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Prospects for the Nuclear Debate in the UK
Published in Andrew Blowers, David Pepper, Nuclear Power in Crisis, 2019
One can discern four ages in the life of civil nuclear power in the UK over the past thirty years. These can be characterised as the age of innocent expectation (1946–1966), the age of doubt (1966–1974), the age of anguish (1974–1981) and the age of public justification (1981 to the present). The first period was associated with the post-war exuberance over nuclear technology. The British led the civil reactor field, and were able to enjoy the apparent luxury but economic burden of designing and constructing individual reactors – the first generation MAGNOX reactors. The second period began with the decision to construct the advanced gas cooled reactor (AGR), the successor to the MAGNOX. But the AGRs were inadequately designed and suffered many delays and cost over-runs. The AGR experience began to raise doubts about the cost effectiveness of nuclear power, and about the accountability of the nuclear industry to public opinion and democratic parliamentary procedures. This point is developed by Fernie and Openshaw in Chapter 5.
Nuclear Energy Security
Published in Maria G. Burns, Managing Energy Security, 2019
This type of reactor is a continuation of the original nuclear reactor designs. Widely used in England, its two main representatives are the advanced gas-cooled reactor (AGR), where enriched uranium is the fuel, and the Magnox, where uranium is used as fuel. Both designs use graphite as a moderator, and carbon dioxide, helium, or another inert gas as the coolant.
Nuclear Power Technologies through Year 2035
Published in D. Yogi Goswami, Frank Kreith, Energy Conversion, 2017
Kenneth D. Kok, Edwin A. Harvego
Gas-cooled reactors were developed and implemented in the United Kingdom. The first generation of these reactors was called “Magnox” and they were followed by the advanced gas-cooled reactor (AGR). These reactors are graphite moderated and cooled by CO2. Magnox reactors are fueled with uranium metal fuel, while the AGRs use enriched UO2 as the fuel material. The CO2 coolant is first circulated through the reactor core and then to a steam generator. The reactor and the steam generators are located in a concrete pressure vessel. As with the other reactor designs, the steam is used to turn the turbine generator to produce electricity. Figure 15.4 shows the configuration for a typical gas-cooled reactor design.
Effect of Cladding Surface Roughness on Thermal-Hydraulic Response of Nuclear Fuel Rod of Advanced Gas-Cooled Reactor
Published in Nuclear Science and Engineering, 2022
Sadek Hossain Nishat, Md. Hossain Sahadath, Farhana Islam Farha
Cooling is a challenging issue in nuclear reactors. Sufficient cooling must be required to properly remove nuclear reactor core heat to produce electric power as well as to prevent fuel cladding damage and core meltdown.9 Many nuclear reactors use liquid water as reactor coolant to sufficiently and safely remove the reactor core heat. But, some nuclear power reactors, for example, MAGNOX, Advanced Gas-Cooled Reactor (AGR), HTGR, etc., use gaseous fluid as the reactor coolant.10 In the U.K.-designed AGRs, carbon dioxide (CO2) is used as the reactor coolant. Because CO2 is a gaseous fluid, its thermal property especially the convective heat transfer coefficient is lower than for liquid fluids. So, a main issue is to enhance the surface heat transfer coefficient of AGR fuel cladding. For this purpose, ribs are placed on the AGR fuel cladding11 to augment or enhance the heat transfer coefficient and to enhance the thermal efficiency of the whole nuclear power plant system.
Fluoride-Salt-Cooled High-Temperature Reactor (FHR) Using British Advanced Gas-Cooled Reactor (AGR) Refueling Technology and Decay Heat Removal Systems That Prevent Salt Freezing
Published in Nuclear Technology, 2019
Charles Forsberg, Dean Wang, Eugene Shwageraus, Brian Mays, Geoff Parks, Carolyn Coyle, Maolong Liu
We describe an FHR design in the early stages of development that uses many of the features of the British advanced gas-cooled reactor (AGR). AGRs have successfully operated for decades in the United Kingdom. In this paper, first, we describe the salt coolant options because it is their ability to deliver high-temperature heat that creates (1) the commercial basis for the FHR and (2) most of the technological challenges of the design. We follow this with a description of the AGR, the refueling technology that enables high-temperature FHR refueling, the proposed FHR design, and the basis for selection of specific design features. This includes the temperature control systems necessary to avoid either freezing of the salt or overheating of the reactor under accident conditions. The system could be built in sizes up to thousands of megawatts thermal. In this paper the term “FHR” refers to characteristics of FHRs in general, and the term “AGR-FHR” refers to characteristics of an FHR that has incorporated features from the AGR.
The evolution of He+ irradiation-induced point defects and helium retention in nuclear graphite
Published in Journal of Nuclear Science and Technology, 2019
Mingyang Li, Chuanqing Shi, Henk Schut, Zhengjun Zhang, Zhengcao Li
Nuclear graphite refers to the graphite intended for application in nuclear reactors. Graphite has been used as the neutron moderator and structural material in several kinds of graphite-moderated nuclear fission reactors, including Magnox, Advanced Gas-cooled Reactor (AGR) and High-Temperature Gas-cooled Reactors (HTGR). The advantages of nuclear graphite include low neutron absorption cross section, high chemical purity, good thermal and mechanical properties at very high temperatures (>1000°C), good compatible with the other materials used in the reactor core, high chemical stability, and inexpensiveness, which enable nuclear graphite to be an important component in nuclear power plants [1–3]. However, Very High-Temperature Reactor (VHTR), which is a Generation IV reactor concept that comprises HTGR but with a potential core outlet temperature of at least 1273 K [4], has set higher demands to nuclear graphite [5–9]. In the moderation process, neutrons collide with carbon nuclei and thereby transfer kinetic energy to carbon matrix, and then the carbon atoms are knocked and will be displaced from their equilibrium positions and thus distort the surrounding lattices [10,11]. As a result of long-term neutron irradiation, the structure and properties of graphite would be seriously harmed, and the safety of nuclear power plants would be threatened by the failure of graphite components. Therefore, it is of great importance to study the response of graphite to radiation damage and to understand the defect behavior in thermal environment.