Advanced Gas-cooled Reactor Research Articles

Multicycle Optimization of Advanced Gas-Cooled Reactor Loading Patterns Using Genetic Algorithms

A genetic algorithm (GA)–based optimizer (GAOPT) has been developed for in-core fuel management of advanced gas-cooled reactors (AGRs) at HINKLEY B and HARTLEPOOL, which employ on-load and off-load refueling, respectively. The optimizer has been linked to the reactor analysis code PANTHER for the automated evaluation of loading patterns in a two-dimensional geometry, which is collapsed from the three-dimensional reactor model. GAOPT uses a directed stochastic (Monte Carlo) algorithm to generate initial population members, within predetermined constraints, for use in GAs, which apply the standard genetic operators: selection by tournament, crossover, and mutation. The GAOPT is able to generate and optimize loading patterns for successive reactor cycles (multicycle) within acceptable CPU times even on single-processor systems. The algorithm allows radial shuffling of fuel assemblies in a multicycle refueling optimization, which is constructed to aid long-term core management planning decisions. This paper presents the application of the GA-based optimization to two AGR stations, which apply different in-core management operational rules. Results obtained from the testing of GAOPT are discussed.

The GCFR revisited

A gas-cooled fast reactor (GCFR) concept cooled by CO 2 and derived from the symbiosis of the British advanced gas-cooled reactor and the prototype fast reactor technologies is reviewed. Application of the reactor would permit the continued growth of plutonium to cease and the stocks of depleted uranium already mined would sustain a large fleet of such reactors indefinitely. The GCFR has high potential for the incineration of the ultra-long-term heavy metal waste arising from fuel processing. Recently, a British-Japanese industrial assessment concluded that the concept shows promise of competing on economic terms with the light-water reactor.

Carbon deposition on stainless steel in oxidising environments

The formation of low-density carbon deposits onto metallic heat transfer surfaces in industrial environments of high carbon activity can have a deleterious effect on thermal efficiency. Often this deposition occurs under reducing conditions and involves carbide (typically of iron) formation as part of the process. It is possible, however, for filamentous carbon deposits to form at much higher oxygen potentials where carbide formation is not favoured thermodynamically. This study considers one such situation by examining the behaviour of a 20Cr25Ni austenitic steel used as fuel cladding in Advanced Gas-cooled nuclear reactors (AGRs). Carbon deposition has been produced in laboratory by exposure at 550°C to CO2 containing 1%CO and 1000 vppm C2H4; the dissociation of the latter additive generates carbon activities much greater than unity. The oxygen potential of this gas is sufficient to oxidise iron and chromium, of the major alloy constituents, but not nickel. The deposition takes place onto substrates that are prepared as electron microscope samples which are subsequently examined by SEM and HREM (high resolution electron microscopy) techniques. The deposits formed consist of solid carbon filaments, approximately 30 nm in diameter, with turbostratic atomic layering. Each filament contains a 10–20 nm metallic nickel particle near its tip. It is argued, and supported by HREM evidence, that such particles are produced as a direct result of the oxidation of the alloy and that prior oxidation is necessary to produce the observed carbon deposition. It is also shown that inhibition of carbon deposition occurs when the oxygen potential of the depositing gas is increased sufficiently to oxidise nickel.

An intelligent system for interpreting the nuclear refuelling process within an advanced gas-cooled reactor

Evaluation of the data produced during the refuelling process in a nuclear power plant is required to ensure proper ‘set-down’ of the fuel assembly, thereby allowing the continued and safe operation of the station. The process of evaluating the data can be time consuming owing to the large amounts of data requiring considerable domain experience and interpretation. This paper presents an intelligent system (IS) to automate the process of data analysis, thereby shortening the evaluation time and providing an explanation of the reasoning behind its conclusions. The intelligent system utilizes a knowledge-based system (KBS), neural network based classification, K-means clustering techniques and rule induction methods to evaluate the data and inform the operator of any errors encountered.

Operational waste removal from Windscale advanced gas-cooled reactor (WAGR)

Operational waste removal from Windscale advanced gas-cooled reactor (WAGR)

Remote material analysis of nuclear power station steam generator tubes by laser-induced breakdown spectroscopy

We report on a novel design 75-m length umbilical fiber-optic LIBS (FOLIBS) system suitable for remotely determining the copper content of 316H austenitic stainless steel superheater bifurcation tubing within the pressure vessels of Advanced Gas Cooled Reactor (AGR) nuclear power stations. The system was deployed during the routine reactor outage programs for Hunterston ‘B’ and Hinkley Point ‘B’ stations during the summer of 1999 and used successfully to determine the copper content of the bifurcations over the range 0.04%<Cu<0.60% (by mass). Measurement times per bifurcation were typically less than 3 min and measurement accuracy was approximately ±25%.

Effect of irradiation creep strain on the coefficient of thermal expansion of AGL IM1-24 and UCAR GCMB nuclear graphite moderator bricks

British Energy Ltd (BE) operates all seven twin advanced gas-cooled reactor (AGR) power stations in the UK. In the AGRs, neutrons from the heat-producing process of fuel fission are moderated by graphite, and both graphite and fuel are cooled by pressurised carbon dioxide. The moderation process, together with corrosion by the coolant, alters the material properties of the graphite, which in turn influences the functionality of the moderator core. As a consequence, it is important to BE to understand the behaviour of irradiated graphite. This paper explains that a materials model of how graphite materials properties evolve with increasing dose and fluctuating temperature was developed.

The assessment of the fire safety case on the advanced gas-cooled reactor power stations / Die Bewertung der Brandsicherheit von fortgeschrittenen gasgekühlten Kernkraftwerken

The paper describes the periodic safety reviews of fire safety carried out on the advanced gas-cooled reactors operated by British Energy Generation. A deterministic approach to the assessment of nuclear safety was adopted for the majority of the reviews; the safety principle and assessment guidelines for this are presented and their application described. Other aspects included within the scope of the reviews were the interaction between fire and other hazards, the use of combustible materials at the power stations sites and fire safety management. Review outcomes are described.

WAGR Decommissioning: The Transport of WAGR Heat Exchangers to Drigg Low Level Waste Repository

The Windscale Advanced Gas-Cooled Reactor (WAGR) operated between 1963 and 1981 and was used as a prototype for experimental research and development related to the Advanced Gas-Cooled Reactor programme. WAGR is now the UK's demonstration project for power reactor decommissioning. A major operation in the decommissioning of the WAGR was the removal of the four heat exchangers (REs) from their concrete bioshields inside the secondary containment sphere and transporting them to Drigg low level waste repository. The issues involved with the transport of the four heat exchangers are described, covering details of the heat exchangers themselves as a transport package. Preparation of the route along public roads is described along with details of the transport vehicles and transport operations.

Use of R5 in plant defect assessment

The R5 defect assessment procedures are widely used to assess components in Nuclear Electric's Advanced Gas-Cooled Reactor plant. In this paper, the use of R5 is illustrated by calculating the creep- fatigue crack growth in a specific component. The resulting crack size is compared with the limiting crack size calculated using R6. The implications of considering ductile tearing in the latter calculation, the possible interaction of tearing with creep crack growth, and the use of leak before break arguments are discussed.

A Review of 14C Waste Arising from the Nuclear Industry in the United Kingdom

We report on a review of 14C production and release from the nuclear industry in the United Kingdom (UK). Using earlier predictions (Bush, White and Smith 1984) it is estimated that a total of 1764 TBq of 14C would have been produced in UK graphite-moderated reactors up to the end of 1995, and 135 TBq of this total would have been released. Release rates, based on reported discharges, of 5.1 and 0.93 TBq GW(e)a-1 have been derived for Magnox reactors and Advanced Gas Cooled Reactors (AGRs), respectively. Using these new figures, cumulative discharges of 14C up to 1995 from reactors have been estimated at 531 TBq. The combination of these reactor discharges and the reported discharges from Sellafield (388 TBq) and Amersham (52 TBq) suggest a total cumulative release of 971 TBq of 14C up to the end of 1995.

The Design, Manufacture and Testing of a Container Module to Satisfy the IAEA Regulations for the Safe Transport of Radioactive Material – 1985 Edition (as Amended 1990), for the Transport of an Advanced Gas Cooled Reactor (AGR) Gas Circulator

Heysham 1 and Hartlepool Nuclear Power Stations share a common design of gas circulator and a requirement was identified to transport a spare circulator between the Stations. The circulators are 6 m long, 2 m diameter and weigh approximately 38 tonnes. The circulator becomes contaminated in use and is classified as suitable for transport as SCO-II. It can therefore be transported within a Type 2 Industrial Package (IP-2), in accordance with the IAEA Transport Regulations. An appropriate package, the Gas Circulator Transportation Module, based on ISO container standards, was designed, manufactured and tested, and a Safety Case prepared addressing the radiological hazards associated with the transport of a gas circulator and demonstrating regulatory compliance. Following the issue of a Certificate of Regulatory Compliance, the container module was used· to transport a gas circulator from Heysham 1 to Hartlepool, paving the way for similar purpose-designed modules for the transport of large components.

The Design, Manufacture and Testing of a Container Module for Transporting an Advanced Gas Cooled Reactor Gas Circulator

Abstract Heysham I and Hartlepool Nuclear Power Stations share a common design of gas circulator and a requirement was identified to transport a spare circulator between the Stations. The circulators are 6 m long, 2 m diameter and weigh approximately 38 tonnes. The circulator becomes contaminated in use and is classified as suitable for transport as SCO-II. It can therefore be transported within a Type 2 Industrial Package (IP-2), in accordance with the IAEA Transport Regulations. An appropriate package, the Gas Circulator Transport Module, based on ISO container standards, was designed, manufactured and tested, and a Safety Case prepared addressing the radiological hazards associated with the transport of a gas circulator and demonstrating regulatory compliance. Following the issue of a Certificate of Regulatory Compliance, the container module was used to transport a gas circulator from Heysham I to Hartlepool, paving the way for similar purpose-designed modules for the transport of large components.

95/04773 Magnox life assurance

This chapter describes the basic design and operation of the different types of nuclear reactor used in the UK to provide electrical power to the National Grid. Generation started at Calder Hall in 1956, followed initially by rather similar Magnox reactors and later by advanced gas-cooled reactors (AGRs). In 1995, a pressurised water reactor (PWR) was commissioned at Sizewell. Issues treated involve the choice of fuel, fuel clad, moderator and coolant and specific concerns relating to effects of neutron irradiation on the brittle fracture of steel reactor pressure vessels (RPVs), cracking in graphite moderator blocks, and both stress-relief cracking and under-clad cracking in PWR RPVs., A brief account is given of ‘Generation IV’ systems and of the fusion reactor.

Quantified risk assessment - a nuclear industry viewpoint

This paper presents a brief summary of the methodology used for the assessment of risk arising from fuel handling and dismantling operations in advanced gas-cooled reactor power stations. The difficulties with and problems arising from such risk assessments are discussed. In particular, difficulties arise from (i) the onerous risk criteria that nuclear plants are expected to satisfy, (ii) the necessary complexity of the plant, (iii) the conflicting requirements for the fault consequence assessments to be bounding but not grossly pessimistic, and (iv) areas of fault frequency assessment which contain possibly subjective considerations such as software and common mode failure.

The key role of heat exchangers in advanced gas-cooled reactor plants

The use of helium as a nuclear reactor coolant has been successfully demonstrated in plants built and operated in the U.K., U.S.A., and Germany. Following the pioneering proof of principle plant, two small power plants were operated for several years and this led to the construction of two commercial power stations. For the next generation of gas-cooled reactors new criteria have been developed, namely, the plants will be smaller, simpler, safer and of lower cost. The base case Modular High-Temperature Gas-Cooled Reactor (MHTGR) utilizes existing technology to offer a tried and proven power generating plant using a conventional steam turbine power conversion system that could be in utility service just after the turn of the century. The capability of the MHTGR to operate at very high temperatures will be exploited early in the next century in the form of advanced variants to meet the needs of the power generation and process industries. A key component in the MHTGR is the heat exchanger, since this is where the reactor thermal energy is transferred to the prime-mover or process system. This paper addresses the various roles that heat exchangers will play in advanced MHTGRs, recognizing that the requirements for the steam cycle, gas turbine (direct- or indirect-cycle), and process heat reactor are unique. Topics include thermodynamic considerations, differing configurations, and construction types; materials (metals, composites, ceramics); germane technology bases; and advanced heat exchanger technologies.

Radiation-Induced Segregation: A Microchemical Gauge to Quantify Fundamental Defect Parameters

AbstractDefect kinetic parameters for radiation-induced grain boundary segregation in austenitic stainless alloys are evaluated by comparing model predictions to measured responses. Heavy-ions, neutrons, and proton irradiations having substantial statistical bases are examined. The combined modeling and measurement approach is useful for quantifying fundamental defect parameters. The mechanism evaluation indicates that vacancy migration energies were 1.15 eV or less and the vacancy formation energy at grain boundaries was 1.5 eV. Damage efficiencies of heavy ions and light-water reactor neutrons were about 0.03. Inferred proton damage efficiencies were about 0.15. Segregation measured in an advanced gas-cooled reactor component was much greater than predicted from those parameters alone.

Scottish Nuclear Limited: The AGR—Past, Present and Future

This article reviews the historical development of the AGR (advanced gas-cooled reactor) in Scotland from its inception to its current successfully established position. It examines where the AGRs will go in the future and concludes with the strategic role of the existing and new plant in the market-led electricity supply industry.

Metal carbonyl decomposition and carbon deposition in the advanced gas-cooled nuclear reactor

Carbon deposition in advanced gas-cooled nuclear reactors has been found to occur preferentially in those regions of the reactor core which receive gas directly from the coolant reprocessing plant. This discovery has led to the hypothesis that such deposition may be catalysed by iron or nickel carbonyls being transported into the reactor core. This article describes experiments involving the decomposition of metal carbonyls in simulated AGR gas mixtures under conditions of temperature and pressure similar to those found within the reactor cores. Nickel carbonyl has been shown to promote significant carbon deposition from the simulated AGR, gas mixtures. The carbons deposited in the laboratory have been compared and contrasted with those found within the AGR. Those gases present which are most susceptible to decomposition under the influence of the decomposition products of nickel carbonyl have been identified, and the influence of AGR fuel cladding material examined. Taken together the results provide a mechanism accounting for the nature and distribution of the carbon deposits occurring in these areas of AGR cores.

Thomas Nelson Marsham, 10 November 1923 - 12 October 1989

Thomas Nelson Marsham was one of that band of distinguished men of strong intellect and strong personality who laid the foundations of a programme of nuclear power generation in this country on a sound technical basis. He joined the new atomic energy enterprise when it was still the responsibility of the Ministry of Supply, transferring to the United Kingdom Atomic Energy Authority on its formation, and eventually retired from the UKAEA as a main board member in 1988. After war-time service in the Merchant Navy, he was trained as a physicist and did his doctoral research in nuclear physics. This training enabled him to master rapidly the basic neutron physics underlying the operation of nuclear reactors but, as his career advanced in the UKAEA, he showed also an excellent grasp of engineering principles. Significantly he was elected to the Fellowship of Engineering in the same year as he was elected to the Royal Society. He attained managerial rank and responsibilities in the expanding AEA at a relatively early age, so his career inevitably mirrors the changing tasks and responsibilities of the Authority itself. However, he will be remembered mainly for two outstanding contributions. The first is to the development of gas-cooled reactors in this country. He played a major role in the successful commissioning of the first generation of gas-cooled reactors at Calder Hall and Chapelcross, and later in the introduction of the more efficient advanced gas-cooled reactors, for which he was a persuasive advocate. His second main interest was the development of the fast-neutron fission reactor, a technology in which he became a national and international authority. He saw the introduction of fast reactors as an essential step towards achieving the long-term gains to the national economy from nuclear power that he believed to be both possible and essential.