Presented in this dissertation is the investigation and development of an adapted lattice physics-to-core simulator two-step procedure based on the SERPENT 2 and NESTLE neutronics codes for the rapid analysis of the Advanced High Temperature Reactor (AHTR). AHTR specific characteristics, such as its longer neutron diffusion length and double heterogeneity of TRISO fuel particles, were taken into consideration when adapting the traditional Light Water Reactor (LWR) lattice to nodal diffusion procedure to AHTR applications. The coarse energy group structure was re-optimized from the traditional LWR 2-group structure to an alternative 4-group structures to address the AHTR specific flux spectrum and neutronics characteristics. A more accurate treatment of the interface between fuel and reflector was implemented using simplified 1-D models along with the application of an Equivalence Theory based Assembly Discontinuity Factor (ADF) adjustment of the resultant few group constants. A similar ADF adjustment was also applied to treat the insertion of control blades to properly account for inter-assembly leakage. The developed two-step procedure was tested against multiple transport based high fidelity reference benchmark models and was deemed to provide reasonably accurate results, with the exception of some peripheral radial power discrepancies which have been attributed to the inadequacy of the 1-D radial reflector model to capture a 1/3 symmetric and cyclic power tilt unique to the AHTR fuel assembly design and core layout. For 2-D and 3-D full core models, eigenvalue agreement was within 130 pcm and power distribution errors within 3.5% Root Mean Squared (RMS) error. The final implementation of this two-step procedure was used to perform a representative neutronic and thermal hydraulic coupled simulation which demonstrated the ability of the developed procedure to perform 3-D full core neutronics calculations with coupling to thermal hydraulic feedback in an extremely expedient manner. This work paves the way to ultimately performing fuel cycle, core / assembly design, and safety margin assessments for the AHTR. Additionally, this procedure greatly reduces the computational expense of performing such simulations and opens the door toward AHTR design optimization.

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

Multi-Physics Modelling (MPM) is an innovative simulation technique that looks very promising for the employment in the field of nuclear engineering as an integrative analysis support in the design development of current and innovative nuclear reactors. This book presents a Multi-Physics Modelling (MPM) approach to the analysis of nuclear reactor core behaviour, developed to study the coupling between neutronics and thermo-hydrodynamics. Reference is made to the Molten Salt Reactor, one of the innovative nuclear systems under development in the framework of the Generation IV International Forum, but the same methodology can be applied to other reactor systems.

This workshop was designed to meet the needs of those currently involved in or are planning a nuclear programme involving research and/or power fission reactors. The workshop had a broad scope including not only fission reactor core calculations, but also safety, fuel management, waste disposal reactor licensing. The lectures and computer exercises covered almost all aspects of the operation of fission reactors.This workshop introduced participants to the methods currently used in fission reactor calculations and to some computer codes in which these methods are used.