Reactor Core Materials Research Articles

High ductility – Good strength stainless steel alloys at the elevated temperature for nuclear reactors core

Reactors materials are the backbone of reactor life extending and making it safer. Accordingly, unabated research continues for the improvement of nuclear reactor materials, especially the reactor core materials. Hence, in the present article, two developed tungsten/molybdenum stainless steel grades MoW1 (0.843 wt% of Mo and 1.801 wt% of W) and MoW2 (0.251 wt% of Mo and 2.433 wt% of W) were produced. AISI316 standard stainless steel was used as a base material. Microstructural features of the produced alloys were investigated using Schaeffler diagram, optical microscopy, SEM-EDS, and X-ray diffraction. The mechanical properties of the produced alloys were studied at room temperature RT and elevated temperatures of 300 and 500 °C. W addition by 1.801 wt% at the expense of Mo (MoW1 steel) strengthened yield strength YS and ultimate tensile strength UTS by 42.344% and 32.022% at room temperature compared to the standard stainless steel AISI316, while a high affinity was found between YS and UTS results for steel containing 2.433 wt% of W (MoW2) and the AISI316. On the other hand, the elongation El% of MoW1 decreased by 20.861% and MoW2 by 17.219% compared to the AISI316. At 300 and 500 °C, the MoW1 stainless steel alloy has higher YS (by 4.865% and 7.865%) and UTS (by 17.580% and 41.077%) than AISI316. While, MoW2 has slightly lower YS (by 7.568% and 3.371%) and higher UTS (by 10.046% and 36.261%) than AISI316. Both of MoW1 and MoW2 has much higher El% than AISI316 at both 300 and 500 °C. As a comparison between the developed stainless steel alloys MoW1 and MoW2 for both 300 and 500 °C, MoW1 gave better YS (by 13.450% and 11.628%), UTS (by 6.8465% and 3.534%), and El% (18.3024% and 18.037%) results than MoW2. Accordingly, depending on the distinctive mechanical properties of the developed MoW1 stainless steel alloy at 300 and 500 °C, it is a strong candidate as an appropriate material for high-temperature conditions of the nuclear reactors core.

“A concise study of radiation-induced lattice strain in fast reactor first-wall materials through coupled computational modeling and experiments”

Radiation-induced defects can alter the crystal structure of materials consequently inducing lattice strain, which can be quantified by examining the full width at half-maximum (FWHM) values of X-ray diffraction (XRD) peaks. In this study, an experimental sample of the alpha manganese was simulated first principally using the LCAO basis set and SGGA-PBE XC functionals. The optimized simulation results were found to be close to the experimental results, with an underestimation error of ∼2.21 %. The same simulation model was then extended to investigate the FWHM variation of XRD peaks as a function of increasing isotropic strain, which is one of the consequences of radiation-induced defects. The study found that the FWHM values increased with increasing isotropic strain, indicating the presence of radiation-induced defects, and resulting lattice strain. To validate the simulation results, we further used the model to simulate high-resolution transmission electron microscopy (HRTEM) diffraction profiles, which were experimentally matched with alpha manganese FFTs identified using HRTEM. The study presents a comprehensive approach to understanding the effects of radiation damage on materials, particularly the unique properties of the alpha manganese structure as a function of isotropic strain. The findings can help develop better models for simulating radiation-induced defects and their effects on materials, which could have significant implications for various technological applications, including developing nuclear reactor core materials.

MODELING OF THE CORIUM AND METALS – COOLERS INTERACTION IN A CORE CATCHER OF A LIGHT WATER REACTOR

The core catcher is one of the mandatory elements of the reactor safety system, which prevents the release of reactor core materials in a severe accident. The core catcher is steel vessel filled with sacrificial materials (SM) and forming a tank where a corium melt coming from the core is formed. The trap is a steel body filled with sacrificial materials (LM) and forming a vessel where a corium bath is formed coming from the core. The melt formed in the core catcher is cooled by heat removal to the cooling water through the shell of the steel vessel, as well as by water supplied directly to the surface of the melt after the dissolution process of the SM in corium (gravitational inversion). The delay in the water supply to the melt is associated with the features of the component structure of corium and its interaction with water (the formation of explosive hydrogen and the possibility of its detonation, as well as the threat of a steam explosion). However, a certain amount of time is spent on the implementation of gravitational inversion, and it is desirable to start the water supply to the melt immediately at the moment when the corium enters the core catcher due to the danger of the system going beyond the permissible limits (the beginning of boiling of uranium dioxide) due to decay heat in the corium. In this regard, the authors have an idea – to use a fusible metal for additional cooling of the surface of the corium in order to organize heat removal and reduce the temperature of the corium in the period before the end of the gravitational inversion process. The article presents the results of modeling the interaction of corium with candidate low-melting metals – coolers. The modeling was conducted using the ANSYS software package. As a result of the conducted work, the time for which each of the considered cooling metals will reach the points of phase transitions of melting and boiling is determined. The analysis of the results allowed us to draw appropriate conclusions about the possible practical implementation of the proposed method of cooling corium.

Strain Localisation and Fracture of Nuclear Reactor Core Materials

The production of prismatic dislocation loops in nuclear reactor core materials results in hardening because the loops impede dislocation motion. Yielding often occurs by a localised clearing of the loops through interactions with gliding dislocations called channeling. The cleared channels represent a softer material within which most of the subsequent deformation is localized. Channeling is often associated with hypothetical dislocation pileup and intergranular cracking in reactor components although the channels themselves do not amplify stress as one would expect from a pileup. The channels are often similar in appearance to twins leading to the possibility that twins are sometimes mistakenly identified as channels. Neither twins nor dislocation channels, which are bulk shears, produce the same stress conditions as a pileup on a single plane. At high doses, when cavities are produced (either He-stabilised bubbles at low temperatures or voids at high temperatures), there can be reduced ductility because the material is already in an equivalent advanced stage of microscopic necking. He-stabilised cavities form preferentially on grain boundaries and at precipitate or incoherent twin/ε-martensite interfaces. The higher planar density of the cavities, coupled with the incompatibility at the interface, results in a preferential failure known as He embrittlement. Strain localisation and inter- or intragranular failure are dependent on many factors that are ultimately microstructural in nature. The mechanisms are described and discussed in relation to reactor core materials.

Microstructural Effects on Irradiation Creep of Reactor Core Materials

The processes that control irradiation creep are dependent on the temperature and the rate of production of freely migrating point defects, affecting both the microstructure and the mechanisms of mass transport. Because of the experimental difficulties in studying irradiation creep, many different hypothetical models have been developed that either favour a dislocation slip or a mass transport mechanism. Irradiation creep mechanisms and models that are dependent on the microstructure, which are either fully or partially mechanistic in nature, are described and discussed in terms of their ability to account for the in-reactor creep behaviour of various nuclear reactor core materials. A rate theory model for creep of Zr-2.5Nb pressure tubing in CANDU reactors incorporating the as-fabricated microstructure has been developed that gives good agreement with measurements for tubes manufactured by different fabrication routes having very different microstructures. One can therefore conclude that for Zr-alloys at temperatures < 300 °C and stresses < 150 MPa, diffusional mass transport is the dominant creep mechanism. The most important microstructural parameter controlling irradiation creep for these conditions is the grain structure. Austenitic alloys follow similar microstructural dependencies as Zr-alloys, but up to higher temperature and stress ranges. The exception is that dislocation slip is dominant in austenitic alloys at temperatures < 100 °C because there are few barriers to dislocation slip at these low temperatures, which is linked to the enhanced recombination of irradiation-induced point defects.

Comparative analysis of film boiling correlations for steam explosion problem

The article considers the existing models and correlations for describing heat transfer during film boiling. Film boiling is an important thermophysical process that determines the course of interaction of the melt of reactor core materials with the coolant, which can potentially occur during a severe accident at a nuclear power plant with a pressurized water-cooled reactor (VVER/PWR). The values of the Nusselt number predicted by these models are compared. A fairly significant dispersion of calculated parameters has been established, making it difficult to unambiguously choose one or another correlation. However, in the range of parameters typical for the interaction of a high-temperature melt of reactor core materials with a coolant, several correlations give quite close values, which allows them to be recommended for use in calculation codes for modeling thermal-hydraulic processes during a severe accident at a nuclear power plant with VVER/PWR type.

Vibration Reduction of Reactor Using Global Magneto-Structural Topology Optimization

The global topology optimization algorithms used in vibration reduction of gapped-iron reactors will achieve the optimal performance under certain constraints, but with a long computational time and poor convergence. To shorten the calculation time, a global magneto-structural topology optimization algorithm (GMS) is proposed, in which the magnetic topology optimization result is used as the initial seed of the structural topology optimization. Firstly, the air gaps areas are set as the optimized area due to the large vibration in this area. And the relative permeability and the young's modulus of the reactor core material in the optimized area are presented as the interpolation function of the defined material density. Secondly, combined with the finite element method (FEM), the gapped-iron core reactor model is optimized, in which the parameters are obtained by measuring the magnetic properties of silicon steel. And the threshold function is presented to eliminate the intermediate density and improve the optimal material boundary. Finally, the stress and vibration displacement curves of the reactor before and after the optimization are calculated. The vibration displacement of the reactor with the optimal topology is reduced by 31% and the number of iterations is reduced by 6 times.

Evaluation of solid-core thermal antimatter propulsion concepts

Energy released during antimatter annihilation can be used to heat a working fluid to directly produce thrust (an “antimatter rocket”) or run a thermodynamic cycle to generate power for conventional electric propulsion systems (antimatter power generation). For a solid reactor core, we evaluate and compare such antimatter propulsion concepts and assess the resulting performance for different propellants (hydrogen, water, methane, and carbon dioxide) and electric propulsion technologies (arcjets, Hall thrusters, and gridded ion thrusters). In contrast to positron–electron annihilation (which only produces gamma rays), complete stopping of relativistic pions released during antiproton–proton annihilation increases the required reactor core size by approximately an order of magnitude, resulting in significantly larger propulsion systems. Because of reactor core material considerations, the maximum temperature and specific impulse achievable in antimatter rockets is limited. By instead operating the core at a lower temperature to drive a closed thermodynamic Brayton cycle, antimatter power generation can run electric propulsion systems with higher specific impulses. Although this reduces the propellant mass needed for a given mission, the combination of thermodynamic cycle efficiency (around 30%) and electric thruster efficiency (typically between 30%–70%) increases the antimatter mass required by one to two orders of magnitude. The lower thrust of electric propulsion systems also leads to longer burn and maneuver times. Considering emerging industry trends, the use of renewable propellants offers large potential benefits to future antimatter propulsion systems that can refuel in space.

Isotopic Inventory and Activity Calculations of Fukushima Daiichi Unit-1 Accident

Nuclear characterization of the spent nuclear fuel in a reactor core is essential, especially in case of severe accidents. The radionuclide inventory and its activity can assist in the management of spent fuel handling, transport, or reprocessing. In this paper, the core of the Fukushima Daiichi Unit-1(FD-U1) accident was modeled using the Monte Carlo code (MCNPX 2.7) linked to the depletion calculation code CINDER'90 and ENDF/B-VII.0 cross-section data library. The isotopic inventory and the activity of the radionuclides for the burned fuel were calculated. The input to the code depends on the previous evolution of the reactor core configurations, dimensions and material of the fuel assemblies, initial uranium enrichment, fuel burn-up, and reactor core operational history. The calculations were validated with experimental measurements which were carried out by the Japan Nuclear Energy Safety Organization (JNES) and verified with published results using ORIGEN2-code by Japan Atomic Energy Agency (JAEA). The masses, activities, specific activities, half-lives, and decay schemes for the actinides and fission products were calculated at the time of the accident and after 50 years of cooling time. The calculations showed that the total activity of the burned fuel in the core at the time of the accident was 9.86E+19Bq and after 50 years was 1.89E+17Bq and the higher inventory concentration in the fuel was dominated by the trans-uranic elements. Also, the specific activity in the core at the time of the accident and after 50 years of cooling time was found to be 1.84E+15Bq/g and 5.86E+12Bq/g, respectively.

CFD Investigation of MHTGR Natural Circulation and Decay Heat Removal in P-LOFC Accident

One of the key postulated accidents in a high-temperature gas-cooled reactor (HTGR) is the pressurized loss of forced circulation (P-LOFC) of the primary loop, which can be triggered by its primary helium circulator trip or turbine trip. If the reactor shutdown cooling system (SCS) fails during a P-LOFC accident, part of the reactor decay heat is absorbed by the reactor core materials and the rest removed by the reactor cavity cooling system (RCCS). In the extended period of P-LOFC accident compound with SCS failure, the core decay heat is supposed to be removed by conduction, natural circulation, convection, and radiation. A three-dimensional (3-D) computational fluid dynamics (CFD) simulation was performed in this research to study the long-term heat removal mechanisms in the General Atomics’ Modular High Temperature Gas-cooled Reactor (MHTGR) design in a P-LOFC accident. The reactor core temperature distribution and flow field were obtained at different decay power levels. The sensitivity of the natural circulation flow to the bypass gap width was investigated. The natural circulation flow intensity is relatively weak but very sensitive to the width of the bypass gaps.

Phase analysis of uranium oxides after reaction with stainless steel components and ZrO2 at high temperature by XRD, XAFS, and SEM/EDX

In the Fukushima Daiichi nuclear power station accident in March 2011, fuel debris was formed when fuel materials reacted with various structural materials in reactor core. Fuel debris retrieval is expected to start in 2021 for decommissioning of the damaged plants. To perform this activity safely, it is necessary to know the properties of the fuel debris. In this study, we investigated the reaction of UO2, stainless steel (SS) components, and ZrO2 at high temperature under oxidizing and reducing conditions, and determined the valence state of uranium in the products, such as CrUO4 and (Fex, Cr1−x)UO4, by X-ray absorption near edge structure (XANES) spectroscopy. Mixed powders of UO2 and SS, Fe, and/or Cr were heated in Ar + 2% O2 (oxidizing condition) or Ar + 10% H2 (reducing condition) at a flow rate of 20 mL/min for 2 h at 1473 or 1673 K. After heat treatment, the phase relation of the products was analyzed by powder X-ray diffraction and scanning electron microscopy with energy-dispersive X-ray spectroscopy. Under reducing conditions, UO2 did not react with SS, Fe, Cr, or ZrO2. In contrast, under oxidizing conditions, UO2 reacted with SS and Cr to form CrUO4 or (Fex, Cr1−x)UO4 at 1473 and 1673 K. When the UO2 and Fe mixture was heated under oxidizing conditions, the Fe2O3 phase coexisted with U3O8 at 1473 K, whereas FeUO4 formed at 1673 K. When the UO2, Fe, and Cr mixtures were heated at 1473 K under the oxidizing condition, the molar ratio of Fe/Cr in the (Fex, Cr1−x)UO4 phase corresponded to the initial molar ratio in the sample. As the iron content of these samples increased, all three lattice parameters of (Fex, Cr1−x)UO4 approached those of FeUO4. XANES spectra revealed that the oxidation state of uranium in CrUO4 and (Fex, Cr1−x)UO4 is pentavalent.

A detailed study on zirconium and its applications in manufacturing process with combinations of other metals, oxides and alloys – A review

Abstract The production and applications of zirconium and based materials are analyzed in this review. The main properties which make the metal attractive in use are its high strength when alloyed with other materials, its outstanding corrosion resistance. This metal preferable in industrial, marine and other environments, good metallurgical workability and superior fatigue resistance. For upcoming originations tough ceramics and its oxide compound, Zirconia, also known as zirconium dioxide, ZrO2 has become a base material. For fiber reinforcement also the ceramics are applicable and recently several latest ceramic fabrication techniques are investigated. This review is a collection of details regarding the physical properties, mechanical properties, fabrication and procurement of zirconium and its based materials. The material zirconium and its alloys have excellent corrosion – radiation-resistance properties and so that it can be used as nuclear reactor core materials such as fuel cladding and guide tubes. The experimental results on manufacturing of ceramics using Aluminum oxide with a regular structure which gives high strengthening properties and the extensive usage in different streams of technology were investigated. With the highly disperse powders new ceramic materials having a high strength were found out.

Measuring the fracture properties of irradiated reactor core graphite

The assessment of the post-cracking behaviour of the reactor core components requires knowledge of the fracture properties, such as the fracture toughness, KIC, and work of fracture, γf, of irradiated graphite. The measurement of these properties is a proven technique for linearly elastic materials but application on small nuclear graphite specimens demands further consideration. The purpose of this work is to develop this technique for use on irradiated graphite specimens whose size and geometry are restricted by the reactor core trepanning or Materials Test Reactor experiments. This paper describes the theoretical basis of the method, the work undertaken to prove the measurement technique and demonstrate its applicability to small irradiated and oxidised graphite specimens. Finally, the paper presents work of fracture values of irradiated graphite trepanned from AGR core bricks; for the first time, the relationship between work of fracture and other graphite properties, such as density, modulus and strength is experimentally demonstrated.

Assessment of a lower head molten pool analysis module using LIVE experiment

The main objective of this work is assessment and improvement of the COUPLE module, the lower head molten pool analysis model included in a reactor analysis code RELAP/SCDAPSIM, by using LIVE test series. This module is used to calculate the heat-up of reactor core material that slumped into the lower head of the reactor vessel. Heat transfer in the radial and axial directions to the wall structures and water surround the debris is considered in the module. In the LIVE test facility, the behavior of reactor pressure vessel in the late in-vessel phase of a postulated severe accident in a nuclear power plant is investigated. Due to the recent interest in in-vessel melt retention through external reactor vessel cooling as a key strategy of the severe accident management, the ability and applicability of the module need to be assessed and improved. The heat flux, wall temperature, crust thickness and molten pool temperature are compared with the experimental data. The code was modified to allow modelling external cooling vessel and an input deck for the LIVE facility was created. The results using other empirical correlations are also compared and, as a result, better agreement of heat flux and wall temperature is obtained.

Enrichment of 57Fe isotope in neutron flux of nuclear reactors observed by Mössbauer spectroscopy

The abundance of 57Fe isotope in nuclear reactor core materials can be considerably enriched by neutron-capture 56Fe(n,γ) reactions. This is demonstrated using the sections of Zr-2.5 wt.%Nb pressure tubes removed from two CANDU* reactors. The tubes contained 0.11 and 0.04wt% Fe and were irradiated for about 10 effective full power years (EFPY) up to ~1026n/m2 fast neutron (E > 1MeV) fluencies. The Mössbauer spectra of 57Fe in irradiated samples indicated up to 10 times larger areas than unirradiated off-cuts from the same pressure tubes. The observed effect is in accord with the values calculated for known thermal neutron-capture cross-sections and resonance capture integrals, neutron flux profiles and spectra, and times of irradiation. The build-up of 57Fe facilitated recording Mössbauer absorption spectra of alloys with minor amount of Fe down to ~ 400ppm, despite intense background radiation emitted by samples. These findings can open new possibilities in post-irradiation studies of alloys used in nuclear reactors and in other objects subjected to large neutron fluencies.

Excitation Function Calculations of Neutron-Induced Reactions of Some Zirconium Target Isotopes

Nuclear structures with high performance materials will be critical for the success of future fusion energy. Zirconium alloys are extensively used for nuclear reactor core materials as fuel cladding and guide tubes due to excellent radiation and corrosion resistant properties. In this paper, the excitation functions of 90Zr(n,2n)89Zr, 90Zr(n,α)87Sr, 91Zr(n,p)91Y and 92Zr(n,p)92Y nuclear reactions were calculated for energies up to 25 MeV. Besides, ALICE/ASH and TALYS 1.6 codes were used for the cross section calculations. The obtained calculations were compared with the experimental data in the literature.

A closed nuclear energy system by accelerator-driven ceramic reactor and extend AIROX reprocessing

For the future energy system, we propose a new closed nuclear energy cycle system, which consists of an accelerator-driven external neutron source, a ceramic reactor and an extend AIROX reprocessing. The attractive features of this system are as follows. (1) The operating mode of the reactor is a combination of subcritical mode and critical mode. Initially, the reactor would be driven by the accelerator external neutron source in subcritical mode. A few years later, the reactor would reach the critical mode, and then would operate for a long time. (2) Nuclear fuels, coolants, and structure materials in the ceramic reactor core are made up of ceramic with excellent thermodynamics properties and neutron performance. Therefore, the ceramic reactor has extremely inherent safety, good breeding performance and high power generation efficiency. (3) Fuel reprocessing uses an extend AIROX reprocessing, which is a simple high-temperature dry process and rarely involved in chemical process. In this reprocessing, only most of fission products are separated. Other isotopes, including uranium isotopes, transuranic nuclides and long-lived fission products, would re-enter the reactor as new fuels. Therefore, this closed nuclear energy system could be known as ADANES, short for Accelerator-Driven Advanced Nuclear Energy System, which can greatly improve the utilization rate of nuclear fuels, enhance the nuclear safety, reduce the nuclear proliferation and become a sustainable and low-carbon energy supply for thousands of years.

The temperature monitoring during the reactor core material irradiation by analyzing the structure of graphite-like boron nitride

Possibility has been shown for determining the temperature of in-core irradiation using a single structural analysis of irradiated boron-nitride ceramics without a time-consuming research stage, including post-irradiation annealing. This is explained by the fact that steady structural states, depending on temperature and not depending on exposure dose, may develop in nanocrystalline graphite-like ceramic boron nitride in high-dose environments. A temperature dependence of the position of the X-ray line (002) has been obtained for the radiation-induced steady structure of ceramics. This dependence can be used for determination of temperature in a fast reactor in a range of 690 to 1870K.It has been found using X-ray structural analysis that reaction-sintered boron-nitride ceramics in the initial and irradiated states comprises two structural components. The first one corresponds to defect-free hexagonal and rhombohedral structures. The X-ray line displacements relate to the second component the nanocrystallites of which contain a great deal of microdeformations stabilized by clusters of vacancy discs.

Мониторинг температуры внутриканального реакторного облучения материалов с помощью анализа структуры графитоподобного нитрида бора

Мониторинг температуры внутриканального реакторного облучения материалов с помощью анализа структуры графитоподобного нитрида бора

Iron Loss Evaluation of Reactor Core With Air Gaps by Magnetic Field Analysis Under High-Frequency Excitation

It is revealed by magnetic field analysis that a difference between an actual reactor loss and a material datasheet loss, which was observed by measurement using an actual reactor with air gaps under high-frequency excitation, is caused by leakage magnetic flux near the air gaps of a reactor. In our previous measurement, an actual reactor loss under high-frequency excitation became large compared with a datasheet loss of the reactor core material. To clarify the reason, we carry out a 3-D magnetic field analysis. As a result, due to the increasing eddy current losses by leakage magnetic flux density near reactor gaps and due to the ununiform distributions of an iron loss caused by leakage magnetic flux, the reactor loss becomes larger than that of the datasheet loss.