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Evacuation Decision–Making at three Mile Island
Published in Andrew Blowers, David Pepper, Nuclear Power in Crisis, 2019
Donald J. Zeigler, James H. Johnson
The pressurised water reactor (PWR) has been one of the most popular reactor designs in the United States. It was also the type of reactor which came within 280 degrees F. of a meltdown at Three Mile Island. In the United Kingdom all reactors are either MAGNOX or Advanced Gas-cooled Reactors. In 1979, however, only nine months after the accident at TMI, the British Government made public its commitment to the PWR as the basis for a major expansion of the nuclear power industry, beginning with the construction of an additional reactor at the Sizewell site along the coast of Suffolk county, England, where there is already an operating MAGNOX reactor (see Chapter 4 of this volume).
The management and regulation of decommissioning wastes
Published in Martin J. Pasqualetti, Nuclear Decommissioning and Society, 2019
With the announcement of 13 July 1988 by the Central Electricity Generating Board (CEGB) that it would not be seeking to renew the operating licence for its twin-reactor Magnox station at Berkeley the era of large-scale decommissioning began in the UK. On current plans the decommissioning of the last two Magnox reactors, at Wylfa, will continue into the twenty-second century. Final dismantling of Heysham 2, the last advanced gas-cooled reactor (AGR) to be brought into service, could last until 2140.
Nuclear reactors and their fuel cycles
Published in R.J. Pentreath, Nuclear Power, Man and the Environment, 2019
A simplified plan of a magnox reactor is shown in figure 4.1. In the earliest types, the reactor pressure vessel was contained within a steel sphere connected by ducts to the steam-generator units. Later designs consist of pre-stressed concrete pressure vessels with an integral arrangement of steam generators; this allowed more than a two-fold increase in reactor size. The reactor core of a magnox reactor is large relative to those of other designs and requires a more or less continuous refuelling process. For this to be done while the reactor is on-load requires a complex fuel-handling system which, operated by remote control, removes spent fuel rods and inserts new ones. A variation of the graphite-moderated reactor, which does not use natural uranium, has been used in the USSR. These reactors have ordinary light water instead of carbon dioxide as coolant; they are therefore referred to as light water graphite reactors (LWGR).
A Machine Learning Method for the Forensics Attribution of Separated Plutonium
Published in Nuclear Science and Engineering, 2022
Patrick J. O’Neal, Sunil S. Chirayath, Qi Cheng
Osborn et al. simulated nine different reactor types: a generic pressurized water reactor (PWR) with 235U enrichments of 2.35, 3.4, and 4.45 at. %; a generic pressurized heavy water reactor (PHWR); the blanket region of a fast breeder reactor (FBR) containing depleted UO2; a generic U.K. MAGNOX reactor; a generic Canadian NRX reactor; a simulation of an irradiation conducted at the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory; and a simulation of an irradiation at the University of Missouri Research Reactor (MURR) (Refs. 3 and 4). The two simulations of irradiations at HFIR and MURR were included because the validation data were produced by performing experimental irradiations at these two reactors, and this necessitated including MCNP simulations in the reactor-type database that mirrored the conditions that created those samples. The HFIR experimental irradiation was conducted on depleted UO2 samples surrounded by a thermal neutron shield (gadolinium), and this was reflected in the MCNP simulations that were incorporated into the database. The use of gadolinium shielding was done to emulate the type of fast neutron flux that might be seen in the blanket region of an FBR. The experimental irradiation at MURR was conducted using natural UO2 samples in a thermal neutron energy spectrum to mimic the environment and fuel composition that might be similar to that of a PHWR. The real-world measurements of the separated Pu samples that resulted from these two experimental irradiations were used to validate the methodology.
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.
Adsorption and Desorption of Tritium in Nuclear Graphite at 700°C: A Gas Chromatographic Study Using Hydrogen
Published in Nuclear Technology, 2019
Ke Deng, Mingjun Zhang, Xijun Wu, Qin Zhang, Guo Yang, Zhaowei Ma, Fei Wei, Guanghua Wang, Wei Liu
IG-110 is a type of isotropic nuclear-grade graphite synthesized by Japan Toyo Graphite Company, Ltd. It is a petroleum coke–based, fine-grained, isomolded nuclear graphite. It has been intensively used in HTGRs (Ref. 15). NG-CT-10 nuclear graphite is manufactured by China Fangda Carbon Company, Ltd., using a similar manufacturing process as for IG-110. NG-CT-10 is a potential moderator and structural material for use in the HTGR and MSR in China.16 NBG-18 nuclear graphite is a pitch coke–based, medium-grained, vibration-molded nuclear graphite. It is synthesized by the German SGL group based on Gilsocarbon nuclear graphite. Gilsocarbon nuclear graphite has been widely used in most gas-cooled reactors in Europe including Magnox, AGR in the United Kingdom, and PBMR in Germany; therefore, the nuclear graphite used in the existing gas-cooled reactors should have similar properties as NBG-18 nuclear graphite. In addition, NBG-18 has been proposed for the PBMR in South Africa.17,18 The concentration of impurities within the three types of nuclear graphite is given in Table II.