Abstract:
A method of determining the bow and twist of a nuclear fuel assembly that utilizes fiber optic shape sensing technology enclosed within a flexible sheath that transmits strain on the interior walls of a control rod guide thimble within the fuel assembly to a fiber cable of the shape sensing technology enclosed within the sheath. The sheath conforms to the interior dimensions of the interior walls of the guide thimble.

Description:
BACKGROUND 
       [0001]    1. Field 
         [0002]    This invention pertains generally to a nuclear reactor fuel assembly and more particularly to a method of determining whether a nuclear fuel assembly has deviated from at least some of its specifications. 
         [0003]    2. Related Art 
         [0004]    The primary side of nuclear reactor power generating systems which are cooled with water under pressure comprises a closed circuit which is isolated and in heat exchange relationship with a secondary side for the production of useful energy. The primary side comprises the reactor vessel enclosing a core internal structure that supports a plurality of fuel assemblies containing fissile material, the primary circuit within heat exchange steam generators, the inner volume of a pressurizer, pumps and pipes for circulating pressurized water; the pipes connecting each of the steam generators and pumps to the reactor vessel independently. Each of the parts of the primary side comprising a steam generator, a pump and a system of pipes which are connected to the vessel form a loop of the primary side. 
         [0005]    For the purpose of illustration,  FIG. 1  shows a simplified nuclear reactor primary system, including a generally cylindrical reactor pressure vessel  10  having a closure head  12  enclosing a nuclear core  14 . A liquid reactor coolant, such as water is pumped into the vessel  10  by pump  16  through the core  14  where heat energy is absorbed and is discharged to a heat exchanger  18 , typically referred to as a steam generator, in which heat is transferred to a utilization circuit (not shown), such as a steam driven turbine generator. The reactor coolant is then returned to the pump  16 , completing the primary loop. Typically, a plurality of the above described loops are connected to a single reactor vessel  10  by reactor coolant piping  20 . 
         [0006]    An exemplary reactor design is shown in more detail in  FIG. 2 . In addition to the core  14  comprised of a plurality of parallel, vertical, co-extending fuel assemblies  22 , for purposes of this description, the other vessel internal structures can be divided into the lower internals  24  and the upper internals  26 . In conventional designs, the lower internals function is to support, align and guide core components and instrumentation as well as direct flow within the vessel. The upper internals restrain or provide a secondary restraint for the fuel assemblies  22  (only two of which are shown for simplicity in this figure), and support and guide instrumentation and components, such as control rods  28 . In the exemplary reactor shown in  FIG. 2 , coolant enters the reactor vessel  10  through one or more inlet nozzles  30 , flows down through an annulus between the vessel and the core barrel  32 , is turned 180° in a lower plenum  34 , passes upwardly through a lower support plate  37  and a lower core plate  36  upon which the fuel assemblies  22  are seated and through and about the assemblies. In some designs, the lower support plate  37  and the lower core plate  36  are replaced by a single structure, the lower core support plate, at the same elevation as  37 . The coolant flow through the core and surrounding area  38  is typically large on the order of 400,000 gallons per minute at a velocity of approximately 20 feet per second. The resulting pressure drop and frictional forces tend to cause the fuel assemblies to rise, which movement is restrained by the upper internals, including a circular upper core plate  40 . Coolant exiting the core  14  flows along the underside of the upper core plate  40  and upwardly through a plurality of perforations  42 . The coolant then flows upwardly and radially to one or more outlet nozzles  44 . 
         [0007]    The upper internals  26  can be supported from the vessel or the vessel head and include an upper support assembly  46 . Loads are transmitted between the upper support assembly  46  and the upper core plate  40 , primarily by a plurality of support columns  48 . A support column is aligned above a selected fuel assembly  22  and perforations  42  in the upper core plate  40 . 
         [0008]    The rectilinearly moveable control rods  28  typically include a drive shaft  50  and a spider assembly  52  of neutron poison rods that are guided through the upper internals  26  and into aligned fuel assemblies  22  by control rod guide tubes  54 . The guide tubes are fixedly joined to the upper support assembly  46  and connected by a split pin  56  force fit into the top of the upper core plate  40 . The pin configuration provides for ease of guide tube assembly and replacement if ever necessary and assures that the core loads, particularly under seismic or other high loading accident conditions are taken primarily by the support columns  48  and not the guide tubes  54 . This support column arrangement assists in retarding guide tube deformation under accident conditions which could detrimentally affect control rod insertion capability. 
         [0009]      FIG. 3  is an elevational view, represented in vertically shortened form, of a fuel assembly being generally designated by reference character  22 . The fuel assembly  22  is the type used in a pressurized water reactor and has a structural skeleton which, at its lower end includes a bottom nozzle  58 . The bottom nozzle  58  supports the fuel assembly  22  on a lower core support plate  36  in the core region of the nuclear reactor. In addition to the bottom nozzle  58 , the structural skeleton of the fuel assembly  22  also includes a top nozzle  62  at its upper end and a number of guide tubes or thimbles  54 , which extend longitudinally between the bottom and top nozzles  58  and  62  and at opposite ends are rigidly attached thereto. 
         [0010]    The fuel assembly  22  further includes a plurality of transverse grids  64  axially spaced along and mounted to the guide thimbles  54  (also referred to as guide tubes) and an organized array of elongated fuel rods  66  transversely spaced and supported by the grids  64 . Although it cannot be seen in  FIG. 3  the grids  64  are conventionally formed from orthogonal straps that are interleafed in an egg crate pattern with the adjacent interface of four straps defining approximately square support cells through which the fuel rods  66  are supported in transversely spaced relationship with each other. In many conventional designs springs and dimples are stamped into the opposing walls of the straps that form the support cells. The springs and dimples extend radially into the support cells and capture the fuel rods therebetween; exerting pressure on the fuel rod cladding to hold the rods in position. Also, the assembly  22  has an instrumentation tube  68  located in the center thereof that extends between and is mounted to the bottom and top nozzles  58  and  62 . With such an arrangement of parts, fuel assembly  22  forms an integral unit capable of being conveniently handled without damaging the assembly of parts. 
         [0011]    As mentioned above, the fuel rods  66  in the array thereof in the assembly  22  are held in spaced relationship with one another by the grids  64  spaced along the fuel assembly length. Each fuel rod  66  includes a plurality of nuclear fuel pellets  70  and is closed at its opposite ends by upper and lower end plugs  72  and  74 . The pellets  70  are maintained in a stack by a plenum spring  76  disposed between the upper end plug  72  and the top of the pellet stack. The fuel pellets  70 , composed of fissile material, are responsible for creating the reactive power of the reactor. The cladding which surrounds the pellets functions as a barrier to prevent the fission by-products from entering the coolant and further contaminating the reactor system. 
         [0012]    To control the fission process, a number of control rods  28  are reciprocally moveable in the guide thimbles  54  located at predetermined positions in the fuel assembly  22 . Specifically, a rod cluster control mechanism  80  positioned above the top nozzle  62  supports the control rods  28 . The control mechanism has an internally threaded cylindrical hub member  82  with a plurality of radially extending flukes or arms  52 . Each arm  52  is interconnected to the control rods  28  such that the control rod mechanism  80  is operable to move the control rods vertically in the guide thimbles  54  to thereby control the fission process in the fuel assembly  22 , under the motive power of control rod drive shafts  50  which are coupled to the control rod hubs  82 , all in a well-known manner. 
         [0013]    As previously mentioned, the fuel assemblies are subject to hydraulic forces that exceed the weight of the fuel rods and thereby exert significant forces on the fuel rods and the fuel assemblies. In addition, there is significant turbulence in the coolant in the core caused by mixing vanes on the upper surfaces of the straps of many grids, which promote the transfer of heat from the fuel rod cladding to the coolant. The substantial flow forces and turbulence can result in severe fretting of the fuel rod cladding if motion of the fuel rods is not restrained. Fretting of the fuel rod cladding can lead to a breach and expose the coolant to the radioactive byproducts within the fuel rods. These same forces can cause vibrations of the fuel assemblies which are restrained by their close proximity to the adjacent assemblies or peripheral core internal hardware. These close tolerances require that the fuel assemblies be manufactured to exacting standards, avoiding any bow or twist which might arise from, for example the welding of the guide thimbles to the grid straps. Any bow or twist may inhibit the insertion or withdrawal of the fuel assemblies from the core. 
         [0014]    Accordingly, a new method is desired that will confirm that these close tolerances are satisfied. 
         [0015]    Furthermore, Such a method is desired that can be carried out expeditiously as not to impede the fuel assembly manufacturing process. 
         [0016]    Further such a method is desired that can also be carried out on spent nuclear fuel assemblies. 
         [0017]    In addition, such a method is desired that can accommodate thimble tubes having varying diameters or thimble tubes of different diameters. 
       SUMMARY 
       [0018]    In accordance with this invention these and other objects are satisfied by a method of determining any distortion along an elongated, axial length of a nuclear fuel assembly having a top nozzle and a bottom nozzle axially spaced from the top nozzle and a thimble tube axially extending between the top nozzle and the bottom nozzle. The method comprises the step of positioning a plurality of strain gauges along a centerline of the thimble tube extending substantially from the top nozzle to the bottom nozzle. Each of the strain gauges is maintained in physical contact with an inside wall of the thimble tube around an inside circumference of the thimble tube at an axial location of the corresponding strain gauge. The output of each of the strain gauges is then transmitted to a remote location. 
         [0019]    Preferably, the strain gauges are fiber optic strain gauges and in one embodiment the strain gauges are enclosed within an outer sheath that extends substantially the axial length of the thimble tube extending substantially between the top nozzle and the bottom nozzle. Desirably, the sheath substantially occupies an entire space between the inside wall of the thimble tube around the entire circumference of the inside wall and the strain gauges at the axial locations of the corresponding strain gauges. In one preferred embodiment the outer sheath is configured to be removable from the strain gauges and replaced with an outer sheath having a different outside diameter to accommodate different thimble tubes having different inside diameters. In one such embodiment the outer sheath has an outside diameter that varies along an axial length of the outer sheath to mate with the inside diameter of the thimble. Preferably, the sheath is flexible. 
         [0020]    In still another embodiment the strain gauges are spaced along the centerline of the thimble tube. In all such embodiments the strain gauges are configured to determine the twist and bow of the nuclear fuel assembly and the strain gauges provide a substantially continuous measurement along the axial length of the thimble tube. 
     
    
     
       BRIEF DESCRIPTION OF THE DRAWINGS 
         [0021]    A further understanding of the invention can be gained from the following description of the preferred embodiments when read in conjunction with the accompanying drawings in which: 
           [0022]      FIG. 1  is a simplified schematic of a nuclear reactor system to which this invention can be applied; 
           [0023]      FIG. 2  is an elevational view, partially in section, of a nuclear reactor vessel and internal components to which this invention can be applied; 
           [0024]      FIG. 3  is an elevational view, partially in section, of a fuel assembly illustrated in vertically shortened form, with parts broken away for clarity; 
           [0025]      FIG. 4  is a schematic cross sectional view of a fiber optic sensing cable and sheath in accordance with one embodiment of this invention; 
           [0026]      FIG. 5  is a schematic cross sectional view of the fiber optic sensing cable and sheath shown in Figure adapted for a different sized thimble tube; and 
           [0027]      FIG. 6  is a perspective view of a top of a fuel assembly skeleton showing a fiber optic sensing cable and sheath being inserted into a thimble tube. 
       
    
    
     DESCRIPTION OF THE PREFERRED EMBODIMENT 
       [0028]    The current method to measure a fuel assembly bow and twist during production and after irradiation is derived from grid envelope measurements. A more direct method of measuring fuel assembly strain induced deformation during manufacture would likely yield a better understanding of the causes and engineering solutions which could avoid such occurrences and/or provide for a cost effective, acceptable fix. Furthermore, such a method that could efficiently be applied to irradiated fuel assemblies could provide information on the acceptability of reinsertion of a fuel assembly into a different core location during refueling or whether special accommodation needs to be made for a spent fuel assembly for storage. Such a method could also advance the state of the art in measuring assembly strain and deflection in a development laboratory. 
         [0029]    Another method to measure fuel assembly distortion is to visually compare a string that is stretched along the length of the assembly to determine the profile of the assembly. This method does not provide accurate information during manufacture and does not readily identify both twist and assembly bow, and is not practical to apply in an irradiated environment. This invention overcomes those limitations. 
         [0030]    A new three-dimensional shape sensing technology has been developed by NASA. This new technology is commonly referred to as fiber optic shape sensing. This technology involves mounting many fiber optic strain gages along the length of a fiber optic cable and using the strain measurements to calculate X-Y-Z coordinates along every point of the fiber optic cable. These products are commercially available from companies like LUNA, Roanoke, Va. and 4DSP, Austin, Tex. A video of the technology can be viewed at: 
         [0031]    http://lunainc.com/growth-area/fiber-optic-shape-sensing/. 
         [0032]    This invention uses this technology in order to accurately measure the profile of a fuel assembly thimble tube. The fiber optic cable would be inserted into a specially designed flexible sheath  84 , one embodiment of which is shown in schematic cross-section in  FIG. 4 , the combined sheath  84  and fiber optic cable being referred to as the strain measurement assembly  88 . The sheath  84  is used to both protect the fiber optic cable  86 , as well as to position the cable in the center of the thimble guide tube  54 . The outer dimensions of the sheath  84  has the same dimensions as the ID of a thimble tube  54 , including the dashpot region in which the inside diameter of the thimble tube  54  narrows. The sheath has to have a sufficient density to transfer the strain on the thimble tube to the fiber optic cable  86 . The sheath  84  may be constructed out of a polyethylene based shielding material or other materials having similar characteristics or other materials having similar characteristics. These materials are commercially available from Shieldwerx in Rio Rancho, N.M. The larger diameter of the sheath  84  is shown by reference character  88  and the smaller diameter by reference character  90 . Note that the dimensions are not drawn to scale and the thickness of the guide thimble walls is exaggerated to highlight the close contact of the sheath  84  with the inner walls of the thimble tube  54 . Multiple interchangeable sheaths are available to accommodate each type of thimble tube, as the inner profile for a 14×14 fuel assembly thimble tube is different than that of a 17×17 fuel assembly thimble tube.  FIG. 5  is a variation of the embodiment shown in  FIG. 4  to illustrate a different sized sheath to accommodate a different sized thimble tube. Like reference characters are used among the several figures to identify corresponding components. The sheath and fiber optic sensor would be inserted through the top of the assembly  22  into an individual thimble tube  54  as shown in  FIG. 6  and would be inserted all the way into the dashpot portion of the thimble tube. A remote data acquisition system would be used to gather the data for future analysis. Potential practical applications of this technology are: 
         [0033]    1. Measuring assembly bow and twist in irradiated fuel assemblies; 
         [0034]    2. Fuel assembly bow and twist inspection for product release instead of the current practice of measuring discreet points on the grid envelope; 
         [0035]    3. Measuring deflection and strain of fuel assemblies in tests performed in a development lab (for instance: the fuel assembly mechanical test, flowing water damping test, etc.). This would potentially replace LVDTs to measure deflection and the welded-on strain gages used to measure strain; and 
         [0036]    4. Measurement of guide tube distortion—the guide tube is the apparatus above the reactor core that houses the core components; this structure is known to distort during manufacturing (after welding) and the fiber optic shape sensing technology may provide a means to measure this distortion. 
         [0000]    The foregoing uses are just examples of the benefits of this invention and other uses may become apparent upon its implementation. 
         [0037]    While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof.