Assessment of neutronic parameter's uncertainties obtained within the reactor dosimetry framework: Development and application of the stochastic methods of analysis
- Service de Physique Experimentale, CEA-CAD/DEN/DER/SPEx, Departement d'Etudes des Reacteurs, 13108 St-Paul lez Durance Cedex (France)
- Service d'Etude des Systemes Innovant, CEA-CAD/DEN/DER/SESI, Departement d'Etudes des Reacteurs, 13108 St-Paul lez Durance Cedex (France)
One of the main objectives of reactor dosimetry is the determination of the physical parameters characterizing the neutronic field in which the studied sample is irradiated. The knowledge of the associated uncertainties represents a significant stake for nuclear industry as shows the high uncertainty value of 15% (k=1) commonly allowed for the calculated neutron flux (E> 1 MeV) on the vessel and internal structures. The study presented in this paper aims at determining then reducing uncertainties associated with the reactor dosimetry interpretation process. After a brief presentation of the interpretation process, input data uncertainties identification and quantification are performed in particular with regard to covariances. Then uncertainties propagation is carried out and analyzed by deterministic and stochastic methods on a representative case. Finally, a Monte Carlo sensitivity study based on Sobol indices is achieved on a case leading to derive the most penalizing input uncertainties. This paper concludes rising improvement axes to be studied for the input data knowledge. It highlights for example the need for having realistic variance-covariance matrices associated with input data (cross sections libraries, neutron computation code's outputs, ...). Lastly, the methodology principle presented in this paper is enough general to be easily transposable for other measurements data interpretation processes. (authors)
- Research Organization:
- American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
- OSTI ID:
- 22039505
- Resource Relation:
- Conference: PHYSOR-2006: American Nuclear Society's Topical Meeting on Reactor Physics - Advances in Nuclear Analysis and Simulation, Vancouver, BC (Canada), 10-14 Sep 2006; Other Information: Country of input: France; 10 refs.
- Country of Publication:
- United States
- Language:
- English
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