Design studies

Presently 19- element natural uranium fuel bundles are used in 220 MW(e) Indian PHWRs[1].

Figure 1 shows the fuel bundle and the element details. The core average design discharge burnup for these bundles is 7000 MW•d/Te U and maximum burnup for the bundle assembly goes upto of 15 000 MW•d/Te U.

Increase in fuel burnup beyond 15 000 MW•d/Te U using higher fissile content materials like slightly enriched uranium, Mixed Oxide and Thorium Oxide in place of natural uranium in fuel elements used in 220 MW(e) PHWRs is being investigated. Performance of these fuel bundles at high burnup is studied. Due to higher fissile content these bundles will be capable of delivering higher burnup than the natural uranium bundles. The maximum burnup studied with these bundles is 30 000 MW•d/Te U.

In PHWR fuel elements no plenum space is available and the cladding is of collapsible type. The additional fission product swelling and gas release due to use of high burnup fuels in PHWRs, needs to be accommodated within the fuel elements taking into account these factors.

a Nuclear Power Corporation of India Limited, Mumbai, India.

b Nuclear Power Corporation of India Limited, Mumbai, India.

c Nuclear Power Corporation of India Limited, Mumbai, India.

d Nuclear Power Corporation of India Limited, Mumbai, India.

e Nuclear Power Corporation of India Limited, Mumbai, India.

f Pressurised heavy water reactors.

g Burnup of MOX bundles is specified in MW•d/Te HE (tonne of heavy element).

FIG. 1. 19–Element fuel bundle for 220 MW(e) reactors.

Studies have been carried out for different fuel element target burnups with different alternative concepts. Modification in pellet shape and pellet parameters are considered.

R&D works for design of full core of Indian 220 MW(e) PHWRs with high burnup fuels are being carried out. Fuel design issues in respect of the higher fissile content have been reviewed [2,3]. Studies on reactor physics characteristics like reactor control, shut down margin, fuel and other systems thermal-hydraulic and material compatibility have been studied. Large scale utilisation of such bundles leads to substantial reduction in the volume of fuel bundles required. The core average discharge burnup increases with this scheme. Due to this, the fuelling rate comes down from 9 bundles / FPD in the case of Natural Uranium core to less number of bundles/FPD in the proposed high burnup cores and so consequently it is planned to go for 2/4 bundle shift scheme instead of 8 bundle shift..

The element power envelope upto the design burnup for different fiffile contents generated by reactor physics calculations are utilized for fuel design. The peak LHR (linear heat rate) of the element is maintained same as current natural U elements, to avoid any thermal hot spots. This has led to increase in residence period corresponding to higher burnups.

2.1 Fuel swelling

Fuel swelling occurs at high burnups. To accommodate higher burnups upto 30 000 MW•d/Te, it is proposed to reduce fuel density by 1 to 2%.

2.2 Residence period

The bundle residence period increases for high burnup fuel. For central channels it is 3 years and for outer most channels it is 5 years compared to 2.5 years presently. This increases oxidation of cladding.

The higher residence period has effect on 1) low cycle fatigue behaviour of fuel cladding and end plate, 2) corrosion and hydriding behaviour of the fuel cladding and end plate 3) fretting damage of fuel bundle and 4) power ramps at higher burnups. The fuel bundle flux depression factors across the elements are higher compared to natural U bundle. The zircaloy corrosion, hydriding and irradiation embrittlement behavior for the bundle for their residence period in core was estimated and is satisfactory for these extended burnups and powers.

2.3 Power ramp effect

The PHWR fuel bundle cladding inner surface is coated with graphite layer to provide resistance to power ramp SCC failures. The proposed 2/4 bundles refuelling shift for high burnup fuels will lead to power ramp on the bundles when bundles in the channel are shifted from 4 to 6^th location in the channel. This happens at a relatively high burnup of about 7500 MW•d/Te U, The ability of graphite coating to provide resistance to power ramp at these burnups is one of the main concern.

2.4 Thermo mechanical analysis

Thermo-mechanical analysis of the fuel element is carried out for the power histories reaching 30 000 MW•d/Te respectively. The fuel material (namely SEU, MOX, ThO2) properties are included in the code data base. The resultant thermo-mechanical parameters, such as fuel temperature, internal fission gas pressure etc, for these high burnup bundles were compared with respect to bundle with current burnups. The fuel element design analysis has been carried out using Fuel Design Analysis code FUDA[4] to check the limiting parameters at higher burnups like fission gas release, internal gas pressures, plenum volume requirements, centre temperature, sheath strains etc,.

The fuel pellet density, grain size and dish depth of the pellet and element LHR parameters, are the one which could be modified in order to limit the fission gas pressure without putting much difficulty in manufacturing. Hence a parametric study was undertaken for these parameters, while keeping the other geometric and operating conditions same as that of the present 19-element fuel bundle used in 220 MW(e) PHWR. The different designs are studied for burnups of 25 000 to 30 000 MW•d/Te. One of the case study is further explained.

Maximum center line fuel temperatures are found to be 2040⁰C and 1670 ⁰C for 19-element fuel bundle at 58 kW/m Linear Heat Rate(LHR) and 22-element fuel bundle at 50 kW/m LHR respectively.

The analysis results are given Table 1. The maximum fission gas pressures reduces with increase in dish depth, with annular pellets and also for low LHR. Decrease in density results in more porosity.

More porosity accommodates more gas. However it also decreases the thermal conductivity, which is found to result in enhanced fuel temperature in present study, and consequent more gas release. The net effect is found to be decrease in fission gas pressure by about 11% with the decrease in density from 10.6 g/cc to 10.5 g/cc. The strain in clad decreases by 17%.

TABLE 1. 19-ELEMENT AND 22-ELEMENT

Sl. No Bundle LHR Variable parameter Burnup Center Fisson Gas Internal gas Power

(KW)

KW/M MW•d/TeHE Temperature (Celsius)

Rel (%) Pressure (Mpa) 19–element fuel bundle peak power element

1 483 58 Normal parameters 25 000 2040 15.3 9.8

2 Double dish pellet 25 000 2040 14.8 5.5

3 Fuel p. grain size 40 μm 25 000 2040 14.7 8.9

4 Low fuel density 25 000 2080 18.4 7.4

5 Pellet central hole 30 000 1950 14.2 6.4

22–element fuel bundle peak power element

6 483 50 Normal parameters 30 000 1670 3.7 6.4

FIG. 2. Element fuel bundle.

The centerline temperature and radial temperature profile are found to be independent of dish depth increment. Fission gas pressure was found to be decreased with dish depth increment on account of more space availability for accommodation of fission gas. Reduction in gas pressure leads to decreased clad strain for increased dish depth pellets.

The studies indicated that, present fuel design is suitable upto 25 000 MW•d/Te U with minor modifications like use of higher grain size, more dish depth etc with decrease in density. For burnups beyond that either annular pellets or the earlier developed 22-element fuel bundle (Ref. 4) with lower LHR, shown in Figure 2 is being considered.