Patent Number: 053176138
Section: summary

This invention relates to a boiling water nuclear reactor having high power density. More specifically, a fuel bundle construction is set forth in which each discrete fuel rod, consisting, for example, of a stack of fuel pellets enclosed in tubular cladding, has its very own concentric fuel channel. The fuel rods and concentric channels are preferably arrayed in a triangular matrix. A core of such fuel bundles has the characteristic that all rods uniformly approach their respective thermal limits enabling higher power density in the reactor. BACKGROUND OF THE INVENTION Problem Statement Boiling water reactor power densities have been limited in the past to less than 56 kilowatts per liter (KW/1), primarily as a result of their original designs. These designs constrain the power outputs of these reactors due to thermal limits and stability considerations. Thermal limits include the maximum linear heat generation rate and the minimum critical power ratio. The maximum linear heat generation rate (MLHGR) is that maximum amount of heat output by a lineal foot of fuel rod. Normal MLHGR rates for a boiling water nuclear reactor are in the range of 12.1 to 14.4 Kw/ft (or, 40 to 47 Kw/m using purely metric units). Simply stated, the MLHGR is a limitation established by the fuel pellet swelling establishing a mechanical interference with the cladding containing the fuel rod. The MLHGR cannot be exceeded at any individual fuel rod within a fuel bundle without potential damage to that particular fuel rod within the fuel bundle. As no individual rod is permitted to be damaged within a fuel bundle, the entire bundle is limited in its performance to maintain the maximum linear heat generation rate in any given fuel rod location. It is to be understood that to the extent a particular bundle constituting part of a reactor core is limited in its output, the entire core is likewise limited. The minimum critical power ratio (MCPR) is the ratio of that level of fuel bundle power at which some point experiences transition from nucleate to film boiling compared to the then present output of the fuel bundle. This ratio is not permitted to be less than a numerical value of one anywhere within an individual fuel bundle. If the limit were to be exceeded at any given location within the fuel bundle, the temperature of the cladding of the fuel rod would rapidly increase due to increased resistance in the heat flow path from the interior of the fuel rod to the exterior of the fuel rod. Potential failure of the particular fuel rod cladding could follow. The concept of a ratio is utilized in establishing limits of critical power within the fuel bundle. The ratio is maintained at a limit where operating conditions--both expected in normal operations and during anticipated abnormal operating occurrences or "transients"--can occur without running the risk of damage to the sealed fuel rods within the reactor. In already designed nuclear reactors, these thermal limits are largely established by the original design. There is, however, a need to increase the power output density of nuclear reactors of new manufacture. Accordingly, the factors relating to the power output densities will be briefly reviewed. Conventional fuel designs will be briefly discussed, especially insofar as they incorporate many heterogenous distributions in their neutron density and related power output. Thereafter, reference will be made to certain new reactor designs. Regarding the factors relating to increasing power densities, vessel sizes are limited in diameter to approximately seven meters, given the desire to continue to use forging to manufacture such vessels in existing manufacturing facilities capabilities. There exists a reluctance to expand vessel fabrication facilities beyond existing size limits under present market realities. Therefore, each reactor vessel is practically limited in its diameter. This requires that the number of fuel bundles within a BWR core is therefore limited. Limitations also exist in establishing the active fuel length of fuel rod bundles since as fuel rod length increases, thermal margins and stability become of concern. The longer the fuel bundle, the greater the possibility of transition boiling unless considerable additional inlet coolant flow is provided. This is aggravated, however, by the higher fuel bundle pressure drop associated with increased length and inlet flowrate. Further, stability at certain power rates requires rods be maintained short. If the boiling length is too long and the two-phase pressure drop too high, thermal-hydraulic, and thermal-hydraulic-nuclear instabilities arise. As a practical matter, the active fuel length is limited to about 12.5 feet (or 3.81 m using metric units). Once it is understood that both vessel diameter and fuel rod length are limited as a practical matter, it becomes clear that the total reactor volume available in any given reactor vessel approaches a limit. Therefore, the practical volume limit for a reactor is about 100,000 liters. When a reactor is built, many costs are fixed and constant regardless of the power output of the installed plant. If the installed plant can have a higher power density, these fixed and constant costs become substantially more efficient. Thus, there is a need for a new fuel design approach with potential to achieve higher power density to reduce the capital costs of nuclear reactors. This will enable any given reactor to have higher power output. The forced circulation boiling water reactor is one alternative reactor that is able to achieve the higher power density requirements. Simply stated, such reactors--by forcing the flow of coolant along internal paths--have the ability to concentrate more power in a given plant location. A new concept boiling water reactor which boils all of the entering coolant to steam and then superheats the steam is another reactor which can benefit from the use of this fuel design. See my copending patent application entitled "STEAM COOLED NUCLEAR REACTOR WITH BI-LEVEL CORE", Ser. No. 07/681,246, filed Apr. 4, 1991, which is incorporated herein by reference. CONVENTIONAL FUEL DESIGNS Conventional fuel designs for boiling water reactors include discrete fuel bundles having groups of vertically upstanding fuel rods supported on a lower tie plate and maintained vertical by an upper tie plate. A channel surrounds the vertically upstanding fuel rods between the tie plates and isolates the fluid flow between the tie plates from the rest of the reactor. Among other things, this arrangement allows predictability of fuel bundle performance down to the fuel bundle level. The chief concern with the design of modern fuel bundles has been nuclear efficiency improvement. Specifically, fuel bundles are designed to extract the maximum energy from the loaded nuclear fuel undergoing fission, typically by striving for uniform fuel rod power levels while minimizing the introduction of neutron absorbing materials in the core. For a number of practical reasons, however, modern fuel bundles utilized in boiling water reactors are not uniform and include heterogenous power outputs on the individual fuel rods within the fuel bundles. For example, the fuel bundle channels are surrounded on the outside by the so-called core bypass volume exterior of the fuel bundle, a volume which is filled with water during operation. For shutdown purposes, the water is displaced by the insertion of control rod blades. Furthermore, nuclear instruments which measure neutron and gamma flux in the core for the purpose of measuring local and global power levels are also located in the water filled bypass region. As a consequence, the fuel rods adjacent to the channel operate in a higher thermal neutron flux due to neutron moderation in the bypass water (which is greater than the moderation provided by the steam/water mixture interior to the bundle) and thus produce more power. To counter this, fuel enrichment is varied relative to these variant fast and slow neutron flux densities, and water rods replace select interior fuel rod locations for adding required fast neutron moderation with the end of maximum power extraction in mind. For at least these reasons, although the fuel bundles are isolated into discrete--and hence predictable increments--the discrete fuel rods within modern fuel bundles are not homogenous relative to one another in their power output. When the fuel rods within any given fuel bundle are not homogenous, this as a practical matters means that some portions of the fuel bundles reach their thermal limits before other portions of the fuel bundles. Those having skill in the art will understand that once a thermal limit is reached any where within a fuel bundle, the other portions of that same fuel bundle, and possibly other fuel bundles in the reactor core, are limited to that power output where the local thermal limit will not be exceeded. STATEMENT OF THE PROBLEM TO BE SOLVED Although modern fuel design has been concerned with nuclear efficiency, when contemplating the appropriate power density for a new nuclear power plant design, there is a tradeoff between nuclear efficiency as it may impact energy utilization and fuel cycle costs--which generally favors low power density--and the fixed plant capital costs incurred at the time of plant construction--which favor high power density to gain improvement in economy of scale. For a high power density plant, certain conventional design constraints in establishing fuel bundle design will be ignored. Since this is the case, the reader will understand that the invention is claimed insofar as departure from these conventional design constraints are concerned. Thus the realization that fuel cycle costs are secondary to plant capital outlay as a practical matter and the concession of nuclear efficiency as a traditional constraint are part of the invention set forth in the following specification. SUMMARY OF THE INVENTION This invention approaches the problem of achieving higher power density by proposing a fuel design which does not necessarily emphasize nuclear efficiency, although some design elements are not inconsistent with this conventional goal either. Instead the design emphasis is on uniformity of fuel rod power output throughout the core. There results a reactor core with discrete fuel bundle units where all rods many uniformly approach their respective thermal limits. With all rods approaching their respective thermal limits with uniformity, it is possible to impart a high power density to the reactor. In the new design, the traditional BWR fuel channel surrounding a group of fuel rods is eliminated. In place of the traditional channel surrounding a group of fuel rods, each fuel rod is surrounded by its own individual cylindrical channel. This individual cylindrical channel on each fuel rod provides thermal hydraulic and heat transfer advantages to enable all fuel rods within the fuel bundle to uniformly approach their own thermal limits. At the same time, the individual fuel rod channels retain the concept of predictably segregating the fuel core into discrete fuel rod units to assure acceptable flow distributions without the need for complex 3-D analyses and multi-assembly thermal hydraulic confirmatory tests. The preferred fuel rod pitch is a triangular pitch between the individual fuel rods as they are discretely surrounded by their own channel. The new triangular geometry provides for more uniform (flat) power distributions within all fuel rods--and hence all groups of fuel rods. Bypass flow is introduced uniformly between the fuel rod channels, rather than heterogeneously in the channel gap and water rods as in present BWR fuel designs. Individual fuel rod channels can be orificed differently, as required, to match inlet flow to fuel rod power output to maintain uniformity between all fuel rods as they approach their respective thermal limits. Gaps between bundles can be eliminated, or at least minimized, yielding even higher power density. Furthermore, for regions of the cores where steam cooling is significant, radiant transfer of heat between the outer surface of the fuel rod and the inner surface of the channel appreciably improves overall heat transfer to the steam coolant (on the order of 5 to 104 of the total heat transferred). Spacers are judiciously applied within the fuel rod channel to limit relative displacement between the channel wall and fuel rod cladding and to serve as turbulent flow promoters.