Technical feasibility of using RU-43 fuel in the CANDU-6 reactors of the Cernavoda NPP
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G. Horhoianu
Abstract
Recovered uranium (RU) is a by-product of many light-water reactor (LWR) fuel recycling programs. A fissile content in the RU of 0.9 to 1.0 % makes it impossible for reuse in an LWR without re-enrichment, but CANDU reactors have a sufficiently high neutron economy to use RU as fuel. The Institute for Nuclear Research (INR) Pitesti has analyzed the feasibility of using RU fuel with 0.9 – 1.1 w% 235U in the CANDU-6 reactors of the Cernavoda Nuclear Power Plant (Cernavoda NPP). Using RU fuel would produce a significant increase in the fuel discharge burnup, from 170 MWh/kgU currently achieves with natural-uranium (NU) fuel to about 355 MWh/kgU. This would lead to reduced fuel-cycle cost and a large reduction in spent-fuel volume per full-power-year of operation. The RU fuel bundle design with recovered uranium fuel, known as RU-43, is being developed by the INR Pitesti and is now at the stage of final design verification. Early work has been concentrated on RU-43 fuel bundle design optimization, safety and reactor physics assessment. The changes in fuel element and fuel bundle design contribute to the many advantages offered by the RU-43 bundle. Verification of the design of the RU-43 fuel bundle is performed in a way that shows that design criteria are met, and is mostly covered by proof tests such as flow and irradiation tests. The most relevant calculations performed on this fuel bundle design version are presented. Also, the stages of an experimental program aiming to verify the operating performance are briefly described in this paper.
Kurzfassung
Wiedergewonnenes Uran (RU) ist ein Abfallprodukt vieler Leichtwasserreaktor-Brennstoffrezyklierungsprogramme. Sein spaltbarer Anteil von 0,9 bis 1,0 % macht es ohne Wieder-Anreicherung für den erneuten Einsatz im LWR untauglich. CANDU-Reaktoren haben hingegen eine ausreichend gute Neutronen-Ökonomie, um RU als Brennstoff verwenden zu können. Das Kernforschungsinstitut in Pitesti (INR) hat die Machbarkeit des Einsatzes von RU-Brennstoff mit 0,9 bis 1,1 Gew.% U235 in den CANDU-6-Reaktoren des Kernkraftwerks Cernavoda analysiert. Der Einsatz von RU-Brennstoff würde zu einem signifikanten Anstieg des Entlade-Abbrands, von derzeit 170 MWh/kgU bei Natururan auf etwa 355 MWh/kgU, führen. Dies hätte niedrigere Brennstoffkreislaufkosten und eine bedeutende Volumen-Reduktion des abgebrannten Brennstoffs über ein Volllast-Betriebsjahr zur Folge. Ein RU-Brennstoffbündel ist unter der Bezeichnung RU-43 an INR Pitesti in der Entwicklung, es befindet sich derzeit in der endgültigen Design-Verifikation. Die anfänglichen Arbeiten konzentrierten sich auf die Design-Optimierung und Abschätzungen zu Sicherheit und Reaktorphysik des RU-43-Elements. Die Änderungen am Brennelement- und Brennstoffbündelauslegung tragen zu vielen Vorteilen des RU-43 bei. Bei der Design-Verifikation wird die Einhaltung der Auslegungskriterien nachgewiesen, sie erfolgt im Wesentlichen durch Durchfluss- und Bestrahlungsversuche. Nachstehend werden die wichtigsten Rechenergebnisse für dieses Brennstoffbündel-Design präsentiert. Darüber hinaus enthält der Beitrag eine kurze Beschreibung des Versuchsprogramms, mit dem seine Betriebseigenschaften nachgewiesen werden sollen.
References
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© 2008, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Analysis of the hydrogen behaviour in compartments of the Ignalina nuclear power plant
- Determination of the RBMK-1500 reactor passport characteristics
- Technical feasibility of using RU-43 fuel in the CANDU-6 reactors of the Cernavoda NPP
- Computational study of moderator flow and temperature fields in the calandria vessel of a heavy water reactor using the PHOENICS code
- Regulatory requirements on level 2 PSA in Germany and their associated potential to improve emergency management
- Dynamic structure response due to reactor cooling piping system failure
- Uncertainty in activation cross-section calculations at intermediate proton energies
- The LTSN solution of the transport equation for one-dimensional cartesian geometry with c = 1
- Application of the UN approximation to the neutron transport equation in slab geometry
- The reflected critical slab problem for one-speed neutrons with strongly anisotropic scattering
- The analytical representation of the fundamental mode in 1-D-geometry for the CANDLE burn-up phenomenon
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Analysis of the hydrogen behaviour in compartments of the Ignalina nuclear power plant
- Determination of the RBMK-1500 reactor passport characteristics
- Technical feasibility of using RU-43 fuel in the CANDU-6 reactors of the Cernavoda NPP
- Computational study of moderator flow and temperature fields in the calandria vessel of a heavy water reactor using the PHOENICS code
- Regulatory requirements on level 2 PSA in Germany and their associated potential to improve emergency management
- Dynamic structure response due to reactor cooling piping system failure
- Uncertainty in activation cross-section calculations at intermediate proton energies
- The LTSN solution of the transport equation for one-dimensional cartesian geometry with c = 1
- Application of the UN approximation to the neutron transport equation in slab geometry
- The reflected critical slab problem for one-speed neutrons with strongly anisotropic scattering
- The analytical representation of the fundamental mode in 1-D-geometry for the CANDLE burn-up phenomenon
