Patent Number: 059600500
Section: summary

FIELD OF THE INVENTION The present invention relates generally to a unique method of determining the fission heat flux for Uranium 235 (U235) fuel-bearing specimens inserted in a test holder of a nuclear reactor, and in particular to a method where the absolute value of the fission flux is obtained without requiring the specimen grams U235/cc, but instead by using experimental data from a thermocouple test specimen. BACKGROUND OF THE INVENTION Prior methods of determining the fission flux within the test specimen generally requires a detailed knowledge of the fuel region properties. This knowledge allows a determination of the conductivity, k, as a function of temperature. The French scientist J. B. J. Fourier in 1822 proposed in Cartesian coordinates the one-dimensional heat transfer equation q=-kA dT/dx, where q=heat rate, T=temperature, A=area, and dx=incremental thickness. This equation in most simple applications leads to .intg.kdT=S, where k must be known as a function of T to solve this equation. While these previous methods depend strongly on stated fuel properties, the method of the invention, as described below, relies more on experimental data, and the assumption of a similar thermal conductivity temperature variation. SUMMARY OF THE INVENTION In accordance with the invention, a unique method of determining the fission flux (usually given in BTU/hr/ft2) is provided which relies on experimental data that can be readily gathered. The data gathered includes the temperature of two thermocouple test specimens (TC's), each TC's gamma scan count ratio, and certain dimensional data. This enables the absolute fission fluxes for the two TC specimens to be obtained, and once this is done, by then using the gamma scan count ratio for the prime specimens in the same holder, the fission flux of the prime specimens can be readily derived. Assuming that the prime specimens have a similar conductivity temperature variation to that of the TC's, then their maximum temperatures can also be determined. As mentioned above, in the prior art methods of determining fission heat flux, the makeup of the fuel region must be known in detail and conductivity equations then need to be formulated. In contrast, the method of the present invention, in many cases when there are reliable measurements, enables the absolute flux and temperature values to be obtained without requiring this kind of detail. The method of the present invention can be used to lend support in confirming answers provided by conventional methods and programs for determining fission flux, to give an indication as to the accuracy of the measurements for particular test specimens, and to provide insight towards developing future fission flux programs. The method of the present invention enables absolute fission fluxes and temperatures to be obtained for U235 fuel-bearing specimens without using conventional techniques. Further, the invention provides that the thermocoupled test specimens are symmetrically placed in a reactor test holder at the same elevation, and, under these circumstances, enables an estimate of the effective grams of U235/cc of the prime specimen remaining to be obtained. Additionally, if the microscopic absorption to fission capture ratio (1+Alpha) is known, the prime specimen U235 fissions remaining can also be determined. It is noted that for the fissions estimate, any significant contributions to the self-shielding from material other than the U235 must also be known. Further, in cases where specimen conductivities vary similarly with temperature, and gamma-scan count ratios relate to the heat source, the present method can also be used to find the magnitude of the heat source. In accordance with a preferred embodiment of the invention, a method is provided for determining the absolute value of fission flux of a prime fuel-bearing specimen containing Uranium 235 inserted into a test holder of a nuclear reactor, the method comprising: inserting into the test holder of the nuclear reactor at least one prime specimen, a plurality of bulk water channels and at least two thermocouple test specimens, the thermocouple test specimens being positioned at the same level in said test holder and comprising first and second outer clads, a central backclad and first and second fuel fillers disposed between respective outer clads and the backclad; determining the temperature of the thermocouple test specimens and the bulk water channels, the gamma scan count ratios for said thermocouple test specimens and for at least one prime specimen, the thicknesses of the outer clads, fuel fillers and backclad of the thermocouple test specimens, and the surface water channel heat transfer coefficient of the thermocouple test specimens; calculating, using the temperatures of said thermocouple test specimens, the ratio of the gamma scan counts of said thermocouple test specimens, the temperature of the bulk water channels; the thicknesses of the outer clad, the fuel fillers and the backclad, and the bulk water channel heat transfer coefficient, the absolute value of the fission heat fluxes for the thermocouple test specimens; and calculating, using the absolute value of the fission heat fluxes for the thermocouple test specimens so determined and the gamma scan ratio for the at least one prime specimen, the absolute value of the fission heat flux for the prime specimen. Other features and advantages of the invention will be set forth in, or apparent from, the following detailed description of the preferred embodiments of the invention.