According to the current international practice of reactor physics and coupled reactor physics/thermal-hydraulics calculations at nodal and pin-wise levels, the group constants (homogenized cross-sections) are generated by deterministic transport codes and parameterized according to burnup, specific isotopes, and state parameters (such as fuel temperature, coolant density, and others).
Recently, however, the group constant generation using high precision Monte Carlo (MC) codes, such as Serpent, has become more common because of the growing demand for calculation accuracy and the increasing computational resources. Using MC methods, it is possible to model complex geometries precisely and solve the neutron transport equation accurately (e.g., continuous-energy nuclear data can be directly used, self-shielding is explicitly computed by the code). A further challenge is that many of the deterministic codes and models have been developed and validated for thermal reactors, and there is only limited experience with fourth-generation reactors. The disadvantage of MC methods is mainly the calculation time required for parameterization and the statistical fluctuations of the calculated low probability reaction rates.
• The work should start by assessing the current MC methods available in the literature. Subsequently, a method for providing the nodal KIKO3DMG code with burnup dependent parameterized group constants has to be developed.
• The elaborated method has to be tested on generation IV reactors (e.g., on the gas-cooled ALLEGRO fast spectrum reactor) and generation II and III reactors (e.g., VVER-440, VVER-1000, and VVER-1200 type reactors).
• The developed method has to be verified by comparing full-core MC and KIKO3DMG calculations covering the possible range of the parameters of group constants. The verification should include sensitivity analyses to quantify, for example, the effect of the number of energy groups on core safety parameters (e.g., reactivity-coefficients, reactivity worth of control rods).
• The burnup processes of the different reactors have to be investigated, including the burnup dependence of core safety parameters, taking into account the thermal-hydraulics feedback effects. The uncertainties of core safety parameters originating from the nuclear data have to be evaluated during the burnup using different nuclear data libraries (e.g., different versions of ENDF/B libraries).
• The KIKO3DMG code, applying the new parametrized group constants, has to be validated on experimental measurements of, for example, VVER type reactors.
• Comparative analyses should be performed between reactor generations regarding the core safety parameters and their uncertainties as a function of the burnup.
• Supporting the core design optimization of the ALLEGRO reactor, coupled ATHLET-KIKO3DMG analysis should be done using the burnup dependent group constants for a chosen reactivity initiated transient of ALLEGRO.