Advanced Gas-cooled Reactor Research Articles

Corrosion Behavior of AGR Simulated Fuels – Evolution of the Fuel Surface

We have prepared a range of Advanced Gas-cooled Reactor (AGR) SIMFUELs at a range of simulated burn ups and, using Raman spectroscopy, have studied the effect of the SIMFUEL dopants on the UO2 crystal structure. We have also studied the effect of deliberately added hydrogen peroxide on the SIMFUEL surface. The intensity of the fundamental U-O stretch (445 cm-1) decreases as the amount of dopant increases in each SIMFUEL burn up composition. A simultaneous increase in the lattice damage (500 – 700 cm-1) peak is observed as the UO2 cubic fluorite lattice structure becomes more distressed and moves towards a tetragonal structure. Exposure to 100 µmol dm-3 H2O2 further decreases the fundamental U-O stretch and increases the lattice damage peak, suggesting that additional point defects are established as the concentration of interstitial oxygen is increased in the lattice via the H2O2 induced corrosion of the SIMFUEL.

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.

Multi-Scale Modelling of Nuclear Reactor Core Graphite

Magnox and Advanced Gas Cooled reactors operated in the UK are cooled by carbon dioxide gas. The graphite moderator bricks in the reactor core lose mass and become more porous during service due to radiolytic oxidation caused by energy deposition, mainly gamma radiation. The microstructure of these graphites comprises filler particles embedded in a graphitised matrix which contains porosity arising from the manufacturing process and subsequent radiolytic oxidation. Computer models have been developed to be representative of the graphite microstructure and these results are used as an input to multi-scale finite element models. These idealised models provide descriptions of the key features that contribute to an understanding of the relationship between the role of filler particles and the amount of porosity on both load-displacement (stress-strain) and fracture characteristics. Two models are considered to span the length-scales: (i) up to millimetre dimensions and (ii) centimetre dimensions. The results are considered with respect to (i) the model length-scale, (ii) the role of the volume percentage of porosity and (iii) the size and distribution of filler particles.

Axisymmetric whole pin life modelling of advanced gas-cooled reactor nuclear fuel

Thermo-mechanical contributions to pellet–clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS’s well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used.

LES of the Flow Inside the Lower Plenum of an Advanced Gas-Cooled Reactor with Conjugate Heat Transfer

The details of flow and conjugate heat transfer in Advanced Gas-cooled Reactors are currently not completely understood because of the reactor complexity and the difficulty to examine the flow while the reactors are in operation. This paper proposes to simulate the flow using Large Eddy Simulation for the turbulence and to compute the conjugate heat transfer by coupling a multi-purpose CFD software with a transient thermal software dedicated to complex solid geometries. Two configurations are used, 1/8th of the full reactor and the full reactor itself. It is shown that simulating 1/8th of the domain only does not allow to reproduce the agitated patterns observed near the centre of the under-dome, where the temperature is the highest.

Characterizing the effects of elevated temperature on the air void pore structure of advanced gas-cooled reactor pressure vessel concrete using x-ray computed tomography

X-ray computed tomography (X-ray CT) has been applied to non- destructively characterise changes in the microstructure of a concrete used in the pressure vessel structure of Advanced Gas-cooled Reactors (AGR) in the UK. Concrete specimens were conditioned at temperatures of 105 o C and 250 o C, to simulate the maximum thermal load expected to occur during a loss of coolant accident (LOCA). Following thermal treatment, these specimens along with an unconditioned control sample were characterised using micro-focus X-ray CT with a spatial resolution of 14.6 microns. The results indicate that the air void pore structure of the specimens experienced significant volume changes as a result of the increasing temperature. The increase in the porous volume was more prevalent at 250 o C. Alterations in air void size distributions were characterized with respect to the unconditioned control specimen. These findings appear to correlate with changes in the uni-axial compressive strength of the conditioned concrete.

Models of coefficient of thermal expansion (CTE) for Gilsocarbon graphites irradiated in inert and oxidising environments

This paper presents the development and validation of an empirical model of radiation effects on coefficient of thermal expansion (CTE) for the Gilsocarbon graphites used in Advanced Gas-cooled Reactors (AGRs). The combined irradiation and oxidation model is based in part on a new model of fast neutron damage in inert environment. The new inert model shows an increase to an “upper shelf” irradiated CTE value at very low dose, then CTE values decrease with increasing dose following a hyperbolic tangent function. The effect of the actual exposure in AGRs is modelled by shifting the inert model in both dose and CTE directions to agree with the CTE measurements on material trepanned from moderator bricks in operating AGRs. The shift in the inert model that is needed to match the trepanned data varies significantly by reactor. The new model predicts randomly-selected validation data that were not used in model fitting as well as it fits the calibration data.

Models of bending strength for Gilsocarbon graphites irradiated in inert and oxidising environments

This paper presents the development and validation of an empirical model of fast neutron damage and radiolytic oxidation effects on bending strength for the moulded Gilsocarbon graphites used in Advanced Gas-cooled Reactors (AGRs). The inert environment model is based on evidence of essentially constant strength as fast neutron dose increases in inert environment. The model of combined irradiation and oxidation calibrates that constant along with an exponential function representing the degree of radiolytic oxidation as measured by weight loss. The change in strength with exposure was found to vary from one AGR station to another. The model was calibrated to data on material trepanned from AGR moderator bricks after varying operating times.

A model of Young’s modulus for Gilsocarbon graphites irradiated in oxidising environments

This paper presents the development and validation of an empirical model of combined fast neutron damage and radiolytic oxidation effects on Young’s modulus for the Gilsocarbon graphites used in advanced gas-cooled reactors. The combined irradiation and oxidation model is based on a model of fast neutron damage in an inert environment and the degree of radiolytic oxidation as measured by weight loss. The change in Young’s modulus with exposure was found to vary from one power reactor to another and with the relative contributions of oxidation and fast neutron damage in data from materials test reactors. The model was calibrated to data from material trepanned from advanced gas-cooled reactor moderator bricks after varying operating times. The model also predicts materials test reactor data that extend beyond the range of currently-available trepanned data. The model is convenient to use because it is an analytic function that eliminates the interpolation on temperature and dose that is used in an earlier model.

Transmission Electron Microscopy Study of Graphite under in situ Ion Irradiation

Graphite is employed as a moderator and structural component in 18 of the UK's fleet of Magnox and Advanced Gas-cooled Reactors (AGRs). During the operational lifetime of a reactor, graphite undergoes complex physical and mechanical property changes including dimensional modification, owing to the effects of temperature, oxidation and irradiation-induced atomic displacements. In order to safely extend the lifetime of the current fleet of AGRs, and also to develop materials for GenIV concepts such as the Very-High-Temperature Reactor (VHTR), it is important to gain a better understanding of the fundamental atomic processes which underpin the behaviour of graphite under current and future operational conditions.This study has focused on the effects of temperature and displacing radiation on the evolution of Mrozowski cracks in highly-orientated pyrolytic graphite (HOPG) using the new Microscope and Ion Accelerator for Materials Investigations (MIAMI) facility. This instrument allows transmission electron microscopy to be performed in situ whilst simultaneously ion irradiating to radiation damage levels typically reached in a reactor. By using this technique, it is possible to explore the development of radiation damage under a range of different conditions continuously from start-to-finish rather than just observing the end-states accessible in ex situ studies.

A Three-Dimensional Methodology for the Assessment of Neutron Damage and Nuclear Energy Deposition in Graphite Components of Advanced Gas-Cooled Reactors

Abstract This paper describes the development of a three-dimensional methodology for the assessment of neutron damage and nuclear energy deposition (or nuclear heating) throughout the graphite cores of the UK’s Advanced Gas-cooled Reactors. Advances in the development of the Monte Carlo radiation transport code MCBEND have enabled the efficient production of detailed fully three-dimensional models that utilise three-dimensional source distributions obtained from Core Follow data supplied by the reactor physics code PANTHER. The calculational approach can be simplified to reduce both the requisite number of intensive radiation transport calculations, as well as the quantity of data output. These simplifications have been qualified by comparison with explicit calculations and they have been shown not to introduce significant systematic uncertainties. Simple calculational approaches are described that allow users of the data to address the effects on neutron damage and nuclear energy deposition predictions of the feedback resulting from the mutual dependencies of graphite weight loss and nuclear energy deposition.

Neutron Flux Measurements in the Side-Core Region of Hunterston B Advanced Gas-Cooled Reactor

Abstract The core restraints of advanced gas-cooled reactors are important structural components that are required to maintain the geometric integrity of the cores. A review of neutron dosimetry for the sister stations Hunterston B and Hinkley Point B identified that earlier conservative assessments predicted high thermal neutron dose rates to key components of the restraint structure (the restraint rod welds), with the implication that some of them may be predicted to fail during a seismic event. A revised assessment was therefore undertaken [Thornton, D. A., Allen, D. A., Tyrrell, R. J., Meese, T. C., Huggon, A.P., Whiley, G. S., and Mossop, J. R., “A Dosimetry Assessment for the Core Restraint of an Advanced Gas Cooled Reactor,” Proceedings of the 13th International Symposium on Reactor Dosimetry (ISRD-13, May 2008), World Scientific, River Edge, NJ, 2009, W. Voorbraak, L. Debarberis, and P. D’hondt, Eds., pp. 679–687] using a detailed 3D model and a Monte Carlo radiation transport program, mcbend. This reassessment resulted in more realistic fast and thermal neutron dose recommendations, the latter in particular being much lower than had been thought previously. It is now desirable to improve confidence in these predictions by providing direct validation of the mcbend model through the use of neutron flux measurements. This paper describes the programme of work being undertaken to deploy two neutron flux measurement “stringers” within the side-core region of one of the Hunterston B reactors for the purpose of validating the mcbend model. The design of the stringers and the determination of the preferred deployment locations have been informed by the use of detailed mcbend flux calculations. These computational studies represent a rare opportunity to design a flux measurement beforehand, with the clear intention of minimising the anticipated uncertainties and obtaining measurements that are known to be representative of the neutron fields to which the vulnerable steel restraint components are exposed.

R.E. Hester and R.M. Harrison (Eds.): Nuclear Power and the Environment

As volume 32 in the RSC series entitled ‘‘Issues in Environmental Science and Technology’’ it follows the general series pattern of editors selecting a range of experienced authors to collate relevant facts that will thoroughly cover the subject matter. Divided into nine chapters, all principal environmental issues in the nuclear industry are dealt with, but there is a distinct UK-centric bias. Each chapter is well referenced, but surprisingly poor cross-indexing makes it difficult to move easily from subject to subject. Globally 436 nuclear power plants have been built, 60 more are under construction and 131 proposed; but now comes the contradiction—the eye-watering estimates for decommissioning, with a prediction of $1 trillion for the USA alone. It is widely assumed that the cost of decommissioning a nuclear generator is the same as its construction cost—but then there is an actual cost, declared in the press but not included in this book. The Windscale advanced gas-cooled reactor (AGR) has just been totally stripped, leaving the building available for reoccupation, at a cost of £110 M—quite a contrast with prediction. It is also claimed the next decommissioning will cost even less. In contrast high level waste (HLW) disposal is in a mess, with most nations continually pushing the problem into the future, exceptions being Sweden and Finland. Sellafield alone has 100 tones of separated plutonium to either use or dispose of; so who will take the decision to store it safely; or better, utilise it as a fuel? The one route clearly not in consideration is to research and build reactors known as accelerator driven systems (ADS) where thorium replaces uranium as the primary fuel. These reactors also consume HL wastes as fuel. Creating what will be in effect a new industrial sector is already well advanced in China and India—but not anywhere in Europe. The lack of a mention of this possible opportunity for European countries is very surprising in a book with this title. With over a century of accumulated knowledge, more is probably known about the effects of radiation on humans than there is about the effects of any other hazard. The average lifetime background dose from natural exposure is 2.2 mSv. In a study of 174,500 workers just 6% had experienced a lifetime dose exceeding 100 mSv, and it was shown statistically that this group had no increase in cancer over that indicated from national rates. There was, however, a slightly elevated risk of leukaemia. In contrast, many European health services restrict mammograms to bi-annual assessment, which have an individual exposure level of 0.4 mSv. This compares to a 0.1 mSv contribution from a 10-h flight, and 16 mSv from a CT heart scan. Bibliography Nuclear Power and the Environment R.E. Hester and R.M. Harrison (Eds.) RSC Publishing, Cambridge, UK, 2011, pp 246 ISBN 978-1-84973-194-2 GBP 65.00

New age nuclear

Since the UK opened its first commercial nuclear power plant at Calder Hall in 1956, nuclear energy has been the cornerstone of the baseload electricity generation. At its peak in 1997 over a quarter of the UK's electricity came from nuclear power, but since that zenith its contribution has declined to today when it delivers just under a sixth of our requirements. There are currently 19 nuclear reactors in operation in the UK at ten plants - seven are Advanced Gas Cooled (AGR) reactors, two are Magnox reactors and the final is a Pressurised Water Reactor (PWR) - the chosen design for future plants. With the final two Magnox plants, along with two more AGRs scheduled to end in 2016, there was a pressing need to shore up electricity generating capacity to avoid shortages. In 2005 Tony Blair's government began an energy review, which led to the 2006 Energy White Paper making a commitment to put the UK on the path to a nuclear renaissance. Despite a change of government, that commitment to nuclear energy is unwavering with the stated aim from Charles Hendry, minister of state for the Department of Energy and Climate Change, to put the UK country back on the nuclear map. However, when a massive earthquake and ensuing tsunami damaged the nuclear plant at Fukushima in Japan, it appeared that the rush to build new nuclear plants in the UK would need to be reappraised. Germany, Italy and Switzerland were quick to put an end to any nuclear aspirations they held, but the UK took its time, assessing the impact of the accident. Mike Weightman, chief inspector of nuclear installations and head of the office of nuclear regulations, was commissioned by the government to report on the implications of the accident on the UK nuclear new build programme. In his final report, released late in 2011, he remained confident that UK nuclear facilities have no fundamental safety weaknesses

Industrial implementation of intelligent system techniques for nuclear power plant condition monitoring

As the nuclear power plants within the UK age, there is an increased requirement for condition monitoring to ensure that the plants are still be able to operate safely. This paper describes the novel application of Intelligent Systems (IS) techniques to provide decision support to the condition monitoring of Nuclear Power Plant (NPP) reactor cores within the UK. The resulting system, BETA (British Energy Trace Analysis) is deployed within the UK’s nuclear operator and provides automated decision support for the analysis of refuelling data, a lead indicator of the health of AGR (Advanced Gas-cooled Reactor) nuclear power plant cores. The key contribution of this work is the improvement of existing manual, labour-intensive analysis through the application of IS techniques to provide decision support to NPP reactor core condition monitoring. This enables an existing source of condition monitoring data to be analysed in a rapid and repeatable manner, providing additional information relating to core health on a more regular basis than routine inspection data allows. The application of IS techniques addresses two issues with the existing manual interpretation of the data, namely the limited availability of expertise and the variability of assessment between different experts. Decision support is provided by four applications of intelligent systems techniques. Two instances of a rule-based expert system are deployed, the first to automatically identify key features within the refuelling data and the second to classify specific types of anomaly. Clustering techniques are applied to support the definition of benchmark behaviour, which is used to detect the presence of anomalies within the refuelling data. Finally data mining techniques are used to track the evolution of the normal benchmark behaviour over time. This results in a system that not only provides support for analysing new refuelling events but also provides the platform to allow future events to be analysed. The BETA system has been deployed within the nuclear operator in the UK and is used at both the engineering offices and on station to support the analysis of refuelling events from two AGR stations, with a view to expanding it to the rest of the fleet in the near future.

Effects of Displacing Radiation on Graphite Observed Using in situ Transmission Electron Microscopy

ABSTRACTGraphite is used as a moderator and structural component in the United Kingdom’s fleet of Advanced Gas-Cooled Reactors (AGRs) and features in two Generation IV reactor concepts: the Very High Temperature Reactor (VHTR) and the Molten Salt Reactor (MSR). Under the temperature and neutron irradiation conditions of an AGR, nuclear-grade graphite demonstrates significant changes to it mechanical, thermal and electrical properties. These changes include considerable dimensional change with expansion in the c-direction and contraction in the a/b-directions. As the United Kingdom’s AGRs approach their scheduled decommissioning dates, it is essential that this behaviour be understood in order to determine under what reactor conditions their operating lifetimes can be safely extended.Two models have been proposed for the dimensional change in graphite due to displacing radiation: the “Standard Model” and “Ruck and Tuck”. The Standard Model draws on a conventional model of Frenkel pair production, point defect migration and agglomeration but fails to explain several key experimental observations. The Ruck and Tuck model has been proposed by M.I. Heggie et al. and is based upon the movement of basal dislocation to create folds in the “graphene” sheets and seeks not only to account for the dimension change but also the other phenomena not explained by the Standard Model.In order to test the validity of these models, work is underway to gather experimental evidence of the microstructural evolution of graphite under displacing radiation. One of the primary techniques for this is transmission electron microscopy with in situ ion irradiation. This paper presents the results of electron irradiation at a range of energies (performed in order to separate the effects of the electron and ion beams) and of combined electron and ion beam irradiation.

Three-dimensional simulation of a simplified advanced gas-cooled reactor fuel element

The present work involves simulations of a simplified three-dimensional representation of the UK fleet of advanced gas-cooled reactors (AGRs) fuel element using a 30° sector configuration. The computations were carried out using the v 2– f formulation which was shown to be one of the most accurate turbulence models in earlier simulations of two-dimensional rib-roughened channels. In the present work main features of the mean flow and heat transfer in the fuel element were identified and discussed. The pressure loss and friction factor were also calculated where good agreement with the experimental correlations was found. Further comparisons were made against simulations of a 2D rib-roughened channel in order to assess the validity and relevance of a ‘2D approximation’ approach. It was shown that although a two-dimensional approach is very useful and economical for ‘parametric studies’, it does not provide an accurate representation of a 3D fuel element configuration, especially for the velocity and pressure coefficient distributions, where large discrepancies were found between the results of the 2D channel and azimuthal planes of the 3D configuration.

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Knowledge-directed characterization of nuclear power plant reactor core parameters

A major life-limiting factor of the UK's Advanced Gas-Cooled Reactors (AGRs) is the condition of the graphite core. Installation of new measurement equipment is difficult and expensive, therefore maximizing the information gained from existing equipment is highly desirable. The main approach to determining the health of an AGR core is through periodic inspections undertaken during planned outages. However, there is the desire to supplement this inspection activity through the analysis of data gathered as part of routine plant operation. One such source of data is measurements taken during refueling and this paper describes knowledge-directed characterization of this refueling data, both spatially across the reactor core and temporally across the operational lifetime of the core. Characterization provides information relating to the current condition of the reactor core and allows suspected ageing trends to be visualized and confirmed. A standard approach for characterizing reactor core data is presented and applied to a variety of different reactor core parameters. The benefit of this approach is that it allows engineers to distill large volumes of refueling data into a readily understandable format in a short period of time. It also allows hypothesized trends relating to the ageing process within the core to be tested and provides supporting evidence for these hypotheses. The trending data is also valuable as it can form the basis of a predictive model of ageing of the reactor core. The ageing process of nuclear graphite is understood from theoretical and experimental viewpoints and this empirical data, gathered from operating reactors, further supports this understanding. This paper represents the initial exploration of using refueling data to construct a predictive model of AGR reactor core ageing.

Non-contact characterisation of conductivity gradient in isotropic polycrystalline graphite using inductance spectroscopy measurements

The two main graphite ageing processes in advanced gas-cooled reactors (AGRs) are fast neutron damage and radiolytic graphite oxidation. These processes change the properties of the graphite core and lead to a reduction in the graphite density, ie a weight loss. Electrical conductivity measurement of the graphite material is among the possible methods for estimating the weight loss and hence provides a measure of the graphite ageing. The inductance spectroscopy technique using eddy current sensors has been implemented in a wide variety of NDT applications, since it enables the possibility of extracting the electrical parameters variation from eddy current measurements and examining more fully the internal material structure. In this paper we present a non-contact eddy current method for determining non-destructively the electrical conductivity gradient through 100 mm-thick graphite sections using inductive spectroscopic measurements collected by a gradiometer. The method employs the commercial finite element software COMSOL to generate the data from a model of a real profile distribution. This data is translated to the profile conductivity of the graphite test object via the solution of an inverse problem. The linear Tikhonov regularisation method and the non-linear regularised Gauss Newton technique are employed in the solution of the inverse problem. During the optimisation, the forward problem is evaluated using COMSOL. Numerical optimisation tests have been carried out for different cases of step conductivity profiles. Initial results using simulated data show that representative estimates of the conductivity profile of the modelled graphite sections have been obtained.