The impact of non-optimized operation of nuclear power plants on the nuclear fuel burn-up
-
H. R. V. de Oliveira
Abstract
The purpose of this paper is to model the additional nuclear power and fuel burn-up needed to compensate non-optimized operation. Technical limitations, such as changes in reactivity, the additional nuclear power needed to compensate deficiencies of the electrical system control, and other related nuclear parameters are evaluated. The equations that define and quantify the additional nuclear power that must be generated by a reactor in order to fulfill the commitment with the national electric system operator are modeled in a simple and precise way.
Kurzfassung
Der Zweck dieser Arbeit ist die Berechnung der zusätzlich benötigten Leistung und des höheren Abbrands von Kernkraftwerken aufgrund eines nicht-optimalen Betriebsmodus durch Reaktivitätswechsel, Mängel im elektrischen Kontrollsystem und anderer nuklearer Parameter. Die Gleichungen, die die zusätzliche Leistung eines Reaktors bestimmen, der in das nationale Versorgungsnetz eingebunden ist, werden in dieser Arbeit auf einfache und präzise Weise ermittelt.
References
1 Kasatkin, A.; Perekalin, M.: Basic Electrical Engineering. 2 ed. Moscow, Peace Publishers, 1966.Search in Google Scholar
2 Fitzgerald, A. E.; Higginbotham, D. E.; GrabelA.: Basic Electrical Engineering. 4 ed. New York, McGraw-Hill, 1975.Search in Google Scholar
3 Duderstadt, J. J.; Hamilton, L. J.: Nuclear Reactor Analysis. New York, John Willey & Sons, 1976.Search in Google Scholar
4 Silva, R. A.; Zimmermann, E.: Development of an Integrated Computer System for Angra I Nuclear Power Plant. In: Proceedings of the Conference on Man-Machine Interface in the Nuclear Industry, pp. 215–223, Tokyo, 1998.Search in Google Scholar
5 Martinez, A. S.; Schirru, R.; Thome, Z. D.: Improving Safety Parameter Display System for Normal Operation. In: Proceedings of the Conference on Man-Machine Interface in the Nuclear Industry, pp. 391–399, Tokyo, 1988.Search in Google Scholar
6 Martinez, A. S.; Oliveira, L. F. S.; Schirru, R.; Thome, Z. D.: A New Concept of Safety Parameter Display System. In: Proceedings of the Seminar on Nuclear Engineering, pp. 407–418, Mexico, 1986.Search in Google Scholar
© 2004, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Improved Dancoff factors for cluster fuel bundles by the WIMSD code
- The impact of non-optimized operation of nuclear power plants on the nuclear fuel burn-up
- Experience with MTR fuel at the HOR reactor
- Comparison of two modern approaches for MBDA calculations of RBMK-type reactors
- Calculation of the energy deposition in the targets from C to U irradiated with intermediate energy protons
- U-Pb accelerator-driven sub-critical systems research at the proton beam of the JINR NUCLOTRON
- Steady state thermal-hydraulic analysis for the ETRR-2 Research Reactor
- Monte Carlo simulation of the ETRR-2 research reactor using the MCNP Code
- Preliminary evaluation of the fixed and fluidized bed nuclear reactor concept using the IAEA-INPRO methodology
- Technical Notes/Technische Mitteilungen
- A clean nuclear blasting technique for the excavation of a new isthmus canal in Central America
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Improved Dancoff factors for cluster fuel bundles by the WIMSD code
- The impact of non-optimized operation of nuclear power plants on the nuclear fuel burn-up
- Experience with MTR fuel at the HOR reactor
- Comparison of two modern approaches for MBDA calculations of RBMK-type reactors
- Calculation of the energy deposition in the targets from C to U irradiated with intermediate energy protons
- U-Pb accelerator-driven sub-critical systems research at the proton beam of the JINR NUCLOTRON
- Steady state thermal-hydraulic analysis for the ETRR-2 Research Reactor
- Monte Carlo simulation of the ETRR-2 research reactor using the MCNP Code
- Preliminary evaluation of the fixed and fluidized bed nuclear reactor concept using the IAEA-INPRO methodology
- Technical Notes/Technische Mitteilungen
- A clean nuclear blasting technique for the excavation of a new isthmus canal in Central America
