VALIDATION OF THE FAST AND THERMAL NEUTRON FLUX PROFILES IN THE ADVANCED TEST REACTOR FOR THE AGR-3/4 EXPERIMENT

摘要

The AGR-3/4 experiment was designed to irradiate TRISO particle fuel in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. Temperature control of the test capsules containing the TRISO particles was important since the goal was to measure diffusion rates of fission product elements through graphite as a function of temperature. During the experiment, differences were observed between temperatures measured by the thermocouples and temperatures calculated with the thermal analysis model. To rule out the possibility that the used ATR driver fuel elements were the root cause of these temperature differences, an experiment was devised to measure the axial fast and thermal neutron flux profiles in the northeast flux trap containing the AGR-3/4 experiment. An important input parameter to the thermal analysis model was the calculated compact heat rates from the MCNP physics model. To verify these calculated heat rates were consistent with the ATR driver fuel element loading and operating lobe power, axial flux wires were inserted into ATR northeast flux trap. The measured fast and thermal neutron axial flux profiles derived from the flux wire activation showed excellent agreement with the calculated flux profiles. This agreement dispelled any concerns that the used ATR fuel elements in the northeast lobe might have introduced unusual axial flux variation. In addition, the agreement provided a validation basis for the physics calculation, specifically, (1) MCNP physics model, (2) Monte Carlo depletion methodology, (3) method used to load the ATR driver fuel, and (4) ATR measured lobe power data used in the calculation.