Extension of lifespan of graphite in fuel blocks of high-temperature gas-cooled reactors as the resource for ensuring design values of nuclear fuel burn-up

O. I. Bulakh, O. K. Kostylev, V. N. Nesterov, E. K. Cherdizov

Результат исследований: Материалы для журналаСтатьярецензирование

Аннотация

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 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 a 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 reduce 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 lifetime 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 chart 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 compliance of the design value of the fuel burnup with expected graphite lifespan.

Язык оригиналаАнглийский
Страницы (с-по)40-52
Число страниц13
ЖурналIzvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika
Том2019
Номер выпуска3
DOI
СостояниеОпубликовано - 2019

ASJC Scopus subject areas

  • Nuclear Energy and Engineering

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