An analysis of in-core flux and temperature data for the evaluation of the performance of an operating reactor

A method of analysis of in-core flux measurements for nuclear reactors is presented. By appropriate data processing, the method determines separately the statistical and deterministic errors present in the measurement. Those errors are carried throughout the analysis for determination of the local power produced, the nuclear hot channel factor F_ΔH^N and the DNB ratio with a required confidence level. An application of the method is given using the flux measurement data obtained for the first core cycle at San Onofre Unit 1. In addition, the method is extended to the processing of the recorded data from in-core thermocouples to obtain the engineering hot channel factors F_ΔH^E and the coolant by-pass fraction. The results of this analysis stresses the importance of accurate processing of nuclear reactor in-core data. Valuable estimates can be made of the core flow distribution and the flow by-pass fraction. These estimates, however, require in-core and loop temperature measurements within an accuracy of ±0·2°F for a corresponding accuracy of 5 per cent on the by-pass flow fraction. This accuracy on temperature measurement is not presently obtainable but should be set as a goal for future pressurised water reactor instrumentation performances.

Accurate knowledge of the power and flow distributions under normal operating conditions does contribute significantly to the safe operation of nuclear power stations. In the selection of in-core instrumentation and its degree of sophistication such an advantage must be considered in the overall economic balance of the plant.