively low, resulting in a high value of CL. Considering the need for an optimum LBB design of the FBR piping system, an experimental approach is adopted to arrive at a realistic estimate of CL. Based on the experiments, a power-law type correlation can be derived to predict the value of CL as a function of ΔP. The evidence from this study points towards the conservatism present in the current FBR LBB design philosophy as per RCC-MRx A16. These observations have several implications for achieving an optimum LBB design criterion for an FBR piping system.

Rankine-vortex model based assessment of CFD methods for simulating the effect of gas entrainment observed in the hot-pool of sodium coold fast breeder reactors

The flow behavior in the hot pool of a compact Sodium Cooled Fast Breeder Reactor (SFR) is investigated in the MICAS facility of the CEA. Computational Fluid Dynamics (CFD) is used to assist the interpretation of the experiments by multi-scale calculations. Special attention is given to the risk of gas entrainment into the reactor core due to the formation of drain-type vortexes at the intermediate heat exchanger intakes. After theoretical considerations based on the Rankine-vortex model, the single effect gas entrainment experiment of Monji is analyzed by different CFD methods. The well-known incapacity of linear RANS models to simulate swirling flow is emphasized and limitations regarding the representation of gas cores by single-phase models are discussed. Then, the flow in the MICAS facility is analyzed by CFD using RANS and LES. The quantitative comparison between experiment and calculation shows that RANS can simulate correctly the global flow patterns in the hot pool. Small-scale drain-type vortexes have not been observed in the RANS calculations. LES can predict the onset of the formation of drain-type vortexes. However, single-phase models cannot simulate stable drain-vortexes.

Studying the effect of assembly homogenization on fuel burn-up for a sodium-cooled fast breeder reactor

Studying the effect of assembly homogenization on fuel burn-up for a sodium-cooled fast breeder reactor

Effect of debris material composition on post accidental heat removal in a sodium cooled fast reactor

Following a core meltdown accident in a pool type sodium cooled fast breeder reactor (SFR), a debris bed is formed over the core catcher. The influence of initial material composition of the debris bed (consisting of fuel and steel) on the post accidental heat removal processes has been investigated in view of reactor safety. A one dimensional transient multiphase thermal model has been developed and employed for this purpose. The model includes sodium boiling, steel melting, fuel melting, slumping of molten particles and formation of molten pool. The overall steel fraction in the debris bed, the extent of steel stratification in the bed due to density difference between the debris particles and the overall porosity of the debris bed are considered as main parameters. It is seen that the assumption of a homogeneous initial mixture of fuel and steel in the debris bed yields a low core catcher plate (CCP) temperature compared to stratified debris beds where steel fraction increases with debris height. In a debris mixture with a low steel content (20% steel and 40% fuel), the peak temperature experienced by the CCP increases only marginally, from 872 K for homogeneous mixture to 900 K for a highly stratified mixture. But, in a debris mixture rich in steel content (30% steel and 30% fuel) the corresponding peak temperatures experienced by CCP are 886 K and 932 K respectively, suggesting that material stratification has significant effect in steel-rich debris mixture. For identical steel stratification profiles, the peak CCP temperature in a steel lean debris mixture is lower than that in a steel rich debris mixture. For 20% overall steel fraction, the peak temperature experienced by CCP is less than the creep limit of 923 K even for the worst stratification. But for 30% overall steel fraction, increasing the extent of steel stratification above a specific value results in CCP peak temperature exceeding the safe limit based on creep failure. Thus, the thermal load on the CCP increases with steel content though the thermal conductivity of steel is much higher than that of the fuel.

Safety Evaluation of Sustainable Uranium Development in China Combined with an Analytical GAN Framework

Uranium is the basic raw material for nuclear energy and is quite highly regarded. Developing a safe supply of uranium is important for safeguarding sustainable nuclear development. The purpose of this study is to evaluate the sustainability of uranium development in China based on dynamic system modeling combined with GAN (Generative Adversarial Network) analysis. We considered eight essential indicators and 42 sub-indicators as part of a detailed quantitative description, and then developed a framework to evaluate and rank China-specific sustainability in light of the quantitative performance of five options for fuel cycle transition scenarios. We began by using KMO sample measurements and the Bartlett Test of Sphericity to determine the suitability of factor analysis and the fitness of the corrected model map and observation data. We then analyzed the roles of different representatives of the decision makers and their impacts on the overall ranking by applying GAN methods from a weighted perspective. Five transition scenarios identified are 1) Pressurized Heavy Water Reactors, 2) Mixed Light Water Reactor + Fast Reactor, 3) Mixed LWR+FR fuel cycle scheme with heterogeneous irradiation, 4) Mixed Pressurized Water Reactor + FR fuel cycle scheme with plutonium recycled directly and repeatedly, and 5) Sodium-cooled fast breeder reactor power plant. The results showed that scenario 1 is the most unsustainable and highly confrontational scenario with a high demand for uranium resources, the lowest sustainability and a high level of antagonism among departments. On the other hand, Scenario 5 requires more advanced technology but exhibits less antagonism among the departments, and thus it largely satisfies the basic requirements for uranium sustainability and low levels of antagonism. In this paper, a safety assessment index system for the uranium supply is computed using a GAN framework. This system plays a crucial role in the sustainable supply and development of uranium, and provides flexibility for coping with the evolution and inherent uncertainties of the necessary technological developments.

Development of In-Service Inspection Rules for Liquid-Metal Cooled Reactors Using the System Based Code Concept

Abstract Inservice inspection rules for liquid-metal cooled plants were historically provided by Section XI, Division 3 of the ASME Boiler and Pressure Vessel Code. However, some parts of the Code, such as acceptance standards for the examination of class 1 and 2 components remained as being in the course of preparation. Although no major revisions were made to Division 3 since the first issue in 1980, a newly developed and published Code Case N-875 now provides alternative examinations to the methods previously contained in Division 3. The Code Case was developed using the System Based Code (SBC) concept pursuing rationalization of codes and standards based on reliability targets throughout a plant's service life. In this paper, an overview of the Code Case is presented. The technical foundation to establish the applicability of these alternative examinations as delineated in the Code Case consists of stage I and II evaluations with compensating individual considerations. Stage I is a structural integrity evaluation without the contribution of inservice inspections, while stage II is evaluation of the detectability of a postulated flaw. Not only conventional direct detection methods, but also indirect detection methods are permitted to be employed through the stage II evaluation. Furthermore, the detailed evaluation procedures are illustrated through the application of the Code Case's evaluation criteria to the primary heat transport piping system of a prototype sodium-cooled fast breeder reactor in Japan, specifically Monju.

Prospects of creation of the two-component nuclear energy system

Possible options of organization of two-component energy system with closed nuclear fuel cycle (CNFC) and new business potential associated with provision of CFC services to foreign customers are examined. Dominating role in the development of nuclear power generation is assigned to VVER reactors with gradually increasing fraction of sodium-cooled fast breeder reactors (FR) incorporated within the joint nuclear fuel cycle operated on MOX-fuel. Components of such energy system perform the following functions: 1. Fast reactors: Generate electric power in base-load mode (possibility of fine tuning of reactor power within limited range (100 – 75 – 100%) is assumed); Utilize waste and/or regenerated uranium for re-fueling power reactors, produce plutonium applicable to the maximum extent for manufacturing MOX-fuel for VVER reactors; Incinerate long-lived highly radioactive wastes – minor actinides separated during reprocessing spent nuclear fuel of FR and VVER reactors. 2. VVER reactors: Generate electricity in compliance with manoeuvrability requirements imposed by the utility company operating the energy system; Utilize MOX-fuel instead of UO2 fuel; Are offered for export with the option of returning SNF back to Russia; Plutonium extracted from VVER spent fuel is used for manufacturing MOX-fuel for SFR. 3. Nuclear fuel cycle facilities: Provide reprocessing of SNF from VVER and SFR with extraction of nuclear materials for recycling; Use depleted or reprocessed uranium and plutonium extracted from spent nuclear fuel for manufacturing MOX-fuel; Provide partitioning of RAW for subsequent utilization of minor actinides and reduction of risks of proliferation of nuclear materials, conditioning and disposal of RAW. Russia possesses capacities for establishing the two-component system with CNFC, as well as the new business approach to rendering CNFC services to foreign customers.

On One Imperfection Estimation Method for Thin Shell Buckling in the Design Code RCC-MR

The thin shell design code RCC-MR is used for sodium-cooled fast breeder reactor components operating at high temperatures. Thin shells from such applications can be designed using linear elastic buckling analysis, following procedures given in RCC-MR. For human safety, such procedures can and should be examined by the broader scientific community. Among such procedures, RCC-MR provides three alternative methods to quantify an imperfection value; and that value is used in subsequent calculations to determine safe loads. Of these methods, the third seems potentially nonconservative for some situations. Here, we examine that third method using detailed numerical examples. These examples, found by trial and error, are the main contribution of this paper. The first example is a nonuniform cylindrical shell closed with a spherical endcap under external pressure. The second is a cylinder with an ellipsoidal head under internal pressure. The third is an L-shaped pipe with an end load. In all three cases, the new computed imperfection quantity is found to be surprisingly small compared to the actual value used for computations (e.g., 25 times smaller), and in two cases, the result is insensitive to the actual imperfection. We explain how the three examples “trick” the imperfection quantification method in three different ways. We suggest that this imperfection quantification method in RCC-MR should be re-examined. The primary value of our paper lies not in new mechanics, but in identifying unexpected ways in which a particular step in shell design using RCC-MR could be potentially nonconservative.

Evaluation of cobalt free coatings as hardfacing material candidates in sodium-cooled fast reactor and effect of oxygen in sodium on the tribological behaviour

The feedback produced by operating Sodium-cooled Fast Reactors (SFRs) has shown the importance of material tribological properties. Where galling or adhesive wear cannot be allowed, hardfacing alloys, known to be galling-resistant coatings, are usually applied on rubbing surfaces. The most used coating is the cobalt-base alloy named Stellite 6® because of its outstanding friction and wear behaviour. Nevertheless, cobalt is an element which activates in the reactor leading to complex management of safety during reactor maintenance and mainly decommissioning. As a consequence, a collaborative work between CEA, EDF and FRAMATOME has been launched for selecting promising cobalt-free hardfacing alloys for the 600 MWe Sodium-cooled Fast breeder reactor project named ASTRID. Several nickel-base alloys have been selected from literature review then deposited by Plasma Transferred Arc or Laser Cladding on 17Cr austenitic stainless steel 316L(N) according to RCC-MRx Code (AFCEN Code). Among the numerous properties required for qualifying their use as hardfacing alloys in SFR, good corrosion behaviour and good friction and wear behaviour in sodium are essential. First results on these properties are shown in this article. Firstly, the corrosion behaviour of all coatings was evaluated through exposure tests in purified sodium for 5000 h at 400 °C. All coatings showed an acceptable corrosion behaviour in sodium. Finally, the friction and wear properties of one alloy candidate, NiCrBSi alloy, were studied in sodium in a dedicated designed facility. The influence of the oxygen concentration in sodium on the friction and wear properties was evaluated.

Experimental Study of Condensation Heat Transfer in Helical Coil Heat Exchanger

Helical coils are used as one of the heat transfer enhancement techniques in many engineering applications such as in nuclear industry, chemical and petrochemical industries, HVAC systems, and many other engineering applications. In nuclear industry, helical coils are used as a method for exchanging heat in large sodium-cooled fast breeder reactor. In comparison with straight-tube heat exchangers, rate of heat transfer for helical coil heat exchanger is significantly high. This is due to secondary flow pattern in planes normal to the main flow direction of fluid in the helical coiled tube. The present work is an experimental investigation on heat transfer characteristics during steam condensing inside a small diameter vertical helical coil tube with the cooling water flowing in the shell in counter flow direction. The effect of mass flow rate and saturation temperature of steam on the heat transfer coefficient is investigated and plotted. The results confirm the significant improvement in the heat transfer coefficient for condensation of steam.

CFD analysis and experimental validation of steady state mixed convection sodium flow

Formation and destruction of thermal stratification can occur under certain flow conditions in the upper plenum of sodium cooled fast breeder reactors (SFR). The flow patterns in the hot sodium pool of SFR upper plenum are very complex and include zones of free and wall-bounded jets, recirculation and stagnation areas. The interaction of sodium flow and thermal stratification has been analyzed experimentally at CEA in the SUPERCAVNA facility. The facility consists of a rectangular cavity with heated side walls. The flow is driven by a wall-bounded cold jet at the bottom of the cavity. Experimental data of the temperature distribution inside the cavity are available for steady-state and transient flow conditions. In the present study, the steady state experiments are analyzed with the CEA CFD reference code TrioCFD and the commercial code FLUENT employing Reynolds Averaged Navier-Stokes (RANS) equations. The SUPERCAVNA modeling is based on preliminary separate effects’ studies of wall-bounded jets and free convection benchmark simulations for low Prandtl number fluids.It was found that a two-dimensional treatment is sufficient to reproduce correctly the measured thermal stratification for steady-state SUPERCAVNA experiments. However, it is necessary to take into account conjugated heat transfer between cavity sodium flow and side-wall heating channel. Using temperature-dependent physical properties was also found to be an important factor in simulating the experiments correctly. Applying the Boussinesq approximation to study the impact of buoyancy on the vertical flow momentum was found to be justified. Turbulence modelling is necessary for a successful simulation of the experiments. For practical reasons, the analysis was restricted to turbulence models of the k-ε family. High Re k-ε models, in either standard, realizable or RNG formulations, do not lead to significant differences in the calculated temperature fields. Non-linear eddy viscosity modelling might improve the quality of the simulation.

APOLLO3 homogenization techniques for transport core calculations—application to the ASTRID CFV core

Abstract This paper presents a comparison of homogenization techniques implemented in the APOLLO3 platform for transport core calculations: standard scalar flux weighting and new flux–moment homogenization, in different combinations with (or without) leakage models. Besides the historical B1 - homogeneous model, a new B-heterogeneous one has indeed been implemented recently in the two/three-dimensional-transport solver using the method of characteristics. First analyses have been performed on a very simple Sodium Fast Reactor core with a regular hexagonal lattice. They show that using the heterogeneous leakage model in association with flux-moment homogenization strongly improves the prediction of k eff and void reactivity effects. These good results are confirmed when the application is done to the fissile assemblies of the more complex CFV (Low Void Effect) core of the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) project of sodium-cooled fast breeder reactor (Generation IV).

Experiments in LEENA facility with modified wire type leak detector layout in large sodium pipelines

Sodium cooled Fast Breeder Reactors (SFRs) are envisaged in the second phase of Indian nuclear power programme. Liquid sodium is used as the coolant in the SFRs due to its favourable nuclear properties and excellent heat transfer properties. Leaks in sodium systems have the potential of being exceptionally hazardous due to the reaction of liquid sodium with oxygen and water vapour in the air. When a sodium leak occurs, the sodium leak rate, the total quantity of sodium leaked and leak detector layout governs the detection time. Other factors to be considered are insulation material packing condition, distance between the leak point and detector, heater layout, pipe geometry, temperature etc. Potential regions of leakage in Fast Breeder Reactor (FBR) sodium circuits are near welds, high stress areas and regions subjected to thermal striping. Early detection of leak is needed for minimizing the quantity of sodium leaked to outside and consequent damages. Three wire type leak detectors (WLDs positioned at 90°, 180° and 270°) working on conductivity principle are used for detecting sodium leak in the large horizontal secondary sodium pipelines of Prototype Fast Breeder Reactor (PFBR). It was found from the upper boundary curve based on LEENA (LEak Experiments in NAtrium) facility experimental results with present PFBR leak detector layout, that a leak rate of 200g/h and above can be detected within 6h. To improve the detection capability, modified leak detector layout was proposed for FBR 800 NB size horizontal secondary circuit pipelines of FBR with seven WLDs positioned at 45°, 90°, 135°, 180°, 225°, 270° and 315° by taking top of pipeline as 0°. It is required to assess the performance of wire type leak detectors along with its layout on large horizontal sodium pipelines and confirm that they are meeting the requirements. LEENA facility was re-commissioned and five sodium leak experiments with modified leak detector layout on large horizontal pipeline were conducted with different sodium leak rates. From the experimental data it was observed that there was considerable reduction in detection time with modified WLD layout. This is due to the reduction in distance between leak point and leak detector (Reduced from 0.63m to 0.31m). From the experiments it was found that a leak rate of 200g/h can be detected in one hour.

On the interdiffusion in multilayered silicide coatings for the vanadium-based alloy V-4Cr-4Ti

To provide protection against corrosion at high temperatures, silicide diffusion coatings were developed for the V-4Cr-4Ti alloy, which can be used as the fuel cladding in next-generation sodium-cooled fast breeder reactors. The multilayered coatings were prepared by halide-activated pack cementation using MgF2 as the transport agent and pure silicon (high activity) as the master alloy. Coated pure vanadium and coated V-4Cr-4Ti alloy were studied and compared as substrates. In both cases, the growth of the silicide layers (V3Si, V5Si3, V6Si5 and VSi2) was controlled exclusively by solid-state diffusion, and the growth kinetics followed a parabolic law. Wagner's analysis was adopted to calculate the integrated diffusion coefficients for all silicides. The estimated values of the integrated diffusion coefficients range from approximately 10−9 to 10−13 cm2 s−1. Then, a diffusion-based numerical approach was used to evaluate the growth and consumption of the layers when the coated substrates were exposed at critical temperatures. The estimated lifetimes of the upper VSi2 layer were 400 h and 280 h for pure vanadium and the V-4Cr-4Ti alloy, respectively. The result from the numeric simulation was in good agreement with the layer thicknesses measured after aging the coated samples at 1150 °C under vacuum.

Determination of in-service inspection requirements for fast reactor components using System Based Code concept

In our previous study (Takaya et al., 2015a), we proposed a new process for determining in-service inspection (ISI) requirements using the System Based Code concept. The proposed process consists of two complementary evaluations, one focusing on structural integrity and the other on plant safety. First, the structural reliability of a specified component is evaluated considering potentially active degradation mechanisms, including those that are not explicitly addressed in the design codes. If the structural reliability meets the requirement, the second evaluation can be conducted, which assesses the detectability of defects before they can grow to an unacceptable size, taking plant safety into account. If there is any feasible way to detect defects, it is adopted as an ISI requirement. Otherwise, a structural integrity evaluation would be required under a sufficiently conservative hypothesis. In other words, if the additional requirements are met, detectability is not an obligation.In this study, the ISI requirements for a reactor guard vessel (RGV) and core support structure (CSS) of a prototype sodium-cooled fast breeder reactor in Japan (Monju) were investigated using the proposed process. Creep-fatigue and fatigue were chosen as the potentially degradation mechanisms for the RGV and CSS, respectively. The Stage I evaluations using the Monte-Carlo method showed that both components had sufficient reliability if these degradation mechanisms were considered. At Stage II, the reliability levels of the components were evaluated assuming initial fully circumferential cracks with a depth equal to 10% of the thickness as additional requirements because there was no available inspection method for the components. It was shown that both components had sufficient reliability even with the additional requirement based on conservative hypothesis. The failure occurrences of these components were practically eliminated. Hence, it was concluded that no ISI requirements were needed for these components.The proposed process is expected to contribute to the realization of effective and rational ISI by properly taking into account plant-specific features.

Ultrasonic high-frequency guided waves for the inspection of shell weld in core support structure of fast breeder reactors

This article reports an ultrasonic high-frequency guided wave based methodology for inspection of shell weld of core support structure of the main vessel in 500-MWe sodium-cooled prototype fast breeder reactor. The high-frequency guided waves are generated in 30-mm-thick plate of core support structure at 2 MHz (frequency–thickness product of ∼60 MHz-mm). These wave modes are fairly nondispersive over considerable distance of propagation through the material. In order to understand the generation and propagation characteristics of these high-frequency guided waves, a series of two-dimensional plane strain finite element models are developed using explicit scheme. The optimum transducer placement for generating efficient high-frequency guided wave is achieved by sweeping the transducer position along the curvature of the main vessel plate with respect to the centerline of the vertical shell support plate. The generated high-frequency guided waves are analyzed and identified with the help of dispersion curves and time–frequency analysis. The simulation results are found to capture the features observed in the experimental measurements rather well, and the time of flight of the reflected waves in the simulation is found to be in excellent agreement with the experimental results. The optimum transducer locations for inspection of a general T-joint configuration with different angles between the web and the base have been studied in detail using finite element simulations. With optimum transducer position, the high-frequency guided waves are also used to inspect the artificially introduced defects in the connecting weld of shell support plate. It has been demonstrated that the high-frequency guided wave could detect defects of 20% wall thickness (6 mm) in the connecting shell weld.