Reactor Structural Materials Research Articles

A Study on Corrosion Product Behavior Prediction for Domestic PWR Primary System by using CRUDTRAN

Radionuclide deposited on the surface of several internal and external systems in a nuclear power plant is created by the activation of corrosion products from nuclear reactor structural materials and fission products. Especially, the constant contact between water and the surface corrodes the inside where primary system makes coolants and corrosion products mixed. Also, these are circulated along the systems. For comparing models, CRUDTRAN, DISER, MIGA-RT and CPAIR codes are analyzed to predict the quantity of radionuclide and corrosion product of primary reactor that are used at the stage of designing. The corrosion products behavior of domestic PWR primary system was predicted by using CRUDTRAN. This study aims to increase the reliability of corrosion product evaluation model by comparing the actual values and calculated values with the data of a Westing House-type Nuclear Power Plant.

Defect microstructural equivalence in molybdenum under different irradiation conditions at low temperatures and low doses

Owing to its advantages such as higher dose rates and shorter time required, ion irradiation, including in-situ ion irradiation of nano-foils on electron microscopes, is being proposed as a surrogate for neutron/reactor irradiation in materials testing for the research of irradiation induced/enhanced degradation of reactor structural materials. This raises a key question: can the microstructural evolution be matched across different irradiation conditions? Here we report our recent study using a rate-theory based cluster dynamics approach on the time dependent defect accumulation in elemental molybdenum under varying ion and neutron irradiation conditions, within a low temperature and low dose regime. The approach solves a large system of coupled differential equations each equation describing the rate of change in the concentration of a particular defect/cluster (defined by its nature and number of constituent point defects) caused by direct production (for primary defects), diffusion flux across spatial grids (for mobile defects, under ion irradiation), and capturing and/or emission interactions with other defects/clusters. Our results suggest that defect evolution under different dose rates can be very well matched in either thin foil ion irradiation or bulk neutron irradiation, by using different irradiation temperatures to counteract the effect of the dose rate variation. Across the two types of irradiation, less perfect matching is obtained mainly due to the strong surface sink effect in the thin foil, but the matching can be improved by combining the temperature shift strategy with selection of a single depth in the foil (as opposed to all depths averaged) for matching, or a greater foil thickness.

Computational and experimental investigations of magnetic domain structures in patterned magnetic thin films

The use of nondestructive magnetic signatures for continuous monitoring of the degradation of structural materials in nuclear reactors is a promising yet challenging application for advanced functional materials behavior modeling and measurement. In this work, a numerical model, which is based on the Landau–Lifshitz–Gilbert equation of magnetization dynamics and the phase field approach, was developed to study the impact of defects such as nonmagnetic precipitates and/or voids, free surfaces and crystal orientation on magnetic domain structures and magnetic responses in magnetic materials, with the goal of exploring the correlation between microstructures and magnetic signatures. To validate the model, single crystal iron thin films (~240 nm thickness) were grown on MgO substrates and a focused ion beam was used to pattern micrometer-scale specimens with different geometries. Magnetic force microscopy (MFM) was used to measure magnetic domain structure and its field-dependence. Numerical simulations were constructed with the same geometry as the patterned specimens and under similar applied magnetic field conditions as tested by MFM. The results from simulations and experiments show that 1) magnetic domain structures strongly depend on the film geometry and the external applied field and 2) the predicted magnetic domain structures from the simulations agree quantitatively with those measured by MFM. The results demonstrate the capability of the developed model, used together with key experiments, for improving the understanding of the signal physics in magnetic sensing, thereby providing guidance to the development of advanced nondestructive magnetic techniques.

High-Temperature Electric and Thermal Conductivity of the Aluminum Alloys SAV-1 and AMG-2

The electric and thermal conductivity of the aluminum alloys AMG-2 and SAV-1, which are used to fabricate the fuel elements of the fuel assemblies of a nuclear reactor, was investigated in the range 290‐490 K. The Debye temperature was calculated and the contribution of the electronic and lattice components to the thermal conductivity of the alloys was determined. It is shown on the basis of an analysis of the mechanisms of electron and phonon scattering and the experimental data that one possible reason for the observed changes of the thermal conductivity in aluminum alloys are the scattering of electrons by phonons and the deviation of the Lorenz number from the theoretical values as a result of inelastic scattering of the electrons. Aluminum alloys distinguished by their physical properties, viz., strength, high plasticity, corrosion resistance, and large store of plasticity, are widely used as a structural material in nuclear reactors of different types. AMG-2 and SAV-1 belong to the group of deformable aluminum alloys in the ternary system Al‐Mg‐Si. Their physical properties depend largely on the ratios of the concentrations of aluminum and the main alloying elements ‐ magnesium and silicon. In addition, different intermellide compounds, for example, AlSiMnFe, AlSiFe, CuAl 2 , and others, can be present in such alloys [1]. The high thermal conductivity and weak susceptibility to radiation damage make it possible to operate a structure based on them in the presence of significant thermal loads, specifically, fabricate protective cladding of fuel elements and different structures in the reactor core that operate in a wide range of radiation fields [2]. All this generates interest in studying these materials. In [3], low-angle neutron scattering was used to study the superatomic structure and it was shown that pores with characteristic radius 5‐50 nm are present in samples of the alloy SAV-1. It was found that irradiation causes an appreciable reduction of the volume fraction of the large scattering structure with radius 40‐50 nm and together with this an increase of the relative amount of the fractions with radius less than 20 nm. The results of neutron studies are correlated with mechanical tests of irradiated alloys and changes in their elemental composition [4]. At the same time the thermal and electrical characteristics of the alloys have not been adequately studied. The thermal conductivity of the alloy SV-1 and its temperature dependence were studied in [5]. However, because there are no data on the electric conductivity the conclusion concerning the relative contribution of possible mechanisms of heat transfer in the experimental temperature range was an estimate. In this connection, in the present work the temperature dependence of the electrical conductivity and the thermal conductivity of the related alloys SAV-1 and AMG-2 were studied in order to obtain a more accurate description of the physical properties of aluminum alloys.

An experience with in-service fabrication and inspection of austenitic stainless steel piping in high temperature sodium system

A creep testing facility along with dynamic sodium loop was installed at Indira Gandhi Centre for Atomic Research, Kalpakkam, India to assess the creep behavior of fast reactor structural materials in flowing sodium. Type 304L austenitic stainless steel was used in the low cross section piping of hot-leg whereas 316LN austenitic stainless steel in the high cross section cold-leg of the sodium loop. The intended service life of the sodium loop was 10 years. The loop has performed successfully in the stipulated time period. To enhance its life time, it has been decided to replace the 304L piping with 316L piping in the hot-leg. There were more than 300 welding joints involved in the integration of cold-leg with the new 316L hot-leg. Continuous argon gas flow was maintained in the loop during welding to avoid contamination of sodium residue with air. Several innovative welding procedures have been adopted for joining the new hot-leg with the existing cold-leg in the presence of sodium residue adopting TIG welding technique. The joints were inspected for 100% X-ray radiography and qualified by performing tensile tests. The components used in the discarded hot-leg were retrieved, cleaned and integrated in the renovated loop. A method of cleaning component of sodium residue has been established. This paper highlights the in-service fabrication and inspection of the renovation.

A way to limit the corrosion in the Molten Salt Reactor concept: the salt redox potential control

The possibility of controlling the salt redox potential thanks to a redox buffer in the Molten Salt Fast Reactor was investigated, the goal was to limit the oxidation of the reactor structural material. Tests were performed in LiF-CaF2 at 850°C on two different redox couples to fix the salt potential, Eu(III)/Eu(II) and U(IV)/U(III), where the first one was used as inactive system to validate the methodology to be applied on the uranium system. A metallic reducing agent (Gd plate for Eu, and U plate for U system) was inserted in the salt, leading to a spontaneous reaction: Eu(III) and U(IV) were then reduced. Eu(III) was fully converted into Eu(II) with metallic Gd, validating the approach. On the U system, the U(IV)/U(III) ratio has to be set between 10 and 100 to limit the core material oxidation: addition of metallic U decreased the concentration ratio from the infinite to 1, showing the feasibility of the salt redox potential control with the U system.

Phase Transformation Behavior and Microstructural Control of High-Cr Martensitic/Ferritic Heat-resistant Steels for Power and Nuclear Plants: A Review

The martensitic/ferritic steels have been used as boiler and turbine materials in power plants, and also been selected as potential materials for structural materials in nuclear reactors. In this paper, the kinetic analysis of the martensite formation and microstructural control of high-Cr martensitic/ferritic steels are reviewed. A modular approach, incorporating Fisher partitioning nucleation and anisotropic growth for impingement, was proposed to describe the martensite formation kinetics under different cooling rates. The kinetic analysis suggested a thermal-activated growth feature occurring during the martensitic transformation of martensitic steels. The microstructure can be tuned by composition optimization and various combinations of heat treatment parameters (temperature, time, severe and minor deformation). For the application in power plant, the potential of boundary-design, refinement of original austenite grain size and the final martensitic lath, pinning effect of stable carbides, in improving the performances of martensitic/ferritic steels at elevated temperatures should be investigated more thoroughly. Furthermore, efforts should be made to explore the effects of retained austenite on the improvement of high-temperature creep strength. For the application of nuclear plants, attempts should also be made to produce Fe powders with uniformly distributed oxide particles by chemical reactions.

The Focused Ion Beam-SEM as a Critical Tool For Nano-scale Characterization of Highly Radioactive Nuclear Fuels

There are unique challenges associated with conducting nanometer and atomic level materials characterization on highly irradiated and/or radiologically contaminated nuclear reactor fuels and structural materials, where contact handling is not feasible with the high levels of ionizing radiation emitted. Historically, particularly for irradiated nuclear fuel, samples could only be prepared remotely in heavily shielded facilities that precluded the fine contact work needed to prepare usable, let alone high quality, transmission electron microscopy thin foils and atom probe tomography (APT) tips. Thus, most published data on irradiated fuels detail micron level characterization (optical and SEM) and little information is available on nanometer and atomic level structure and chemistry.

Thermal Diffusivity of Reduced Activation Ferritic/Martensitic Steel Determined by the Time Domain Photoacoustic Piezoelectric Technique

The thermal diffusivity of reduced activation ferritic/martensitic steel (CLF-1), which is recognized as the primary candidate structural material for the test blanket module of the international thermal-nuclear experimental reactor, has been studied by the time-domain (TD) photoacoustic piezoelectric (PAPE) technique. The TD PAPE model based on a simplified thermoelastic theory under square-wave modulated laser excitation is presented, relating the TD PAPE signal to the modulation frequency, thermal diffusivity, and other material parameters. Thermal diffusivities of reference samples such as copper and nickel were measured and analyzed, by which the validity of the technique is verified. The thermal diffusivity of the CLF-1 sample was measured to be \(8.2\,\text {mm}^{2}{\cdot }\text {s}^{-1}\), which is at a medium level among the ordinary steel materials (\(3\,\text {mm}^{2}{\cdot }\text {s}^{-1}\) to \(14\,\text {mm}^{2}{\cdot }\text {s}^{-1})\) and has decent heat-dissipation ability. The results show that the TD PAPE technique can provide a fast and economic way for the investigation of the thermophysical properties of fusion reactor structural materials.

Structure of Carbon-Coated Graphite and its Application in Nuclear Industry

High Temperature Gas-cooled Reactor is one of the candidates for the next generation advanced nuclear energy systems. As high temperature gas cooled reactor neutron moderator, reflector and structural material, nuclear graphite is very important in the nuclear industry to assure the safe operation of the reactor. The oxide reactor of nuclear graphite will change its internal pore structure and properties, such as mechanics properties, thermodynamic performance, radiant resistance, and so on. In this study, graphite coated with carbon was prepared by liquid impregnation method. Infrared spectral radiometry (FTIR), scanning electron microscope (SEM) and thermal gravity analysis (TGA) methods were used to characterize the structure of the carbon-coated graphite. Neutron absorption test method was also used to test the resistance to radiation. The results show that when the heat treatment time was 6h with the heat treatment temperature was 500¡æ, the material had the best morphology and had the largest neutron adsorption area (4.63 mb). The results suggest that this carbon-coated graphite has very good application prospect in nuclear industry.

Interaction of carbon–vacancy complex with minor alloying elements of ferritic steels

Interstitial carbon, dissolved in bcc matrix of ferritic steels, plays an important role in the evolution of radiation-induced microstructure since it exhibits strong interaction with vacancies. Frequent formation and break-up of carbon–vacancy pairs, occurring in the course of irradiation, affect both kinetics of the accumulation of point defect clusters and carbon spatial distribution. The interaction of typical alloying elements (Mn, Ni, Cu, Si, Cr and P) in ferritic steels used as structural materials in nuclear reactors with a carbon–vacancy complex is analyzed using ab initio techniques. It is found that all the considered solutes form stable triple clusters resulting in the increase of the total binding energy by 0.2–0.3eV. As a result of the formation of energetically favourable solute–carbon–vacancy triplets, the dissociation energy for vacancy/carbon emission is also increased by ∼0.2–0.3eV, suggesting that the solutes enhance thermal stability of carbon–vacancy complex. Association of carbon–vacancy pairs with multiple solute clusters is found to be favorable for Ni, Cu and P. The energetic stability of solute(s)–carbon–vacancy complexes was rationalized on the basis of pairwise interaction data and by analyzing the variation of local magnetic moments on atoms constituting the clusters.

Nuclear data for neutron calculations of the reactors

Nuclear data on nuclear reaction cross sections with gammaquanta escape in the interaction of fast neutrons with the reactor structural materials are important for neutron-physical calculations of reactors. These data are necessary to estimate the gamma fields in the reactor core and calculate radiation protection. In addition, the study of nuclear reaction cross sections are needed to study different mechanisms of nuclear reactions and features of the excited states of nuclei and their decay.The results of studying nuclear data on nuclear reaction cross sections on cadmium in the energy range from 2 to 8 MeV are given in the paper. The values of cross sections are recovered from the corresponding amplitude instrument spectra, and their errors are estimated. The experimental results are compared with the results of theoretical calculations, performed within the framework of the statistical Hauser-Feshbach model, taking into account particle preequilibrium escape.The obtained results are included in the international experimental nuclear data library and can be used for neutron-physical calculations of reactors. The investigated theoretical models can be used for estimating nuclear data, which are missing in the international experimental nuclear data libraries.

Reconstruction of distribution functions of fast ions and runaway electrons in fusion plasmas using gamma-ray spectrometry with applications to ITER

Gamma-ray spectrometry on ITER can provide information both on confined fusion alpha particles for optimization of plasma heating and runaway electrons, which is important for safe reactor operations. For the purpose of deconvolution of gamma-ray spectra recorded in fusion plasma experiments the DeGaSum code has been developed. The code can be applied for processing of both spectra of monoenergetic gamma rays, which are born in nuclear reactions produced by alpha particles and other fast ions, and continuous bremsstrahlung spectra generated by runaway electrons in the MeV range in the plasma and reactor structure materials. Gamma-ray spectrometer response functions and bremsstrahlung spectra generated by electrons in the MeV energy range are calculated and used in the DeGaSum code. The deconvolution of the discrete spectra allows the identification of nuclear reactions, which give rise to gamma rays, and the calculation of their intensities. By applying the code for continuous hard x-ray spectra, the runaway electron energy distribution can be inferred. It can provide the maximal energy of runaway electrons with accuracy, which satisfies the ITER project requirements. The code has been used for processing of spectra recorded in JET experiments. An application of the deconvolution technique for gamma-ray emission measurements on ITER is discussed.

Radiation Resistance of Structural Materials of Nuclear Reactors on Irradiation with High-Energy Hydrogen and Helium Ions

Basic principles of determination of the radiation resistance of structural materials of nuclear reactors with implantation of high-energy hydrogen and helium atoms have been presented. The parameters of the process of implantation of light irons have been calculated. By scanning-electron-microscopy, optical-microscopy, and interference methods, the authors have studied the surface structure of samples of steel-3, stainless steel, and D16 alloy immediately after irradiating them with hydrogen and helium atoms with an energy of 200 to 400 keV in the range of doses from 1016 to 3 · 1017 ions/cm2 and after annealing these samples thermally at temperatures from 300 to 550°C. Threshold blistering doses for all the studied materials and annealing temperatures for visualizing structural defects have been determined.

Wetting Angle Measurement of Liquid PbBi on RAFM Steel Surface Based on Ellipse Fitting

The fusion power is considered an ideal energy for the future due to abundant, green and safety. Wetting properties between the liquid metal and the structural materials in the liquid blanket of the fusion reactor related to the heat exchange efficiency and the irradiated plasma resistance, and directly related to the service life of the reactor structural materials. This article describes the wetting angle measurement experimental procedure of liquid lead-bismuth (PbBi) at reduced activation ferritic/martensitic steel (RAFM steel) surface based on the high-temperature static sessile drop method under the protection of high purity argon, including the images enhancement processing obtained by CCD camera, using ellipse fitting method to calculate the wetting angle, and were measured at 350-600°C. The experimental results show that the wetting angles of liquid PbBi and RAFM steel are greater than 140°, non-wetting.

An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR).This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation.The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution.Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions.The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

An examination of the use of the Modified Jogged-Screw model for predicting creep behavior in Zircaloy-4

The alloy Zircaloy-4 is used primarily as a structural material in nuclear reactors. A thorough understanding of the thermal creep response of such materials is essential to enable accurate forecasting of material behavior during both service and long-term storage. To this end, the Modified Jogged-Screw model, originally developed to predict steady-state creep rates in γ-TiAl, has been applied to Zircaloy-4. This work provides a more complete database for model validation through creep testing of several Zircaloy-4 specimens over a range of stress and temperature conditions. Bright-field scanning transmission electron microscopy was used to observe the substructure and quantify the model parameters important to the Modified Jogged-Screw model. These observations were used to confirm the applicability of the proposed model and assess the sensitivity of the model parameters to test conditions. The increased availability of statistics for model parameters as a result of substructural observation will elucidate the strengths and weaknesses of the model as previously proposed.

Grain boundary engineering for structure materials of nuclear reactors

Grain boundary engineering (GBE), primarily implemented by thermomechanical processing, is an effective and economical method of enhancing the properties of polycrystalline materials. Among the factors affecting grain boundary character distribution, literature data showed definitive effect of grain size and texture. GBE is more effective for austenitic stainless steels and Ni-base alloys compared to other structural materials of nuclear reactors, such as refractory metals, ferritic and ferritic–martensitic steels, and Zr alloys. GBE has shown beneficial effects on improving the strength, creep strength, and resistance to stress corrosion cracking and oxidation of austenitic stainless steels and Ni-base alloys.

Current understanding of radiation-induced degradation in light water reactor structural materials

Current phenomenological knowledge and understanding of mechanisms are reviewed for radiation embrittlement of reactor pressure vessel low alloy steels and irradiation assisted stress corrosion cracking of core internals of stainless steels. Accumulated test data of irradiated materials in light water reactors and microscopic analyses by using state-of-the-art techniques such as a three-dimensional atom probe and electron backscatter diffraction have significantly increased knowledge about microstructural features. Characteristics of solute clusters and deformation microstructures and their contributions to macroscopic material property changes have been clarified to a large extent, which provide keys to understand in the degradation mechanisms. However, there are still fundamental research issues that merit study for long-term operation of reactors that requires reliable quantitative prediction of radiation-induced degradation of component materials in low-dose rate high-dose conditions.

HETEROGENEOUS MICROSTRUCTURE AND TEXTURE EVOLUTION DURING FABRICATION OF Zr--Sn--Nb ZIRCONIUM ALLOY SHEETS

Zirconium alloys are widely used as fuel cladding and structural materials for nuclear reactors due to the low neutron absorption cross-section, good corrosion resistance and acceptable mechanical properties. These properties are greatly dependent on microstructural and textural features, such as grain morphology, grain size, crystallographic texture and distribution of precipitates. It is necessary to understand microstructure and texture evolution during fabrication in order to optimize the manufacturing process and to improve the service performance. In this work, microstructure and texture evolution during fabrication of Zr-Sn-Nb new zirconium alloy sheets are investigated using XRD, SEM-ECC, TEM and EBSD. The results show that the random texture formed by β quenching transforms into tilt basal texture after hot rolling. The basal texture keeps stable during the following