Andrew Grief

Development of Local Heat Transfer Models for Safety Assessment of Pebble Bed High Temperature Gas-Cooled Reactor Cores

This paper presents a model developed for determining fuel particle and fuel pebble temperatures in normal operation and transient conditions based on multi-scale modelling techniques. This model is qualified by comparison with an analytical solution in a one-dimensional linear steady state test problem. Comparison is made with finite element simulations of an idealised “two-dimensional” pebble in transient conditions and with a steady state analytical solution in a spherical pebble geometry. A method is presented for determining the fuel temperatures in the individual batches of a multi-batch recycle refuelling regime. Implementation of the multi-scale and multibatch fuel models in a whole-core CFD model is discussed together with the future intentions of the research programme.© 2008 ASME

Development of Local Heat Transfer and Pressure Drop Models for Pebble Bed High Temperature Gas-Cooled Reactor Cores

This paper describes pressure drop and heat transfer coefficient predictions for a typical coolant flow within the core of a pebble bed reactor (PBR) by examining a representative group of pebbles remote from the reflector region. The three-dimensional steady state flow and heat transfer predictions utilized in this work are obtained from a computational fluid dynamics (CFD) model created in the commercial software ANSYS FLUENT™. This work utilizes three RANS turbulence models and the Chilton-Colburn analogy for heat transfer. A methodology is included in this paper for creating a quality unstructured mesh with prismatic surface layers on a random arrangement of touching pebbles. The results of the model are validated by comparing them with the correlations of the German KTA rules for a PBR.Copyright

A Methodology for Accident Analysis of Fusion Breeder Blankets and Its Application to Helium-Cooled Lead–Lithium Blanket

“Fusion for Energy” (F4E) is designing, developing, and implementing the European helium-cooled lead-lithium (HCLL) and helium-cooled pebble-bed test blanket systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and Idaho National Laboratory efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to the ITER accident analyses. The methodology phases are illustrated in this paper by its application to the EU HCLL TBS using both MELCOR and RELAP5 codes.

Development of Local Heat Transfer Models for the Safety Assessment of Prismatic Modular High Temperature Gas-Cooled Reactor Cores

This paper presents a model developed for determining fuel particle and fuel block temperatures of a prismatic core modular reactor during both normal operation and under fault conditions. The model is based on multi-scale modeling techniques and has been qualified by comparison with finite element simulations for both steady state and transient conditions. Further, a model for determining the effective conductivity of the block fuel elements — important for heat removal in loss of flow conditions — is presented and, again, qualified by comparison with finite element simulations. A numerical model for predicting conduction heat transfer both within and between block fuel elements has been developed which, when coupled with the above multi-scale model, allows simulations of whole cores to be carried out whilst retaining the ability to predict the temperatures of individual coolant channels and individual coated particles in the fuel if required. This ability to resolve heat transfer on length scales ranging from a few meters down to a few microns within the same model is very powerful and allows a complete assessment of the fuel and structural temperatures within a core to be made. More significantly, this level of resolution facilitates interactive coupling with neutronics models to enable the strong temperature/reactivity feedbacks, inherent in such cores, to be resolved correctly.Copyright

Development of Local Heat Transfer Models for the Safety Assessment of High Temperature Gas-Cooled Reactor Cores—Part II: Prismatic Modular Reactors

This paper extends the work of Part I to be applicable to prismatic block fuel elements and presents a model developed for determining fuel compact and fuel block temperatures of a prismatic core modular reactor. The model is applicable both in normal operation and under fault conditions and is an extension of the multiscale modeling techniques presented in Part I. The new model has been qualified by comparison with finite element simulations for both steady-state and transient conditions. Furthermore, a model for determining the effective conductivity of the block fuel elements-important for heat removal in loss of flow conditions—is presented and, again, qualified by comparison with finite element simulations. A numerical model for predicting conduction heat transfer both within and between block fuel elements has been developed, which, when coupled with the above multiscale model, allows simulations of whole cores to be carried out, while retaining the ability to predict the temperatures of individual coolant channels and individual coated particles in the fuel if required.