Patent Number: 051679060
Section: description

BEST MODE FOR CARRYING OUT THE INVENTION The present invention will be best understood by reference to FIGS. 1 and 2 which are, respectively, vertical and horizontal cross sections of portions of a pressurized water reactor utilizing displacer rods of the present invention. The overall construction of such nuclear reactors will be known to those skilled in the art. Also, the general construction of these reactors is given in numerous of the above-cited patents, for example U.S. Pat. No. 4,716,006 that is incorporated herein by reference to teach the overall construction of pressurized water reactors. FIG. 1 is a cut-away drawing of a single fuel element 10 of such a reactor. It contains, for example, a plurality of elongated metal-clad fuel rods 12 held in suitable upper and lower grids 14, 16. The fuel element 10 also has appropriately spaced guide tubes 18 to be used for regular axially movable control rods (not shown). A portion of these guide tubes 18 is used to support displacer rods 20 to be discussed in connection with FIGS. 3A, 3B, 4 and 5. It will be understood that a given nuclear reactor will have a plurality of these fuel elements 10, and that the number of fuel rods 12, their spacing and their fuel loading may vary depending upon the position of that fuel element in a specific nuclear reactor. Water is caused to flow through the interstices between the fuel rods 12 and the displacer rods 20, with this water (in this type of reactor) providing both the moderation of the neutrons and the cooling of the fuel elements. A transverse cross section of a portion of a typical nuclear fuel element of FIG. 1 showing the positional relationship of the fuel rods 12 and the displacer rods 20 is shown in FIG. 2. Also indicated is a typical instrument thimble 21 for this fuel element. It can be seen from these figures that the displacer rods 20 exclude the water from a portion of the fuel element 10. As discussed above, this is required to overcome the excess reactivity of the nuclear reactor. In most of the prior art, these displacer rods are completely withdrawn at selected positions and times so as to increase the quantity of the moderator within the fuel elements. According to the present invention, however, the displacer rods 20 are not moved. Referring now to FIGS. 3A and 3B, one embodiment of a displacer rod 20 is illustrated that does not require any movement in order to regulate the excess reactivity of the reactor. The various components thereof are illustrated as enlarged in order to better understand the construction. In this particular embodiment, there is a support sleeve 22 which is perforated as at 24 such that the fluid moderator of the reactor can penetrate into the interior of the sleeve. Within the sleeve 22 there can be axially-spaced support plates 26 which, in turn, can also be provided with perforations as at 28. Supported by these plates 26 are elements 30 of a material that has a selected sacrificial (dissolution, volatilization, sublimation, etc.) rate in the fluid moderator such that the volume of the elements 30 decreases at a rate to increase the proportion of the fluid moderator within the reactor at a rate sufficient to compensate for the changing excess reactivity of the reactor. In a reactor having water as the moderator, the elements 30 can be fabricated from one of the slowly-dissolving aluminum alloys, for example. As will be understood from a table referenced hereinafter, several materials are available with slow dissolution rates that can be used for this purpose. Further, there will be materials known to those versed in corrosion art that will have satisfactory "dissolution" rates. Since there are vertical temperature and neutron flux gradients in this type of reactor, and since the change in excess reactivity is non-linear, different alloys can be used along the length of a given displacer rod 20 to achieve the desired dissolution rate at the specific temperature and flux of those locations and thereby provide the desired rate of change of the volume of the displacer rods to correct for temperature and radiation effects. Other methods for control of the rate of displacer rod volume change relate to the construction of the displacer rod (such as changes in the components thereof described with respect to FIGS. 3A and 3B, below). This control of dissolution, volatilization, sublimation, etc. can be effected by controlling the surface-to-volume ratio of any sacrificial component of the displacer rods. For example, various shape configurations are envisioned, such as providing passageways through a generally cylindrical body 30" (see FIG. 4). Other methods of effecting a change in the surface-to-volume ratio will be known to persons skilled in the art. In FIG. 3A, the elements 30 of sacrificial material are depicted as pellets for convenience of illustration. Typically, these pellets would have a diameter of about 0.8 in. In other embodiments they can be configured as spherical or multiple spheres 30''' that provide further control of the change in the relative surface areas and the compositions of the units (see FIG. 5). Thus, the present invention is not to be limited by the physical configuration of the "sacrificial" material within the displacer rods 20 or by the physical arrangement or presence of the containment sleeves 22. FIG. 3B illustrates the general structure of the displacement rods (depicted in FIG. 3A) after substantial operation of a nuclear reactor into which these rods have been inserted. This shows that a portion of each of the elements 30' of sacrificial material has been removed; however, there is substantial structural integrity to the elements so that fragments are not incorporated into the flowing coolant/moderator. Not all reactors will require that the sacrificial material be contained in a sleeve. It may be in the form of rods or a series of unclad pellets with, for example, some form of stiffener to prevent entrapment of material in the flowing coolant. The particular choice of structure will depend upon the particular reactor environment. The desired compositional change along the length of the rods can be achieved by conventional powder-metallurgy techniques, for example. Although the structure illustrated in FIG. 3A is initially designed for use in a pressurized water reactor where the moderator is the flowing water coolant and the sacrificial material dissolves in the water, the same principal can be applied to nuclear reactors where it is desired to gradually change (increase) the ratio of a fluid (liquid or gas) moderator with respect to the quantity of unburned nuclear fuel. This fuel moderator can be either a liquid or a gas. Typical of such moderators are liquid sodium or helium. However, the concept of using a sacrificial material is not limited to these named moderators, but is applicable to all fluid moderators. In these instances, the material of the sacrificial material is chosen to provide a desired rate of dissolution, volatilization or sublimation in the fluid of choice. It is preferred that the reaction at the displacer rod result in no fine materials. For example, in a liquid-cooled reactor it is desired that the sacrificial material result in an ionic form within the fluid. Since some filtering of the fluid moderator can be accomplished in normal operation of the reactor and since most reactors provide for a daily filtration, an ion exchange bed can be added to remove the results of the dissolution. In the case of gas-cooled reactors, it is desired that the sacrificial material be volatilized or sublimed. In this case, appropriate gaseous separation techniques would be applied. For water cooled and moderated nuclear reactors an average corrosion rate of about 1.3 mg/cm.sup.2 -day will be required, although a range of about 0.5 to about 2.5 mg/cm.sup.2 -day is envisioned for the various types of nuclear reactors and for the non-linear change in the excess reactivity. This assumes a normal operating life of the reactor at 18 to 24 months. The actual rate will depend upon the temperature and pressure of the system. For a PWR operating at 580 degrees F and a pressure of 2250 psi, the average corrosion rate to achieve the necessary change in volume will be about 1 mg/cm.sup.2 -day. Listed in the single Table are several aluminum alloys with the published corrosion (dissolution) rate in water. It can be seen that there are numerous of these alloys that will provide the rates for the reactors currently of interest. Thus, knowing the operating temperature and pressure, the excess reactivity that is to be controlled, and the corrosion rates in the fluid of the reactor, an improved displacer rod can be constructed that will automatically adjust in volume as the nuclear fuel is burned. Of course, persons knowledgeable in corrosion, volatilization and sublimation are aware of other materials that exhibit rates in this range. In this manner, the fuel efficiency of the reactor will be increased in a much less costly manner than taught by the prior art. Although certain constructions and materials are discussed herein for illustration, these are not given as a limitation of the present invention. Rather, the invention is to be limited only by the appended claims or their equivalents when read together with a detailed discussion of the invention. TABLE ______________________________________ CORROSION RATES OF ALUMINUM ALLOYS IN AQUEOUS SOLUTIONS TREATMENT TEST SOL. RATE ALLOY TEMP. pH Other mg/cm.sup.2 -day REF. ______________________________________ 1100 H14 0.255 1 3004 H34 0.306 1 4043 H14 0.248 1 5005 H34 0.276 1 5050 H34 0.258 1 5052 H34 0.268 1 5154 H34 0.241 1 5454 O 0.257 1 5454 H34 0.253 1 5456 O 0.282 1 5083 O 0.347 1 5083 H34 0.277 1 5086 H34 0.322 1 2014 T6 0.477 1 2024 T3 0.756 1 2024 T86 0.596 1 2024 TB1 0.536 1 6061 T4 0.028 1 6061 T6 0.312 1 7075 T6 0.509 1 7079 T6 0.469 1 1199 1.15 2 5154 H38 1.04 2 5454 H34 1.11 2 5457 H34 1.05 2 5456 O 2.18 2 5456 H321 0.12 2 5083 O 1.11 2 5086 O 1.07 2 M388 500 5 22.4 3 M388 500 6.7 28 3 M388 422 300PSIG 9.35 3 X8001 500 5.5 92.3 4 X8001 600 5.5 240 4 X2219 550 9 100 4 198X 600 5.5 33 4 ______________________________________ REFERENCES: 1. ASM, "METALS HANDBOOK", Ninth Ed, Vol 13, pp 599. Weathering data for 1.27 mm thick Al alloys after 7 years exposure 2. ASM, "METALS HANDBOOK", Ninth Ed, Vol 13, pp 605. Summary of data from 10 years seawater exposures 3. C. R. BREDEN, N. R. GRANT, "SUMMARY OF CORROSION INVESTIGATIONS ON HIG TEMPERATURE ALUMINUM ALLOYS", Argonne National Laboratory, February 1960. Flow of water (7 fps) 4. N. R. GRANT, "SUMMARY OF CORROSION INVESTIGATIONS OF HIGHTEMPERATURE ALUMINUM ALLOYS", Argonne National Laboratory, September 1961. Flow of water (7 fps) 5. ASM ENGINEERING BOOKSHELF, "SOURCE BOOK ON SELECTION AND FABRICATION O ALUMINUM ALLOYS", American Society for Metals, 1978, pp. 9-11.