NESC9597 FLODIS.
FLODIS, Thermal Response of FSV HTGR Core
NAME OR DESIGNATION OF PROGRAM, COMPUTER, DESCRIPTION OF PROGRAM OR FUNCTION, METHOD OF SOLUTION, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, TYPICAL RUNNING TIME, UNUSUAL FEATURES OF THE PROGRAM, RELATED AND AUXILIARY PROGRAMS, STATUS, REFERENCES, MACHINE REQUIREMENTS, LANGUAGE, OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED, OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHORS, MATERIAL, CATEGORIES
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| Program name | Package id | Status | Status date |
|---|---|---|---|
| FLODIS | NESC9597/01 | Arrived | 27-APR-2001 |
Machines used:
| Package ID | Orig. computer | Test computer |
|---|---|---|
| NESC9597/01 | IBM 360 series |
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3. DESCRIPTION OF PROGRAM OR FUNCTION
FLODIS was developed to analyse shutdown transients for the Fort St. Vrain high-temperature gas- cooled reactor (HTGR) core. The program is a lumped node representation for the 37 refueling regions in the active core, the side reflector blocks, the gas annulus between the core barrel and the prestressed concrete reactor vessel (PCRV) liner, and the PCRV cooling system. Heat conduction in all three coordinate directions and to the coolant is modeled. The calculation uses the specified operating conditions for the reactor at power to determine appropriate loss coefficients for the variable orifices in each refueling region. Iterative techniques are used to determine the distribution of coolant flow as a function of time during the transient. Both forced and natural convection flows can be calculated. FLODIS can be adapted to other HTGR systems with minor modifications.
FLODIS was developed to analyse shutdown transients for the Fort St. Vrain high-temperature gas- cooled reactor (HTGR) core. The program is a lumped node representation for the 37 refueling regions in the active core, the side reflector blocks, the gas annulus between the core barrel and the prestressed concrete reactor vessel (PCRV) liner, and the PCRV cooling system. Heat conduction in all three coordinate directions and to the coolant is modeled. The calculation uses the specified operating conditions for the reactor at power to determine appropriate loss coefficients for the variable orifices in each refueling region. Iterative techniques are used to determine the distribution of coolant flow as a function of time during the transient. Both forced and natural convection flows can be calculated. FLODIS can be adapted to other HTGR systems with minor modifications.
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4. METHOD OF SOLUTION
The mathematical model consists of two parts -
the flow distribution and the temperature distribution. Navier- Stokes equations are applied to the flow problem. The temperature distribution differential equations are integrated using a Crank- Nickolson technique. The resulting algebraic equations are solved with a Gauss-Seidel iterative procedure.
A rectangular grid is superimposed on the reactor to define the mesh spaces. There are 142 mesh spaces in the active core, 146 in the side reflector, and 52 in the gas annulus between the core barrel and PCRV liner. Axially, there are 20 mesh spaces for a total of 6800 representing the entire system.
The mathematical model consists of two parts -
the flow distribution and the temperature distribution. Navier- Stokes equations are applied to the flow problem. The temperature distribution differential equations are integrated using a Crank- Nickolson technique. The resulting algebraic equations are solved with a Gauss-Seidel iterative procedure.
A rectangular grid is superimposed on the reactor to define the mesh spaces. There are 142 mesh spaces in the active core, 146 in the side reflector, and 52 in the gas annulus between the core barrel and PCRV liner. Axially, there are 20 mesh spaces for a total of 6800 representing the entire system.
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10. REFERENCES
- D.D. Paul
Depressurization Accident Analysis for the Fort St. Vrain Reactor, ORNL/NUREG/TM-58, December 1976.
- D.D. Paul
Depressurization Accident Analysis for the Fort St. Vrain Reactor, ORNL/NUREG/TM-58, December 1976.
NESC9597/01, included references:
- D.D. Paul:FLODIS: A Computer Model to Determine the Flow Distribution and
Thermal Response of the Fort St. Vrain Reactor
ORNL/TM-5365 (June 1976).
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NESC9597/01
source program mag tapeFLODIS PSource program SRCTPtest-case data mag tapeSample problem DATTP
test-case data mag tapeControl information DATTP
report ORNL/TM-5365 (June 1976) REPPT
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- G. Radiological Safety, Hazard and Accident Analysis
- H. Heat Transfer and Fluid Flow
Keywords: convection, reactor cores, thermal conduction.