Graphite For High Temperature Gas-cooled Reactors Research Articles

Experimental study of friction coefficient of graphite for high-temperature gas-cooled reactors

High-temperature gas-cooled reactors, as highly regarded fourth-generation reactors, extensively utilize graphite materials. It is necessary to perform a study of the characteristics of the graphite materials aiming at optimizing the structural design and neutron physics calculation for the reactor. Thus, the investigation of the graphite friction coefficient is a crucial aspect of the research. This paper presents experiments on the graphite friction coefficient conducted using a specially designed experimental device that closely replicates the operational conditions of high-temperature gas-cooled reactors. The experimental data on the graphite friction coefficients under various atmospheric and temperature conditions are obtained. The graphite friction coefficient under a helium atmosphere demonstrates an initial increase followed by a decrease with rising temperature, reaching a maximum value of 0.17 at 200 °C. However, beyond 400 °C, the friction coefficient stabilizes gradually and fluctuates within the range of 0.03 to 0.06. Under a nitrogen atmosphere, the graphite friction coefficient exhibits the same temperature-dependent pattern as observed in helium. The peak value likewise occurs at 200 °C, measuring 0.23. At temperatures below 400 °C, the graphite friction coefficients in a nitrogen atmosphere surpass those in helium. While beyond 400 °C, the friction coefficients in a nitrogen atmosphere likewise fluctuate within the range of 0.03 to 0.06.

Long-term corrosion of HTGR graphite with SiC coating

When creating a Kazakhstani high-temperature gas-cooled reactor (HTGR), a type of fuel with improved graphite and its gradient protective coating based on silicon carbide (SiC) will be used. So, the purpose of this work is determination the rate and features of the long-term high-temperature corrosion of reactor graphite and its SiC coating at a temperature of 1200K and a pressure of 200-250 kPa in helium with an admixture of water and oxygen in concentrations close to the normal operation of an HTGR; development a numerical model of a corrosion experiment with the convection method of supplying the working mixture. As a result, four successive test stages lasting 30 days each (the total duration of corrosion was 120 days), the weight loss of pure graphite samples (without SiC coating) range from 1.8% to 10.4%, depending on the manufacturer. In general, the average corrosion rate of unprotected graphite corresponds to the literature data. The use of a SiC coating over graphites reduces the corrosion rate of samples by a factor of 100-500. The developed model made it possible to describe both the high-temperature corrosion processes of HTGR graphite and the dynamics of the mass transfer of all gas components throughout the entire volume of the installation during the experiment.

Modeling the Processes of Hydrogen Isotopes Interactions with Solid Surfaces

The investigation of the mechanisms and dynamics of hydrogen isotopic interaction with solid surfaces (metals, ceramics, graphites, eutectics) in temperature and pressure ranges is important not only for the correct prediction of each isotope’s evolution but also for substantiation of the safe operation of hydrogen-facing structural materials. The interaction of the hydrogen isotopes mix with the surface of solid metal or liquid eutectics is a complicated multistage H-D-T-O-solid interacting process depending on material property, environment, and the solid’s surface parameters. To better understand the mechanisms of hydrogen isotopes interchange at a solid surface and to identify the limiting stages in the sorption-desorption processes, a reactor experiment of neutron irradiation was conducted with lithium-containing eutectics as tritium-generating media under the external flow of hydrogen. This work presents the model and results of its application to fitting the experimental results of tritium yield from the lithium-lead eutectics Pb83Li17 under thermal neutrons irradiation at the IVG.1M reactor in Kazakhstan. The elaborated model and the approach used were also applied to the simulation of high-temperature gas-cooled reactor graphite corrosion in water vapors.

Research on microstructure and oxidation resistant property of ZrSi2-SiC/SiC coating on HTR graphite spheres

ZrSi2-SiC/SiC coating was prepared on the surface of high temperature gas-cooled reactor (HTR) matrix graphite spheres by two-step pack cementation and sintering process. The microstructure, oxidation resistance and thermal shock resistance properties of the as-prepared coatings with different original powder mixtures were investigated. Results show that dense microstructure of the ZrSi2-SiC/SiC coating and continuous ZrSiO4-SiO2-ZrO2 glass phase generated during the oxidation process were the key factors for the outstanding thermal properties. When the mole ratio of Zr:Si:C reaches 1:7:3 in the second pack cementation powders, the coated graphite spheres have optimum oxidation resistant ability. The weight gain is only 0.6wt% after 15 times thermal shock tests and 0.12wt% after isothermal oxidation test at 1500°C for 20h in air. The oxidation resistant mechanism of the coating was also discussed. The dense inner SiC layer and the outer glass layer generated during the oxidation process could protect the ZrSi2-SiC/SiC coating from further oxidation.

The mechanical behavior and reliability prediction of the HTR graphite component at various temperature and neutron dose ranges

In a pebble-bed high temperature gas-cooled reactor (HTR), nuclear graphite serves as the main structural material of the side reflectors. The reactor core is made up of a large number of graphite bricks. In the normal operation case of the reactor, the maximum temperature of the helium coolant commonly reaches about 750°C. After around 30 years’ full power operation, the peak value of in-core fast neutron cumulative dose reaches to 1×1022ncm−2 (EDN). Such high temperature and neutron irradiation strongly impact the behavior of graphite component, causing obvious deformation. The temperature and neutron dose are unevenly distributed inside a graphite brick, resulting in stress concentrations. The deformation and stress concentration can both greatly affect safety and reliability of the graphite component. In addition, most of the graphite properties (such as Young's modulus and coefficient of thermal expansion) change remarkably under high temperature and neutron irradiations. The irradiation-induced creep also plays a very important role during the whole process, and provides a significant impact on the stress accumulation. In order to simulate the behavior of graphite component under various in-core conditions, all of the above factors must be considered carefully. In this paper, the deformation, stress distribution and failure probability of a side graphite component are studied at various temperature points and neutron dose levels. 400°C, 500°C, 600°C and 750°C are selected as the core-side temperature of the graphite component, while the range of neutron dose is 0–1×1022ncm−2 (EDN). It is shown in the numerical analysis that temperature strongly affects the histories of both stresses and failure probabilities as fast neutron dose increases. In addition, the behavior of graphite brick at 750°C is observed obviously different from that at other temperatures.

The Stress and Reliability Analysis of HTR’s Graphite Component

The high temperature gas cooled reactor (HTR) is developing rapidly toward a modular, compact, and integral direction. As the main structure material, graphite plays a very important role in HTR engineering, and the reliability of graphite component has a close relationship with the integrity of reactor core. The graphite components are subjected to high temperature and fast neutron irradiation simultaneously during normal operation of the reactor. With the stress accumulation induced by high temperature and irradiation, the failure risk of graphite components increases constantly. Therefore it is necessary to study and simulate the mechanical behavior of graphite component under in-core working conditions and forecast the internal stress accumulation history and the variation of reliability. The work of this paper focuses on the mechanical analysis of pebble-bed type HTR's graphite brick. The analysis process is comprised of two procedures, stress analysis and reliability analysis. Three different creep models and two different reliability models are reviewed and taken into account in simulation. The stress and failure probability calculation results are obtained and discussed. The results gained with various models are highly consistent, and the discrepancies are acceptable.

Development of an Evaluation Model for the Thermal Annealing Effect on Thermal Conductivity of IG-110 Graphite for High-Temperature Gas-Cooled Reactors

The thermal conductivity of graphite components used as in-core components in high-temperature gascooled reactors (HTGRs) is reduced by neutron irradiation during reactor operation. The reduction in thermal conductivity is expected to be reversed by thermal annealing when the irradiated graphite component is heated above its original irradiation temperature. In this study, to develop an evaluation model for the thermal annealing effect on the thermal conductivity of IG-110 graphite for the HTGRs, the thermal annealing effect evaluated quantitatively at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. Moreover, the thermal conductivity of IG-110 graphite was calculated by using a modified thermal resistance model considering the thermal annealing effect. The following results were obtained. (1) The thermal annealing effect on the thermal conductivity of IG-110 graphite could be evaluated quantitatively and a thermal annealing model was developed based on the experimental results at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. (2) The thermal conductivities of IG-110 graphite calculated by using the modified thermal resistance model considering the thermal annealing effect showed good agreement with experimental measurements. This study has shown that it is possible to evaluate the annealed thermal conductivity of IG-110 graphite by using the modified thermal resistance model at irradiation temperatures of 550–1150°C and irradiation fluences of up to 1.5 dpa.

The stability of SiC coating and SiO 2/SiC multilayer on the surface of graphite for HTGRs at normal service condition

The stability of SiC coating in helium with a low concentration of O 2, CO 2, and H 2O is a key factor for their application in improvement of oxidation resistance of graphite for high temperature gas-cooled reactors (HTGRs). Through thermodynamic analysis, it is found that the influence factor controlling the critical temperature of passive oxidation for SiC is partial pressure of active gas in helium; the critical temperature of passive oxidation for SiC increases with the partial pressure of O 2, CO 2, and H 2O, SiC is prone to undergo active oxidation in He–CO 2 and He–H 2O system. SiO 2/SiC multilayer coating can improve the oxidation resistance of graphite at higher temperature than SiC coating does under normal operation condition for HTGRs.

Compressive Impact Strength of High Temperature Gas-cooled Reactor Graphite

To investigate the effect of strain rate on fracture behavior for coarse grained nuclear graphite, PGX, a hydraulic servo type impact testing machine has been constructed and compressive impact strength test was performed at various strain rates up to more than 100 (1/s). From the results, the following conclusions were derived.(1) Compressive impact strength of graphite increases with increasing of strain rate in the range of 10-3 to 100 (1/s).(2) Compressive impact strength decreases drastically for strain rates more than 100 (1/s).(3) Compressive impact strength dose not depend on specimen volume.

Structural graphite for high-temperature gas-cooled reactors

A new type of nuclear graphite has been developed, consisting of noncalcined petroleum coke with coal-tar pitch binder. This material has strength properties that are about twice those of conventional graphites. The irradiation resistance of the new graphite is good over the temperature range of 800 ° to 1400 °C to a fluence of 1.3 × exp 26 n/m 2(0.18 MeV).

Leakage Flows in High-Temperature Gas-Cooled Reactor Graphite Fuel Elements

Leakage Flows in High-Temperature Gas-Cooled Reactor Graphite Fuel Elements

Highly Dense Graphite Matrix: A New Material for the Conditioning of Radioactive Wastes

In continuation of 20 yr of experience with high-temperature gas-cooled reactor graphite fuel element fabrication (molded fuel spheres and monolithic block fuel elements), NUKEM introduced a new graphite-based material with an inorganic binder produced by molding. A mixture of nickel and sulfur is used as the inorganic binder for natural graphite powder. The fabrication is facilitated by the low temperature of 500/sup 0/ C in generating a chemically and thermally stable nickel sulfide. The newly developed material is suitable as a matrix for the fixation of high-active waste and as a corrosion-resistant layer of the steel supporting tube for final disposal of spent fuel elements. The material of natural graphite/nickel sulfide is distinguished by high density, high corrosion resistance, a low leaching rate, good thermal conductivity, an appropriate coefficient of thermal expansion, and high thermal stability due to the high melting point of nickel sulfide (790/sup 0/ C). Currently, a pilot plant for fabrication of containers for spent fuel elements is being erected. Additionally, development work is in progress to obtain basic design data for a high-level waste fixation facility.