Numerical Simulations of Single Flow Element in a Nuclear Thermal Thrust Chamber (original) (raw)

Numerical Study of Single Flow Element in a Nuclear Thermal Thrust Chamber

Journal of Aeronautics & Aerospace Engineering, 2015

The objective of this study was to develop an efficient and accurate computational methodology to predict detailed thermo-fluid environments of a single flow element in a hypothetical solid-core nuclear thermal thrust chamber assembly. Several numerical and multi-physics thermo-fluid models, such as chemical reactions, turbulence, conjugate heat transfer, porosity, and power generation, were incorporated into an unstructured-grid, pressure-based computational fluid dynamics solver used in this investigation. A secondary objective was to develop a porosity model for simulation of the whole solid-core nuclear thermal engine without resolving thousands of flow channels inside the solid core. Detailed numerical simulations of a single flow element with different power generation profiles were conducted to investigate the root cause of a phenomenon called mid-section corrosion that severely damaged the flow element assembly of early solid-core reactors. Under the assumptions employed in this effort and for the first time, the result demonstrated flow choking in the flow element. The possibility of flow choking in part of the flow element indicated a potential coolant mass flow imbalance, which could lead to a high local thermal gradient in coolant-starved flow elements and possibly the eventual mid-section corrosion.

Development of an Efficient CFD Model for Nuclear Thermal Thrust Chamber Assembly Design

The objective of this effort is to develop an efficient and accurate computational methodology to predict both detailed thermo-fluid environments and global characteristics of the internal ballistics for a hypothetical solid-core nuclear thermal thrust chamber assembly (NTTCA). Several numerical and multi-physics thermofluid models, such as real fluid, chemically reacting, turbulence, conjugate heat transfer, porosity, and power generation, were incorporated into an unstructuredgrid, pressure-based computational fluid dynamics solver as the underlying computational methodology. The numerical simulations of detailed thermo-fluid environment of a single flow element provide a mechanism to estimate the thermal stress and possible occurrence of the mid-section corrosion of the solid core. In addition, the numerical results of the detailed simulation were employed to fine tune the porosity model mimic the pressure drop and thermal load of the coolant flow through a single flow element. The use of the tuned porosity model enables an efficient simulation of the entire NTTCA system, and evaluating its performance during the design cycle.

Multiphysics Thrust Chamber Modeling for Nuclear Thermal Propulsion

2006

The objective of this effort is to develop an efficient and accurate thermo-fluid computational methodology to predict environments for a solid-core, nuclear thermal engine thrust chamber. The computational methodology is based on an unstructured-grid, pressure-based computational fluid dynamics formulation. A two-pronged approach is employed in this effort: A detailed thermo-fluid analysis on a multi-channel flow element for mid-section corrosion investigation; and a global modeling of the thrust chamber to understand the effect of heat transfer on thrust performance. Preliminary results on both aspects are presented.

Multiphysics Computational Analysis of a Solid-Core Nuclear Thermal Engine Thrust Chamber

Journal of Propulsion and Power, 2010

The objective of this effort is to develop an efficient and accurate computational heat transfer methodology to predict thermal, fluid, and hydrogen environments for a hypothetical solid-core, nuclear thermal engine -the Small Engine, In addition, the effects of power profile and hydrogen conversion on heat transfer efficiency and thrust performance were also investigated. The computational methodology is based on an unstructured-grid, pressure-based, all speeds, chemically reacting, computational fluid dynamics platform, while formulations of conjugate heat transfer were implemented to describe the heat transfer from solid to hydrogen inside the solid-core reactor. The computational domain covers the entire thrust chamber so that the afore-mentioned heat transfer effects impact the thrust performance directly. The result shows that the computed core-exit gas temperature, specific impulse, and core pressure drop agree well with those of design data for the Small Engine. Finite-rate chemistry is very importaot io predictiog the proper energy balance as naturally occurring hydrogen decomposition is endothermic. Locally strong hydrogen conversion associated with centralized power profile gives poor heat transfer efficiency and lower thrust performance. On the other hand, uoiform hydrogen conversion associated with a more uoiform radial power profile achieves higher heat transfer efficiency, and higher thrust performance.

ICMIEE 18-218 Thermal Hydraulics Simulation of Fuel SubAssembly for 1200 MWe Nuclear Power Reactor

2018

This study illustrates the turbulent flow simulation of coolant water through the three sub-channels of a fuel sub-assembly at a pressure around 16 MPa. The geometry details of the fuel rods, coolant sub-channels and operating parameters are similar to those of Rooppur Nuclear Power Reactor under construction in Bangladesh. The fuel sub-assembly is modeled using seven fuel rods where k-ε turbulence model is used for turbulent flow simulation. The effect of turbulent flow on temperature, velocity, pressure drop, friction factor and Nusselt number in interior, edge and corner sub-channels have been discussed for various axial locations (z =0–45Dh). Thermal hydraulic properties of the coolant water are studied for safety analyses such as: i) Hot spot in coolant channel and ii) Departure from Nucleate Boiling (DNB

A Multiphysics Simulation Suite for Sodium Cooled Fast Reactors

EPJ Web of Conferences, 2021

A simulation suite has been developed to model reactor power distribution and multiphysics feedback effects in Sodium-cooled Fast Reactors (SFRs). This suite is based on the Finite Element Method (FEM) and employs a general, unstructured mesh to solve the Simplified P3 (SP3) neutron transport equations. In the FEM implementation, two-dimensional triangular elements and three-dimensional wedge elements are selected. Wedge elements are selected for their natural description of hexagonal geometry common to fast reactors. Thermal feedback effects within fast reactors are modeled within the simulation suite. A thermal hydraulic model is developed, modeling both axial heat convection and radial heat conduction within fuel assemblies. A thermal expansion model is included and is demonstrated to significantly affect reactivity. This simulation suite has been employed to model the Advanced Burner Reactor (ABR) benchmark, specifically the MET-1000. It has been demonstrated that these models s...

Thermal Hydraulic Analysis in Reactor Vessel Internals Considering Irradiation Heat

2015

The present study is to evaluate flow and temperature distributions in the APR1400 reactor vessel internals (RVIs) considering irradiation heat using the computational fluid dynamics code. The analysis is performed using a simplified reactor core model and the one-quarter geometry model that is postulated to be the flow symmetry condition, to calculate more effectively. The axial power density at the beginning of cycle (BOC), middle of cycle (MOC) and end of cycle (EOC) conditions is used as inputs to the CFD analysis, and velocity and temperature distributions are calculated on the four cross-sections of the reactor vessel (RV): lower support structure (LSS), core shroud (CS), upper part of the core and hot/cold leg center line. This method established the effect of a variation of the flow according to BOC, MOC and EOC conditions. The results of each model show the similar patterns in the flow distributions, however, the little difference appears in the temperature distributions un...

Preliminary numerical analysis of the flow distribution in the core of a research reactor

2020

The thermal-hydraulic safety analysis of research reactors establishes the safety criteria to ensure the integrity of the fuel elements in the reactor core. It assures that all core components are being adequately cooled during operation. It is necessary to know if the average mass flow rate (and their standard deviation) among the fuel assemblies are enough to cool the power generated during operation. Once satisfied such condition, it allows the calculation of the maximum heat flux transferred from fuel assemblies to the coolant, and if the maximum cladding temperatures are below the limits set by the safety criteria. Among the objectives, this study presents a methodology for a preliminary threedimensional numerical analysis of the flow distribution in the core of the IEA-R1 research reactor, under steady state condition. For this, the ANSYS-CFX commercial code was used to analyze the flow dynamics in the core, and to visualize the velocity field. It was possible to conclude that...