Analysis of safety parameters for the MINERVE reactor
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A. E. Ateya
Abstract
The present study aims to show the effect of different cross section libraries on MINERVE reactor safety parameters. The MCNP5 calculation model of the MINERVE reactor facility is used to determine core and safety parameters such as axial and radial fission rate distributions, control rod worth and spectral indices. Different neutron spectra were achieved by changing the experimental lattice within the MINERVE reactor. MINERVE provides a large experimental basis for the improvement of the cross section databases. The current study calculates these parameters by MCNP5 code using three different cross section libraries (the continuous energy cross sections of the ENDFB-VI, T16-2003, and ENDF/B-VII.1 libraries). Neutronic calculations were performed for R1UO2, R1MOX, R2UO2, and R2MOX core configurations, representative of a LWR loaded with UO2, mixed oxide matrix, over moderated UO2 and over moderated MOX, respectively. The study aims to determine the most suitable cross section library to be used with MCNP5 reactor calculation code for light water reactor fuel lattice. The MCNP5 results were compared with the experimental results.
Kurzfassung
Ziel dieses Beitrags ist es, den Einfluss verschiedener Wirkungsquerschnitts-Bibliotheken auf die Sicherheitsparameter des MINERVE Reaktors zu untersuchen. Das MCNP5-Rechenmodell der MINERVE-Reaktoranlage wird verwendet zur Bestimmung von Sicherheitsparametern wie z. B. axiale und radiale Spaltratenverteilung, Steuerstabgüte und spektralen Indizes. Verschiedene Neutronenspektren wurden durch Änderung des experimentellen Gitters innerhalb des MINERVE-Reaktors erzeugt. MINERVE bietet eine große experimentelle Basis für die Verbesserung von Wirkungsquerschnitts-Datenbanken. In der derzeitigen Studie werden diese Parameter mit Hilfe des MCNP5-Codes unter Verwendung von drei verschiedenen Wirkungsquerschnitts-Bibliotheken (ENDFB-VI, T16-2003 und ENDF/B-VII.1) berechnet. Neutronenphysikalische Simulationsrechnungen wurden für R1UO2, R1MOX, R2UO2, und R2MOX-Kernkonfigurationen durchgeführt. Ziel war es, die am besten geeignete Wirkungsquerschnitts-Bibliothek für die MCNP5-Berechnungen von Brennstoffgittern in Leichtwasserreaktoren zu bestimmen. Die MCNP5-Rechenergebnisse wurden mit experimentellen Ergebnissen verglichen.
References
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© 2015, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Improved data evaluation methodology for energy ranges with missing experimental data
- Importance weighting of local flux measurements to improve reactivity predictions in nuclear systems
- Safety assessment of fuel cycle facilities following the lessons learned from the accident at the Fukushima-Daiichi nuclear power plant
- The ultimate response guideline simulation and analysis by using TRACE for Lungmen ABWR nuclear power plant
- Development of a hydrogen diffusion gothic model of MARK III-containment
- Analysis of safety parameters for the MINERVE reactor
- Nondestructive radioactive tracer technique in performance evaluation of organic based ion exchange materials Purolite NRW-4000 and Duolite A-378
- Neutronic performance of (ReprocessedU/Th)O2 fuel in CANDU 6 reactor
- Modelling study on production cross sections of 111In radioisotopes used in nuclear medicine
- Study of 60Co as gamma source in backscatter gamma densitometers
- Determination of activity concentration of natural and artificial radionuclides in sand samples from mediterranean coast of Antalya in Turkey
- Technical Notes
- Numerical solution of diffusion equation to study fast neutrons flux distribution for variant radii of nuclear fuel pin and moderator regions
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Improved data evaluation methodology for energy ranges with missing experimental data
- Importance weighting of local flux measurements to improve reactivity predictions in nuclear systems
- Safety assessment of fuel cycle facilities following the lessons learned from the accident at the Fukushima-Daiichi nuclear power plant
- The ultimate response guideline simulation and analysis by using TRACE for Lungmen ABWR nuclear power plant
- Development of a hydrogen diffusion gothic model of MARK III-containment
- Analysis of safety parameters for the MINERVE reactor
- Nondestructive radioactive tracer technique in performance evaluation of organic based ion exchange materials Purolite NRW-4000 and Duolite A-378
- Neutronic performance of (ReprocessedU/Th)O2 fuel in CANDU 6 reactor
- Modelling study on production cross sections of 111In radioisotopes used in nuclear medicine
- Study of 60Co as gamma source in backscatter gamma densitometers
- Determination of activity concentration of natural and artificial radionuclides in sand samples from mediterranean coast of Antalya in Turkey
- Technical Notes
- Numerical solution of diffusion equation to study fast neutrons flux distribution for variant radii of nuclear fuel pin and moderator regions
