The grand harbour hotel

Wednesday 11^th May 2016
07h30: Breakfast (Boardwalk Restaurant)
08h00: Conference Service Desk Opens, Speedwell Room
08h20: Announcements Mayflower Suite
PROJECT BLACKSTONE Chairman: Prof. G. Peter Matthews
08h30: Materials Test Reactor Programme for the Investigation of Graphite Irradiation Creep and Annealing

T.O. van Staveren, J.A. Vreeling and A.J. de Koning

This paper provides an update on the graphite irradiation creep materials test reactor programme that NRG executes to support the Advanced Gas-Cooled Reactor life- time extension program of EDF Energy. The key aim of this experiment is to generate data on the creep and annealing behaviour of (radiolytically oxidised) graphite. Four irradiation stages of the graphite irradiation creep programme have been completed.
The ACCENT creep experiment consists of a stack of modules in which paired stressed and un-stressed graphite samples are placed. Samples are divided over six modules that are loaded in one capsule for irradiation in a high flux position of the HFR. A stress of 5 or 10 MPa is applied to half of the samples by gas-filled bellows. For phase 3 of ACCENT, a dedicated anneal module is used to irradiate previously stressed and their un-stressed reference specimen. The samples are characterised between each irradiation stage.
Both pre-irradiated and virgin graphite are used as source material from which the samples are machined. The pre-irradiated samples were radiolytically oxidised in an AGR reactor core and in the High Flux Reactor (HFR) in Petten, the Netherlands.
This presentation will provide an update on the ACCENT project, highlighting key results and the future steps that will be taken as part of the project.
08h50: Blackstone MTR Phase 3: Characterisation of Graphite Material Property Changes at Heysham 2 and Torness AGR Power Stations

B.C. Davies, M. Brown, P.C. Matthews, M.R. Bradford, I. Foreshaw and J. Reed

Project Blackstone is EDF Energy’s MTR experiment that provides information on material property changes due to fast neutron irradiation and radiolytic oxidation ahead of the operating AGR reactors. The project is delivered by NRG in the Netherlands and has been ongoing since around 2006. The primary objective is to produce and characterise gilsocarbon graphite with the high weight loss and fast neutron fluence that are expected at AGR end of life conditions. Phase 2 was completed in 2015 and the results will be used in analyses that support AGR graphite core safety cases.
Heysham 2 and Torness are the youngest AGR stations in the fleet and are expected to operate to a later date than the other AGRs. Only a limited amount of material from these stations was investigated in Blackstone Phases 1 and 2. This paper will present EDF Energy’s objectives for Blackstone Phase 3 and the intended use of the data to support the operation of these two stations. A summary of the project scope and schedule of work will be provided.
INSPECTION Chairman: Dr. Mark Bradford
09h10: A Success Story: In-Situ Robotic Graphite Moderator Block Repair

D. Gurin and E. Petit de Mange

Diakont has successfully conducted a programme of development, qualification, and field operation of specialised robotic tooling which inspected and repaired the graphite moderator blocks and fuel channels of the RBMK reactor Units 1 and 2 at the Leningrad Nuclear Power Plant. Cracking and deformation had occurred in these blocks, resulting in deformation of the fuel channels. Remediation was required for continued safe reactor operation and for facility license extension. The output of the 2012-2015 program was that Diakont safely and successfully performed the following inspection and repair functions remotely, all while the reactor remained fuelled:

• Crack removal through horizontal and vertical longitudinal variable-azimuth block cutting;

• Drilling of calibrated holes which unloaded internal stresses, stopping crack growth;

• Edge-squaring of the block mating faces;

• Mechanical realignment and straightening of the fuel channels;

• Generation of compensating loads upon the graphite stack;

• Collection and processing of the resulting radioactive graphite swarf and dust;

• Rad-hardened laser-visual inspection of block positioning and dimensions;

• Multi-axis gyroscopic measurement of fuel channel spatial curvature.
As a result of this repair programme, the properties of the graphite stack and the reactor component alignment in both reactor units have been confirmed restored, allowing for continued safe operation and life extension.
09h30: Estimating Fuel Channel Bore from Fuel Grab Load Trace Data

3. 
   Berry, D. Pattison, G.M. West, S.D.J. McArthur and A. Rudge
   

Detailed measurements of the graphite core fuel channels are made by specialist inspection equipment during planned outages, typically every 18 months to 3 years. The bores of the graphite fuel bricks are obtained during these inspections and are used to provide important information about the health of the core. Additionally, less detailed online monitoring data is obtained much more frequently during refuelling events, called the fuel grab load trace (FGLT), which can be also used to infer the health of the graphite core. This paper describes the process of creating a model which isolates a component of the refuelling data and maps it directly to dimensional measurements of fuel channel bore. The model is created from a combination of the theoretical understanding of the physical interactions of the fuel stringer during refuelling events and several years of refuelling and inspection data to estimate suitable model parameters. Initially the model created was a coarse estimation of FGLT to fuel bore dimension but through refinements a much more accurate model has been created. An application of this model is shown through a case study of a recent outage where estimations were made on refuelling data and were compared to previously unseen inspection data.

09h50: Automated Video Processing and Image Analysis Software to Support Visual Inspection of AGR Cores

Paul Murray, Graeme West, Kristofer Law, Stephen Buckley-Mellor, Graeme Cocks and Chris Lynch

Remote visual inspection of fuel channels in Advanced Gas-cooled Reactor (AGR) cores provides nuclear operators with an understanding of the condition of the UK’s fleet of nuclear power plants. During planned, periodic outages, specialist inspection tools equipped with video cameras and other sensors are manipulated inside fuel channels selected for inspection and a video of the entire channel bore is recorded for each. If cracks are observed in this process, a montage of the entire crack region needs to be: produced, analysed and sentenced (classifying the crack morphology, location, orientation and size) before the station is returned to service – provided it is safe to do so.
At the present time, the video analysis and crack montage production is done manually by an expert team of inspection engineers. In line with this process, bespoke image stitching software named “ASIST” (Automated Software Image Stitching Tool) has been trialled in the last 12 months and evaluated using data from: Dungeness, Hunterston B, Hinkley Point B, Heysham 1 and Torness outages. The software is now almost ready to replace the manual process and will provide higher quality images with 100% channel visualisation properties in a fraction of the time taken by the current approach.
This paper provides a summary of the ASIST evaluation undertaken in the last year. It also describes recent research endeavours aiming to provide ASIST with: crack detection techniques; keyway locating algorithms and methods to compute Structure-from-Motion which will facilitate the extraction of 3D depth information directly from the 2D video footage.
10h10: In-situ Testing of Nuclear Graphite at High Temperature

Dong Liu, Bernd Gludovatz, Claire Acevedo, Harold Barnard, Bryan Roebuck, Martin Kuball and Robert O. Ritchie

This paper provides an update on the most recent advances made in the high temperature in situ tests of Gilsocarbon graphite including high temperature tomography, high temperature Raman spectroscopy mapping and thermal-electro mechanical tests.
Synchrotron radiation x-ray computed micro-tomography was undertaken at the Advanced Light Source of the Lawrence Berkeley National Laboratory. A unique in situ ultrahigh temperature tomography rig that permits real-time investigation of damage evolution under load at temperatures up to 2000°C was adopted. Gilsocarbon specimens with dimensions similar to those from trepanned samples were tested, typically 4x4x20 mm, by a three-point bending configuration. Both plain and notched specimens were studied for strength and fracture toughness at room temperature, 650°C and 1000°C, respectively. A full X-ray tomography scan was performed at each step as the specimens were incrementally loaded. These tests have revealed for the first time the three-dimensional deformation and fracture of the nuclear graphite at temperature. In general, the strength and fracture toughness of Gilsocarbon graphite was found to increase with temperature. To understand the underlying physical mechanisms, high temperature Raman spectroscopy mapping (40x40 µm) of the graphite surface were undertaken in a hot cell (with Ar atmosphere) at the Centre for Device Thermography and Reliability, University of Bristol, at room temperature and at 800°C. We found indications that the amplitude of residual strain reduces at high temperature; this could be one of the contributing factors to the nominal high strength measured at temperature. Electro-thermal uniaxial tension tests on Gilsocarbon graphite at the National Physical Laboratory indicate that the resistance of the material changes with temperature and applied load. Results from the three techniques will be discussed with respect to the understanding of the high temperature behaviour of nuclear graphites.
10h30: Coffee break (Mayflower Foyer)
10h50: Characterisation of Gilsocarbon Graphite by Pycnometry and Surface-Area Analysis

K.L. Jones, G.M. Laudone and G.P. Matthews

The oxidation and consequent weight loss within Gilsocarbon graphite depends on pore-level processes that are not explicitly described within the current EDF finite element model (FEAT-DIFFUSE6). In particular, the probability of de-activation of oxidising species generated by the ionising radiation depends on pore size. The characterisation of pore sizes is challenging for several reasons, including the large range of size scales from micropores (<2nm) to macropores (>50nm), the topological complexity at each scale, and the small size of Blackstone samples that have been trepanned and irradiated in the Petten materials test reactor. A further complication is that some pores are closed, with no access to the flowing carbon dioxide gas and inhibitors including methane, and therefore have different oxidation characteristics.

Plymouth University has taken a multi-faceted approach to the measurement of pore sizes based on pycnometry, surface area measurement coupled with Grand-Canonical Monte-Carlo modelling, mercury porosimetry and void network modelling. This presentation focusses on the first two of these, both of which have required specialised apparatus because of the small size of the samples. The results are discussed within the context of current postulates about the change in void size with ageing, specifically pore erosion and coalescence.

11h10: Advances in Electronic Speckle Pattern Interferometry for the Measurement of the Coefficient of Thermal Expansion of Nuclear Graphite

John Dinsdale-Potter, Paul Ramsay and Nassia Tzelepi

The Coefficient of Thermal Expansion (CTE) gives a key insight into the thermomechanical behaviour of graphite representing the graphite components’ response to temperature change and providing an indication of the strains experienced in the reactor core. Both are of these properties are integral to reactor life predictions.

This paper details the development progress and objectives of the Electronic Speckle Pattern Interferometry (ESPI) CTE programme. The main aims of the project are to improve the throughput of CTE measurements of irradiated samples and to confirm the accuracy of the existing Post Irradiation Examination (PIE) method using dilatometers. The paper discusses the inherent challenges of measuring small volumetric changes with non-compliant sample geometry and presents the results of the validation programme.

11h30: Super Articulated Control Rod for Ageing AGR Graphite Reactor Core

Hassan Salih and Neil McLachlan

The design intent of the Advanced Gas Cooled Reactor (AGR) graphite core is for the core channels to remain sufficiently straight that the movements of the control rods, to regulate and shut down the reactor, will not be impeded. Due to the irradiation induced ageing effects the core channels will undergo a degree of distortion, which could ultimately impair the movements of the control rods.
The HPB/HNB AGR reactors have Articulated Control Rods (ACR) comprising eight sections with seven articulating joints. Increasing the number of sections, and the articulating joints, will lead to a more flexible control rod that should negotiate a distorted channel more easily. This is the design concept of the Super-Articulated Control Rod (SACR).
The paper will demonstrate the benefits arising from the adoption of the SACR design. Solid 3D Finite Element analyses, using ABAQUS/Explicit, were carried out to simulate the contact interaction of the SACR with the bore of a distorted core channel. The FE results were validated against data from a corresponding test rig. It was found that the SACR can negotiate a far more distorted core channels in comparison with the original ACR design.
11h50: Assessment of Graphite Material Property Evolution

1. 
   Moore, T. Kirwan, M. Joyce, B.C. Davies and M.R. Bradford
   

The Core Component Condition Assessment leg of the Advanced Gas-cooled Reactor (AGR) safety case requires prediction of the structural integrity of core components. In turn, these predictions require a further model to describe the evolution of graphite material properties in response to long term reactor service. The methods to develop and calibrate this material model have been significantly extended in recent years to make better use of the volume of data from both material test reactors and inspections of the AGRs themselves.

This paper will present the most recent description of material behaviour inferred from channel bore distortion data, including ovalisation, for both the general AGR fuel brick population, as well as techniques for obtaining further information from the measurements of specific, repeat inspected, components.
12h10: Lunch (Boardwalk Restaurant)
13h00: Coaches depart for Beaulieu
17h00: Coaches depart from Beaulieu

19h30: Drinks Reception (Mayflower Foyer)

20h00: Conference Banquet (Mayflower Suite)

Dress: Business casual
Thursday 12^th May 2016

REMINDER: ROOMS MUST BE VACATED BY 10am: secure luggage storage at Main Reception or Conference Office (Speedwell Room)
07.30: Breakfast (Boardwalk Restaurant)
08.00: Conference Service Desk Opens, Speedwell Room
DIFFUSE 6 Chairman: Mr Matthew Bamber
08.30: Revisiting AGR Graphite Weight-Loss Predictions: A First-Principles Review of the Basis and Application of the FEAT-DIFFUSE Code

A. J. Wickham

With the operation of the oldest UK Advanced Gas-Cooled Reactor reaching 40 years, and the onset of limited through-block cracking in graphite core components, the increasing difficulty of predicting precisely the extent and distribution of radiolytic weight loss within individual components and around the core potentially stands in the way of further extensions of their operating lives.
Weight loss, equating to a loss of density and a potential reduction in strength, interacts strongly with changes in thermal, mechanical and physical properties arising from fast-neutron irradiation, all of which play a role in determining stress distribution and potential cracking within the components.
This presentation examines the application of the predictive code FEAT-DIFFUSE to the four different reactor designs, ‘stripping down’ the predictive process to its basic input parameters, seeking to identify uncertainties, assumptions and unknowns in the underlying radiation chemistry, the evaluation of radiation dose, the diffusion and permeation of reactor-coolant (reactant) gas through the component porosity, and features which might cause individual reactors to deviate from the general behaviour.
A combination of modern statistical treatment of inspection data and an ability to ‘tune’ theoretical models on the basis of a comparison of prediction of the next sampling results with their actual data has so far allowed continuing operational safety cases to be made with a sufficient margin of safety. Now, as the reactors approach the last phase of their operational lives, it is time to re-examine the underlying basic research and to derive a competent understanding of the uncertainties which remain in the predictions in order to manage the concluding phase of power generation with the appropriate margin of safety.
08h50: Updating the FEAT-DIFFUSE6 Brick Boundary Conditions for Graphite Weight-Loss Predictions at Dungeness B

J. Petherick

A review of trepanning data for Dungeness B revealed a slight shortfall in predicted radial weight loss profile. This lead to a structured reassessment being carried out on the pressure and coolant boundary conditions that are applied within the FEAT-DIFFUSE6 finite element calculations of weight loss. This included a review of the reactor drawings and flow network calculations combined with a parametric study of the weight loss profile predictions with the optimum conditions defined using “goodness of fit” techniques. The development results in a significant improvement in radial weight loss profile prediction at Dungeness B. This paper presents an overview of the process that has been followed and of the improvement in the predictions for trepanned samples.
CRACKING (2) Chairman: Ms. Nassia Tzelepi
09h10: Shaking Table Testing of an Advanced Gas Cooled Reactor Core Model with Cracked Array Configurations

L. Dihoru, O. Oddbjornsson, M. Dietz, T. Horseman, P. Kloukinas, E. Voyagaki, A. J. Crewe, C. A. Taylor and A. G. Steer

The consequences of the ageing of the Advanced Gas Cooled Reactors (AGRs) need to be assessed for their continued safe and reliable operation. This includes consideration of their seismic behaviour. Differential irradiation shrinkage and thermal stress changes in graphite components cause cracking that affect the core geometry and structural functionality of its keying system. The ageing issues for seismic resilience are investigated using large complex computational models. To extend their validation envelope, a physical model using quarter sized components in which various patterns and percentages of simulated cracked bricks can be embedded has been developed by the University of Bristol. This model has been used to provide experimental data that can be used for direct comparison with computer model predictions. This paper presents the physical model and the experimental design associated with the work. Intact and two cracked array configurations have been tested on a shaking table. The displacement and the acceleration responses inside the array have been measured for various dynamic input magnitudes and orientations. Relevant test results of component displacements, channel profiles and cracked brick separation are presented that show that the model rig is capable of exploring array responses to dynamic inputs in a consistent and robust way.
09h30: Investigations into Prompt Secondary Cracking in AGR Graphite Fuel Bricks using the eXtended Finite Element Method

O.A. Booler, I.J. Slater, S. Baylis and A.G. Steer

Assessments of the stability of keyway root cracks in AGR graphite fuel bricks have shown that the energy released by the fracture of the final ligament during the propagation of a primary axial crack was a significant proportion of the total energy released. The resultant kinetic energy from such an event causes stress waves that could constructively interfere in the opposite keyway corner, which may lead to prompt secondary cracking and consequential structural performance issues within the graphite core. To demonstrate this phenomenon numerically, the dynamic response of the final fracture event must be taken into consideration, followed by secondary crack propagation using the eXtended Finite Element Method (XFEM). However, in ABAQUS the XFEM framework is only available for static analyses; hence issues arise when taking into consideration the dynamic effects from the primary crack. Here, a method for simulating a dynamic stress field in static equilibrium is presented by applying d’Alembert’s principle and the conservation of linear momentum. The implications of the dynamic mechanisms on prompt secondary cracking are also discussed.
09h50: A Small-Specimen Notched Strength Test of Gilsocarbon

M.S.L. Jordan, S. Wilkinson, M. Haverty, A. Tzelepi, D. Nowell and T.J. Marrow

During inspections of Advanced Gas-cooled Reactors (AGRs), measurements of the physical and mechanical properties of the moderating graphite core are routinely conducted. Notches, such as the stress-concentrating keyways, weaken components, but notch sensitivity is not measured in these inspections because there is no established methodology for small test specimens of graphite. Current structural integrity assessments and forward simulations assume that radiolytic-oxidation has no effect on graphite notch sensitivity. A notched 6 x 6 x 19 mm beam test geometry is proposed, loaded by four-point bending, to measure notch sensitivity in graphite small specimens; this non-standard geometry requires a rigorous validation process, given that the microstructural scale (i.e. filler particle size) is greater than one sixth of the sample’s smallest dimension.
Notch geometry effects from the fracture testing of over 100 virgin graphite samples are reported, with further validation and a deeper mechanistic understanding developed by combining finite element modelling, X-ray tomography (XCT) and digital volume correlation (DVC). As part of the methodology validation for reactor-extracted material, an initial assessment of the baseline error from an XCT-DVC analysis has been performed on tomographs of trepanned radiolytically-oxidised Gilsocarbon at different weight losses.
10h10: Fracture of AGR Graphite using Bromine Intercalation

W. Bodel, P. Martinuzzi, B. Davies, P. Mummery and A. Steer

While recreating external loading of AGR components within laboratory conditions is relatively straightforward, a method of recreating the internal stresses experienced by AGR fuel bricks during their operational lifetime is a more desirable and representative alternative. Irradiating specimens for subsequent analysis is tricky and would likely involve access to a materials test reactor. Previous attempts at generating internal stresses using thermal means have been restricted by difficulties in generating temperature gradients large enough to result in fracture, therefore an alternative is desirable.
This work applies internal stresses using intercalation (the introduction of guest ions/molecules between atomic planes in layered host structures); in this case intercalation of bromine into graphite. During intercalation, the graphite layers are forced apart by the presence of the guest ions, simulating the swelling that occurs in graphite under significant levels of neutron irradiation.
X-ray Computerised Tomography (XCT) is carried out on samples of nuclear grade Gilsocarbon graphite which are immersed in a bromine atmosphere at the Manchester X-ray Imaging Facility (MXIF). Previous work focused on fast scans to observe crack initiation in small specimens; present work is concentrated on larger specimens with more detailed scanning conditions.

10h30: Coffee break (Mayflower Foyer)

10h50: Feedback from Breakout Sessions Chairman: Prof. Peter Flewitt

Reports from each of the breakout sessions, and further discussion.
CRACKING (3) Chairman: Mr. Alan Steer
11h35: The Application of 3D Solid Finite Element Modelling and the Effect of Cracked Bricks in Support of Hinkley Point B and Hunterston B Safety Cases

S. Bazell and N. McLachlan

As the AGR graphite core ages, the components of the core change shape, lose weight and the fuel bricks are postulated to crack. As a consequence, the geometry of the core will alter from the original design; therefore, there is a need to determine the distorted geometry in order to demonstrate the safety functions of the core.
SALCOR is a whole core modelling and analysis methodology, using 3D solid finite elements in the ABAQUS Finite Element Analyses (FEA) software, which has been developed to model the graphite components of an AGR core. A User Defined Material is used to model the graphite material properties, whilst Field Variables are used to apply the relevant dose etc.
A set of FEA models, representative of regions of up to four Hinkley Point B and Hunterston B layers and including Singly Cracked Bricks and Doubly Cracked Bricks, have been built and analysed to investigate the effects of cracked bricks. In particular, the analysis addresses the question of what happens to the different keys when the bricks have cracked and whether the keys can disengage.
This paper presents some of the analysis work undertaken in support of the Hinkley and Hunterston Safety Cases.
11h55: SALCOR Solid Modelling and Analysis Methodology for an Ageing AGR Core Progress and Benefits

Hassan Salih and Neil McLachlan

The Advanced Gas Cooled reactor (AGR) graphite core is made up of many large and small graphite bricks, arranged in vertical columns, and keyed together via an axial and radial keying system. The AGR core graphite components exhibit dimensional changes when subjected to irradiation, and hence the geometry of the core will alter from its original design.

The distorted geometry of the core is usually predicted using a stick and spring elements modelling approach, which has been validated against test rig data. An alternative methodology (SALCOR) based on FE solid modelling has been developed. The aim is to minimise the conservatism inherent in the stick and spring methodology and to account for simplified features and core behaviour.

The paper will outline the progress and the methods developed to enable the application of SALCOR to arrangements of the HPB/HNB reactor core. This will include solid modelling of the core components, assembly into core arrays of different sizes, simulation of the opening of axially singly, shear and separation of doubly cracked fuel bricks. The approach uses a UMAT with field variables, contacting surfaces with friction and a numerical stabilisation scheme of the ABAQUS solution.
12h15: SALCOR Simulation of Local AGR Core Region with Multiple Singly and Doubly Cracked Fuel Bricks and Potential Key Disengagement

Hassan Salih and Neil McLachlan

The SALCOR (Solid Analysis of Loaded CORe) methodology is aimed at increasing the understanding and reducing the uncertainty of behaviour of ageing AGR graphite cores. This is facilitated by use of solid modelling for the shapes of the core components, and a UMAT with irradiation and weight loss field variables for defining the dimensional changes. The SALCOR methodology, in which core component interactions is modelled explicitly using contacting surfaces, is valid before, during and post potential key disengagement from a fuel brick keyway.
SALCOR is used to clarify the issue of potential key disengagement as predicted by an AGRIGID analysis of HPB/HNB whole core with multiple axially singly (SCB) and doubly (DCB) cracked fuel bricks. This is done via sub-modelling of the problematic region using the solid modelling and analysis methodology of SALCOR.
The sub-model, with 31 SCBs and DCBs is analysed for the combined loading of gravity, irradiation dimensional changes and SCBs opening using the UMAT, gas differential pressure and boundary displacements from the AGRIGID analysis. The results from the study and the use of the pictorial visualisation of the SALCOR model demonstrated that local key disengagement is unlikely to occur.

12h35: Lunch (Boardwalk Restaurant)

13h20: GCORE Validation for Large Arrays of Radially Keyed Bricks: Comparison of the Experimental Measurements and Computational Predictions

H. Riley, S. Corrigan, T. Horgan, H. Graham and A.G. Steer

To ensure safe and reliable operation of the UK’s Advanced Gas-cooled Reactors (AGRs), seismic qualification of the graphite cores is required to demonstrate safe shut down during an infrequent seismic event. Seismic assessments of the AGR cores use a finite element modelling approach called GCORE. The GCORE methodology has undergone validation based on dynamic response tests with small arrays of intact keyed bricks and static loading tests on large arrays containing simulated cracked bricks. To extend the validation of the GCORE methodology to the dynamic response of large arrays with cracked bricks, a shaking table experiment with a multi-layer keyed array (MLA) test rig with quarter-sized components based on late in life core geometry is been undertaken at the University of Bristol.
This paper compares GCORE predictions against the measured dynamic responses for arrays containing intact and simulated double axially cracked bricks. In particular, the effects of cracked bricks on the dynamic response of the array and their propensity to separate, which could in turn make key disengagement a possibility, is investigated. The comparisons demonstrate that GCORE delivers reasonably accurate predictions for simulated seismic excitations covering a wide range of input magnitudes and frequency contents.
13h40: GCORE Seismic Analysis of AGR Cores: Development of Modelling and Analysis Methods for Later in Life Core Conditions

G. Cowell, J. Fleet, T. Horgan, I. Houghton, N. Reid, A.G. Steer and B. Wedd

As the graphite cores of the Advanced Gas-cooled Reactors (AGRs) age, they experience radiation induced weight loss and shrinkage. These act to weaken the graphite bricks and cause bricks to crack. As this happens there is the possibility that the graphite core behaviour could alter to challenge the fundamental nuclear safety requirements for the AGRs. To ensure safe and reliable operation of the UK’s AGRs, seismic qualification of the graphite cores is required to demonstrate safe shut down during an infrequent seismic event. Seismic assessments of the AGR cores use a finite element modelling approach called GCORE.
The GCORE modelling method has been developed to represent more accurately core seismic behaviour for later in life conditions. These are predicted to include a large number of single axially cracked bricks with varying sizes of crack opening. As the core ages and components weaken, the GCORE analyses predict that multiple components in the graphite keying system will be overloaded during a seismic event. The various methods introduced into the GCORE analysis methodology to deal with the predicted cracking extents and seismically induced damage are presented in this paper.
14h00: Cracked Graphite Brick Interaction Analysis: An Application to the UK AGR Graphite Core

H. Teng, I.J. Slater, C.J. Jones, N. McLachlan and J. Wright

In some of the UK’s Advanced Gas-cooled Reactors (AGR), some cracking of the graphite core components has been observed, as the components continue to age with irradiation. It is essential to understand the cracking behaviour, particularly keyway root cracking, to help underwrite the safety cases for operation of the cores to their ultimate lifetimes, and enable EDF Energy to take informed lifetime investment decisions. A 3D 3x3x3 Finite Element (FE) array model (3 layers, 3 rows, and 3 columns) named as the Cracked Brick Neighbourhood Array (CBNA) has been recently developed for this purpose.
The CBNA model has been successfully used to investigate the interactions for a number of cracked brick configurations in the HPB/HNB graphite core. A number of analyses with different brick geometries and crack locations/insertion times are carried out to provide results including the contact loads on the end-face key/key way, the stresses at the bore, the keyway roots, and the crack opening etc., so that the likely modes and sequences of cracking in the presence of a singly axially cracked brick or multiple cracks can be established.

This paper presents indicative results obtained from the CBNA model for the following reactor operating conditions:

• At-power

• Shutdown

and some discussions are provided.
14h20: Modelling Crack Initiation on AGR Graphite Bricks

3. 
   Leguillon, V.X. Tran, P. Massin, P. Martinuzzi and D. Geoffroy
   

During the operating life of an Advanced Gas-cooled Reactor, the core, made of thousands of graphite bricks, is subjected to internal stresses arising because of dimensional changes due to irradiation. It may lead to crack initiation and propagation especially in keyway corners.

Using a matched asymptotic approach, a novel tool for predicting crack initiation in notches has been developed in the University Pierre and Marie Curie Paris 6 (France). This method takes into account both stress and energy conditions. Depending on the stress state and the geometry, crack initiation is mostly driven by one or the other. It is based on a critical value of the leading generalised stress intensity factor at the notch which arises to be the relevant parameter to predict the crack onset. They are extracted from a finite element solution and then compared to the critical value provided by the coupled criterion as a function of the material toughness and tensile strength. The method also determines the length of the initial crack as well as its direction. This method has been validated and compared on various experiments on brittle materials with many different notches. It has also been tested on experiments performed on non-irradiated graphite bricks.

14h40: SOLFEC Validation for Large Arrays of Radially Keyed Bricks: Comparison of Experimental Measurements and Computational Predictions

2. 
   Cannell, S. Brasier and N. McLachlan
   

To ensure safe and reliable operation of the UK’s Advanced Gas-cooled Reactors (AGRs), seismic qualification of the graphite cores is required to demonstrate safe shut down during an infrequent seismic event. Seismic assessments of the AGR cores use a finite element modelling approach called GCORE, modelling the behaviour of the graphite core components. The GCORE model represents graphite components using rigid elements and the interactions between keying systems using springs, loose keys are not modelled as bodies. As such, whilst GCORE is capable of predicting the possibility of key disengagement, the methodology is not currently capable of predicting behaviours following key disengagement.

The experimental computational code, SOLFEC, is designed to simulate multi-body contact dynamics of large systems. It provides the capability to explicitly model components within a graphite core during a seismic event and predict the consequence component interactions beyond the capability of GCORE. Since SOLFEC is an experimental computational code, examples of its usage and existing validation is limited. As such, this paper focuses on the validation of SOLFEC predictions against the measured dynamic responses for large arrays containing intact and simulated double axially cracked bricks produced during rig tests at the University of Bristol.
15h00: Effects of Singly Cracked Bricks on the Deformation of Neighbouring Channels in an AGR Graphite Core

M. Treifi, B.J. Marsden, G.N. Hall and P.M. Mummery

The graphite core of an Advanced Gas-cooled Reactor (AGR) changes over time due to aging caused by fast neutron damage. This may lead to cracking of some of the core components. Fortunately, the graphite core is tolerable to cracking, but this tolerability is not expected to be unbounded. A singly cracked fuel brick is expected to start opening after stress reversal (this is the period when the stresses at bore and periphery change sign). Consequently, cracked fuel bricks may start to interact with neighbouring components in ways that were not likely to have been accounted for in the original design. This interaction may lead to possible channel distortion, which, if excessive, could limit the safe operation of the reactor. Therefore, it is paramount from a safety point of view to understand the effect of cracking on the core integrity under different static and dynamic operational conditions. This would help to ensure that the fundamental safety requirements are upheld for continuous operation of a reactor that contains limited damage to its graphite core. In this paper, the effect of the opening of single cracks in fuel graphite bricks on the deformation of neighbouring channels is investigated via numerical simulation under normal operational (static) conditions. Different numerical models are presented to explore possible consequences of crack opening on channel distortion and core integrity.
15h20: Predicting Keyway Root Cracking Behaviour

R. Gray, M. Joyce, B.C. Davies and M.R. Bradford

Current methods predict that the fuel moderator bricks within the Advanced Gas-cooled Reactor (AGR) core will develop cracks from their radial keyway roots in late life due to the evolution of differential and thermal strains. Due to variability of both the local irradiation environment and material properties, not all bricks will undergo cracking at a common time; rather it is expected that cracking will be a progressive process through the reactor core. The AGR safety case is tolerant to a specified fraction of cracked bricks, and therefore it is important that appropriate predictions of the onset time and subsequent rate of cracking are available.
One of the key challenges to providing appropriate predictions of this behaviour is to appropriately describe the variability and covariance of material properties and local reactor environment. This paper presents recent developments in these areas using a time efficient route such that the contributions of each variability source can be investigated.
15h40: Summing Up, Conference Close and Coffee


ACKNOWLEDGEMENTS and SPONSORS
The organising committee and the British Carbon Group would like to thank the following organisations for financial support for this meeting

• 
  EdF Energy Generation Ltd
  
• 
  AMEC Foster Wheeler
  
• 
  Atkins Global
  
• 
  Frazer Nash Consultants
  

The British Carbon Group (registered charity 207890) is affiliated to The Royal Society of Chemistry, The Institute of Physics and The Society of Chemical Industry.
To join The British Carbon Group, or for more information, please contact any one of the Officers or Committee Members attending all or part of the meeting:
Dr. Peter Minshall; Prof. Malcolm Heggie; Dr. Norman Parkyns; Prof. Tony Wickham; Ms. Nassia Tzelepi
Small Print
Whilst every care has been taken in the preparation of this information, no responsibly for errors or any consequences thereof can be accepted by The British Carbon Group, EdF Energy Generation Ltd, The Grand Harbour Hotel Ltd., Southampton, or by any named individual.
The British Carbon Group is registered in accordance with The Data Protection Act 1998 and subsequent amendments. No personal data will be shared with any third party without permission (see also registration form) and no card data will be retained beyond the conclusion of the meeting either in electronic or paper form. Card payments are handled through WorldPay on behalf of The British Carbon Group and The National Westminster Bank.
The British Carbon Group acts only as agent for the reservation of personal accommodation. Any issues relating to your hotel accommodation must be taken up directly with the hotel. No responsibility can be accepted by The British Carbon Group for loss or damage to personal items during the conference, and cars are parked at the hotel or elsewhere at your own risk.

Directory: edf
edf -> Uncertainty Quantification for Numerical Model Validation summer school cea-edf-inria 2011 June 27th–July 8th, 2011, cea cadarache, Saint Paul-lez-Durance, France
edf -> Breathe in: experience. Breathe out: poetry