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2025 ANS Annual Conference
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Chicago, IL|Chicago Marriott Downtown
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High-temperature plumbing and advanced reactors
The use of nuclear fission power and its role in impacting climate change is hotly debated. Fission advocates argue that short-term solutions would involve the rapid deployment of Gen III+ nuclear reactors, like Vogtle-3 and -4, while long-term climate change impact would rely on the creation and implementation of Gen IV reactors, “inherently safe” reactors that use passive laws of physics and chemistry rather than active controls such as valves and pumps to operate safely. While Gen IV reactors vary in many ways, one thing unites nearly all of them: the use of exotic, high-temperature coolants. These fluids, like molten salts and liquid metals, can enable reactor engineers to design much safer nuclear reactors—ultimately because the boiling point of each fluid is extremely high. Fluids that remain liquid over large temperature ranges can provide good heat transfer through many demanding conditions, all with minimal pressurization. Although the most apparent use for these fluids is advanced fission power, they have the potential to be applied to other power generation sources such as fusion, thermal storage, solar, or high-temperature process heat.1–3
Byoung-Uhn Bae, Jae-Bong Lee, Yu-Sun Park, Jong-Rok Kim, Seok Cho, Kyoung-Ho Kang
Nuclear Technology | Volume 207 | Number 5 | May 2021 | Pages 680-691
Technical Paper | doi.org/10.1080/00295450.2020.1796078
Articles are hosted by Taylor and Francis Online.
To investigate thermal-hydraulic phenomena during an intermediate-break loss-of-coolant accident (IBLOCA) and evaluate the effect of a direct vessel injection (DVI) line break, an integral effect test using the Advanced Thermal-hydraulic Test Loop for Accident Simulation (ATLAS) test facility was conducted as the B3.2 test item of the international cooperation project Organisation for Economic Co-operation and Development (OECD)–ATLAS Project Phase 2 (ATLAS-2) (OECD-ATLAS2). The initial and boundary conditions for the test were determined referring to the Advanced Power Reactor 1400 MW(electric) (APR1400) as a prototype with three-level scaling methodology. A single-failure criterion was applied to the operation of the safety injection pump (SIP), and four safety injection tanks (SITs) were available to cool down the reactor coolant system. In the test result, as the break nozzle was located at the DVI line, the clearance of the upper downcomer could make an effective flow path of the steam toward the break and quench the reactor core. Maximum cladding temperature was measured before clearance of the upper downcomer. Coolant inventory in the reactor pressure vessel was maintained due to the safety injection without any further core heatup. So, it was proved that the current design of the safety systems in APR1400 had a sufficient long-term cooling capability with a single SIP during a DVI line break IBLOCA. The ATLAS test data were utilized to evaluate the prediction capability of the thermal-hydraulic system code Multi-dimensional Analysis of Reactor Safety KINS Standard (MARS-KS) for a DVI line break IBLOCA scenario. The calculation result with the uncertainty propagation analysis using the PArallel computing Platform IntegRated for Uncertainty and Sensitivity analysis (PAPIRUS) toolkit proved that major phenomena such as uncovery of the core or intermittent injection of the SIT flow could be reasonably predicted by the MARS-KS code.