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Title: Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor

The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 900 degrees C or operational fuel temperatures above 1250 degrees C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR's higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Engineering and Environmental Laboratory (INEEL) has investigated this event for the past three yearsmore » for the HTGR. However, the computer codes used, and in fact none of the world's computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertaninty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.« less
Authors:
; ;
Publication Date:
OSTI Identifier:
911001
Report Number(s):
INEEL/EXT-04-02459
TRN: US0704335
DOE Contract Number:
DE-AC07-99ID-13727
Resource Type:
Technical Report
Research Org:
Idaho National Laboratory (INL)
Sponsoring Org:
DOE - NE
Country of Publication:
United States
Language:
English
Subject:
21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; COMPUTER CODES; ENERGY CONVERSION; FISSION PRODUCT RELEASE; GAS COOLED REACTORS; HYDROGEN PRODUCTION; LOSS OF COOLANT; NATURAL CONVECTION; OXIDATION; POWER GENERATION; PROCESS HEAT; REACTOR CORES; SAFETY; SAFETY ANALYSIS; VALIDATION; High Temperature Gas-Cooled Reactor