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Schultz, Richard

Publications and source records attributed to Schultz, Richard.

6 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Investigation of Abnormal Heat Transfer and Flow in a VHTR Reactor Core

The main objective of this project was to identify and characterize the conditions under which abnormal heat transfer phenomena would occur in a Very High Temperature Reactor (VHTR) with a prismatic core. High pressure/high temperature experiments have been conducted to obtain data that could be used for validation of VHTR design and safety analysis codes. The focus of these experiments was on the generation of benchmark data for design and off-design heat transfer for forced, mixed and natural circulation in a VHTR core. In particular, a flow laminarization phenomenon was intensely investigated since it could give rise to hot spots in the VHTR core.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Transient Mixed Convection Validation for NGNP

The results of this project are best described by the papers and dissertations that resulted from the work. They are included in their entirety in this document. They are: (1) Jeff Harris PhD dissertation (focused mainly on forced convection); (2) Blake Lance PhD dissertation (focused mainly on mixed and transient convection). This dissertation is in multi-paper format and includes the article currently submitted and one to be submitted shortly; and, (3) JFE paper on CFD Validation Benchmark for Forced Convection.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

WAVE PROPAGATION in the HOT DUCT of VHTR

In VHTR, helium from the reactor vessel is conveyed to a power conversion unit through a hot duct. In a hypothesized Depressurized Conduction Cooldown event where a rupture of the hot duct occurs, pressure waves will be initiated and reverberate in the hot duct. A numerical model is developed to quantify the transients and the helium mass flux through the rupture for such events. The flow path of the helium forms a closed loop but only the hot duct is modeled in this study. The lower plum of the reactor vessel and the steam generator are treated as specified pressure and/or temperature boundary to the hot duct. The model is based on the conservation principles of mass, momentum and energy, and on the equations of state for helium. The numerical solution is based on the method of characteristics with specified time intervals with a predictor and corrector algorithm. The rupture sub-model gives reasonable results. Transients induced by ruptures with break area equaling 20%, 10%, and 5% of the duct cross-sectional area are described.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

USING CFD TO ANALYZE NUCLEAR SYSTEMS BEHAVIOR: DEFINING THE VALIDATION REQUIREMENTS

A recommended protocol to formulate numeric tool specifications and validation needs in concert with practices accepted by regulatory agencies for advanced reactors is described. The protocol is based on the plant type and perceived transient and accident envelopes that translates to boundary conditions for a process that gives the: (a) key phenomena and figures-of-merit which must be analyzed to ensure that the advanced plant can be licensed, (b) specification of the numeric tool capabilities necessary to perform the required analyses—including bounding calculational uncertainties, and (c) specification of the validation matrices and experiments--including the desired validation data. The result of applying the process enables a complete program to be defined, including costs, for creating and benchmarking transient and accident analysis methods for advanced reactors. By following a process that is in concert with regulatory agency licensing requirements from the start to finish, based on historical acceptance of past licensing submittals, the methods derived and validated have a high probability of regulatory agency acceptance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

HTTF Core Stress Analysis

In accordance with the need to determine whether cracking of the ceramic core disks which will be constructed and used in the High Temperature Test Facility (HTTF) for heatup and cooldown experiments, a set of calculation were performed using Abaqus to investigate the thermal stresses levels and likelihood for cracking. The calculations showed that using the material properties provided for the Greencast 94F ceramic, cracking is predicted to occur. However, this modeling does not predict the size or length of the actual cracks. It is quite likely that cracks will be narrow with rough walls which would impede the flow of coolant gases entering the cracks. Based on data recorded at Oregon State University using Greencast 94F samples that were heated and cooled at prescribed rates, it was concluded that the likelihood that the cracks would be detrimental to the experimental objectives is small.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗