This invention relates to fuel rods for fuel assemblies used in light water nuclear reactors.

Fuel assemblies for light water nuclear reactors comprise an array of cylindrical rods termed fuel rods. Each fuel rod contains a stack of pellets made of a fissile material such as uranium, especially uranium 235.

The reaction which produces energy in a nuclear reactor is the neutron-induced fission of uranium atoms. The fission reactor rate has a spatial distribution which is affected by the neutron flux distribution. Because the distribution of fission-inducing neutrons is approximately cosinosoidal along the axis of the fuel rods, the depletion of fissile fuel, or burnup, distribution along the length of a fuel rod is non-uniform, with the ends of the rod receiving less burnup than the center. This is inefficient utilization of the uranium 235 in the fuel rod ends.

One well-known improvement is to add pellets comprising naturally enriched uranium (.711 weight percent U-235 in U-238) or depleted uranium (containing less than .711 wt.% U-235) at the rod ends. These pellets are called axial blanket pellets and function to generate plutonium by the absorption of neutrons by the uranium 238. Because plutonium is fissile, the blanket pellets improve the burnup distribution somewhat.

A fast reactor fuel pin having end portions containing fertile material is known from FR-A-1 565 786, wherein end regions of the fissile material have reduced diameter zones.

It is the principal object of the present invention to provide a fuel rod design which accomplishes an improved axial burnup distribution.

With this object in view, the present invention resides in an improved fuel rod for a fuel assembly for a light water nuclear reactor which fuel rod has an axial stack of pellets, said stack of pellets having a central region including fuel pellets of a first enrichment of fissile fuel and having at least at one axial end of said rod a region of blanket pellets of a second enrichment lower than said first enrichment and containing fertile species, characterized in that adjacent said central region, there are provided regions of low density pellets having a total fuel density, as expressed in units ^qf fuel mass per unit of axial length, lower than the density of said fuel pellets of said central region.

With this arrangement high uranium utilization is achieved since lower fuel density at the ends of the rod favorably distributes the neutron flux and burnup of fuel.

The invention will become more readily apparent from the following description of a preferred embodiment thereof shown, by way of example only, in the accompanying drawings, wherein:

• Figure 1 is a graph of a typical prior art and altered axial burnup distribution;
• Figure 2 is a schematic of a fuel rod; and
• Figure 3 is a schematic of a low density fuel pellet.

A high uranium utilization fuel rod is provided which improves the neutron flux distribution and consequently the resulting burnup distribution approximately as illustrated by Figure 1 in which curve 1 is representative of the axial burnup which occurs by the prior art and curve 2 is the axial burnup which occurs with the new design. The advantage of the improvement is the greatly decreased differential in burnup between the rod middle and rod ends from curve 1 to curve 2.

The high uranium utilization fuel rod is an improvement over even the axial blanket concept. An axial blanket is still employed, but low density fuel pellets 3 are also included on the ends of the fuel rod 4 as shown in Figure 2. Fuel rod 4 in Figure 2 comprises a stack of pellets within a clad tube 8, which stack of pellets has two regions of "standard pellets" 6, two regions of "blanket pellets" 5 (one at each end of rod 4), and two "reduced density" pellet regions 3. Figure 2 also shows an "optional" standard pellet region 7.

The low density fuel pellets 3 could be hollow annular pellets, (9 in Figure 3) lower density pellets, smaller pellets or other concept designed to reduce the uranium loading per unit length. The U-235 enrichment of these low density pellets 3 would typically be the same as the standard pellets 6 and 7 of the same rod but would not have to be.

Figure 3 is a schematic of a hollow center annular pellet 9. This embodiment, and the other concepts named above, achieves a lower uranium density per axial inch along rod 4 as compared to the density in regions 6 and 7.

The characteristics of the pellets used in the various regions are summarized in Table I.

The improvement in utilization of uranium in the upper and lower ends of the fuel rod upon inclusion of low density pellets 3 is due to changes in the usage of neutrons occasioned by the change in the local ratio of fuel (uranium) to moderator (hydrogen); a ratio termed by nuclear engineers, the H/U ratio. Fewer neutrons leak out of the fuel region without engaging in fission reactions with a decreased H/U ratio.

A preferred embodiment of the invention is described in Table II.