Reactor Graphite Research Articles

Experience and Application Possibilities of Scanning Devices for Monitoring Radiation Fields in Shut-Down Uranium–Graphite Reactors

Experience and Application Possibilities of Scanning Devices for Monitoring Radiation Fields in Shut-Down Uranium–Graphite Reactors

SELECTING OF MATRIX COMPOSITION FOR IMMOBILIZATION OF IRRADIATED FUEL OF THE IGR REACTOR

Impulse Graphite Reactor (IGR) is a unique nuclear installation namely research reactor with the core consisting of uranium-graphite fuel elements with 90% enrichment by U235. Possible immobilization of the first core taken from the reactor in 1967 is studied here as part of the IGR reactor conversion to low-enriched uranium fuel. Irradiated fuel is related to the category of mediumactive waste according to the content of beta-emitters. Expected value of highly-enriched uranium fuel (HEU) burnup rate was less than 1 % mass. The experts from the Institute of Atomic Energy (IAE) proposed the way how to immobilize irradiated IGR reactor fuel into the cement matrix. First, the fuel is suggested to dilute with depleted uranium in order to lower the enrichment up to the <20 mass. % by U235. The properties of the matrix should meet the engineering criteria ruled by international and national standards. So, according to aforesaid, The IAE team is facing an actual problem of selection the matrix and understanding its characteristics.The work was aimed at selection of composition of cement mortar for making robust matrix to immobilize irradiated IGR reactor fuel.The paper describes techniques and results of the time identification (start) of solidification, gradual volume change, water separation, velocity and uniform cement mortar.

Preliminary analysis of the irradiation deformation of a typical molten salt reactor graphite component

Nuclear graphite commonly serves as the moderator and constructs the coolant flowing channel in liquid fuel molten salt reactors. Under irradiation and high temperature, graphite components will experience obvious deformation due to irradiation creep, thermal expansion, dimensional change, elastic deformation, and thermo-mechanical properties change. The lifespan of graphite components is a limiting parameter of liquid fuel molten salt reactors, determining by the deformation of the graphite components. An improved approach based on the finite element method for graphite deformation analysis was proposed in this paper. Then this method was utilized to analyze a 10-year deformation history of a typical graphite component. The findings show that the stress-strain field of the graphite component significantly varies with the increasing irradiated neutron fluence. The dimensional change plays a primary role in influencing the deformation of the graphite components, determining the overall deformation trend. The irradiation creep plays a secondary role in influencing the deformation, alleviating the deformation of the graphite components to prolong the graphite lifespan. The improved method, which considers various factors like irradiation creep, thermal expansion, dimensional change, and elastic deformation, tends to provide more accurate results compared to the traditional method that focuses solely on the dimensional changes. Over 10 years, the contraction deformation of graphite components results in a reduction in graphite volume of up to 3.87%. The space for graphite contraction is filled with molten salt, resulting in a maximum increase of 22.49% in fuel volume. The alteration in volume of both graphite and fuel leads to a reduction in reactivity of up to 840 pcm.

Identification of the regularities of the thermohydraulic processes of the cooling system of an experimental device based on a mathematical model

The article is devoted to the study of a nitrogen supply system for cooling an experimental device designed for testing in a pulsed reactor with a graphite moderator. To analyze emergency situations caused by disruption of the nitrogen supply system in the cooling paths of the fuel assembly cover and the experimental device power housing, a mathematical model has been developed to describe thermodynamic processes in a system of interconnected tanks that simulate individual sections of the experimental device hydrodynamic cooling system in a pulsed graphite reactor and implemented in the mathematical interactive package for engineering calculations in MATLAB/Simulink. To study the normal operation of the system and analyze possible emergencies, the cooling system of the experimental device was presented in the form of groups of interconnected sections of the hydrodynamic system-tanks, with specified volumes and hydraulic resistances of pipelines connecting these tanks. Verification of the model, calculations of the experimental device cooling system normal operation mode, calculations of emergency situations caused by a violation of the nitrogen supply system normal operation in the cooling paths of the fuel assembly cover were carried out, in particular, it was found that for the cooling system of the experimental device WF-2, the breakdown of the gearbox P-12 can cause depressurization of the fuel assembly cover and an increase in pressure in the power housing, the opening of the safety valve prevents exceeding the pressure limit in the housing, the device maintains integrity even in emergency situations, without being subjected to destruction of the power elements of the structure at a pressure of up to 8 MPa

Testing of a heterogeneous fuel rod in the research Impulse graphite reactor

The paper presents the results of an experiment conducted in the IGR (Impulse Graphite Reactor) to study the behavior of a heterogeneous fuel rod under loss-of-flow conditions. A heterogeneous fuel rod is one of the options under consideration for the layout of a core of fast reactors. The fuel rod consists of two zones of enriched fuel, separated by a fuel zone with low content of uranium-235.During the experiment, conditions were provided that correspond to the expected operating conditions of a fuel rod in a fast reactor at a nominal power, including the level and distribution of energy release throughout the height of a fuel rod and the temperature of an ambient sodium.The experiment results allowed evaluating the state of fuel pellets subjected to thermal impact, typical for the initial phase of an Unprotected Loss-of-Flow (ULOF), in which there is a decrease in heat removal from fuel rods, heating of fuel pellets up to the melting temperature and destruction of the fuel cladding.The methodological approaches and obtained experimental data will be used in preparation and conduction of experiments with model heterogeneous fuel assembly (FA) when simulating the ULOF.

Application of ion implantation as a tool to study neutron induced morphological changes in HOPG and RBMK-1500 reactor graphite

Based on the RBMK-1500 (rus.: peaктop бoльшoй мoщнocти кaнaльный, PБMК; reaktor bolshoy moshchnosti kanalnyy, “high-power channel-type reactor”) reactor irradiation conditions, ion implantation was used as a tool to study neutron induced morphological changes in both highly oriented pyrolytic graphite (HOPG) and nuclear grade RBMK graphite. Graphite samples were implanted with 180 keV 14N+ ions at the fluences of 1.0 × 1016 ions/cm−2 and 2.5 × 1016 ions/cm−2. To study temperature effects on both ion migration and structural disorder level in the graphite matrix, the implantation procedures were carried out under different temperature conditions – room temperature (RT) and 500 °C. Subsequently, the distribution profiles of the implanted 14N+ ions were obtained by using secondary ion mass spectroscopy (SIMS) technique, while the microstructural properties of the graphite were evaluated by Raman spectroscopy. Implantation induced primary displacement damage profiles and implanted ion profiles were evaluated theoretically by using GEANT4 10.6 and SRIM-2013 codes. The profiles of implanted nitrogen obtained by SIMS technique were found to be in good agreement with theoretical ones. The surfaces of HOPG samples displayed macroscopic damage in form of fractures after 14N+ ion implantation at 500 °C; however, in case of RBMK graphite, the sample surfaces remained without visually observable changes. The Raman spectra showed an increase of sp3-related content and formation of amorphous carbon in both HOPG and RBMK graphite samples; the latter exhibits higher structural disorder at microscopic level and stronger amorphization than HOPG.

Approbation of a New Ion-Plasma Technology for Reactor Graphite Deactivation

Approbation of a New Ion-Plasma Technology for Reactor Graphite Deactivation

Experimental opportunities and main results of the impulse graphite reactor use for research in safety area

The paper presents main technical characteristics of the Impulse Graphite Reactor (IGR), which determine its unique experimental opportunities to research behavior of the fuel rods and fuel assemblies in transient and accident operation conditions. A brief review of main experimental programs implemented at the IGR in support of safety of various nuclear reactor operation was conducted. Obtained results became new and unique ones in many aspects and enabled determining important fuel rod parameters from the reactor safety and performance point of view, data on interaction process between core material melt and coolant, as well as on its controlled moving throughout special channels. The gained experience and competency of the personnel, developed methodological approaches to tests, technical equipping of the reactor allow now successfully implement new experimental programs, manage actual scientific issues and expand international cooperation in enhancing nuclear power safety.

14C AMS measurement of gas diluted reactor graphite using a Gas Injection System

For the final disposal of reactor graphite a quantitative characterization of its radioactivity is needed. In this work an emphasis is placed on the isotope 14C, an isotope accessible to AMS measurements. Reactor graphite can be highly activated and therefore difficult to measure without the risk of contaminating the AMS setup. Using an Elemental Analyzer (EA), a Gas Injection System (GIS) and the 6 MV AMS system of the Institute for Nuclear Physics, University of Cologne, we automated the measurement process for gaseous, diluted samples. The presented work shows the first steps to verify the possibility of gas dilution and provides a comparison to samples with known concentrations of 14C. In the future, our system will provide a high sample throughput of gas-diluted probes with high activities.

Feasibility study mixed oxide fuel tests in the impulse graphite reactor

The article presents the results of computational studies to substantiate the possibility of testing the fuel of the Multipurpose hYbrid Research Reactor for a High-tech Applications reactor (MYRRHA) in the Impulse Graphite research Reactor (IGR). A scheme for conducting experiments is proposed, which allows simulating the operating conditions of MYRRHA fuel elements in a wide range of changes in their thermohydraulic parameters, including the conditions of accidents with the destruction of fuel pellets, blocking the flow area and stopping the coolant flow. A method for thermohydraulic calculations has been developed, implemented in the software environment of the computational fluid dynamics (CFD) code. A computational model of the experimental device is developed, optimized according to the spatial partitioning grid. A turbulence model for determining the parameters of convective heat and mass transfer in a eutectic lead-bismuth alloy is selected and substantiated.

A surrogate machine learning model for advanced gas-cooled reactor graphite core safety analysis

A surrogate machine learning model was developed with the aim of predicting seismic graphite core displacements from crack configurations for the advanced gas-cooled reactor. The model was trained on a dataset generated by a software package which simulates the behaviour of the graphite core during a severe earthquake. Several machine learning techniques, such as the use of convolutional neural networks, were identified as highly applicable to this particular problem. Through the development of the model, several observations and insights were garnered which may be of interest from a graphite core analysis and safety perspective. The best performing model was capable of making 95% of test set predictions within a 20 percentage point margin of the ground truth.

Radiocarbon generation and atmospheric release assessment from nuclear power plant with RBMK type reactors

The article provides the results of the estimation of 14C nuclide generation and release from the systems and structures of RBMK-1500 reactors operated at Ignalina NPP. The main circulation circuit and gaseous circuit of this type of reactor are the primary sources of 14C generation and airborne release, evaluated to be 0.491 and 0.367 TBq per one calendar year (10.5 months of effective reactor operation at 1350 MWe power), respectively. The analysis of the operation specifics of these systems showed that they don’t provide an effective barrier to contain 14C due to the 14CO2 chemical form of radiocarbon in RBMK reactors, i.e., 100% release (0.858 TBq/year) have been evaluated available for atmospheric release. These results have been compared with the atmospheric releases evaluated from the 14C activity concentration measured in Scots pine tree (Pinus Sylvestris L.) ring samples by single stage accelerator mass spectrometer (SSAMS, NEC, USA) at Vilnius Radiocarbon facility, resulting to airborne release value of 1 TBq/year for one RBMK-1500 reactor. Slightly higher value reproduced from the tree ring samples indicates that the reactor graphite stack might be also a source of 14C releases of about 0.14–0.24 TBq/year. Based on similar design features the results of the study can be easily projected to the RBMK-1000 reactors rescaling 14C source term according to the lower reactor power.

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.

Problems of solid radioactive waste management on the example of the radioactive isotope 14C

The management of irradiated graphite waste continues to attract the attention of specialists from a number of related industries, in connection with the technical and economic component, along with the fact that solid radioactive waste entering the environment in the process of disposal acquires biogenic character. The environmental safety of solid radioactive waste management is associated with the possibility of biologically significant 14C to enter the human body.
 Resulting from the development of the nuclear industry, 14C has become one of the causes of global and local pollution. The consequences of thermonuclear explosions, operation of nuclear power plants, as well as nuclear fuel cycle enterprises, isotope production for the production of drugs labeled with 14C, research institutions are considered as sources of 14C.
 The world output of accumulated reactor graphite is estimated at 250,000 tons. In the Russian Federation, the total volume is approximately 60,000 tons of graphite. The share of accumulated irradiated graphite in Russia comprises 24% of the total volume of accumulated graphite.
 Having studied and analyzed the statistical data of scientific research, using the methods of generalized data processing, systematization and comparative analysis, we have taken the liberty to make a general conclusion about modern approaches and the relevance of solving the problem of SRW management on the example of the 14C radioactive isotope.

Thermodynamic simulation of radioactive graphite combustion in water vapor

The paper analyzes the gas phase, the behavior uranium, the calculation of thermophysical properties during heating of nuclear reactor graphite containing various actinides. The study was carried out by thermodynamic modeling using the TERRA software package in the range from 423 to 2123 K at a constant pressure of 0.1 MPa. Up to a temperature of 623 K, the main components of the gas phase are water vapor, carbon dioxide, methane and hydrogen. Above 1023 K, the gas phase contains predominantly carbon dioxide and monoxide. It was revealed that in the range of 423-923 K uranium is present in the form of condensed compounds - uranium dioxide and uranyl chloride. Up to a temperature of 1623 K, uranium is present in the form of condensed uranium dioxide and calcium uranate. At temperatures above 2023 K, condensed uranium compounds are converted into gaseous and ionized compounds of uranium trioxide and dioxide. The thermophysical characteristics change unevenly up to a temperature of 923 K.

Actinides and fission products in reactor graphite after loss-of-flow accident

Samples of irradiated graphite from a uranium-graphite reactor contaminated with fuel debris from loss-of-flow accidents, are investigated. Peculiarities of spatial distribution, speciation, and kinetics of leaching by aqueous solution of actinides and fission products are studied at scales from the whole reactor stack down to sub-mm graphite fragments. The main fraction of fuel debris is associated with aluminosilicate-based phases decorating cracks and other defects of damaged graphite details. Elemental composition of actinides and fission products depends on both irradiation and leaching conditions. Significant differences are observed between leach kinetics of various radionuclides; integral leached fraction varies between ~5% for 244Сm to ~90% for 134,137Cs. For some radionuclides sudden massive release due to uneven dissolution of the carrier phases is observed. Disposal of irradiated graphite with significant fraction of fuel debris requires dedicated treatment aimed either at decontamination or at conversion of the radionuclides to insoluble well-fixed form.

CAD Approach to Control High-Temperature Processes in SiC Technology

Manufacturing silicon carbide semiconductor devices may require high-temperature treatment in closed graphite reactors. This makes temperature control of processed SiC substrates difficult, since their temperature cannot be measured directly. As the monitoring of the SiC sample temperature is critically important for proper process flow, an indirect method involving the use of the CAD approach has been developed. A numerical model of a furnace reactor was created on the basis of the commercial ANSYS package, allowing for the simulation of thermal fields under given heat-dissipation conditions in the modeled area and in the presence of gaseous and liquid media participating in heat exchange and transport. Obtained simulation results remain very consistent with the reference temperature measurements of selected areas of the reactor. The model acts as an accurate tool for temperature distribution verification during the high-temperature annealing of and diffusion of dopants for silicon carbide.

Defect formation and bending properties in graphite under He atom implantation investigated by molecular dynamics method

The formation and evolution of radiation defects and their effects on bending property of a single graphite under low energy helium atom implantation have been studied at atomic scale through molecular dynamics simulations. Various simple interstitial and vacancy defects are observed during the implantation process, which is a function of incident energy and cross section of helium-carbon interaction. Large defects are observed after continuous helium atom implantations. The effects of incident energy on the density, size and distribution of defects are analyzed in detail. Furthermore, the kinetic evolution process of radiation defects at different annealing temperatures have also been simulated, including the migration, aggregation, transformation, coalescence, and self-healing. The large defect clusters, crossing several carbon layers, have been confirmed to be the primary reasons that affect the bending property of graphite. All these results provide a new understanding for further applications of graphite in nuclear reactors.

The effect of radiolytic oxidation on the elastic properties of nuclear graphite measured by electronic speckle pattern interferometry

• Three-point bend measurement data characterised by ESPI and inverse modelling. • Static Young's modulus , Poisson's ratio and strength determined. • Measurements performed on 218 irradiated graphite beams taken from UK AGR stations. • Graphite samples (6 × 6 × 19 mm) cover dose range to 21.1 DPA and weight loss to 46%. • Modulus and strength correlate well with radiolytic weight loss. The Young's modulus, Poisson's ratio and flexural strength of 218 beams of radiolytically oxidised nuclear graphite have been determined by combining electronic speckle pattern interferometry, ESPI, and inverse finite element modelling techniques. The graphite was extracted from three operational advanced gas cooled reactor stations where it had been subjected to fast neutron irradiation and radiolytic oxidation resulting in mass loss. The 6 × 6 × 19 mm beam samples had densities ranging from 1063 – 1840 kg.m −3 , equivalent to weight losses up to 46%, with cumulative neutron doses of up to 161 × 10 20 n.cm −2 equivalent Dido nickel dose, or 21.1 displacements per atom, significantly extending the datasets available in the open literature to higher levels of both dose and oxidation. In situ ESPI measurements were collected during three-point bend tests, creating full field displacement maps of the evolving deformation to which simulations were matched by chi-squared minimisation of the elastic properties. Potential loading abnormalities were accounted for in the fitting process to ensure the simulations were representative, with total uncertainties in the determined Young's modulus and Poisson's ratio values of ± 4.5% and ± 0.08, respectively. Consistent trends were observed in the behaviour of the material from the different stations, with strong positive dependences on density for both the flexural strength and Young's modulus. The strength and Young's modulus remained well correlated even to high levels of oxidation mass loss. Conversely, the Poisson's ratio was practically invariant with oxidation with mean values of 0.21 ± 0.03, 0.24 ± 0.05 and 0.25 ± 0.05 for the three stations. The combined ESPI-finite element measurement approach is demonstrated as robust for irradiated small specimens, making it suitable for monitoring graphite in future reactors. The collected data support the safety assessments of the current reactors, enabling predictions of the responses to progressively higher mean core mass loss.

A proposed approach for the evaluation of consequences of a large aircraft crash accident at an RBMK type reactor site during decommissioning

Nuclear power plants with RBMK reactors were designed in late sixties - early seventies and were built entirely in the European part of the Russian Federation, the Lithuanian Republic and the Ukrainian Republic. The plants in the Lithuanian and the Ukrainian Republics are in various stages of decommissioning. As currently planned, the most of RBMK reactors still operating in the Russian Federation will end their operation in this decade.The reactor building, spent nuclear fuel and radioactive waste storage facilities, built at the time of construction of RBMK plants are not specially designed to withstand the direct impact of an aircraft crash. There were no such regulatory requirements at the time the plants were designed and, no such requirements were introduced by regulatory bodies later on. But for now, in the aftermath of the September 11, 2001 terrorist attack on the United States, the event of an aircraft crash on a nuclear facility can no longer be considered as extremely unlikely.This paper considers the state of a nuclear power plant with the RBMK type reactor(s) during certain decommissioning and dismantling stages and discusses the approach for the evaluation of radiological consequences to the public in the event of a large aircraft crash on potentially hazardous objects of the plant. The damage to spent nuclear storage facilities, the reactor graphite stack and storage facilities for the combustible waste may cause a significant release of radionuclides to the environment. Key components of the approach include the evaluation of a structural damage to the facilities, the evaluation of a fire propagation and impact, the evaluation of a spent nuclear fuel cooling capability in storage pools, the evaluation of an airborne radionuclide release, the evaluation of a radionuclide dispersion in the environment and the evaluation of the public exposure. A set of widely used software and methods for the performance of the specific tasks is discussed. The paper also describes how the proposed approach was applied for the evaluation of radiological consequences from a Boeing 767-400 type aircraft crash at the Ignalina NPP site where a plant with two RBMK-1500 type reactors is decommissioned.The paper does not address modern radioactive waste and spent nuclear fuel management and storage facilities that might be built at RBMK sites as well. As a rule, assessment and mitigation of an aircraft crash accident consequences is an integral part of design of these new facilities.