Advanced Gas-cooled Reactor Research Articles

United Kingdom nuclear reactor design and operation.

This paper will consider briefly two main points. First, why one should feel confident that UK nuclear plants can be operated safely. This will be done by highlighting a number of basic principles of safe reactor design to show what they mean and how they are applied. Second, the case in which matters went so disastrously wrong, at Chernobyl, will be examined relative to the points considered in the first part of the paper on UK reactors. The main features of nuclear plant design can be considered by reference to an advanced gas-cooled reactor (AGR) (Fig. 1). The core is an assembly of graphite bricks acting as the neutron moderator. It contains vertical channels in which the fuel elements are stacked. Each fuel element is an assembly of rods made up of pellets of uranium dioxide in a stainless steel can. Heat is produced by fission of the uranium atoms in the UO2, and this process gives rise to the radioactive species which, with the UO2, are contained in the stainless steel can. Heat is taken away by a flow of pressurised carbon dioxide gas driven up the channels by circulators. The hot CO2 is passed into the boilers, flowing over tubes containing water which is turned into steam to drive the turbine generators. The carbon dioxide is retained at pressure inside the pressure vessel, a massive pre-stressed concrete structure.

Research for Power

Fluid flows occur throughout the power generation process: from the flow of steam through turbines to the dispersion of hot gases from power station stacks; from the flow of molten metal in a weld pool to the generation of wind power; from the circulation of high-pressure CO2 in advanced gas-cooled reactors to the discharge of warm water to the sea. These flows may be compressible or incompressible, singlephase or multi-phase; they may be driven by buoyancy forces or they can be influenced by applied magnetic fields.

The dynamic response of multi-layers AGR core brick arrays

The graphite core of the Advanced Gas Cooled Reactor (AGR) core has been seismically qualified using a non-linear computer program AGRCOR. This complex three-dimensional code has been verified by comparing predicted responses with exact solution and also with AMP2D code. Using the contact input parameters derived from an extensive test program, the code was validated by comparing predicted responses with measurements of dynamic models of various complexity, one (36 bricks), two (63 bricks) and four (117 bricks) test assemblages. The impact loads generated in the key/keyway and between adjacent bricks due to postulated earthquake input were used to assess the seismic capability of the core by relating these loads to some form of cumulative damage criteria obtained from combined test and analysis.

Uptake of [<sup>35</sup>S] Carbonyl Sulphide and [<sup>14</sup>C] Carbon Dioxide by Crops in the Vicinity of an Advanced Gas-Cooled Reactor

Uptake of [<sup>35</sup>S] Carbonyl Sulphide and [<sup>14</sup>C] Carbon Dioxide by Crops in the Vicinity of an Advanced Gas-Cooled Reactor

Modelling and parametric analysis of an advanced gas-cooled reactor

This paper describes a parameter identification scheme that operates on a mathematical model of one of the CEGB's Commercial Advanced Gas-cooled Reactors. Parameters in a 10th-order non-linear deterministic model are varied to fit experimental data from two rod movement transients. Two sets of results are analyzed, at low and high power operation of the Hinkley Point B CAGR plant. The values obtained for one of the parameters, the fuel-temperature coefficient of reactivity, are compared with those obtained from the standard CEGB analysis procedure and found to agree within the estimated error bounds.

High-temperature resistant materials and structural ceramics for use in high-temperature reactor and fusion reactor plants-requirements for modern physico-chemical analysis

Following a brief description of primary energy requirements and the available resources, the application of advanced high-temperature gas-cooled reactors (HTR) for nuclear process heat and the technological aspects of Tokamak fusion reactors are reported. In both areas, the development of materials, in general, and materials characterization using modern physico-chemical methods, in particular, are crucial. The investigations for the HTR are concentrated on the optimization and on the verification of the long time behaviour of the materials. In contrast, the fusion reactor materials work is concerned with fundamental development stages and with plasma-wall interactions.

Background to BS 5500 and its use in the design of AGR plants

BS 5500 is a British Standard to be applied during the design and construction of unfired fusion welded pressure vessels and was first published in 1976. This is a master pressure vessel standard whose requirements are constantly under review and reflect the general experience and technical developments in the pressure vessel industry and the relevant research in various academic establishments. The paper describes the format of the standard and the general requirements and shows how these have been applied to the Advanced Gas Cooled Reactors in the UK including an indication where additional requirements have been met to achieve the desired vessel integrity.

Application of distortion technique for model validation to nuclear power plant - AGR scatter plug

The distortion technique has been applied to validate quantitatively a dynamic model of a scatter plug as used on advanced gas-cooled reactors from comparison with recorded rig results, A brief outline of the technique is given which illustrates the distortion required for the constant parameters within the model and the assessment of model adequacy. Initially these assessments indicated that the model was not adequate to explain the test results. This led to significant improvements of the model and better under. standing of the model and the tests, Eventually, after several improvements in the model, the quantified criterion inherent in the distortion technique was satisfied indicating model adequacy.

Nuclear tribology: a personal perspective

For some 25 years, the Central Electricity Generating Board (CEGB) has been using nuclear reactors as the heat source for the safe, reliable, and economic production of electricity. Prior to, and during, this period many challenges have been addressed in many technical disciplines, and tribologists have often been involved in development work. This paper discusses examples of nuclear engineering where the author has had some personal involvement. The subjects will embrace aspects of tribology, including lubrication, friction and wear, and of production, ie safety, reliability and economy. The intention is to discuss basic aspects of reactor tribology and the strategies pursued in addressing questions. Possible lessons for the future will also be considered. As a ‘personal perspective’, the author concentrates on the advanced gas-cooled reactor (AGR).

Acoustic emission from irradiated nuclear graphite

Abstract Measurements of acoustic emission (AE) from a range of four unirradiated nuclear graphites during three-point bend tests are reported. Results are in agreement with the trends found in earlier work using different AE apparatus. The technique is applied to the testing of small beam specimens cut from irradiated Civil Advanced Gas-cooled Reactor (CAGR) graphite fuel sleeves after discharge from the reactor. The AE information is explained by considering separately the known changes in graphite microstructure that occur in the reactor due to radiolytic oxidation and fast neutron irradiation. Coarsening of the material due to radiolytic oxidation increases the total number of AE events and the proportion of events of low amplitude. Fast neutron irradiation increases the fracture stress and makes the stress-strain curve more linear. As a consequence, the number of AE events is reduced along with the proportion of events of low amplitude.

Seismic response of the advanced gas cooled reactor core

The seismic response of the Advanced Gas Cooled Reactor (AGR) core has been calculated using an idealisation having several hundred thousand degrees of freedom. The graphite bricks which make-up the core are idealised as rigid masses, whilst contact spring elements are used to represent the load transmissions or impacts that can take place between the bricks. The necessary input information for the contact spring elements (i.e. stiffness, damping and friction) has been obtained by test work. Whilst the dynamic analysis of the core itself is highly non-linear, the supporting steel structures are linearly elastic. Consequently, the dynamic characteristics of the supporting structures are evaluated with the non-linear core structure uncoupled and are then used with the non-linear core data in a step-by-step explicit time-history. This paper discusses the development of the analytical model, the theory and the computer codes used. Results from some of the parameter studies, sensitivity studies and predictions of core response to earthquakes are presented.

AGR v PWR The debate continues

When the Central Electricity Generating Board (CEGB) invited tenders and placed a contract for the Advanced Gas Cooled Reactor (AGR) at Dungeness B in 1965 - preferring it to the Pressurised Water Reactor (PWR) — the AGR was lamentably ill developed. The effects of the decision were widely felt, for it took the British nuclear industry off the light water reactor highway of world reactor business and up an idiosyncratic private highway of its own, excluding it altogether from any material export business in the two decades which followed. Yet although the UK may have made wrong decisions in rejecting the PWR in 1965, that does not mean that it can necessarily now either correct them, or redeem their consequence, by reversing the choice in 1985. In the 20 years since 1965 the whole world economic and energy picture has been transformed and the national picture with it. Picking up the PWR now could prove as big a disaster as rejecting it may have been in 1965.

The mechanical properties of irradiated pitchcoke graphite

Small specimen measurements of dynamic Young's Modulus E, flexural strength S and critical stress intensity factor K Ic of pitchcoke graphite irradiated to a high fuel burn-up in a Civil Advanced Gas-cooled Reactor are reported. Previous data indicate that the fractional change in S due to fast neutron damage alone follows the corresponding fractional change in √ E and due to radiolytic oxidation alone approximately follows the corresponding changes in E. No comparative measurements of K Ic have previously been made. Thus, up to now it has been assumed that the fractional changes in K Ic are identical to the corresponding fractional changes in S. The new K Ic data support the fast neutron damage relation but suggest that K Ic is reduced to a slightly lesser extent than E or S with radiolytic oxidation. A comparison with a previous thermal oxidation study is made.

Aspects of fretting wear of sprayed cermet coatings

Two experimental fretting programmes which investigated aspects of fretting wear of sprayed cermet coatings are reviewed. These programmes were conducted in support of components used in the advanced gas-cooled reactor. It is speculated that the results from these programmes are compatible with a simple two-stage wear model. This model assumes that an initial wear process occurs which is dominated by an interlocking and removal of asperities. Such a phase will be dependent on the superficial contact areas and possibly the interfacial load, but the latter aspect is not considered. This initial wear is of very short duration and is followed by a mild, oxidative, wear mode. Coatings data are also compared with those for structural steels. In short-term low temperature tests it appears that structural steels have comparable performance with the cermet coatings but it is argued that this is an artefact of the wear process. However, at high temperatures (600 °C) wear of stainless steel could not be determined, the specimens showing a net weight gain. It is concluded that for in-reactor fretting applications cermet coatings will have advantages over structural steels at low temperatures. Even in high temperature regions some operation at low temperatures is experienced and consequently cermet coatings may be useful here also.

The impact fretting wear of a nuclear reactor component

In the pressurized CO 2 advanced gas-cooled reactor several components are subject to gas-flow-induced vibration and wear. The impact fretting wear behaviour of two such components, which form a Nimonic PE16-stainless steel pair, was studied in the present work. Experiments were carried out to assess the wear rate dependence on test duration, temperature and gas composition and to simulate the pitted scar morphology of the ex-reactor components. The impact wear behaviour is shown to be dependent on all these parameters and a transition from a severe to a mild form of wear is identified at a temperature in the range 250–450 °C, the exact temperature depending on the impacting conditions and gas composition. Below the transition temperature the specific wear rate is approximately 20 times greater than that which occurs above it. Wear scars resembling the pitted morphology of those found on the ex-reactor Nimonic components have only been obtained in tests in which the Nimonic specimens were impacted against a pregalled stainless steel surface.

The Thermal–Hydraulic Design of Main Feed Water Pump Suction Systems for Large Thermal Power Plant

British domestic design practice is described for boiler feed pump suction systems associated with high level deaerator/feed water tanks in large modern power plant. The objective of thermal–hydraulic design is the provision of adequate pressure to boiler feed water pump suctions under all transient and steady state conditions to ensure continuity of water flow and reliable pump operation. Methods are available for the determination of pressure decay in feed tanks during load transients, but the successful application of mathematical relationships between pressure decay and pressure rise in suction pipework depends on the correct identification of operational circumstances. Progress in recent years has been on the identification of circumstances which produce governing conditions and has taken design practice beyond the stage established by early workers. Developed practice has been successfully applied by the Central Electricity Generating Board and its main plant contractors to forty-nine 500 MW units in coal- and oil-fired stations and to twenty-three 660 MW units in coal, oil and advanced gas-cooled reactor stations, which at 39 680 MW, constitutes the major proportion of plant in England and Wales. Aspects of plant behaviour and operation crucial to successful system design are featured. Recommendations are made for plant layout, pipework routing, tank outlet and pipework sizing, net positive suction head (NPSH) determination etc.

Pyrolysis of nuclear reactor circulator oil in carbon dioxide: II. Long term degradation studies

The thermal degradation of Advanced Gas cooled nuclear Reactor (AGR) circulator oil in carbon dioxide has been studied using degradation temperatures up to 650°C and for degradation periods up to 4 h. The degradation products were measured and identified using gas chromatography (GC) and GC combined with mass spectrometry. The following conclusions were drawn: 1. The production of lower molecular weight hydrocarbons at temperatures below 525°C results from primary decomposition reactions. At higher temperatures the net rate of formation falls as they are converted to other species. The production of higher hydrocarbons is complex. 2. The relative proportions of the different species vary with temperature, e.g. propene forms 19% of the total at 540°C, but only 2% at 400°C and 650°C. 3. The proportion of alkenes falls and the proportion of aromatics rises with both increasing temperature and increasing time. 4. A lower proportion of aromatic species results from degradation in helium compared with degradation in carbon dioxide. 5. The reactions are too complex to allow the data to be fitted to a rate equation.

UKAEA and CEGB Development Programmes

The best method of identifying and solving the problems that might arise while decommissioning a nuclear reactor is to pursue a real project. The Windscale advanced gas-cooled reactor (WAGR) project has already highlighted several topics and thus indicated the direction in which the development programme should go. It is recognized that the concrete pressure vessels, which are not part of the WAGR structure, present special problems and so most of the CEGB work to date has been aimed in that direction. Both programmes are described.

Decommissioning Windscale Advanced Gas-Cooled Reactor

This paper describes the UKAEA project to decommission the Windscale advanced gas-cooled reactor. The plan is to take down the reactor itself, the heat exchangers and the sphere in which they are all contained. A special remote-handling machine is being developed which will cut up the radioactive items, and a facility is being built to package the waste for storage and disposal.

The radiolysis of simple gas mixtures—III. the production of “trace organics”

Carbon dioxide based gas mixtures, similar to those used as coolants in the Advanced Gas-cooled Nuclear Reactors, have been radiolyzed at the comparatively low dose rate of 3 Gy s -1 using γ-radiation from a 60Co source. The concentrations of certain C 4 and C 5 alkanes, ethanal and propanone, produced by radiolysis, have been determined under different conditions of temperature, pressure, gas composition, and surface area. Observations showed that these compounds were produced in concentrations of the order of 1/1000 of the methane concentration and that the production of these “trace organics” varied in a complex way with the experimental parameters. Mechanisms for the production of these organics have been proposed. Some unidentified products of radiolysis were observed and suggestions are made as to their nature.