OECD MCCI Project: Category 2 Coolability Engineering Enhancement Tests (Final Report)

Farmer, M. T.; Lomperski, S.; Aeschlimann, R. W.; Kilsdonk, D. J.

doi:10.2172/1177571

Cited by 4 publications

(13 citation statements)

References 14 publications

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“…As shown in Table 1-2, a total of three tests were carried out to provide data for these two approaches. The results are summarized below under these same two general headings; refer to [5] for detailed descriptions of the test facilities and experiment results for the various tests.…”

Section: Category 2 Test Resultsmentioning

confidence: 99%

“…As shown in Table 1-2, one large scale water-cooled basemat integral effect test was conducted in this area to provide data on the cold-crucible method for melt stabilization, and two separate effect tests that utilized the SSWICS test facility were conducted to provide data on the melt fragmentation method for melt stabilization [5]. The specific objective of the Water-Cooled Basemat (WCB-1) test was to provide prototypic data on the transient evolution and stabilization of a core melt in a generic water-cooled core catcher design for advanced plant applications.…”

Section: Crust Breachmentioning

confidence: 99%

“…Other key data recorded during the test included temperatures within test section structural sidewalls, melt/crust upper surface temperature, off gas temperature, and pressures at various locations within the system. Measurement uncertainties for the various instruments utilized in the tests are provided in [5].…”

Section: Apparatusmentioning

confidence: 99%

“…The tanks were positioned to provide initial net driving heads of 0.05, 0.1, 0.15, and 0.2 bar. The driving head accounted for the effect of the melt hydrostatic pressure on the elevation evaluation [5]. The tanks contained sufficient water inventory to completely quench the melt to saturation so that makeup water would not be needed during the test.…”

Section: Sswics-12 Apparatusmentioning

confidence: 99%

“…Four nozzles were cast within the basemat and positioned so that they were roughly equidistant from the liner and neighboring partitions. The top of each nozzle was flush with the basemat surface and each was filled with a specialty design porous concrete [5] to provide a path for water flow up into the melt while preventing melt flow downward into the nozzle. A leak-tight 0.05 mm-thick stainless steel diaphragm was welded over the top of each nozzle to serve as a temporary seal to prevent water flow from the feed line up into the thermite during the vessel preheat phase of the experiment.…”

Section: Sswics-12 Apparatusmentioning

confidence: 99%

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OECD MCCI-2 Project. Final Report

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et al. 2010

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Section: Category 2 Test Resultsmentioning

confidence: 99%

Section: Crust Breachmentioning

confidence: 99%

Section: Apparatusmentioning

confidence: 99%

Section: Sswics-12 Apparatusmentioning

confidence: 99%

Section: Sswics-12 Apparatusmentioning

confidence: 99%

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OECD MCCI-2 Project. Final Report

Farmer

Lomperski

Kilsdonk

et al. 2010

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Assessment of corium concrete interaction and impact of water ingression mechanism using top flooding strategy for a small PWR with MELCOR 2.2

Hengra

Song²,

Kim³

et al. 2023

Annals of Nuclear Energy

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Light Water Reactor Sustainability Program, U.S. Efforts in Support of Examinations at Fukushima Daiichi-2017 Evaluations

Farmer

2017

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Although the accident signatures from each unit at the Fukushima Daiichi Nuclear Power Station (NPS) [Daiichi] differ, much is not known about the end-state of core materials within these units. Some of this uncertainty can be attributed to a lack of information related to cooling system operation and cooling water injection. There is also uncertainty in our understanding of phenomena affecting: a) in-vessel core damage progression during severe accidents in boiling water reactors (BWRs), and b) accident progression after vessel failure (ex-vessel progression) for BWRs and Pressurized Water Reactors (PWRs). These uncertainties arise due to limited full scale prototypic data. Similar to what occurred after the accident at Three Mile Island Unit 2, these Daiichi units offer the international community a means to reduce such uncertainties by obtaining prototypic data from multiple full-scale BWR severe accidents.Information obtained from Daiichi is required to inform Decontamination and Decommissioning activities, improving the ability of the Tokyo Electric Power Company Holdings, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This document, which has been updated to include FY2017 information, summarizes results from U.S. efforts to use information obtained by TEPCO Holdings to enhance the safety of existing and future nuclear power plant designs. This effort, which was initiated in 2014 by the Reactor Safety Technologies Pathway of the Department of Energy Office of Nuclear Energy Light Water Reactor (LWR) Sustainability Program, consists of a group of U.S. experts in LWR safety and plant operations that have identified examination needs and are evaluating TEPCO Holdings information from Daiichi that address these needs. Each year, annual reports include examples demonstrating that significant safety insights are being obtained in the areas of component performance, fission product release and transport, debris end-state location, and combustible gas generation and transport. In addition to reducing uncertainties related to severe accident modeling progression, these insights are being used to update guidance for severe accident prevention, mitigation, and emergency planning. Furthermore, reduced uncertainties in modeling the events at Daiichi will improve the realism of reactor safety evaluations and inform future D&D activities by improving the capability for characterizing potential hazards to workers involved with cleanup activities.Highlights in this FY2017 report include new insights with respect to the forces required to produce the observed Daiichi Unit 1 (1F1) shield plug endstate, the observed leakage from 1F1 components, and the amount of combustible gas generation required to produce the observed explosions in Daiichi Units 3 and 4 (1F3 and 1F4). This report contains an appendix with a list of examination needs that was updated after U.S. experts reviewed recently obtained information from examinations at Daiichi. A...

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OECD MCCI Project: Category 2 Coolability Engineering Enhancement Tests (Final Report)

Cited by 4 publications

References 14 publications

OECD MCCI-2 Project. Final Report

OECD MCCI-2 Project. Final Report

Assessment of corium concrete interaction and impact of water ingression mechanism using top flooding strategy for a small PWR with MELCOR 2.2

Light Water Reactor Sustainability Program, U.S. Efforts in Support of Examinations at Fukushima Daiichi-2017 Evaluations

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