Evaluation of NSRR reactor characteristics using a core transient behavior simulation code EXCURS-NSRR

The Nuclear Safety Research Reactor(NSRR) in Japan Atomic Energy Research Institute (JAERI) is a modified TRIGA-ACPR(Annular Core Pulse Reactor) which was constructed in 1975 in order to investigate the fuel behavior mainly under reactivity initiated accident (RIA) conditions. This reactor generates very sharp pulse power with the maximum of 23GW by rapid reactivity insertion of the maximum of 4.7$, and has capability to simulate a power burst in RIAs of power reactors. Fuel failure mechanisms and the fuel failure threshold in RIAs have been investigated through irradiation of test fuel rods in the NSRR. The control system and the operation data acquisition system of the NSRR were modified in 1989. By the modification, the controlled high power operation with various power shape became possible. Also on-line data acquisition of reactor data such as reactor power, regulating rod position, and so on, became possible by the modification. Evaluation of reactor characteristics became easy and accurate by detailed comparison of the time history of operation data and calculation results. Authors have tried to evaluate some parameters or constants of reactor characteristics by using a core transient behavior simulation code EXCURS-NSRR which consists of one point reactor kinetic equations and thermal equations ofmore » driver fuel and coolant. Especially in a relatively small and slow pulse power generation with reactivity insertion of less than 1$, evaluation of feedback reactivity coefficient and relation between reactivity insertion and regulating rod position was conducted. This article presents the evaluation results by the comparison between obtained reactor data and parametric calculation results. (author)« less