Graphite Bricks Research Articles

ADDITIONAL CHARACTERISATION OF THE THERMAL NEUTRON PILE AT THE NATIONAL PHYSICAL LABORATORY, UK.

As part of its wide-ranging neutron metrology capabilities, the National Physical Laboratory in the UK has a thermal neutron facility in which accelerator-produced neutrons are moderated within a large assembly or pile of graphite bricks. The neutron field has previously been well characterised in terms of the fluence rate and energy spectrum at various irradiation positions. However, recent changes to the structure (e.g. enlarging the central irradiation cavity) have prompted a renewal and extension of this work. We have also used Monte Carlo modelling to improve our understanding of the pile's performance.

Finite element modelling of multilayer Advanced Gas-cooled Reactor bricks and creep interaction

The structural integrity of nuclear graphite bricks is important in Advanced Gas-cooled Reactors (AGRs) as they not only provide moderation but channels for fuel, cooling and control rods. AGR graphite moderator bricks are subjected to fast neutron irradiation and radiolytic oxidation during reactor operation, leading to component dimensional and material properties changes. With irradiation ageing the graphite components deform in the axial and radial directions causing changes to bore diameter and brick length. Furthermore, deformation at the brick end faces (called dishing) can lead to the formation of axial gaps between the bricks which can then lead to fuel channel bowing. However, it is believed that irradiation creep reduces the potential size of these axial gaps and retains bricks in contact to some extent. Therefore, as the reactors age it is important to understand the nature of this contact behaviour between bricks in the fuel channels. This paper focuses primarily on the finite element modelling of contact behaviour between bricks and the effects of irradiation-induced creep on contact conditions and hence, ovality of the brick bore. Multilayer fuel brick models and a brick with a rigid body on the top surface have been modelled. The results show that multilayer models are required to understand the contact conditions between the bricks throughout the life of a reactor.

Statistical modelling of graphite brick weight loss in Advanced Gas Cooled Reactors

Demonstrating the physical integrity of graphite cores in UK Advanced Gas Cooled Nuclear Reactors operated by EDF Energy (formerly British Energy) is essential to the demonstration of their continued safe operation. The cores contain around 3000 graphite bricks which are subject to cracking and weight loss due to oxidation. If too much graphite is lost by oxidation the core will cease to act as an efficient moderator for neutrons. Periodic inspections are made of parts of the core to provide information on the state of a sample of the bricks. Statistical models are routinely used to help understand the evolution of brick cracking in the reactors. In this paper details are given of the use of mixed effects statistical models for weight loss to predict what will be seen at reactor inspections. By making blind predictions of core behaviour and then comparing these with observations from inspections, the predictive performance of the models has been quantified and they are then used to produce long-term forecasts of core behaviour. A key feature of the models is the need to represent system variability at different scales. Being able to forecast core behaviour enables reactor lifetimes to be estimated which has major economic implications.

Modelling fracture of aged graphite bricks under radiation and temperature

The graphite bricks of the UK carbon dioxide gas cooled nuclear reactors are subjected to neutron irradiation and radiolytic oxidation during operation which will affect thermal and mechanical material properties and may lead to structural failure. In this paper, an empirical equation is obtained and used to represent the reduction in the thermal conductivity as a result of temperature and neutron dose. A 2D finite element thermal analysis was carried out using Abaqus to obtain temperature distribution across the graphite brick. Although thermal conductivity could be reduced by up to 75% under certain conditions of dose and temperature, analysis has shown that it has no significant effect on the temperature distribution. It was found that the temperature distribution within the graphite brick is non-radial, different from the steady state temperature distribution used in the previous studies [1,2]. To investigate the significance of this non-radial temperature distribution on the failure of graphite bricks, a subsequent mechanical analysis was also carried out with the nodal temperature information obtained from the thermal analysis. To predict the formation of cracks within the brick and the subsequent propagation, a linear traction–separation cohesive model in conjunction with the extended finite element method (XFEM) is used. Compared to the analysis with steady state radial temperature distribution, the crack initiation time for the model with non-radial temperature distribution is delayed by almost one year in service, and the maximum crack length is also shorter by around 20%.

Dynamic fracture effects on remote stress amplification in AGR graphite bricks

An initial investigation, using a newly developed tool within Code_Aster known as the eXtended Finite Element Method with cohesive elements (XCZM), has been undertaken to investigate the dynamic stress state of a 2D graphite brick geometry under the loading of a dynamically propagating crack. This was with the aim of investigating the phenomenon known as ‘prompt secondary cracking’ (PSC) and any effects geometrical features may have on the remote stress state of the brick. It was seen that the PSC mechanism could be modelled, with methane holes possibly having an effect on this phenomenon by altering the initial static stress state.

Dismantling of the graphite pile of Latina NPP: Characterization and handling/removal equipment for single brick or multi-bricks

This work describes the issues related the dismantling of graphite piles of the 1st generation gas cooled reactor of Latina NPP (Italy).The retrieval of the graphite is a strategic matter for the decommissioning of this type of plant: in this study were described and analysed the current approaches used to access the core and to perform the remote and dry extraction of graphite bricks from the top.Based on these data, the removal of the graphite of Latina NPP will be planned; the extraction of the graphite will be carried out layer by layer by means of a dedicated remote controlled handling systems. This equipment will be duly designed according to the nuclear, physical and mechanical constraints of the graphite piles in core. In doing that the issues regarding the irradiated graphite have been also analysed by FEM code, especially those related to the core geometry and to the proposed technique of hooking the graphite bricks by a ‘gripper’ tool inside the axial channel.Data on fresh nuclear grade and irradiated graphite, used for the numerical simulations, were obtained by means of experimental tests, which were carried out on samples extracted from the reactor, and from theoretical models.The results obtained could support the final design of proper lifting and gripper tools and handling equipment, for single brick or multi-bricks, and to implement waste management strategy for the graphite.

Analytical and numerical study of graphite IG110 parts in advanced reactor under high temperature and irradiation

Structural design of nuclear power plant project is an important sub-discipline of civil engineering. Especially after appearance of the fourth generation advanced high temperature gas cooled reactor, structural mechanics in reactor technology becomes a popular subject in structural engineering. As basic ingredients of reflector in reactor, graphite bricks are subjected to high temperature and irradiation and the stress field of graphite structures determines integrity of reflector and makes a great difference to safety of whole structure. In this paper, based on assumptions of elasticity, side reflector is regarded approximately as a straight cylinder structure and primary creep strain is ignored. An analytical study on stress of IG110 graphite parts is present. Meanwhile, a finite element procedure for calculating stresses in the IG110 graphite structure exposed in the high temperature and irradiation is developed. Subsequently, numerical solution of stress in IG110 graphite structure is obtained. Analytical solution agrees well with numerical solution, which indicates that analytical derivation is accurate. Finally, influence of temperature, irradiation dimensional change, creep and Poisson’s ration on stress field, variation of stress against time and deformation of structure are discussed.

Modelling 3D crack propagation in ageing graphite bricks of Advanced Gas-cooled Reactor power plant

In this paper, crack propagation in Advanced Gas-cooled Reactor (AGR) graphite bricks with ageing properties is studied using the eXtended Finite Element Method (X-FEM). A parametric study for crack propagation, including the influence of different initial crack shapes and propagation criteria, is conducted. The results obtained in the benchmark study show that the crack paths from X-FEM are similar to the experimental ones. The accuracy of the strain energy release rate computation in a heterogeneous material is also evaluated using a finite difference approach. Planar and non-planar 3D crack growth simulations are presented to demonstrate the robustness and the versatility of the method utilized. Finally, this work contributes to the better understanding of crack propagation behaviour in AGR graphite bricks and so contributes to the extension of the AGR plants’ lifetimes in the UK by reducing uncertainties.

Spatial variability in the coefficient of thermal expansion induces pre-service stresses in computer models of virgin Gilsocarbon bricks

In this paper, the authors test the hypothesis that tiny spatial variations in material properties may lead to significant pre-service stresses in virgin graphite bricks. To do this, they have customised ParaFEM, an open source parallel finite element package, adding support for stochastic thermo-mechanical analysis using the Monte Carlo Simulation method. For an Advanced Gas-cooled Reactor brick, three heating cases have been examined: a uniform temperature change; a uniform temperature gradient applied through the thickness of the brick and a simulated temperature profile from an operating reactor. Results are compared for mean and stochastic properties. These show that, for the proof-of-concept analyses carried out, the pre-service von Mises stress is around twenty times higher when spatial variability of material properties is introduced. The paper demonstrates that thermal gradients coupled with material incompatibilities may be important in the generation of stress in nuclear graphite reactor bricks. Tiny spatial variations in coefficient of thermal expansion (CTE) and Young's modulus can lead to the presence of thermal stresses in bricks that are free to expand.

Finite element modelling of the effect of temperature and neutron dose on the fracture behaviour of nuclear reactor graphite bricks

Graphite moderator bricks used within many UK gas-cooled nuclear reactors undergo harsh temperature and radiation gradients. They cause changes in material properties of graphite over extended periods of time. Consequently, models have been developed in order to understand and predict the complex stresses formed within the brick by these processes. In this paper the effect of irradiation temperature and neutron dose on the fracture characteristics, crack initiation and crack growth are investigated. A finite element (FE) mechanical constitutive model is implemented in combination with the damage model to simulate crack growth within the graphite brick. The damage model is based on a linear traction–separation cohesive model in conjunction with the extended finite element method for arbitrary crack initiation and propagation. Results obtained have showed that cracks initiate in the vicinity of the keyway fillet of the graphite brick and initiation time accelerates with higher temperatures and doses.

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.

Experimental and numerical researches on collision of graphite bricks

Graphite bricks have important applications in high temperature gas-cooled reactors (HTGRs). The core of HTGR is a pebble-bed surrounded by graphite bricks, so vibrations and collisions between graphite bricks caused by external excitation have important influence on structural stability of the core. The experiment is aimed at studying the collisions between two bricks, using linear track test and optics acquisition system to obtain the collision characteristics of graphite bricks. Experimental results show the contact time decreases while the coefficient of restitution rises up with collision speed increasing. The results data are well-distributed and basic rules are credible. In the experiment, the impact surfaces did not make contact at the same time, as the specimens would twirl gently to adjust posture, and perfect flat collision was difficult to achieve. We simulated the collision process by a two-dimensional numerical model which had a small collision angle between the two contact surfaces. The simulation results fit well with experimental data and reveal the main influence factor of the collision characteristics. We have also discussed factors of small collision angle in the experiment, the dispersion degree of the contact time data in different speed ranges and the relationship between the coefficient of restitution and the speed. Discussion on the posture adjustment in a flat collision is an important part in the paper, and the analysis of that is meaningful for the future dynamic simulation of the core structure.

Subsurface cracks in AGR graphite bricks

Sub-surface cracks present difficulties for eddy current methods as the skin effect causes currents to flow preferentially near the surface. This work examines graphite as a case study for the qualitative assessment of sub-surface cracks and employs multi-frequency eddy current techniques. The research has particular relevance to Advanced Gas-Cooled Reactors (AGR׳s) as during their operation there is the potential for cracks to develop within moderator bricks. This work reports that subsurface slots of 18% brick thickness can be detected. This work confirms these results with a parametric 3D finite element study.

Combining models of behaviour with operational data to provide enhanced condition monitoring of AGR cores

Installation of new monitoring equipment in Nuclear Power Plants (NPPs) is often difficult and expensive and therefore maximizing the information that can be extracted from existing monitoring equipment is highly desirable. This paper describes the process of combining models derived from laboratory experimentation with current operational plant data to infer an underlying measure of health. A demonstration of this process is provided where the fuel channel bore profile, a measure of core health, is inferred from data gathered during the refuelling process of an Advanced Gas-cooled Reactor (AGR) nuclear power plant core. Laboratory simulation was used to generate a model of an interaction between the fuel assembly and the core. This model is used to isolate a single frictional component from a noisy input signal and use this friction component as a measure of health to assess the current condition of the graphite bricks that comprise the core. In addition, the model is used to generate an expected refuelling response (the noisy input signal) for a given set of channel bore diameter measurements for either insertion of new fuel or removal of spent fuel, providing validation of the model. This benefit of this work is that it provides a greater understanding of the health of the graphite core, which is important for continued and extended operation of the AGR plants in the UK.

Modelling Crack Growth within Graphite Bricks due to Irradiation and Radiolytic Oxidation

During operation, nuclear graphite structures are exposed to gradients of neutron radiation and temperature. They are also prone to radiolytic oxidation, subsequent porosity changes and weight loss. These processes cause significant changes in the material properties of graphite and create complicated stress fields within graphite bricks. Significant efforts have been made to analyse these stresses over recent years. In this paper, the influence of stresses within a typical graphite brick, during operation, on crack growth is studied. A finite element (FE) mechanical constitutive model is implemented in ABAQUS as a user-subroutine (UMAT) in conjunction with a damage model to simulate crack growth within the graphite brick. The damage model is a linear traction separation model and it is applied with XFEM enrichment to model crack initiation and growth. The results indicate that the cracks initiate in the vicinity of the keyway roots where the maximum principal stress is at its peak. Complete failure was not observed from the simulation of the graphite brick under 30 years of irradiation conditions with a maximum brick temperature of 500°C. Computed stresses and crack lengths were found to be largely dependent on the FE mesh refinement and the results of mesh sensitivity studies are also presented in this paper.

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.

Degradation of graphite in gas cooled reactors due to radiolytic oxidation

Magnox reactors employ pile grade A (PGA) graphite as a moderator. Reactor cores are constructed typically of twelve to thirteen layers of graphite bricks. Fuel channels (FC) are in the centre of all bricks and interstitial channels (IC) at the centre of the corners of every second set of four bricks. The reactor core is cooled by carbon dioxide, the temperature of graphite core increases from 250°C at the bottom to 360°C at the top of the core. The neutron dose increases progressively with the operating time of the reactor. The graphite core looses mass as a result of radiolytic oxidation. The process is dependent on both total energy deposition and temperature which correlates with core height. Fast neutron dose accumulates at the same rate as the total energy deposited and is readily available. The reduction of density of moderator graphite increases the porosity and in turn changes both the physical and mechanical properties of graphite. The mechanical properties and density of graphite are measured either on samples installed in the reactor prior to service or trepanned from graphite bricks. The data obtained on these samples are interrogated using probability modelling to establish trends with increasing service life. Results of the analyses are illustrated in the paper.PGA graphite is an aggregate of coarse needle coke filler particles within a matrix of fine coke flour particles mixed with pitch binder. The bricks are fabricated in the green condition by extrusion of dry calcinated coke impregnated with liquid pitch binder and then graphitized at 2800°C. This produces a polygranular aggregate with orthotropic properties. The strength properties of graphite are measured using different types of tests. The most commonly used tests involve bending, uniaxial and diametral compression. The initiation and propagation of cracks was investigated to improve understanding of strength behaviour. Cracking was examined on macro-scale using optical microscopy and on micro-scale using focused ion beam, FIB. In addition, digital image correlation was used to investigate the initiation of cracking. It was shown that highly localized strain regions are formed on the tensile surface of beams loaded in bending. One of the strained regions develops into a process zone which initiates cracking when reaching a length of 2–3mm. Crack propagation then occurs rapidly along an irregular path.

Effect of horizontal cross flow on the heat transfer form the moderator bricks in the Advanced Gas-cooled Reactor: A CFD study

The paper reports an investigation of the effect of the horizontal cross flow on the temperature of the moderator brick in the UK Advanced Gas-cooled Reactor (AGR) using computational fluid dynamics (CFD) with a conjugate heat transfer model for the solid and fluid. The commercial software package ANSYS Fluent is used to study the correlation between the flow structure and heat transfer which ultimately determine the temperature within the system. The CFD model comprises the full axial length of one-half of a typical fuel channel (assuming symmetry) and part of the neighbouring channels on either side. Two sets of simulations have been carried out, namely, one with cross flow and one without cross flow. The effect of cross flow has subsequently been derived by comparing the results from these two groups of simulations. The study shows that a small cross flow can have a significant effect on the cooling of the graphite brick, causing the peak temperature of the brick to reduce significantly. Two mechanisms are identified to be responsible for this. Firstly, the small cross flow causes a significant redistribution of the main axial downward flow and this leads to an enhancement of heat transfer in some of the small clearances, and an impairment in others although overall, the enhancement is dominant leading to a better cooling. Secondly, the cross flow makes effective use of the small clearances between the key/keyway connections which increases the effective heat transfer area, hence increasing the cooling. Under the conditions of no cross flow, these areas remain largely inactive in heat transfer. The study shows that the cooling of the moderator is significantly enhanced by the cross flow perpendicular to the main cooling flow.

Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite.

State of the Art of the Ignalina RBMK-1500 Safety

Ignalina NPP is the only nuclear power plant in Lithuania consisting of two units, commissioned in 1983 and 1987. Unit 1 of Ignalina NPP was shut down for decommissioning at the end of 2004 and Unit 2 is to be operated until the end of 2009. Both units are equipped with channel-type graphite-moderated boiling water reactors RBMK-1500. The paper summarizing the results of deterministic and probabilistic analyses is developed within 1991–2007 by specialists from Lithuanian Energy Institute. The main operational safety aspects, including analyses performed according the Ignalina Safety Improvement Programs, development and installation of the Second Shutdown System and Guidelines on Severe Accidents Management are discussed. Also the phenomena related to the closure of the gap between fuel channel and graphite bricks, multiple fuel channel tube rupture, and containment issues as well as implication of the external events to the Ignalina NPP safety are discussed separately.