Core management and calculation tools for the WWR-M research reactor in Ukraine

Full text: The WWR-M research reactor in Kiev, Ukraine is used for neutron physics and materials research, radioisotope production and neutron transmutation doping of silicon. The total power of the reactor is restricted by 10 MW. The number of channels for experiments, isotope production and neutron transmutation doping of silicon is not too large. Moreover, the current design and requirements to the control rod system do not allow the channels to be near the center of the core. Core loading patterns satisfying all the safety constraints and fuel requirements, fuel types used, number of fuel assemblies in the core and discharged fuel burnup should be determined by the optimization procedure using computer codes. To provide high reliability and safety of the reactor, computer codes based on adequate models should be applied. Diffusion approximation is invalid for neutronics modeling of the WWR-M reactor because of neutron streaming phenomenon in beam tubes. The Monte-Carlo method and other high-order approximations of the neutron transport equation can be used for such calculations but they are very computationally intensive. The iterational hybrid method combining diffusion model with higher approximations of the neutron transport equation has been developed for neutronics calculation of the WWR-M reactor. This technique has been examined by comparing its results for 1-D and 2-D test problems to solutions obtained for the same problems using the Monte-Carlo and high-order discrete ordinate methods. High efficiency and accuracy of the iterational hybrid method and possibility of its application for neutronics calculation of the WWR-M research reactor using a personal computer has been proved. The code VICA based on this method has been developed for 3-D neutronics calculation of the WWR-M reactor. The results of calculation were shown to be consistent with the results of measurement. The total number of measurement data was about one hundred while average deviation of the calculated and measured values of the thermal neutron flux was about 7%. Errors in calculation of the effective multiplication factor for various core loading patterns were less than 0.5%. The code was examined also by the analysis of such research reactor accident as a damage of a large neutron beam tube located near the core with accompanying release of positive reactivity due to penetration of water. The penetration of water into the void tube was simulated by loading an ampoule filled up by water. The measured value of reactivity was 0.09 βeff while the calculated value was 0.12 βeff. A very effective way to make a research reactor more effective is to optimise its core layout, increasing neutron flux densities in the reactor channels and decreasing its fuel expenditures. Although various heuristic approaches exist and help to generate improved core loading patterns, they can't provide high quality of the solutions obtained. WWR-M reactor core loading patterns, fuel types used, number of fuel assemblies in the core and discharged fuel burnup are optimized using the code PORT. The optimization problem is to maximize Σipiφi - CF, where pt and φi are the 'neutron price' and neutron flux in the i-th irradiation channel, respectively, and CF is the feed fuel cost. The optimal core loading pattern, fuel types used, number of fuel assemblies of each type and their discharged burnups are determined under the constraints on the maximum allowed fuel burnup and temperature of fuel surface. Number and types of fuel assemblies as well as discharged fuel burnup are evaluated primarily for simplified models using an optimization algorithm based on backward diffusion calculation and linear programming. To determine the best placement of real set of fuel assemblies in the core for realistic model, an algorithm based on the successive mixed-integer linear programming and backward diffusion calculation has been developed. To specify and control input data and review the results of neutronics and thermal-hydraulics calculation in convenient form, a graphical interface for automation and visualization of core design has been developed. Information for all core reloads is recorded in database, thus whole history of each fuel assembly can be reviewed and analyzed. (author)