Patent Number: 055725600
Section: summary

INTRODUCTION 1. Field of the Invention The present invention relates to fuel assemblies for nuclear reactors, and in particular, to a fuel assembly for a boiling water reactor having fuel rods with varying pitch. 2. Background In the nuclear reactors intended for the generation of power, nuclear fuel assemblies are customarily of the rod type in which elongated nuclear fuel rods are supported or positioned at their lower and upper ends between lower and upper tie plates respectively and which are arranged in closely spaced parallel arrays in generally square configurations. As is well known, each fuel assembly for a boiling water type of water cooled reactor is typically enclosed by an outer channel, usually square, which confines the coolant which enters that fuel assembly to that particular fuel assembly until it exits the assembly at the top of the reactor core. The coolant passing through the fuel assembly consists of a mixture of liquid water and steam. At the bottom entrance of the fuel assembly, the coolant is liquid water having a temperature at/or slightly below its saturation temperature. As coolant flows upward through the assembly, power is transferred from the fuel rods to the coolant, steam is produced, and the fraction of steam in the coolant is increased. At the top of the fuel assembly, the coolant which has been heated by the fuel rods is primarily steam. As a result of a high volume fraction of steam in the upper region of the reactor core, the upper region of the core becomes under-moderated and over-enriched due to the presence of too few hydrogen atoms compared to the number of fissionable uranium or plutonium atoms. As a consequence, less than optimum uranium utilization results. Many attempts have been made in the prior art to increase the amount of moderator in the upper portions of a fuel assembly to improve neutronic efficiency. A commonly used design has been to incorporate one or more water rods, inner water channels or other coolant moderator flow conduits or paths within which single phase liquid water as the coolant moderator flows from the bottom of the assembly toward the top at a rate sufficient to prevent boiling of this flow. However, there are drawbacks to the designs which include such features. For example, a reduction in critical heat flux (CHF) performance occurs because the coolant moderator flow that must be supplied to the water rods/channels to preclude boiling inside these rods/channels occurs at the expense of the coolant moderator flow available for cooling the nuclear fuel rods. It is known in boiling water reactor fuel designs to include within fuel assemblies part-length fuel rods in order to mitigate the over-enriched and under-moderated conditions in the upper region of the core. Accordingly, some of the fuel rods in a fuel assembly are truncated at some intermediate elevation in the core. This leaves an unfilled coolant channel above that elevation. By providing a truncated fuel rod, several important benefits are achieved. For example, there is a neutronic advantage in increasing the amount of fuel in the bottom of the core as compared to the top of the core. A more axial uniformity in water to fuel ratio is thereby achieved with an associated improvement in fuel cycle costs, increased shut-down margin, reduced pressure drop (principally because of increased flow area, but decreased wetted surface also reduces the pressure drop), and increased core stability because the pressure drop reduction occurs at the top part of the bundle where two phase pressure drops are most significant. However, by including part length fuel rods, the amount of fuel in the fuel assembly is decreased. In addition to the upper region being under-moderated and over-enriched, a further problem in typical boiling water reactors is that the central region along the axes of the fuel assemblies may be under-moderated and over-enriched. In order to increase the amount of moderator so as to improve neutron moderation and economy, an elongated central water channel is provided which forms a centrally disposed path for the flow of moderator/coolant along the length of, but physically separated from, the fuel rods. The central water channel can have any cross-sectional area and/or geometry, positioned centrally and symmetrically within the outer channel, or asymmetrically displaced from the central axis within the outer channel, and can be oriented around its central axis so that its walls which extend the length of the assembly are either parallel or non-parallel to the walls of the outer channel. The central water channel can have a square cross-sectional area as described for example in U.S. Pat. No. 4,913,876 or an array of circular tubes or water rods extending along the length of the fuel assembly. Alternatively, the cross-sectional area of the central water channel is a cruciform and divides the rod array into quadrants as described for example in U.S. Pat. Nos. 4,478,786 and 4,795,608. Sufficient liquid coolant is circulated through the central channel to keep the contained coolant largely or completely in the liquid phase. The liquid moderator inside the water channel(s) not only increases moderation in the center of the assembly, but also increases moderation in the upper regions of the assembly. The presence of liquid as contrasted to gaseous moderator in the central region of the fuel assembly increases the nuclear performance of the assembly by providing a greater number of hydrogen atoms which functions, in part, to slow down neutrons and thereby increase the likelihood of further fissions. Another important attribute of a central water channel is that the void coefficient of reactivity is less negative. By having the void coefficient of reactivity less negative, reactor stability improves by reducing the coupling between core reactivity and core moderator thermal hydraulic conditions. The moderator in the fuel assembly that is not within the central water channel and which is termed active coolant surrounds the nuclear fuel rods and is heated by means of conduction/convection. As reactivity increases, heating of the active coolant is increased. Increased heating of the active coolant results in greater steam void formation and a reduction in moderation. The increase in voids and reduced moderation results in reduced reactivity. Heating of the coolant/moderator which is within the central water channel is relatively small and is largely unaffected by the heat released from the fuel rods. Thus, an assembly with a central water channel has a greater fraction of moderator in the core that does not become void when reactivity increases. There is thus less decrease in reactivity to steam void formation. Regardless of the particular configuration, each central water channel within a fuel assembly has an inlet disposed towards the bottom to allow subcooled liquid water to enter the central water channel and an outlet towards the top. The inlet subcooling and the flow rate inside the central water channel are such that the coolant which flows up inside the central water channel does not experience any significant boiling. The objective of incorporating such internal water channels is to increase the amount of liquid water within the fuel assembly and thus achieve increased neutron moderation in the center and top parts of the fuel assembly. The primary benefits of this increased moderation are improved fuel utilization and improved stability (e.g., less tendency towards coupled nuclear/thermal-hydraulic oscillations). In order to increase the size of the center water channel and to thereby increase the moderation in the upper portion of the core, designs have been utilized in which the upper portions of selected fuel rods are removed in order to accommodate an expanding central water channel as described for example in U.S. Pat. No. 4,957,698 and U.S. Pat. No. 4,968,479 or in which the upper portion of selected fuel rods are removed and replaced with water rods in fluid communication with the center water channel as described for example in U.S. Pat. No. 5,255,300. However, such designs which eliminate some or portions of some fuel rods thereby decrease the power generated by the fuel assembly. Similarly, if the diameter of the fuel rods adjacent the center water channel was decreased in order to accommodate the increasing size of the center water channel, the power generated by the assembly would decrease. It would thus be an advantage over the prior art to have a fuel assembly with a central water channel that varies in cross-sectional area and/or shape while maintaining the diameter of the fuel rods as well as retaining the same number of fuel rods as in an assembly having a central water channel with a uniform cross-sectional area and thereby not decreasing the amount of power capable of being generated by the fuel assembly. SUMMARY OF THE INVENTION In accordance with one aspect of the invention, a nuclear fuel assembly for boiling water reactors is provided comprising a plurality of elongated nuclear fuel rods; a lower tie plate for positioning the bottom ends of the plurality of nuclear fuel rods in an array having an at least one pitch; an upper tie plate for positioning the top ends of the plurality of nuclear fuel rods; an outer channel surrounding the plurality of nuclear fuel rods for conducting coolant/moderator about the plurality of nuclear fuel rods from the bottom of the assembly toward the top of the assembly; a spacer for providing support of the fuel rods over the length of the assembly and located between the upper and lower tie plates for positioning the fuel rods in a second array having a second at least one pitch, said fuel rods passing through apertures in the spacer and being retained in spaced apart relationship by said spacer; an inner channel having at least one wall for conducting coolant/moderator through the inner channel from the bottom of the assembly toward the top of the assembly.