Reactor kinetics codes are used to calculate assembly averaged neutron flux and power distributions in a reactor core under transient conditions. The UK PANTHER code developed by British Energy is a typical example (Hutt, 1996). It includes a neutron diffusion neutronics model, coupled with a 1D thermal hydraulics model for the core region. The code has also been coupled with the RELAP5 system thermal hydraulics code, see below, to provide a neutronic/primary circuit modelling capability. The code can perform reactor calculations, fuel management studies and safety transient analysis. It can also be used for on-line calculation support. Other codes include: RAMONA, PARCS (Joo, 1998), SIMULATE-K, CORETRAN and SAPHYR.

Whole-core events, such as macroscopic temperature changes cause global power changes and these can be modelled adequately with point-kinetics models. These models require as input, reactivity coefficients, the effective delayed neutron fraction, generation time and control rod worths. However, in some transient conditions such as rod ejection or control rod drop, localised events occur that require multi-dimensional neutron kinetics analysis with codes of the type mentioned above. One-, two — and three-dimensional models require neutronic input parameters such as assembly averaged neutron cross­sections and delayed neutron fractions that are obtained from the nuclear data codes. The neutronics codes typically model energy groups condensed into two energies. Delayed neutron fractions would usually be modelled on a nodal basis.

A review of applicable existing thermal gas cooled reactor experience and previous gas cooled fast reactor projects is given by Mitchell et al. (2001). Gas reactor physics methodologies have been established, e. g. WIMS and PANTHER for application in the UK gas reactor industry. Gas reactor physics methodologies are being extended to high — temperature reactor applications with pebble bed fuels, taking advantage of already existing experience. The fuel and core design for gas cooled fast reactors are necessarily different, e. g. the graphite pebble bed concept cannot be used because graphite is a moderator and also because of the fast neutron core reactivity sensitivity to geometry variation. The fast gas reactor core will probably be based on more conventional LMFBR design using MOX or UOX steel clad pellets, e. g. as in the ETGBR design.

A range of computer codes has been developed for fast reactor neutronics (IAEA-TECDOC-1083, 1999). These include codes based on classical diffusion theory, transport theory and Monte Carlo methods. Within the European fast reactor community, the European reactor analysis optimised system (ERANOS) code system has been developed. This system embodies a modular system of codes not only for performing neutronic system design calculations but also for experimental analysis of critical facilities.

Undoubtedly, more research will be needed to develop reactor physics methodologies for the evolutionary plants and certainly for the more innovative concepts. However, there already exists substantial pool of experience on which to build.