NEA-0447 EUREKA.
EUREKA, Reactivity Transients in LWR from Control Rod, Coolant Flow, Temperature
NAME OR DESIGNATION OF PROGRAM, COMPUTER, NATURE OF PHYSICAL PROBLEM SOLVED, METHOD OF SOLUTION, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, TYPICAL RUNNING TIME, FEATURES, AUXILIARIES, STATUS, REFERENCES, MACHINE REQUIREMENTS, LANGUAGE, OPERATING SYSTEM OR MONITOR UNDER WHICH PROGRAM IS EXECUTED, OTHER RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHOR, MATERIAL, CATEGORIES
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| Program name | Package id | Status | Status date |
|---|---|---|---|
| EUREKA | NEA-0447/01 | Tested | 01-JUL-1976 |
Machines used:
| Package ID | Orig. computer | Test computer |
|---|---|---|
| NEA-0447/01 | IBM 370 series | IBM 370 series |
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4. METHOD OF SOLUTION
Coupled neutronic and thermal-hydrodynamic response is calculated for multi-regions in the core through one-point neutron kinetics. The core considered in the model can be divided into up to five regions (channels) radially and 20 regions (21 nodes) axially. Each subdivision gives its own power generation rate and reactivity feedback weighting factor for an application to point kinetics calculation. Each channel is represented by a fuel rod which contains three materials (fuel meat, gas gap and cladding) and associated coolant and is subdivided into up to 25 nodes in radial direction. There are provisions to calculate thermal hydrodynamics of channels and temperature profiles in fuels. The transient power is determined by a numerical solution of the one point kinetics equation. The temperature distributions in each fuel are computed on the basis of a one-dimensional heat conduction equation. Local heat generation is determined from reactor power and specified weighting factors. Heat transfer correlations at the clad-coolant interface are employed in each of the various heat transfer regimes, such as convection, nucleate, transition and film boiling. The hydrodynamics in each coolant channel are solved through a one-dimensional conservation equation of mass, moment and energy.
Coupled neutronic and thermal-hydrodynamic response is calculated for multi-regions in the core through one-point neutron kinetics. The core considered in the model can be divided into up to five regions (channels) radially and 20 regions (21 nodes) axially. Each subdivision gives its own power generation rate and reactivity feedback weighting factor for an application to point kinetics calculation. Each channel is represented by a fuel rod which contains three materials (fuel meat, gas gap and cladding) and associated coolant and is subdivided into up to 25 nodes in radial direction. There are provisions to calculate thermal hydrodynamics of channels and temperature profiles in fuels. The transient power is determined by a numerical solution of the one point kinetics equation. The temperature distributions in each fuel are computed on the basis of a one-dimensional heat conduction equation. Local heat generation is determined from reactor power and specified weighting factors. Heat transfer correlations at the clad-coolant interface are employed in each of the various heat transfer regimes, such as convection, nucleate, transition and film boiling. The hydrodynamics in each coolant channel are solved through a one-dimensional conservation equation of mass, moment and energy.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM
The thermal hydrodynamics model used has caused numerical instability in certain situations, i.e. transient boiling occurred seriously in coolant channels. There are also many uncertainties in the areas of transient heat transfer correlations and void volume production.
The thermal hydrodynamics model used has caused numerical instability in certain situations, i.e. transient boiling occurred seriously in coolant channels. There are also many uncertainties in the areas of transient heat transfer correlations and void volume production.
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10. REFERENCES
- M. Ishikawa, Y. Kuge, N. Ohnishi, E. Takeuchi and Y. Kanbayashi:
'EUREKA, A Computer Code for Uranium-Oxide Fuelled, Water Cooled Reactor Kinetics Analysis.'
JAERI 1235 (September 1974).
- M. Ishikawa et al.:
'Fast Transient Analysis for Light Water Power Reactor by EUREKA, Coupled Nuclear Thermal Hydrodynamic Kinetic Code'
JAERI 1201 (January 1971).
- M. Ishikawa, N. Ohnishi, Y. Kuge, E. Takeuchi and Y. Kanbayashi:
'Analysis of Moderator Feedback Effects on Rapid Power Burst by EUREKA'
NEA Computer Program Library Newsletter no. 19 (June 1975).
- M. Ishikawa, Y. Kuge, N. Ohnishi, E. Takeuchi and Y. Kanbayashi:
'EUREKA, A Computer Code for Uranium-Oxide Fuelled, Water Cooled Reactor Kinetics Analysis.'
JAERI 1235 (September 1974).
- M. Ishikawa et al.:
'Fast Transient Analysis for Light Water Power Reactor by EUREKA, Coupled Nuclear Thermal Hydrodynamic Kinetic Code'
JAERI 1201 (January 1971).
- M. Ishikawa, N. Ohnishi, Y. Kuge, E. Takeuchi and Y. Kanbayashi:
'Analysis of Moderator Feedback Effects on Rapid Power Burst by EUREKA'
NEA Computer Program Library Newsletter no. 19 (June 1975).
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NEA-0447/01
| File name | File description | Records |
|---|---|---|
| NEA0447_01.001 | SOURCE PROGRAM (F4) EBCDIC | 5134 |
| NEA0447_01.002 | ASSEMBLER ROUTINE EBCDIC | 81 |
| NEA0447_01.003 | SAMPLE PROBLEM INPUT DATA | 60 |
| NEA0447_01.004 | SAMPLE PROBLEM PRINTED OUTPUT | 5084 |
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- F. Space - Time Kinetics, Coupled Neutronics - Hydrodynamics - Thermodynamics
- G. Radiological Safety, Hazard and Accident Analysis
Keywords: LWR reactors, accidents, control elements, coolants, hydrodynamics, kinetics, reactivity, temperature distribution, thermodynamics.