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Title: THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.
Authors:
Publication Date:
OSTI Identifier:
6626209
Report Number(s):
ORNL-5951
ON: DE84016680
DOE Contract Number:
AC05-84OR21400
Resource Type:
Technical Report
Resource Relation:
Other Information: Portions are illegible in microfiche products. Original copy available until stock is exhausted
Research Org:
Oak Ridge National Lab., TN (USA)
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; COMPUTER CODES; T CODES; HTGR TYPE REACTORS; HEAT TRANSFER; HYDRAULICS; ENERGY TRANSFER; FLUID MECHANICS; GAS COOLED REACTORS; GRAPHITE MODERATED REACTORS; MECHANICS; REACTORS 210300* -- Power Reactors, Nonbreeding, Graphite Moderated