Thermal hydraulic assessment of a new design of PWR fuel assembly
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Abstract
The recycling of the spent fuel of nuclear reactors is the important aspects in the fuel cycle technology. One of the recycling of the spent fuel is reuse of spent PWR (pressurized water reactor) fuel in CANDU reactor after re-fabricated into CANDU (CANda deuterium uranium) fuel bundles; this technology called Direct Use of Spent PWR Fuel In CANDU Reactor (DUPIC). This technology was modified however; it introduced a new design for the PWR fuel assembly. The new fuel can be transferred directly to CANDU reactors without processing, only after a certain cooling time. In this study the evaluation from thermal hydraulic point of view is done. A Westinghouse PWR fuel assembly was selected as a reference. Thermal hydraulic simulation for reference and new design fuel occurred by using RELAP5 (Reactor Excursion and Leak Analysis Program) code, then comparing the both results. The hottest rod was considered a worst condition in the reactor so the model simulated it rather than whole core. The new design achieves a good results comparing with a reference one.
Kurzfassung
Das Recycling der abgebrannten Brennelemente von Kernreaktoren ist ein wichtiger Aspekt in der Brennstoffkreislauftechnik. Eines der Recyclingverfahren für abgebrannte Brennelemente ist die Wiederverwendung abgebrannter DWR-Brennstoffe (Druckwasserreaktor) im CANDU-Reaktor, nachdem sie in ein CANDU-Brennstoffbündel (CANda deuterium uranium) umgebaut wurden; diese Technologie heißt Direct Use of Spent PWR Fuel In CANDU Reactor (DUPIC). Diese Technologie wurde modifiziert, indem ein neues Design für das PWR-Brennstoffbündel eingeführt wurde. Der neue Brennstoff kann erst nach einer gewissen Abkühlzeit ohne weitere Bearbeitung direkt in die CANDU-Reaktoren überführt werden. In dieser Studie wird die Bewertung aus thermohydraulischer Sicht durchgeführt. Als Referenz wurde ein Westinghouse PWR Brennelement ausgewählt. Die thermohydraulische Simulation für Referenz- und neue Designbrennstoffe erfolgte unter Verwendung des RELAP5-Codes (Reactor Excursion and Leak Analysis Program) und anschließendem Vergleich der beiden Ergebnisse. Dabei wurde das Hauptaugenmerk auf die Berechnung des heißesten Stabs gelegt. Das neue Design erzielt gute Ergebnisse im Vergleich zu einem Referenzmodell.
References
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© 2018, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Technical Contributions/Fachbeiträge
- Porosity evolution study in irradiated UO2 fuel based on fuel matrix swelling
- Data collection assessment for the human performance analysis in nuclear installations
- Abnormal control rod withdrawal analysis for innovative research reactor using PARET-ANL codes
- Improving CANDU performance by using uranium – thorium mixed oxide fuel
- Computation of gamma radioactivity of natural rocks in the vicinity of Antalya province and its effect on health
- Calculation of liquid waste discharge limits for routine discharge of Tehran research reactor
- Thermal hydraulic assessment of a new design of PWR fuel assembly
- Simulation of error in suppression of Xenon oscillations in a WWER-1000 nuclear reactor
- A novel dual-molality densitometer for gauging in annular two phase flows using radial basis function
- Design and construction of toroidal field coil for Taban device
Articles in the same Issue
- Contents/Inhalt
- Contents
- Technical Contributions/Fachbeiträge
- Porosity evolution study in irradiated UO2 fuel based on fuel matrix swelling
- Data collection assessment for the human performance analysis in nuclear installations
- Abnormal control rod withdrawal analysis for innovative research reactor using PARET-ANL codes
- Improving CANDU performance by using uranium – thorium mixed oxide fuel
- Computation of gamma radioactivity of natural rocks in the vicinity of Antalya province and its effect on health
- Calculation of liquid waste discharge limits for routine discharge of Tehran research reactor
- Thermal hydraulic assessment of a new design of PWR fuel assembly
- Simulation of error in suppression of Xenon oscillations in a WWER-1000 nuclear reactor
- A novel dual-molality densitometer for gauging in annular two phase flows using radial basis function
- Design and construction of toroidal field coil for Taban device
