Showing 85 results for Author: Rike Bostelmann
Sep, 2026
ORNL Report
This technical memorandum compiles the contributions presented at the 2026 R-Matrix Workshop on Methods and Applications, held July 13–17, 2026, at Oak Ridge National Laboratory. The workshop brought together researchers working in phenomenological R-matrix theory, experimental nuclear physics, nuclear-data evaluation, uncertainty quantification, and computational methods,…
Aug, 2026
ORNL Report
Aug, 2026
Journal
As both industry and government move toward meeting their decarbonization goals, nuclear power is increasingly being considered due to its legacy of providing clean and reliable power. Pebble bed (PB) high-temperature gas-cooled reactors (HTGRs) are a generation-IV reactor design being considered because of their past operational experience, intrinsic safety features, and…
Jul, 2026
Conference Paper
Nuclear data are an essential component of the foundation on which all modeling and simulation methods and tools are relying upon, from the front end to the back end of the nuclear fuel cycle. In this study, the impact of uncertainties in nuclear data is investigated for a representative molten chloride fast reactor, for several important metrics, including eigenvalue, rea…
Jul, 2026
Conference Paper
Jul, 2026
Conference Paper
Sensitivity coefficients calculated with Monte Carlo codes are widely used for nuclear data uncertainty quantification in the modeling and simulation of complex 3D reactor systems. This study systematically compares sensitivity coefficients and associated statistical uncertainties for the multiplication factor and fuel temperature reactivity across multiple Monte Carlo cod…
Jul, 2026
Conference Paper
This work introduces recently implemented capabilities in the SCALE code system’s TRITON reactor physics sequence that improve molten salt reactor (MSR) modeling: (1) a volume balancing option, which automatically balances the volumes of the fed material with a corresponding material removal to maintain fixed mixture volumes in the neutron transport model and accurate dens…
Jul, 2026
Journal
This paper demonstrates the applicability of the SCALE code system to TRistructural-ISOtropic (TRISO)-fueled heat-pipe microreactors through depletion, transportation criticality, and shielding analyses of a generic reference design. The study supported U.S. Nuclear Regulatory Commission code-readiness efforts for advanced non-light-water reactor technologies and is inte…
Jul, 2026
Conference Paper
Fluid fuel molten salt reactor (MSR) concepts may achieve continuous-at-power operation with continuous or batch fresh fuel feed to counter reactivity loss from fuel burnup and fission product buildup. Accurate modeling and simulation of MSRs requires consideration of online fuel feed during depletion as well as a capability to determine the required feed amount to maintai…