The GCFR System is considered within the GIF programme (The US Generation IV Implementation Strategy, 2003; Figure 12.3). It is a good candidate for electricity

production as well as actinide management. It may also be a candidate for hydrogen production. It is based on a closed fuel cycle and therefore offers a more sustainable fuel cycle option. The reference plant is 288 MWe. However, it requires significant advances in fuel and materials.

The GFR is a logical progression from VHTR gas-cooled systems described above and earlier European and US gas reactor technology. It is therefore seen as a longer term deployment option.

13.9.4.1 TOP Accident in a Sodium-Cooled ADS. Since the sodium-cooled ADS is initially sub-critical, fast or medium ramp rates can be accommodated for sizeable reactivity insertions. For example, calculations (Wider, 1997) of $170 per second for a total insertion of $2.65 at a sub-criticality of —$3 and $6 per second for a total insertion of up to $3 for a sub-criticality of —$5 showed no significant power excursion. For a critical reactor, the power would be 2000 times the nominal power.

For slower ramp rates, e. g. at $0.1$ per second, for a similar total insertion of about $3 at a sub-criticality of —$3 per second, a failure of a single channel occurred. At a sub­criticality of —$5, failure also occurred but later, and it did not occur at all for a — $10 sub-criticality. This compares with the critical reactor case in which pin failures in 1 out of 10 channels occur.

Thus for fast ramp rates without scram, the ADS behaves in a benign manner, for slow ramp rates there may be limited core damage at a later stage for insufficient sub-criticality or failure to scram the beam.

13.9.4.2 TOP and RIA Accidents in a Gas-Cooled ADS. Regarding fast and medium ramps, the gas-cooled fast ADS would behave similarly to the sodium-cooled ADS, i. e. benignly. For slower ramp rates, there could be rather more pin failures, because fuel dispersal may be less and, therefore provide less negative feedback.

13.9.4.3 TOP and RIA Accidents in a Thermal Molten Salt-Fuel Mixture. A thermal ADS would also act benignly under fast or medium ramp insertions. This system would also probably show an advantage for slower insertions compared with the fast systems above, because with a fluid system, the pins would probably not fail.

15.2. PLANT LIFE EXTENSION

The technical issues associated with plant ageing centre around the ageing of mechanical, electrical and materials ageing of plant components, particularly concretes and steels (Govaerts, 2001). The EC is funding a major research programme on this issue and a selection of some of the on-going projects is summarised below (Table 15.1). Utility practices for the safe management of nuclear power plant ageing in the EU are given in

Table 15.1. EC Research in nuclear fission energy (1998-2002)

Plant life extension and management Severe accident management Evolutionary concepts

Waste and spent fuel management

and disposal

Partitioning and transmutation Decommissioning of nuclear installations

Safety and efficiency of
future systems

Risk assessment and management Monitoring and assessment of occupational

exposure

Off-site emergency management Restoration and long-term management of

contaminated environments

EUR 19843 (2001). The phenomena include thermal fatigue and stress corrosion, and relate to the thermal and mechanical loads to which the components are subjected. Chemical factors may also be an issue.

There may also be practical factors that present a range of difficulties in presenting a case for life extension; e. g. hardware and software may become obsolete, original suppliers may no longer be able to supply replacements, etc. Computer codes may become outdated and no longer supported by developers. Rules and standards may change. Knowledge may reside in staff who have retired or about to retire, documentation may not be adequate without the presence of experienced original authors. Modern non-destructive testing (NDT) methods may be able to identify defects that had not previously been observed, but also in a positive sense, may be able to confirm the absence of defects (Table 15.2).

Sub-channel analysis is performed to determine the safety margin to boiling in peak rated channels in LWR assemblies. The flow and heat transfer distributions inside a fuel assembly can be analysed by sub-channel codes. The usual reason for analysis is to demonstrate compliance with the ‘Departure from Nucleate Boiling Ratio’ (or DNBR) requirements. The codes calculate the DNB from various channel-averaged parameters. Well-known sub-channel codes are COBRA and VIPRE (Table 16.4).

In such codes, two-phase flow is normally treated via a 3D flow model, which is coupled to a 1D model for fuel rods of different ratings. A detailed model of the heat transfer between the surface of the cladding and the coolant is included. The critical heat flux is calculated with a correlation.

Most fuel bundle designs are complex and it is necessary to consider the effect of such

support and promote mixing. There is a need to improve sub-channel codes to take account of these features and ensure, in particular, that void distributions are adequately modelled.

New fuel vendors supply correlations for their individual fuel rod designs. These are developed for fresh fuel and generally do not include the effects of burn-up, so their adequacy for highly irradiated fuel needs to be established. In highly irradiated rods, the surface may be significantly oxidised with different thermal-hydraulic performance characteristics. A particular issue may be different boiling characteristics and any influence on critical heat flux needs to be established.