Patent Number: 047598964
Section: description

DESCRIPTION OF THE PREFERRED EMBODIMENTS Referring now to the drawings where similar features of the invention are designated by the same reference numerals among the various figures of the drawings. FIG. 1 illustrates a type of pressurized light water nuclear reactor 10 to which the present invention may be adapted. The invention, however, is not to be limited to such a reactor 10 which is being described primarily for purposes of description and explanation of the invention. A pressure vessel 11 houses a nuclear core 12 which is structurally supported therein by a set of components which are often referred to as the reactor internals (not completely shown). The nuclear core typically comprises a plurality of fuel assemblies 14, square in cross section, and stacked side by side in a parallel array. The nuclear core 12 has a resulting shape which approximates that of a right circular cylinder with the periphery being irregular or stepped when viewed in cross section due to the square configuration of the fuel assemblies 14. This may be partially seen in FIG. 2. The reactor internals include a core barrel 15 which separates the reactor coolant flow entering the pressure vessel 11 (through nozzle 16) from the reactor coolant exiting the pressure vessel 11 (through nozzle 17). In this manner, the reactor coolant flow may be directed down the outside of core barrel 15, turn 180.degree. and flow up through the nuclear core 12. The reactor internals also serve to make the transition from the irregular shape of the core periphery to the circular shape of the core barrel 15. Typically, vertical stainless steel plates 18 are positioned against the irregular core pheriphery. The vertical plates are supported by a plurality of horizontally positioned former plates 19 which are bolted to the vertical plates 18. The former plates 19 are in turn, bolted to the core barrel 15. The space between the horizontal former plates 19 is filled with water which flows in the same direction through the nuclear core 12. Since this flow is core bypass flow, it is desirable to maintain the flow at a minimum value yet sufficient to cool the former plates 19, vertical plates 18 and core barrel 15. Pressure vessel 20 is typically made from steel plates which are welded together in the axial and/or circumferential directions as represented by the axis A--A and/or B--B in FIG. 1 which respectively represent the centerlines of such welds. It is to be understood that the locations of such welds are not necessarily fixed relative to the nuclear core 12. However, once a reactor is fitted with a nuclear core, then the location of the pressure vessel welds are fixed relative to that core and any other core later loaded into that pressure vessel throughout the lifetime of that reactor. Viewed differently, for any given reactor, the orientation of the irregular core periphery is fixed relative to the welds (A--A or B--B) joining the plates making up the pressure vessel. In the example shown in the drawings, horizontal weld 21 is located approximately at core midplane and vertical weld 22 is located opposite the corner 23 of fuel assembly 14". As illustrated, welds 21 and 22 are thus exposed to the most severe fast flux condition. In actual practice this may not be the case; there may be only one weld or such welds may be exposed to the least severe fast flux condition. It is necessary that the location of the actual welds of a pressure vessel be determined relative to the core location and orientation and relative to the power and fast neutron flux distribution (horizontal and vertical) output by a particular core. Still referring to FIG. 2, irregular line C--C in combination with plates 18 circumscribe an arbitrary area of the core 25 which for purposes of providing an example for the description of the invention is to be deemed to materially contribute to the fast neutron flux to which weld 22 is exposed. As can be seen, area 25 encompasses parts of fuel assemblies 14', 14", 14'", and 14"". Appropriate nuclear calculations are required to be performed to determine the actual contribution to the fast neutron flux from each fuel rod in the peripheral core area and to determine the amount of displacer rods and/or nuclear poisoning needed to reduce the fast flux at the welds of the pressure vessel 11 to an acceptable level. For further description of the invention, it will be assumed that the calculations show that core peripheral area 25 described by line C--C needs to be provided with displacer rods and/or poisoned and/or provided with nuclear reflective material. The displacer rods which may comprise solid or sealed hollow tubes fit within openings within the fuel assemblies which contain a nuclear moderator (in the described reactor 10, the moderator comprises the light water reactor coolant.) By displacing the moderator within area 25, less moderating of the fast neutrons (produced by the fissioning of the nuclear fuel) occurs, which causes less fission to occur and, therefore, causes less fast neutrons to be produced. In this manner, weld 22 is exposed to less fast neutrons. If a nuclear poison rod is used, it will have the beneficial effect of the displacer rods and absorb some of the slow or moderated neutrons which are available for fissioning; and, thereby, additionally reduce the production of the fast neutrons which additionally reduces the number of fast neutrons which would be absorbed by weld 22. The nuclear poison would also absorb some of the fast neutrons to which the weld 22 would be otherwise exposed. If a reflector rod is used, it will provide the beneficial effect of the displacer rod and will reflect some neutrons back into the core and away from weld 22. Any of the techniques or a combination of such techniques may be employed with equal effectiveness consistent with and as constrained by the aforementioned nuclear calculations. FIG. 3 illustrates, in the lower portion thereof, a partial cross section of the core 12 including core area 25 taken along the grids 29 of fuel assemblies 14', 14", 14'" and 14"". Grids 29 typically are positioned at various locations along the length of fuel assembly 14 and serve to space and support a plurality of parallel arranged fuel rods 26 and control rod guide thimbles 27 at appropriate distances from each other so as to allow the reactor coolant to circulate in heat transfer relationship with fuel rods 26. Such grids 29 are well known in the art. As can be seen, guide thimbles 27 comprise hollow tubes which are attached to grid 29 and are spaced or distributed over the cross section thereof. Fuel assemblies 14 are thus comprised of fuel rods 26, guide thimbles 27 and grids 29. Guide thimbles 27 serve to guide the movement of control rods which fit within guide tubes 27. Control rods primarily serve to control the power output of the nuclear reactor 10, and are well known in the art. Typically, control rods comprise pellets of a nuclear absorbing material stacked end on end within hollow tubes and sealed at the ends. A plurality of such tubes are then attached to a central hub which is connected to a control rod drive mechanism. Since any fuel assembly 14 may be placed at any location within core 12, it is common practice to have each fuel assembly 14 equipped with the same number of guide tubes 27 and that they be positioned at the same location within the fuel assembly. In this manner, the position and operation of the control rod assemblies are not constrained by any fuel assembly 14; and, as explained, any fuel assembly 14 can be placed anywhere in the core 12. Typically, control rod assemblies are not located at the core pheriphery, but the fuel assemblies 14 at the core periphery are equipped with guide tubes 27. Hence, the guide tubes 27 in the fuel assemblies 14 at the core periphery are not used. The centermost opening 33 in fuel assembly 14 is typically provided for purposes of instrumentation and accordingly, is not available for one of the rods of the present invention. The dots 28 in FIG. 2 within core area 25 represent the location of unused guide tubes 27 in fuel assemblies 14. Referring again to FIG. 3, a displacer/absorber/reflector rod assembly 30 is shown therein. Displacer/absorber/reflector rod assembly 30 comprises a plurality of displacer/absorber/reflector rods 31 arranged parallel to each other and attached to hub 32. The arrangement of displacer/absorber/reflector rods 31 on assembly 30 coincides with the location of guide tubes 27 in core area 25. Thus, each of rods 31 fit within guide tubes 27 of core area 25. Rods 31 may comprise a displacer rod in the form of a hollow sealed tube made, for example, from zirconium alloy or stainless steel. Or, rods 31 may comprise a displacer/reflector rod made, for example, from a solid material such as stainless steel, or any other suitable reflector material, or may comprise a sealed hollow tube filled with a reflector material such as zirconia, or any other suitable reflector material. Or, rods 31 may comprise a displacer/absorber rod made as conventional control rod absorber rods are made and using the same materials: for example, boron carbide, cadmium-indium-silver, hafnium, etc. Or, rods 31 may comprise a combination of such displacer rods containing absorbing materials and reflecting materials appropriately positioned along the length of rod 31. Such absorbing and/or reflecting materials may extend the full length of rods 31. Or rods 31 may comprise a portion of reflector and/or absorber material appropriately located between particular core axial positions with the remainder of the rod 31 being substantially nonabsorbing and/or nonreflecting. Or, rods 31 may have any desired length. In essence, rod assembly 30 may be precisely tailored in its makeup as determined by the nuclear calculations previously mentioned. The various combinations are limitless. Also. any core area may be provided with one or more rod assemblies 30. The displacer/absorber/reflector rod assembly 30 may be positioned within its designed area 25 during core loadings or reloadings and simply left there during subsequent reactor operation. Rod assembly 30 may be conventionally sandwiched between the core upper and lower support plates (not shown) so as to retain the same in its assembled position during reactor operation. In carrying out the method of the invention, the following procedure may be used. The relationship of the core 12 relative to the horizontal B--B and/or vertical A--A welds on the pressure vessel 11 are documented. Calculations and/or measurements of the fast neutron flux (greater than one million electron volts or other agreed upon value) are ascertained at the location of the welds. Using whatever criteria is desired, by which the value of the neutron flux is to be reduced, calculations are made which determine the size and location of the core peripheral area which is to be provided with one or more rod assemblies 30 and to determine whether reflection and/or absorption is required in addition to the displacer rod effect and over what length and at what axial location. Based on such calculations, the precisely tailored rod assemblies are fabricated and loaded into position at the desired core location. The reactor is thereafter normally operated at rated power. Measurements may be made during reactor operation to verify the design and placement of the displacer/absorber/reflector rods and if any changes are deemed warranted, they may be carried out as per the above-described procedure. While the invention has been described, disclosed, illustrated and shown in certain terms or certain embodiments or modifications which it has assumed in practice, the scope of the invention is not intended to be nor should it be deemed to be limited thereby and such other modifications or embodiments as may be suggested by the teachings herein are particularly reserved especially as they fall within the breadth and scope of the claims here appended.