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ANS > Publications > Journals > Nuclear Technology > Volume 156 |
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Enhancement of Safety Analysis Capability for a CANDU-6 Reactor Using RELAP-CANDU/SCAN Coupled Code System Volume 156 · Number 2 · November 2006 · Pages 159-167 Technical Paper · Reactor Safety  Manwoong Kim, Hyun-Koon Kim, Hho-Jung Kim, Su Hyon Hwang, In Seob Hong, Chang Hyo Kim ANS Store: Purchase Article The purpose of this study is the development and verification of the coupled code system SCAN and RELAP-CANDU for transient analysis of a Canada deuterium uranium (CANDU) reactor. For this purpose, a spatial kinetics calculation module is developed and implemented in SCAN, a three-dimensional (3-D) CANDU-pressurized heavy water reactor neutronics design and analysis code. Then, a dynamic linked library of the SCAN code is generated for the integration with RELAP-CANDU. The RELAP-CANDU code has been developed for best-estimate transient simulation of CANDU reactor coolant systems based on the RELAP5 code. The SCAN code is a 3-D neutronic calculation code, which is composed of both unified nodal methods based on coarse-mesh finite difference method solutions to the time-dependent two-group diffusion equations. To verify the reliability of the coupled code system RELAP-CANDU/SCAN, the 40% reactor inlet header break accident, the 100% reactor outlet header break accident, and the pump suction pipe break are analyzed. The proposed coupled thermal-hydraulic and neutronic analyses methodology shows that there is an important margin in the traditional accident analysis. |
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