Abstract

High-temperature gas-cooled reactor (HTGR) is one of promising candidates for new generation of nuclear power reactors. This type of nuclear reactor is characterized with the following principal features: highly efficient generation of electricity (thermal efficiency of about 50%); the use of high-temperature heat in different production processes; reactor core self-protection properties; practical exclusion of reactor core meltdown in case of accidents; the possibility of implementation of various nuclear fuel cycle options; reduced radiation and thermal effects on the environment, forecasted acceptability of financial performance with respect to cost of electricity as compared with alternative energy sources. The range of output coolant temperatures in high-temperature reactors within the limits of 750–950 °C predetermines the use of graphite as the structural material of the reactor core and helium as the inert coolant. Application of graphite ensures higher heat capacity of the reactor core and its practical non-meltability. Residence time of reactor graphite depends on the critical value of fluence of damaging neutrons (neutrons with energies above 180 keV). In its turn, the value of critical neutron fluence is determined by the irradiation temperature and flux density of accompanying gamma-radiation. The values of critical fluence for graphite decrease within high-temperature region of 800–1000 °C to 1·1022 – 2·1021 cm–2, respectively. The compactness of the core results in the increase of the fracture of damaging neutrons in the total flux. These circumstances predetermine relatively low values of lifespan of graphite structures in high-temperature reactors. Design features and operational parameters of GT-MHR high-temperature gas-cooled reactor are described in the present paper. Results of neutronics calculations allowing determining the values of damaging neutron flux, nuclear fuel burnup and expired lifespan of graphite of fuel blocks were obtained. The mismatch between positions of the maxima in the dependences of fuel burnup and exhausted lifespan of graphite in fuel blocks along the core height is demonstrated. The map and methodology for re-shuffling fuel blocks of the GT-MHR reactor core were developed as the result of analysis of the calculated data for ensuring the matching between the design value of the fuel burnup and expected total graphite lifespan.

Highlights

  • Development and implementation of new generation of nuclear reactors recognized in pursuance with IAEA classification as the safest reactor types are the promising direction of development of nuclear industry

  • GT-MHR reactor (Gas Turbine – Modular Helium Reactor) is one of the promising candidates aspiring for leadership among the new generation of high-temperature gas-cooled reactors (HTGR) nuclear reactors satisfying safety requirements put forward in the XXI century

  • In order to estimate service life of graphite it is necessary to determine the distribution of graphite irradiation temperature over the reactor core, because it is the decisive factor influencing the value of critical fluence of damaging neutrons

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Summary

Introduction

Development and implementation of new generation of nuclear reactors recognized in pursuance with IAEA classification as the safest reactor types are the promising direction of development of nuclear industry. One of such types of advanced nuclear reactors are the high-temperature gas-cooled reactors (HTGR) (Grebennik et al 2008). Research conducted in the field of investigation of functionality of reactor-grade graphite in Russia is directed towards finding the solution of the main problem of objective interpretation of already existing experience of graphite operation in commercial and power uranium-graphite reactors for the purpose of application of graphite in the design of advanced nuclear reactors of HTGR type. The second problem in order of importance is the development of principles for handling graphite irradiated in uranium-graphite reactors

GT-MHR reactor
Design and main performance parameters of GT-MHR reactor
Temperature of irradiation and service life of graphite
Conclusion

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