Management of Ageing Processes in Graphite Reactor Cores: 309 - Rilegato

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ISBN 10: 0854043454 ISBN 13: 9780854043453

Casa editrice: Royal Society of Chemistry, 2007

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The Special Publications Series is a collection of books produced from the proceedings of international symposia. Editors bring together contributions from authorities in the field and the books provide snapshots of the latest developments of that field.

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Informazioni sull?autore

The Author Gareth Neighbour is a Senior Lecturer in Engineering at the University of Hull and is an accepted international authority on the phenomenon of irradiation creep in nuclear graphite. He has acted on various International Scientific Advisory Committees and was a member of the Organising Committee for the premier international carbon conference, Carbon 2006.

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Management of Ageing in Graphite Reactor Cores

By Gareth B Neighbour

The Royal Society of Chemistry

Copyright © 2007 The Royal Society of Chemistry
All rights reserved.

ISBN: 978-0-85404-345-3

Contents

Part A – Designs & Operation, 
The Management of Magnox Graphite Reactor Cores to Underwrite Continued Safe Operation A T Ellis and K M Staples, 3, 
AGR Core Design, Operation and Safety Functions A G Steer, 11, 
An Overview of British Energy's Graphite Core Assessment Methodology M R Bradford, 19, 
UK Regulatory Strategy for Management of Graphite Ageing in Gas-Cooled, Graphite-Moderated Reactors G B Heys, 25, 
Part B – Advances in the Fundamental Knowledge of Graphite Behaviour under Irradiation, 
The Effects of Thermal Annealing on the Mechanical Properties of PGA Graphite M P Metcalfe and J F B Payne, 35, 
Development of a Model of Dimensional Change in AGR Graphites Irradiated in Inert Environments E D Eason, G N Hall and B J Marsden, 43, 
Relating Measurements of Mechanical Properties of Nuclear Graphites to Reactor Conditions. A Review of the Effects of Temperature and Pressure B McEnaney, 51, 
Flexural Strength of Graphite M Novovic, G B Heys and P Bowen, 59, 
Observations of Strain Localisation and Failure in Nuclear Graphite M R Joyce and T J Marrow, 67, 
Evaluation of the Brazilian Disc Test in Fracture Strength Measurements of Nuclear Graphite H Li, S L Fok and B J Marsden, 75, 
Irradiation Damage in Graphite from First Principles M Heggie, I Suarez-Martinez, G Savini, A El-Barbary, C Ewels and R Telling, 83, 
Development of a Code to Predict Fundamental Material and Fracture Properties of Nuclear Graphite C Lynch and G B Neighbour, 91, 
Modelling Graphite Ageing: Black Art or Forensic Science? M A Davies and M R Bradford, 100, 
A Holistic, Structurally- Based, Approach to Modelling Graphite Properties M R Bradford and A G Steer, 108, 
X-ray Tomographic Observations Applied to Porosity Models for the Thelmal Properties of Oxidised Nuclear Graphite L Babout, J Marrow, P Mummery and B J Marsden, 116, 
Part C - Advances in the Understanding of Graphite-Component Behaviour and its Assessment, 
Crack Propagation Resistance and Damage Mechanisms in Nuclear Graphite A Hodgkins, J Marrow, P Mummery, A Fok and B J Marsden, 127, 
Graphite Irradiation Testing at the NRG Petten J G van der Laan and J A Vreeling, 134, 
The Development of Measurement Techniques for Mechanical Properties Applicable to Small Reactor Graphite Samples N Tzelepi, M Brown, P C Matthews, J F B Payne and M P Metcalfe, 142, 
Techniques for Measuring the Properties of Unirradiated and Irradiated AGR Graphite S D Preston and W E Ellis, 150, 
Modelling of Nuclear Graphite under Irradiation Conditions D K L Tsang and B J Marsden, 158, 
Predicting the Stresses and Deformations of Irradiated Graphite Moderator Bricks C Jones, 167, 
Explicit Modelling of Graphite Core Components During a Reactor Thermal Transient C D Elcoate, G M Davis and J F B Payne, 175, 
Part D - The Consequences of Graphite Deterioration: Whole-Core Behaviour, 
PBMR Graphite Core Structures Fitness for Purpose Management Strategy M A Davies, M N Mitchell and P J Venter, 185, 
Seismic Modelling of an AGR Nuclear Reactor Core B Kralj, S Humphreys and B Duncan, 193, 
Measurement of AGR Graphite Fuel Brick Shri nkage and Channel Distortion A Cole-Baker and J Reed, 201, 
Investigation of Degradation of AGR Graphite Core Geometry Using a Whole Core Scale Model E Castro, 209, 
The Feasibility of Introducing Core Stiffening Devices and Diverse Shutdown Systems at Hinkley Point B / Hunterston B and Hartlepool / Heysham I M W Davies, 216, 
An Integrated Architecture for Graphite Core Data Analysis G Jahn, S McArthur and J Reed, 224, 
Graphite Core Condition Monitoring Through Intelligent Analysis of Fuel Grab Load Trace Data G M West, G J Jahn, S D J McArthur and J Reed, 232, 
Application of Whole Core Modelling Methodology to Life Extension of AGR Reactor Graphite Cores N McLachlan, D J Shaw and M H S Salih, 240, 
A Strategy for Monitoring Bore Cracking in AGR Graphite Cores P R Maul and P C Robinson, 248, 
Development and Validation of Whole Core Modelling Methodology for AGR Graphite Cores D J Shaw, M H S Salih and N Mclachlan, 256, 
Method for Determining if Control Rods and Fuel Stringer can be Inserted / Removed from an AGR Core A Thomson, M Ripley and N Mclachlan, 264, 
Part E - Lessons Learned from Management of Ageing Plant, 
Experience of Ageing Processes in Nuclear Power Reactors B Eyre, 273, 
The Importance of Understanding of Basic Principles in the Management of Ageing Plant M Burdekin, 282, 
An Overview and Summary of the Technical Papers in Parts B, C and D P E J Flewitt, B McEnaney, B J Marsden and G B Neighbour, 291, 
Author Index, 297, 
Subject Index (by Keywords), 298, 

CHAPTER 1

The Management of Magnox Graphite Reactor Cores to Underwrite Continued Safe Operation

A T Ellis and K M Staples

British Nuclear Group, Reactor Sites, Berkeley, Gloucestershire, UK. Email: alun.t.ellis@magnox.co.uk

Abstract

The first generation of UK Nuclear Power generating plant consists of Magnox reactors that are graphite moderated and CO2 gas cooled. Of the tranche of these power stations built over the period from the late 1950's to the early 1970's four sites remain in operation. It was recognised within the original design that there would be a loss of graphite, due to oxidation, during periods of operation. This oxidation of the PGA graphite occurs by a radiolytic process induced by ionising radiation and results in a degradation of physical, chemical and mechanical properties of the graphite. It is these changes that have to be accommodated in the arguments to demonstrate safe operation to the end of the declared life of these remaining reactors.

Following a review of the background this paper will describe the key challenges that are addressed and the processes adopted within British Nuclear Group, Reactor Sites, to provide this demonstration via three inter-related safety cases. These three safety cases consider core integrity, the long term graphite transient following a depressurisation fault and reactor shutdown reactivity margins. Attention will be given to the reactor core integrity issues where the safety case invokes a multi-legged approach to demonstrate the required high integrity.

Keywords

Magnox reactors, graphite reactor core, safety case

INTRODUCTION

The UK nuclear industry developed rapidly with twenty six Magnox reactors, at eleven power station sites commissioned between 1956 and 1972. All Magnox reactors are graphite moderated, and CO2 gas cooled. All except the final four reactors were constructed with steel reactor pressure vessels; the final four featuring pre-stressed concrete vessels (Figure 1). In general these reactors have provided reliable electrical power generation over their operating lives. However they have now entered a planned closure and decommissioning phase. Eight (of the twenty six) reactors, managed by British Nuclear Group for the Nuclear Decommissioning Authority, are still operational although each has a planned date between 2006 and 2010 when generation will cease.

It was recognised during the original reactor design that graphite cores would be subject to weight loss due to the interaction of the graphite with oxidising species produced by radiolysis of the carbon dioxide coolant, leading to degradation of physical, chemical and mechanical properties. The effects of such degradation on continued safe operation of the reactors, until cessation of generation at each plant, are considered on a regular basis by plant monitoring and review of the associated safety cases.

This paper will explain a typical reactor core design, describe the basis of each of the three individual, but inter-related, safety cases underwriting continued safe operation and conclude with comments on the key challenges.

REACTOR CORE DESIGN

The core of each reactor is an assembly of graphite bricks containing vertical channels for fuel, control devices, specimens and passages for coolant flow. The Oldbury core build dimensions are; overall height: 9.75m and radius: 6.8 to 7.2m (depending on reflector thickness) with an active core height: 8.6m and radius: 6.4m. Brick height is 81.3cm and either 17.1cm or 22.1 cm width. The structure is carried on a steel diagrid via support plates and a steel restraint structure assists in maintaining it in position and shape around the circumference (Figure 2). Disruption to the arrangement of the graphite structure during service could lead to channel flow impairment (and ultimately fuel melt) and restriction of control rod entry.

The distribution of graphite bricks in the core structure is divided into two parts; the central moderator, or active core incorporating fuel and interstitial channels and, surrounding this, the reflector. The structure comprises top reflector bricks and a number of moderator bricks; the bottom bricks also acting partly as a bottom reflector. The cross section of the bricks varies with each station design but vertical keys were commonly used to interlock the columns in the later station designs (Figure 3). The keys allow relative movement, radially and vertically, between adjacent columns to accommodate the effects of thermal transients and irradiation induced dimensional changes in graphite bricks. Each core, dependent on design, is provided with a number of thermocouples to measure graphite material and channel outlet gas flow temperatures. The core restraint structure is also provided with monitoring thermocouples. Typically, core inlet gas temperatures are of the order of 225 °C whilst outlet gas temperatures are approximately 370 °C.

Two types of graphite were used in core construction; Pile Grade A (PGA) and Pile Grade B (PGB). PGA has a higher density and a lower neutron capture cross-section than the PGB material and forms the moderator. Top, bottom and side reflector blocks are a mixture of PGA and PGB materials dependent on the specific reactor design. Both materials were manufactured by a similar method using petroleum coke; a by-product produced during petroleum processing, for the filler material. This coke was ground, mixed with a coal-based binder pitch and then heated and extruded in a hydraulic press to form 'green bricks' which were baked, at ~1000 °C for several days, to a hard brittle material. The extrusion process created a material with anisotropic mechanical properties. To reduce the porosity and thereby increase the density of PGA material, these bricks were then impregnated with coal-tar at a high temperature (~1000 °C) and pressure before the final baking process at 2800 °C. The resulting PGA material has a virgin porosity of ~20%. The individual bricks were finally machined before being used in the construction of the reactor core.

REACTOR CORE SAFETY CASES

Continued safe operation of each reactor core requires that; core structural integrity, the ability to manage reactivity for 'shut down' and long term 'hold down' purposes, and the effects of reactor pressure vessel depressurisation and subsequent air ingress to the core are addressed. British Nuclear Group addresses these issues in three inter-related site specific safety cases for each power station site as described below.

Core Integrity Safety Case

Graphite ageing arises from both radiolytic oxidation and fast neutron damage. Best, Stephen and Wickham (1985) reviewed graphite oxidation and presented a revised model to describe the available data. Their model explains that irradiation of carbon dioxide gas contained in graphite pores produces an 'active species'. The 'active species' entities may either recombine or diffuse to the walls of the pores in the graphite and react with it to produce carbon monoxide, thus enlarging the pores and reducing the graphite density (Figure 4).

The purpose of the graphite is to moderate high energy neutrons and in doing so some carbon atoms are displaced from their lattice position. This leads to graphite shrinkage (and the consequent development of internal stresses because of differential effects), changes in material properties and the accumulation of Wigner, or stored, energy which can be released when the core is heated above its normal operational temperatures.

In addition to its role as a moderator, the functional requirements for the graphite reactor core at the Magnox power stations is maintenance of both the geometry of the control rod channels and reactor core flow geometry. If the geometric integrity of the graphite core were to be seriously challenged, by distortion or widespread brick cracking and displacement of major fragments, insertion of the control rods could be affected. Additionally, through wall brick cracking, if accompanied by significant opening of the crack, could permit substantial flow by-pass from the associated fuel channel, degrading the cooling of the fuel in normal operation and during faults. To ensure a robust high integrity argument is provided for continued operation the safety case for each reactor is constructed around four legs. The four legs are inspection, monitoring, structural integrity assessment and consequences.

Inspection Leg – Forewarning of Severe Cracking

Each reactor is shutdown on a biennial basis for maintenance and inspection. During these outages all reasonably practicable steps are taken to confirm the integrity of the core graphite, and to provide either warning of brick cracking or other significant deterioration in the core. The activities undertaken during each reactor outage, including Norebore and TV viewing campaigns, are termed inspections. Channels to be inspected, which are a small portion of the total core complement, are selected on the basis of a targeted (higher risk), repeat (previous indications) and speculative inspection strategy and the results reviewed against defined acceptance criteria.

In a Norebore inspection, a device is passed down a fuel channel, providing continuous measurements of two perpendicular channel diameters, and of the departure of the channel from being straight and vertical. TV visual inspection of fuel and interstitial channels are undertaken using a sophisticated conical mirror camera enabling almost the entire wall area of a channel to be viewed at a single pass. This unit also incorporates a side viewing camera capable of providing high resolution crack gape images to 0.1 mm, for investigative purposes.

Monitoring Leg – Deterioration of the Graphite and Impact on the Overall Condition

On-line monitoring provides information about the reactor conditions at power and gives forewarning of significant brick damage. Material condition monitoring is provided by the removal of graphite samples from the reactor during outages. Graphite samples, typically 12 mm diameter x 20 mm in depth, are trepanned from the brick walls during the inspection programme. Additional capability is provided by special graphite samples installed in the reactor at start of life. These samples, some of which are removed periodically from the reactor, are used for both mechanical and chemical property measurements and provide a monitor of the condition of the graphite following in-service exposure. The data are incorporated into the overall data set and are used for safety case review and inputs to the revised structural integrity argument.

Confirmatory evidence of the retention of the design core geometry is routinely provided by monitoring for (i) free entry and movement of the control rods, (ii) the absence of anomalous channel gas outlet temperatures or graphite moderator temperatures and (iii) the absence of abnormal restrictions during refuelling. During normal plant operation 'freedom of movement' checks are undertaken on certain control rods to detect core distortion. Control rod checks are taken following shutdown, unplanned or planned and during start-up and during power operation. Following unplanned/planned shutdowns the operator immediately confirms that all control rods have fully inserted into the core. At statutory outage periods all control rods are tested during the first few hours of the shutdown and again on completion of maintenance. These tests are analysed for evidence of adverse trends in insertion time.

At power a daily calculation of maximum assessed fuel can temperature includes a full temperature survey. Any temperature indications which do not behave as expected will prompt an investigation. All fuel element, channel gas outlet and graphite temperatures are recorded shortly after shutdown. Regular monitoring of these outputs is then instituted (hourly if practicable). The temperatures are examined for indications which are not consistent with expectations based on previous experience. During start up fuel element and channel gas outlet (CGO) temperatures are measured to identify whether any individual channel is operating at anomalous higher temperatures.

Any difficulties that arise during refuelling operations are reviewed for potential channel damage/distortion issues. A burst fuel can detection system is installed which continually monitors all channels in the core. Damage to fuel could occur either by core distortion and mechanical damage to the fuel, or by overheating of fuel caused by channel blockage or channel bypass flow. If this were to occur, it would be readily detected by the system, prompting the appropriate response from the operator.

Structural Integrity Assessment Leg

The structural integrity assessment leg of the safety case addresses the increase in weight loss of the graphite during service and the maintenance of its functional capability in the unlikely event that crack-like defects are produced. A review of the existing safety case for the functional capability of the core with respect to structural integrity issues is undertaken. This takes into account the output from the plant monitoring programme. This work is aimed primarily at demonstrating that the likelihood of brick cracking is small. It also provides a consideration of the possibility that graphite shrinkage might lead to keyway pinching and thereby prevent the as-designed core movement provided by the key/keyway system. This leg comprises three main elements:

• Assessment of the changes either predicted or measured in physical and mechanical properties of the graphite, including weight loss.

• Calculation of the various components of loads and stresses arising in the graphite as a result of individual internal and external mechanisms.

• Assessment of the integrity against a suitable failure criterion.

The failure criterion adopted is a simple ratio of applied combined stress to material failure stress. This ratio is termed a fractional strength value, or utilisation. For an analysis at the best estimate level, a fractional strength value of 1.0 is considered to correspond to a 50% probability of localised cracking. However, due to a number of approximations the calculated utilisations are conservative and regarded as indicative of the risk of failure rather than definitive.

Consequences Leg– Consideration of Postulated Brick Cracking and Keyway Pinching

An understanding of likely modes of brick failure leads to a view on the probability of significant cracking. It is necessary to assess the potential consequences in the hypothetical event that cracked bricks are present in the cores. Consideration is also given to the potential effects on fuel can temperatures, due to coolant flow bypassing, of more onerous but hypothetical cracks. The potential for impairment of control rod movement is considered since should there be more than one crack in a brick, there is the potential for a detached fragment of graphite to either impede control rod entry or impair fuel cooling. The potential for keyway pinching to impede the entry of control rods also forms part of the argument.

Reactor Shutdown Reactivity Margin Safety Case

When a reactor trips, automatically due to a fault signal (unplanned) or manually by the operator (planned), all available control rods fall into the core, under gravity, to shut the reactor down. The reactor safety case demonstrates that the associated reactivity loss is always sufficient to shut-down the reactor, with a significant excess known as the shutdown margin (SDM). It is also necessary to demonstrate a hold-down margin, ensuring that the reactor will not unintentionally become re-critical following shutdown. The following factors will affect the reactivity of a shut-down Magnox reactor viz, core temperature, xenon and other reactor poisons, removal of control rods, removal of flux flattening absorbers, refuelling, presence of air and presence of moisture. Following a shutdown, the reactivity of the core initially falls due to increases in the concentration of xenon-135 which continues to be produced by radioactive decay because it is no longer removed by neutron absorption. After about 12 hours, however, the rate of production of xenon-135 drops and the core free reactivity (the reactivity neglecting control rods and absorbers) will eventually rise above its at-power value. Clearly, the removal of control rods and absorbers from a shutdown core will result in reactivity increase, and because of this, Operating Rules place restrictions on their removal. Similarly, Identified Operating Instructions are used to control refuelling activities, including the movement of fuel and absorbers into and out of the reactor cores (Western, D. J. and Corkerton, P.A. 1998)

(Continues...)

Excerpted from Management of Ageing in Graphite Reactor Cores by Gareth B Neighbour. Copyright © 2007 The Royal Society of Chemistry. Excerpted by permission of The Royal Society of Chemistry.
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Dettagli dell'edizione

• Editore: Royal Society of Chemistry
• Data di pubblicazione: 2007
• Lingua: Inglese
• ISBN 10: 0854043454
• ISBN 13: 9780854043453
• Rilegatura: Copertina rigida
• Numero di pagine: 296
• Redattore: Neighbour Gareth B.
• Contatto del produttore: non disponibile
• Persona responsabile: non disponibile