CORTAP: a coupled neutron kinetics-heat transfer digital computer program for the dynamic simulation of the high temperature gas cooled reactor core

Technical Report · Fri Jan 07 04:00:00 EST 1977

DOI:https://doi.org/10.2172/7315582· OSTI ID:7315582

Cleveland, J C

CORTAP (Core Transient Analysis Program) was developed to predict the dynamic behavior of the High Temperature Gas Cooled Reactor (HTGR) core under normal operational transients and postulated accident conditions. CORTAP is used both as a stand-alone component simulation and as part of the HTGR nuclear steam supply (NSS) system simulation code ORTAP. The core thermal neutronic response is determined by solving the heat transfer equations for the fuel, moderator and coolant in an average powered region of the reactor core. The space independent neutron kinetics equations are coupled to the heat transfer equations through a rapidly converging iterative technique. The code has the capability to determine conservative fuel, moderator, and coolant temperatures in the ''hot'' fuel region. For transients involving a reactor trip, the core heat generation rate is determined from an expression for decay heat following a scram. Nonlinear effects introduced by temperature dependent fuel, moderator, and coolant properties are included in the model. CORTAP predictions will be compared with dynamic test results obtained from the Fort St. Vrain reactor owned by Public Service of Colorado, and, based on these comparisons, appropriate improvements will be made in CORTAP.

• Research Organization: Oak Ridge National Lab., TN (USA)
• DOE Contract Number: W-7405-ENG-26
• OSTI ID: 7315582
• Report Number(s): ORNL/NUREG/TM-39
• Country of Publication: United States
• Language: English

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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
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Graphite Moderated
AFTER-HEAT
C CODES
COMPUTER CODES
ENERGY TRANSFER
GAS COOLED REACTORS
GRAPHITE MODERATED REACTORS
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KINETICS
REACTOR COMPONENTS
REACTOR CORES
REACTOR KINETICS
REACTOR SHUTDOWN
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SCRAM
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