Normal Operation

Many of these data are proprietary to fuel vendors and the data are not publicly available. Relative fuel bundle performance in regard to grids and safety margin is an important differentiator across different designs. Nevertheless, some data are available from international bodies, e. g. the NEA data bank maintains an international fuel performance experiments (IFPE) database which contains a wide range of fuel performance data (NEA Annual Report, 2002).

15.3.1 Transient/Accident Conditions Argonne National Laboratory (ANL). The USNRC, with the co-operation of EPRI, are sponsoring an experimental test programme at ANL (Fuel Safety Criteria

Table 15.3. Fuel behaviour


Experimental programmes

Normal operation Transient conditions

Vendor proprietary programmes, IFPE ANL, CABRI, Halden, NSRR

Fuel Safety Criteria Technical Review (2000), Bassette (2000), Papin and Schmitz (1999), Wiesenack (1997), Fujishiro and Ishijima (1994), Fuketa (1999).

Technical Review, 2000; Bassette, 2000) to determine the behaviour of high burn-up fuel under simulated LOCA conditions. Another primary objective is to provide data on the mechanical properties of high burn-up cladding for analysing the transients of interest in safety case analysis.

Secondary objectives are to develop a methodology for estimating fuel behaviour under LOCA conditions that can be applied to different cladding types of similar properties. There are also benchmark tests on fresh cladding to determine low-burn-up properties on modern day clads and to check for consistency with earlier results.

Three types of tests are being conducted — oxidation tests, to develop and validate kinetics models; quenching tests to evaluate current acceptance criteria or for establishing new criteria; and also structural response tests to establish whether coolable geometry or control rod insertion could be affected by external mechanical loads. CABRI. The CABRI reactor is managed by the Institute for Radiological and Nuclear Safety (IRSN), previously the Institute for Protection and Nuclear Safety (IPSN) in France. The main purpose of the facility is to test the response of fuel rods under reactivity-initiated accident (RIA) conditions (Fuel Safety Criteria Technical Review, 2000; Papin and Schmitz, 1999). It consists of a water pool driver reactor and a sodium-cooled experimental loop. Fuel rods are subjected to pulses of a few tens of milliseconds, and the timescales of the rod temperature transient are very fast, representative of fast reactor RIA conditions. Tests have been conducted for both high burn-up UO2 and MOX fuel. Where a fuel failure occurred in a MOX rod it is observed that events are more energetic than for the UO2 case. It is thought that enhanced fission gas migration to the grain boundaries, combined with higher porosity in the Pu rich region of the MOX fuel, results in the greater fuel dispersion and coolant ejection in the MOX case.

The CABRI facility is being modified to include a water loop to create more representative PWR conditions. This will enable temperature transients on timescales several orders of magnitude to be examined, more representative of RIA timescales in a PWR. The thermal-hydraulic conditions are likely to have a much greater influence on events than in the sodium-cooled case. The CABRI water loop programme is being co-ordinated in the international community by OECD. A schematic is shown in Figure 15.1. Halden. The Halden project in Norway has been running for many years and addresses many facets of fuel performance. It is co-ordinated by the OECD and is supported by various national and industrial bodies in about 20 countries. The Halden test facility incorporates high resolution and advanced instrumentation on rod thermal response, fission product gas release and pellet clad interaction (PCI) (Fuel Safety Criteria Technical Review, 2000; Wiesenack, 1997).


Figure 15.1. CABRI water loop. Source: IRSN (2002).

With the increasing emphasis on the performance of high burn-up fuel, rodlets of irradiated fuel from commercial reactors have been fabricated and tested for both normal and transient conditions. Burn-ups in ranges as high as 50-80 MWdkg-1 are included in recent test programmes. Tests can also be carried out with water loops, if appropriate.

The facility has also been used for testing the performance of MOX fuel and is a prime source of data for such advanced fuels. NSRR. The Japanese Atomic Energy Research Institute (JAERI) in Tokai operates the Nuclear Safety Research Reactor (NSRR) for investigations of fuel rod performance under transient RIA conditions (Fuel Safety Criteria Technical Review, 2000; Fujishiro et al., 1994; Fuketa, 1999). The reactor is an annular water pool type reactor.

The recent programme has focussed on MOX fuel. For fresh fuel, the response of MOX fuel was found to be consistent with the behaviour of UO2 fuel. Comparisons were made under similar test conditions. No effects of plutonium agglomerates were observed.

At increasingly higher burn-ups, the MOX fuel exhibited more fuel swelling and higher gas release than UO2 fuel. These phenomena are consistent with the results from the CABRI programme.

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