European Project "Supercritical Water Reactor-Fuel Qualification Test": Overview, Results, Lessons Learned, and Future Outlook

Mariana Ruzickova, Ales Vojacek, Thomas Schulenberg, Dirk C. Visser, Radek Novotny, Attila Kiss, Csaba Maraczy, Aki Toivonen

    Research output: Contribution to journalArticleScientificpeer-review

    3 Citations (Scopus)

    Abstract

    The supercritical water reactor (SCWR) is one of the six reactor concepts being investigated under the framework of the Generation IV International Forum (GIF). One of the major challenges in the development of a SCWR is to develop materials for the fuel and core structures that will be sufficiently corrosion resistant to withstand supercritical water conditions and to gain thermal-hydraulic experimental data that could be used for further improvement of heat transfer predictions in the supercritical region by numerical codes. Previously, core, reactor, and plant design concepts of the European high-performance light water reactor (HPLWR) have been worked out in great detail. As the next step, it has been proposed to carry out a fuel qualification test (FQT) of a small-scale fuel assembly in a research reactor under typical prototype conditions. Design and licensing of an experimental facility for the FQT, including the small-scale fuel assembly, the required coolant loop with supercritical water, and safety and auxiliary systems, was the scope of the recently concluded project "Supercritical Water Reactor-Fuel Qualification Test" (SCWR-FQT) described here. This project was a collaborative project cofunded by the European Commission, which took advantage of a Chinese-European collaboration, in which China offered an electrically heated out-of-pile loop for testing of fuel bundles. The design of the facility, especially of the test section with the fuel assembly, and the most important results of steady-state and safety analyses are presented. Material test results of the stainless steels considered for the fuel cladding are briefly summarized. Finally, important outcomes and lessons learned in the "Education and Training" and "Management" work packages are presented.
    Original languageEnglish
    Article number011002
    Number of pages10
    JournalJournal of Nuclear Engineering and Radiation Science
    Volume2
    Issue number1
    DOIs
    Publication statusPublished - 2016
    MoE publication typeA1 Journal article-refereed

    Keywords

    • SCWR
    • stress corrosion
    • corrosion
    • thermohydraulics
    • test loop

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