Izvestiya vuzov. Yadernaya Energetika

The peer-reviewed scientific and technology journal. ISSN: 0204-3327

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

9/30/2019 2019 - #03 Physics and technology of nuclear reactors

Bulakh O.I. Kostylev О.К. Nesterov V.N. Cherdizov E.K.

DOI: https://doi.org/10.26583/npe.2019.3.04

UDC: 621.039.531

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 см–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.

References

  1. Grebennik V.N., Kukharkin N.E., Ponomarev-Stepnoy N.N. High-Temperature Gas-Cooled Reactors are an Innovative Direction in the Development of Nuclear Energy. Moscow. Energoatomizdat Publ., 2008, 136 p. (in Russian).
  2. Mochalov A.M., Najmushin A.G., Nesterov V.N., Pugachev D.K. Determination of the accumulation rate of Wigner’s stored energy in a graphite moderator. Izvestia Vysshikh Uchebnykh Zawedeniy. Yadernaya Energetika. 2015, no. 4, pp.101-111 (in Russian).
  3. Cyganov A.A., Hvostov V.I., Komarov E.A., Kotlyarevskij S.G., Pavlyuk A.O., Shamanin I.V., Nesterov V.N. Problems of utilization of reactor graphite of shut down industrial uranium-graphite reactors. Izvestiya Tomskogo politekhnicheskogo universiteta. 2007, no. 2, pp. 94-98 (in Russian).
  4. Goncharov V.V., Burdakov N.S., Virgil’ev Yu.S., Karpuhin V.I., Platonov P.A. The Effect of Irradiation on Graphite Nuclear Reactors. Moscow. Atomizdat Publ., 1978, 272 p. (in Russian).
  5. Platonov P.A., Chugunov O.K. Radiation damage of graphite and the problem of extending the service life of graphite stack in high-power pressure tube reactor. Proc. of the VII-th International Conference on Reactor Materials. Dimitrovgrad, 2003, pp. 95-114 (in Russian).
  6. Ran F., Adamantiades A., Kenton J., Brown Ch. Nuclear Energy Directory. Moscow. Energoatomizdat Publ., 1989, 752 p. (in Russian).
  7. Mohammadkhani F., Shokati, N., Mahmoudi, S.M.S., Yari, M., Rosen, M.A. Exergoeconomic assessment and parametric study of a Gas Turbine-Modular Helium Reactor combined with two Organic Rankine Cycles. Energy. 2014, v. 65, pp. 533-543.
  8. Sahin S., Erol O., Sahin H. M. Investigation of a gas turbine-modular helium reactor using reactor grade plutonium with Th-232 and U-238. Progress in nuclear energy. 2016, v. 89, pp.110-119.
  9. Kodochigov N., Sukharev Yu., Marova E., Mitenkova E., Novikov N. Features of calculation of temperature reactivity coefficient in the GT-MHR reactor. – In: III-rd Intern. Topical Meeting on High Temperature Reactor Technology. Johannesburg, South Africa, 1–4 Oct., 2006, Paper C00000173, CD. 9, pp. 31-42.
  10. Evaluation of high temperature gas cooled reactor performance: Benchmark analysis related to the PBMR-400, PBMM, GT-MHR, HTR-10 and the ASTRA critical facility, IAEA-TECDOC-1694, International Atomic Energy Agency, Vienna (2013), р. 688.
  11. Chen F., Dong Y., Zheng Y., Shi L. Benchmark Calculation for the Steady-State Temperature Distribution of the HTR-10 under Full-Power Operation. Journal of Nuclear Science and Technology. 2009, v. 46, no. 6, pp. 572-580.
  12. La Bar M.P., Shenoy A.S., Simon W.A., Campbell E.M. The Gas Turbine Modular Helium Reactor. Nuclear News. Oct. 2003, p. 28.
  13. Gas turbine power conversion systems for modular HTGRs. IAEA-TECDOC-1238, International Atomic Energy Agency, Vienna (2001), 209 p.
  14. Nesterov V.N. Ensuring the design value of the burn-up depth of nuclear fuel of high-temperature gas-cooled reactors by the efficiency of graphite. Izvestia Vysshikh Uchebnykh Zawedeniy. Yadernaya Energetika. 2013, no. 2, pp. 133-142 (in Russian).
  15. Golovackij A.V., Nesterov V.N., Shamanin I.V. The organization of the iterative process in the numerical reconstruction of the neutron spectrum in a breeding system with a graphite moderator. Izvestiya vuzov. Fizika. 2010, v. 53, no. 11, iss. 2, pp. 10-14 (in Russian).
  16. Shamanin I.V., Bedenko S.V., Nesterov V.N., Lucik I.O., Prec A.A. Solution of the system of multigroup neutron transport equations in subcritical systems. Izvestiya vuzov. Yadernaja energetika. 2017, no. 4, pp. 38-49 (in Russian).
  17. Golovackij A.V., Nesterov V.N., Shamanin I.V. Effect of composition and burnout of nuclear fuel on the effective value of the damaging neutrons flux in the GT-MGR reactor. Izvestiya Tomskogo politekhnicheskogo universiteta. 2010, no.4, pp. 14-18 (in Russian).
  18. Bajbakov D.F., Godovyh A.V., Martynov I.S., Nesterov V.N. Influence of the nuclide composition of the fuel load on the multiplying and reproducing properties of the active zone of the KLT-40S reactor installation. Izvestiya vuzov. Yadernaya ehnergetika. 2016, no. 2, pp. 99-111 (in Russian).
  19. Sedov A.A., Frolov A.A. Computational study of the influence of some systematic factors on the temperature of the fuel in an ultrahigh-temperature gas reactor with prismatic fuel assemblies. VANT. Ser. Fizika yadernykh reaktorov. 2010, no. 3, pp. 80-90 (in Russian).
  20. Bojko V. I., Gavrilov M. P., Koshelev F. P., Mescheryakov V. N., Nesterov V.N., Ratman A.V., Shamanin I.V. Estimation of the service life of graphite fuel blocks of the GT-MGR reactor. Izvestiya Tomskogo politekhnicheskogo universiteta. 2005, no.5, pp. 81-84 (in Russian).
  21. Golovackij A.V, Nesterov V.N., Shamanin I.V. Optimum graphite operating temperature to ensure the design depth of nuclear fuel in the GT-MGR reactor. Izvestiya Tomskogo politekhnicheskogo universiteta. 2011, no.2, pp. 71-76 (in Russian).

critical fluence reactor-grade graphite damaging neutrons graphite lifespan nuclear fuel burnup GT-MHR HTGR

Link for citing the article: Bulakh O.I., Kostylev О.К., Nesterov V.N., Cherdizov E.K. 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. Izvestiya vuzov. Yadernaya Energetika. 2019, no. 3, pp. 40-52; DOI: https://doi.org/10.26583/npe.2019.3.04 (in Russian).