PSR-0411 MORECA.
MORECA, Simulating Modular High-Temperature Gas Cooled Reactor Core Heatup
NAME OR DESIGNATION OF PROGRAM, COMPUTER, DESCRIPTION OF PROGRAM OR FUNCTION, METHODS, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, TYPICAL RUNNING TIME, STATUS, REFERENCES, HARDWARE REQUIREMENTS, LANGUAGE, SOFTWARE REQUIREMENTS, NAME AND ESTABLISHMENT OF AUTHOR, MATERIAL, CATEGORIES
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| Program name | Package id | Status | Status date |
|---|---|---|---|
| MORECA | PSR-0411/01 | Arrived | 03-MAY-2002 |
Machines used:
| Package ID | Orig. computer | Test computer |
|---|---|---|
| PSR-0411/01 | IBM PC |
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3. DESCRIPTION OF PROGRAM OR FUNCTION
MORECA was developed for analyzing postulated long-term core heatup scenarios for which active cooling systems used to remove afterheat following the accidents can be assumed to be unavailable. Simulations of long-term loss-of-convection accidents, both with and without depressurization of the primary coolant, have shown that maximum core temperatures stay below the point at which any significant fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. MORECA models the U.S. Department of Energy reference design of a standard modular high-temperature gas-cooled reactor (MHTGR). It is based on the ORNL ORECA code system.
MORECA was developed for analyzing postulated long-term core heatup scenarios for which active cooling systems used to remove afterheat following the accidents can be assumed to be unavailable. Simulations of long-term loss-of-convection accidents, both with and without depressurization of the primary coolant, have shown that maximum core temperatures stay below the point at which any significant fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. MORECA models the U.S. Department of Energy reference design of a standard modular high-temperature gas-cooled reactor (MHTGR). It is based on the ORNL ORECA code system.
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PSR-0411/01, included references:
- S.J BallMORECA: A Computer Code for Simulating Modular High-Temperature Gas-Cooled
Reactor Core Heatup Accidents
NUREG/CR-5712, ORNL/TM-11823 (October 1991)
- RSICC
README.TXT
(May 14,1999)
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15. NAME AND ESTABLISHMENT OF AUTHOR
Contributed by: Radiation Safety Information Computational Center
Oak Ridge National Laboratory
Oak Ridge, Tennessee, U. S. A.
Developed by: Oak Ridge National Laboratory, Oak Ridge, Tennessee
through the Energy Science and Technology Software Center,
Oak Ridge, Tennessee, USA
Contributed by: Radiation Safety Information Computational Center
Oak Ridge National Laboratory
Oak Ridge, Tennessee, U. S. A.
Developed by: Oak Ridge National Laboratory, Oak Ridge, Tennessee
through the Energy Science and Technology Software Center,
Oak Ridge, Tennessee, USA
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PSR-0411/01
PROCAL.DBF DBASE databaseCAHE2.F FTN Subroutine
DIDMOD.PRG DBASE program
FLOTI3.F FTN Subroutine
GOODOM.F FTN Subroutine
MHTS2.DAT Sample Problem Input
MOREC.OUT Sample Problem Output
MORECX.F Main FTN Program
MORECX.MAK Old Make File
MORECX.EXE Executable
ORECOM.DBF DBASE database
OREP.PRG DBASE program
OREVAR.DBF DBASE database
OVAR.NDX DBASE database index
PPCAL.PRG DBASE program
PRESS3.F FTN Subroutine
PROC.NDX DBASE database index
BOTTE4.F FTN Subroutine
PROCAL.DBT DBASE database memo
RADFA4.F FTN Subroutine
RCCSP.F FTN Subroutine
SUB350.F FTN Subroutine
Keywords: reactor safety.