Lessons learned from hydrogen generation and burning during the TMI-2 event

Henrie, J.O.; Postma, Albert

doi:10.2172/6237032

Cited by 21 publications

(13 citation statements)

References 13 publications

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“…However, because of their slow sampling rate, recorded peak temperatures from the RTDs were less than 93°C, which was much lower than the 650°C estimates obtained assuming peak pressures measured in the reactor building. 84,83 Physical damage to organic materials substantiated that temperatures exceeded 232°C, which was also much higher than available data from the RTDs. Hence, data were evaluated as "Qualified" for all times except at temperature peaks.…”

Section: Statusmentioning

confidence: 80%

“…Several sources of data were available to estimate the reactor building pressure during the TMI-2 accident. 83,84 There were two pressure transmitters associated with the strip chart recorder for which data were recorded continuously, and six reactor building pressure switches. In addition, pressure transmitters measured the reactor building pressure as a steam generator reference pressure, and these data were recorded on the reactimeter every 3 seconds.…”

Section: Building Pressurementioning

confidence: 99%

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TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

Rempe¹,

Knudson²

2013

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Section: Statusmentioning

confidence: 80%

Section: Building Pressurementioning

confidence: 99%

TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

Rempe¹,

Knudson²

2013

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“…Because temperature measurements are key to understanding the TMI-2 accident, the survivability and performance of the 16 Resistance Temperature Detectors (RTDs) in the reactor building air handling system were investigated as part of the TMI-2 Data Qualification effort. As described within this section, evaluations 24,25 included analyzing data collected during and after the accident, observations of damage to materials within the containment, and data from in-situ tests conducted after the accident. In addition, comparisons were made with other data obtained from the containment and engineering analyses performed using such data.…”

Section: Containment Temperaturementioning

confidence: 99%

Qualification of Daiichi Units 1, 2, and 3 Data for Severe Accident Evaluations - Process and Illustrative Examples from Prior TMI-2 Evaluations

Rempe

Knudson

2014

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The accidents at the Three Mile Island Unit 2 (TMI-2) Pressurized Water Reactor (PWR) and the Daiichi Units 1, 2, and 3 Boiling Water Reactors (BWRs) provide unique opportunities to evaluate instrumentation exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during the TMI-2 accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This review focused on the set of sensors deemed most important by post-TMI-2 instrumentation evaluation programs. Instrumentation evaluation programs focused on data required by TMI-2 operators to assess the condition of the reactor and containment and the effect of mitigating actions taken by these operators. In addition, prior efforts focused on sensors providing data required for subsequent forensic evaluations and accident simulations.

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“…Because of the magnitude of the deflagration event that occurred in the containment building in the TMI-2 accident, there has been considerable interest in hydrogen combustion in the severe accident research community [59]. A major deflagration event in the containment could potentially cause a pressure spike great enough to challenge the integrity of the containment walls.…”

Section: Hydrogen Burnmentioning

confidence: 99%

Development and application of the dynamic system doctor to nuclear reactor probabilistic risk assessments.

Kunsman¹,

Aldemir

Rutt

et al. 2008

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This LDRD project has produced a tool that makes probabilistic risk assessments (PRAs) of nuclear reactors-analyses which are very resource intensive-more efficient. PRAs of nuclear reactors are being increasingly relied on by the United States Nuclear Regulatory Commission (U.S.N.R.C.) for licensing decisions for current and advanced reactors. Yet, PRAs are produced much as they were 20 years ago. The work here applied a modern systems analysis technique to the accident progression analysis portion of the PRA; the technique was a system-independent multi-task computer driver routine.Initially, the objective of the work was to fuse the accident progression event tree (APET) portion of a PRA to the dynamic system doctor (DSD) created by Ohio State University. Instead, during the initial efforts, it was found that the DSD could be linked directly to a detailed accident progression phenomenological simulation code-the type on which APET construction and analysis relies, albeit indirectly-and thereby directly create and analyze the APET. The expanded DSD computational architecture and infrastructure that was created during this effort is called ADAPT (Analysis of Dynamic Accident Progression Trees). ADAPT is a system software infrastructure that supports execution and analysis of multiple dynamic event-tree simulations on distributed environments. A simulator abstraction layer was developed, and a generic driver was implemented for executing simulators on a distributed environment.As a demonstration of the use of the methodological tool, ADAPT was applied to quantify the likelihood of competing accident progression pathways occurring for a particular accident scenario in a particular reactor type using MELCOR, an integrated severe accident analysis code developed at Sandia. (ADAPT was intentionally created with flexibility, however, and is not limited to interacting with only one code. With minor coding changes to input files, ADAPT can be linked to other such codes.) The results of this demonstration indicate that the approach can significantly reduce the resources required for Level 2 PRAs. From the phenomenological viewpoint, ADAPT can also treat the associated epistemic and aleatory uncertainties.This methodology can also be used for analyses of other complex systems. Any complex system can be analyzed using ADAPT if the workings of that system can be displayed as an event tree, there is a computer code that simulates how those events could progress, and that simulator code has switches to turn on and off system events, phenomena, etc.Using and applying ADAPT to particular problems is not human independent. While the human resources for the creation and analysis of the accident progression are significantly decreased, knowledgeable analysts are still necessary for a given project to apply ADAPT successfully.This research and development effort has met its original goals and then exceeded them.4

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Lessons learned from hydrogen generation and burning during the TMI-2 event

Cited by 21 publications

References 13 publications

TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

Qualification of Daiichi Units 1, 2, and 3 Data for Severe Accident Evaluations - Process and Illustrative Examples from Prior TMI-2 Evaluations

Development and application of the dynamic system doctor to nuclear reactor probabilistic risk assessments.

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