Johnson, Richard W | Engineering Papers

DOE OSTI2012

MODELING STRATEGIES TO COMPUTE NATURAL CIRCULATION USING CFD IN A VHTR AFTER A LOFA

A prismatic gas-cooled very high temperature reactor (VHTR) is being developed under the next generation nuclear plant program (NGNP) of the U.S. Department of Energy, Office of Nuclear Energy. In the design of the prismatic VHTR, hexagonal shaped graphite blocks are drilled to allow insertion of fuel pins, made of compacted TRISO fuel particles, and coolant channels for the helium coolant. One of the concerns for the reactor design is the effects of a loss of flow accident (LOFA) where the coolant circulators are lost for some reason, causing a loss of forced coolant flow through the core. In such an event, it is desired to know what happens to the (reduced) heat still being generated in the core and if it represents a problem for the fuel compacts, the graphite core or the reactor vessel (RV) walls. One of the mechanisms for the transport of heat out of the core is by the natural circulation of the coolant, which is still present. That is, how much heat may be transported by natural circulation through the core and upwards to the top of the upper plenum? It is beyond current capability for a computational fluid dynamic (CFD) analysis to perform a calculation on the whole RV with a sufficiently refined mesh to examine the full potential of natural circulation in the vessel. The present paper reports the investigation of several strategies to model the flow and heat transfer in the RV. It is found that it is necessary to employ representative geometries of the core to estimate the heat transfer. However, by taking advantage of global and local symmetries, a detailed estimate of the strength of the resulting natural circulation and the level of heat transfer to the top of the upper plenum is obtained.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Bypass Flow Computations using a One-Twelfth Symmetric Sector For Normal Operation in a 350 MWth VHTR

Significant uncertainty exists about the effects of bypass flow in a prismatic gas-cooled very high temperature reactor (VHTR). Bypass flow is the flow in the gaps between prismatic graphite blocks in the core. The gaps are present because of variations in their construction, imperfect installation and expansion and shrinkage from thermal heating and neutron fluence. Calculations are performed using computational fluid dynamics (CFD) for flow of the helium coolant in the gap and coolant channels along with conjugate heat generation and heat transfer in the fuel compacts and graphite. A commercial CFD code is used for all of the computations. A one-twelfth sector of a standard hexagonal block column is used for the CFD model because of its symmetry. Various scenarios are computed by varying the gap width from zero to 5 mm, varying the total heat generation rate to examine average and peak radial generation rates and variation of the graphite block geometry to account for the effects of shrinkage caused by irradiation. The calculations are for a 350 MWth prismatic reactor. It is shown that the effect of increasing gap width, while maintaining the same total mass flow rate, causes increased maximum fuel temperature while providing significant cooling to the near-gap region. The maximum outlet coolant temperature variation is increased by the presence of gap flow and also by an increase in total heat generation with a gap present. The effect of block shrinkage is actually to decrease maximum fuel temperature compared to a similar reference case.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2011

Undesirable Flow Behavior in a Proposed Validation Data Set

The next generation nuclear plant (NGNP), whose development is supported by the U. S. Department of Energy, will be a very high temperature reactor (VHTR). The VHTR is a single-phase helium-cooled reactor that will provide helium at up to 1000 C. The prospect of a coolant at these temperatures circulating in the reactor vessel demands that careful analysis be performed to ensure that excessively hot spots are not created and that sufficient mixing of the coolant is obtained. Computational fluid dynamics (CFD) coupled with heat transfer will be used to perform the desired analyses. However, primarily because of the imperfect nature of modeling turbulent flow, any CFD calculations used to perform nuclear reactor safety analysis must be validated against experimental data. Experimental data have been taken in a scaled section of the lower plenum of a prismatic VHTR at the matched index of refraction (MIR) facility at the Idaho National Laboratory. These data were taken with the intent that they be examined for use as validation data. A series of investigations have been conducted to assess the MIR data. Issues that have already been examined include the extent of the required computational domain, the outlet boundary condition, the inlet data and the effect of the turbulence model. One of the jets that flow into the model impacts on a wedge, which represents a portion of a hexagonal graphite block that lines the inner wall of the lower plenum. The nature of the flow below this particular jet is such that a randomly varying recirculation zone is created. This recirculation zone is seen to change in size, causing a relatively long-time scale of motion or disturbance of the flow in the model. It is concluded that such a feature is undesirable in a validation data set, firstly because of its apparent random nature and, secondly, because to obtain an appropriate long-time average would be impractical because of the compute time required. It is predicted computationally that by eliminating the first of the four inlet jets into the scaled model, the resulting recirculation zone is rendered stable.

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DOE OSTI2011

The Role of Data Transfer on the Selection of a Single vs Multiple Mesh Architecture for Tightly Coupled Multiphysics Applications

Data transfer from one distinct mesh to another may be necessary in any number of applications, including prolongation operations supporting multigrid solution methods, spatial adaptation, remeshing, and arbitrary Lagrangian-Eulerian (ALE) and multiphysics simulation. This data transfer process is also referred to as remapping, rezoning and interpolation. Intermesh data transfer has the potential to introduce error into a simulation; the magnitude and importance of which depends on the transfer scenario and the algorithm used to perform the transfer. For a transient analysis, data transfer may occur many times during a simulation, with possible error accumulation at each transfer. The present study develops selected scenarios that illustrate data transfer error and how it might impact an analysis. This study examines remapping error by using static analytical functions to compare various remapping schemes. It also investigates the significance and nature of data transfer error for a simple multiphysics system involving a transient coupled system of partial differential equations. It concludes that remapping error can be significant both for static functions and for coupled multiphysics systems. Aggregate error is shown to be a function of remapping scheme, mesh coarseness, nature of the remapped function and mesh disparity. In cases of extreme mesh disparity, this study shows that remapping can lead to excessive error and even to solution instability. Further, this work motivates that remapping error should be included in the estimation of numerical error, if data transfer is employed in a numerical simulation.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2010

Next Generation Nuclear Plant Methods Technical Program Plan

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope of the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2010

Next Generation Nuclear Plant Methods Technical Program Plan

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope of the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.

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