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Benchmark and demonstration of the CHD code for transient analysis of fast reactor systems
Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
Uppsala Univ, Uppsala, Sweden.;Univ Calif Berkeley, Dept Nucl Engn, Berkeley, CA 94720 USA..
2017 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 109, p. 712-719Article in journal (Refereed) Published
Abstract [en]

In this paper the dynamic thermal hydraulic fast reactor simulation code CHD is presented. The code is built around a scriptable object-oriented framework in the programming language Python to be able to flexibly describe different reactor geometries including thermal-hydraulics models of an arbitrary number of coolant channels as well as pumps, heat-exchangers and pools etc. In addition, custom objects such as the Autonomous Reactivity Control (ARC) system for enhanced passive safety are modeled in detail. In this paper we compare the performance of the CHD code with other similar fast reactor dynamics codes using a benchmark study of the European Sodium cooled Fast Reactor (ESFR). The results agree well, both qualitatively and quantitatively with the code benchmark. In addition, we demonstrate the code's ability to simulate the long-term asymptotic behavior of a neutronically shut down reactor in an unprotected loss of flow scenario using a model of the Advanced Burner Reactor (ABR). (C) 2017 Elsevier Ltd. All rights reserved.

Place, publisher, year, edition, pages
Elsevier, 2017. Vol. 109, p. 712-719
Keywords [en]
Fast reactor, Thermal-hydraulics, Point-kinetics, Transient, Passive safety, ULOF
National Category
Energy Engineering
Identifiers
URN: urn:nbn:se:uu:diva-339705DOI: 10.1016/j.anucene.2017.05.031ISI: 000418211500071OAI: oai:DiVA.org:uu-339705DiVA, id: diva2:1177806
Available from: 2018-01-26 Created: 2018-01-26 Last updated: 2018-01-26Bibliographically approved

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Hellesen, C.

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