A thermodynamic approach for advanced fuels of gas-cooled reactors

https://doi.org/10.1016/j.jnucmat.2005.04.041Get rights and content

Abstract

For both high temperature reactor (HTR) and gas cooled fast reactor (GFR) systems, the high operating temperature in normal and accidental conditions necessitates the assessment of the thermodynamic data and associated phase diagrams for the complex system constituted of the fuel kernel, the inert materials and the fission products. A classical CALPHAD approach, coupling experiments and thermodynamic calculations, is proposed. Some examples of studies are presented leading with the CO and CO2 gas formation during the chemical interaction of [UOx/C] in the HTR particle, and the chemical compatibility of the couples [UN/SiC], [(U, Pu)N/SiC], [(U, Pu)N/TiN] for the GFR system. A project of constitution of a thermodynamic database for advanced fuels of gas-cooled reactors is proposed.

Introduction

Within the frame of the Generation IV project, CEA is interested in two types of gas-cooled reactors [1]:

  • the high temperature reactor (HTR) whose development is mainly based on the knowledge achieved in the seventies. The current needs are related to a higher operating temperature for the fuel and the structural materials;

  • the gas cooled fast reactor (GFR) for electricity production, whose fuel cycle is optimised to recycle actinides and to minimise the waste production. No complete reference system exists. The needs are considerable and concern the choice of the core and structural materials, the passive safety system capabilities and the recycle techniques.

In both cases, the fuel may operate at about 1000–1200 °C in normal conditions and may reach 1600–1700 °C in case of accident.

For the HTR concept, classical TRISO particles are considered (Fig. 1) [2], [3]. The fuel kernel, made of pure UO2 or of a mixture of UO2 and UC2, is surrounded by several successive coating layers:

  • a graphite buffer layer that accommodates the noble and CO, CO2 gas release as well as fuel swelling;

  • an inner pyrolytic carbon layer that protects the SiC layer from chemical reactions with both fuel and fission products;

  • a silicon carbide layer that decreases solid fission product diffusion out of the particle and enhances the mechanical behaviour of the system;

  • an outer pyrolytic carbon layer that protects the SiC layer and contributes to the particle mechanical strength; in case of failure of the SiC layer, this layer must play the role of barrier towards fission product diffusion. ZrC is also considered to replace SiC.

For the GFR system, the criteria for the choice of the core materials are the following ones: a high volume fraction of actinide, a geometry and thermal properties that allow a fast cooling, gas temperatures ranging from 400 to 850 °C (1200 °C–1600 °C in accidental conditions), a mechanical strength towards gas pressure and the fission product retention [4]. (U, Pu)C carbides and (U, Pu)N nitrides are candidates for the fuel kernel because of their high actinide density, and their elevated decomposition temperature and thermal conductivity. Several fuel forms are considered: composite ceramic–ceramic fuel (cercer) with closely packed fuel kernels or fibers, advanced fuel particles with large fuel kernels and thin coatings or ceramic clad, solid-solution metal (cermet) fuels.

The choice of the fuel coating materials has to meet some constraints regarding fast neutron damage, mechanical behaviour, thermal properties … but also chemical compatibility with the fuel kernel from 1000 °C to about 2000 °C. The most promising materials for core structures are ceramics such as SiC, ZrC, TiC, TiN, ZrN … that may be inert towards the fuel kernel [4]. In case of a chemical interaction between the fuel kernel and the ceramic, an intermediate layer of a material could be added to play the role of barrier.

Section snippets

Needs

For the HTR system, the thermomechanical behaviour of the particle is function of CO, CO2 and fission product release which gas pressures must be well known to calculate the stresses on the inner pyrolytic carbon, SiC and outer pyrolytic layers as well as of the chemical state of fission products and their diffusion coefficients which influence thermal conductivity, creep and melting point of the fuel [5]. Thermodynamic calculations or empirical laws are commonly used to estimate the CO and CO2

Thermodynamic approach

A usual coupling of experiments and thermodynamic calculations is proposed by using the CALPHAD method. In this method, the Gibbs energy functions of all solid, liquid and gas phases are assessed on the basis of the available experimental data (both phase diagrams and thermodynamic data). The CALPHAD method presents several advantages: it allows to calculate thermodynamic equilibria and associated phase diagrams for complex materials containing a lot of elements from the extrapolation of binary

Examples

The first example concerns the HTR fuel assembly when the two following ones lead with GFR’s system.

Conclusion

For both HTR and GFR systems, the high fuel operating temperature requires the assessment of both thermodynamic data and phase diagrams for the [fuel kernel + inert materials + fission products] system, specially for carbide and nitride fuels. A thermodynamic database associated to a Gibbs energy minimizer code is a necessary tool to help the fabrication process and the fuel design as well as to understand and predict the physicochemical behaviour of such complex systems in normal and long duration

References (32)

  • G.K. Miller et al.

    J. Nucl. Mater.

    (2003)
  • A. Heiss

    J. Nucl. Mater.

    (1975)
  • J.F.A. Hennecke et al.

    J. Nucl. Mater.

    (1971)
  • R. Lorenz et al.

    J. Inorg. Chem.

    (1969)
  • T.M. Besmann et al.

    J. Chem. Thermodyn.

    (1982)
  • C. Guéneau et al.

    J. Nucl. Mater.

    (2002)
  • P.Y. Chevalier et al.

    J. Nucl. Mater.

    (2001)
  • P.E. Potter

    J. Nucl. Mater.

    (1972)
  • H.J. Seifert et al.

    J. Alloys Compd.

    (2001)
  • H. Bernard

    J. Nucl. Mater.

    (1989)
  • F. Carré, NEA workshop on ‘R&D needs for current and future nuclear systems’, 6–8 November 2002,...
  • C. Than et al.

    Nucl. Des. Eng.

    (2002)
  • N. Chauvin, in: The 2004 Frédéric Joliot and Otto Hahn Summer School, August 25–September 3, 2004, Cadarache,...
  • M. Phelip, G. Degeneve, M. Pelletier, F. Michel, P. Guillermier, The ATLAS HTR Fuel Simulation Code: Objectives,...
  • A. Petti, T.J. Dolan, G.K. Miller, R.L. Moore, W.K. Terry, A.M. Ougouag, C.H. Oh, H. D. Gougar, Report...
  • R. Schram, R. Konings, W. Rijinsburger, ‘TBASE CONSULT Manual’, The Netherlands Energy Research Foundation ECN, updated...
  • Cited by (37)

    • Experimental investigation and thermodynamic modeling of the U–Nb system

      2021, Journal of Materials Science and Technology
      Citation Excerpt :

      Uranium is an important material for nuclear fuel. However, it cannot be directly used in a reactor owing to its poor mechanical properties and corrosion resistance [1–3]. Its properties such as corrosion and oxidation resistance, ductility, and strength can be improved by adding Nb, Ti, Mo and Zr since the addition of these elements enhances the stability of the high-temperature bcc γ-U phase.

    • Thermodynamic assessments of the U–Nb–Mo and U–Nb–Cr ternary systems

      2021, Calphad: Computer Coupling of Phase Diagrams and Thermochemistry
      Citation Excerpt :

      It has been widely considered that U-based nuclear fuels are good candidates for next-generation reactors due to their high thermal conductivity and high melting points [1,2]. The investigations of U-based nuclear fuels focus not only on the uranium ceramics, but also on the U-based metallic alloys [3–5]. Extensive studies of the U-X binary systems have been carried out for developing new nuclear materials which could have low-enriched uranium to achieve the decrease of proliferation risks [6,7].

    • Actinide Alloy Phase Diagram

      2020, Comprehensive Nuclear Materials: Second Edition
    View all citing articles on Scopus
    View full text