FOEHN: The critical experiment for the Franco-German High Flux Reactor
- Kernforschungszentrum Karlsruhe GmbH (Germany)
A critical experiment for the Franco-German High Flux Reactor was carried out in the French reactor EOLE (CEN Cadarache). The purpose of the experiment was to check the calculation methods in a realistic geometry and to measure effects that can only be calculated imprecisely (e.g. beam hole effects). The structure of the experiment and the measurement and calculation methods are described. A detailed comparison between theoretical and experimental results was performed. 30 refs., 105 figs.
- Research Organization:
- Kernforschungszentrum Karlsruhe (Germany). Instltut fuer Neutronenphysik und Reaktortechnik
- DOE Contract Number:
- AC05-84OR21400
- OSTI ID:
- 7399183
- Report Number(s):
- ORNL/tr-91/23; KFK-1064; ON: DE92000159
- Resource Relation:
- Related Information: Translation of the German report KFK-1064
- Country of Publication:
- Germany
- Language:
- English
Similar Records
MCNP analysis of the FOEHN critical experiment
Sensitivity and uncertainty analysis applied to the JHR reactivity prediction
The Reactivity Temperature Coefficient Analysis in Light Water Moderated UO{sub 2} and UO{sub 2}-PuO{sub 2} Lattices
Technical Report
·
Fri Oct 01 00:00:00 EDT 1993
·
OSTI ID:7399183
+1 more
Sensitivity and uncertainty analysis applied to the JHR reactivity prediction
Conference
·
Sun Jul 01 00:00:00 EDT 2012
·
OSTI ID:7399183
+3 more
The Reactivity Temperature Coefficient Analysis in Light Water Moderated UO{sub 2} and UO{sub 2}-PuO{sub 2} Lattices
Journal Article
·
Thu May 15 00:00:00 EDT 2003
· Nuclear Science and Engineering
·
OSTI ID:7399183
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
EOLE REACTOR
CRITICALITY
GEOMETRY
GRENOBLE REACTOR
BORON
CALCULATION METHODS
CONTROL ELEMENTS
DELAYED NEUTRONS
DESIGN
DIFFUSION
EXPERIMENTAL DATA
MEASURING METHODS
NEUTRON FLUX
NEUTRON REFLECTORS
POWER DISTRIBUTION
REACTIVITY
REACTOR CONTROL SYSTEMS
REACTOR CORES
REACTOR INTERNALS
REACTOR KINETICS
THEORETICAL DATA
THERMAL NEUTRONS
VOIDS
BARYONS
CONTROL SYSTEMS
DATA
ELEMENTARY PARTICLES
ELEMENTS
FERMIONS
FISSION NEUTRONS
HADRONS
HEAVY WATER COOLED REACTORS
HEAVY WATER MODERATED REACTORS
INFORMATION
KINETICS
MATHEMATICS
NEUTRONS
NUCLEONS
NUMERICAL DATA
RADIATION FLUX
REACTOR COMPONENTS
REACTORS
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SEMIMETALS
TANK TYPE REACTORS
TEST REACTORS
220600* - Nuclear Reactor Technology- Research
Test & Experimental Reactors
220100 - Nuclear Reactor Technology- Theory & Calculation
22 GENERAL STUDIES OF NUCLEAR REACTORS
EOLE REACTOR
CRITICALITY
GEOMETRY
GRENOBLE REACTOR
BORON
CALCULATION METHODS
CONTROL ELEMENTS
DELAYED NEUTRONS
DESIGN
DIFFUSION
EXPERIMENTAL DATA
MEASURING METHODS
NEUTRON FLUX
NEUTRON REFLECTORS
POWER DISTRIBUTION
REACTIVITY
REACTOR CONTROL SYSTEMS
REACTOR CORES
REACTOR INTERNALS
REACTOR KINETICS
THEORETICAL DATA
THERMAL NEUTRONS
VOIDS
BARYONS
CONTROL SYSTEMS
DATA
ELEMENTARY PARTICLES
ELEMENTS
FERMIONS
FISSION NEUTRONS
HADRONS
HEAVY WATER COOLED REACTORS
HEAVY WATER MODERATED REACTORS
INFORMATION
KINETICS
MATHEMATICS
NEUTRONS
NUCLEONS
NUMERICAL DATA
RADIATION FLUX
REACTOR COMPONENTS
REACTORS
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SEMIMETALS
TANK TYPE REACTORS
TEST REACTORS
220600* - Nuclear Reactor Technology- Research
Test & Experimental Reactors
220100 - Nuclear Reactor Technology- Theory & Calculation