Thermal Interactions during the Three Mile Island Unit 2 2-B Coolant Pump Transient

Kuan, P.; Anderson, James L.; Tolman, E.L.

doi:10.13182/nt89-a27690

Cited by 6 publications

(3 citation statements)

References 2 publications

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“…An (about 2 mm) occurred in the SFD-ST test as a analysis of the increase in system pressure result of a minimum of power cycling and a slow measured during the B-loop pump transient in heatup in a steam-rich environment. The the TMI-2 accident and the associated energetics minimum power cycling produced relatively suggests that up to 32% of the hydrogen coarse crack patterns, and the slow heatup generated in the accident may have been enhanced oxidation prior to zircaloy melting and produced during this event (Kuan et al 1989). resulted in oxygen-rich zircaloy melts that were Hydrogen peaks following reflood have also 85 NUREG/CR-6160…”

Section: Nureg/cr-6160mentioning

confidence: 99%

“…assessment of possible mechanisms and the Other medium-and low-volatility fission energetics associated with the damage to the products (Sb and Ru) are largely absent from the upper core support plate in the TMI-2 core ceramic melt phase, but are found concentrated suggests that the damage likely occurred during in metallic phases within the melt. Small, but the B-loop pump transient as a result of high-significant, amounts (2% -15%) of the volatile temperature steam and hydrogen generated by fission products I and Cs remain in the ceramic zircaloy-steam reaction (Kuan et al 1989).…”

Section: Nureg/cr-6160mentioning

confidence: 99%

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Summary of important results and SCDAP/RELAP5 analysis for OECD LOFT experiment LP-FP-2

Coryell¹

1994

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This report summarizes significant technical findings trom the LP-FP-2 Experiment sponsored by the Organization of Economic Cooperation and Development (OECD). It was the second, and final, fission product experiment conducted in the Loss-of-Fluid Test (LOFT) facility at the Idaho National Engineering Laboratory. The overall technical objective of the test was to contribute to the understanding of fuel rod behavior, hydrogen generation, and fission product release, transport, and deposition during a V-sequence accident scenario that resulted in severe core damage. An 11 by 11 test bundle, comprised of 100 prepressurized fuel rods, 11 control rods, and 10 instrumented guide tubes, was surrounded by an insulating shroud and contained in a specially designed central fuel module, that was inserted into the LOFT reactor. The simulated transient was a V-sequence loss-of-coolant accident scenario featuring a pipe break in the low pressure injection system line attached to the hot leg of the LOFT broken loop piping. The transient was terminated by reflood of the reactor vessel when the outer wall shroud temperature reached 1517 K. With sustained fission power and heat from oxidation and metal-water reactions, elevated temperatures resulted in zircaloy melting, fuel liquefaction, material relocation, and the release of hydrogen, aerosols, and fission products. A description and evaluation of the major phenomena, based upon the response of on line instrumentation, analysis of fission product data, postirradiation examination of the fuel bundle, and calculations using the SCDAP/RELAP5 computer code, are presented.

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Section: Nureg/cr-6160mentioning

confidence: 99%

Section: Nureg/cr-6160mentioning

confidence: 99%

Summary of important results and SCDAP/RELAP5 analysis for OECD LOFT experiment LP-FP-2

Coryell¹

1994

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“…If the core has been uncovered, has dried out, and portions of it heated up lO more than 1700 K, then addition of water may, as it did at TMI-2 (starting at minute 174 of the accident), cause massive zirconium burning, hydrogen generation, and the creation of a large porous rubble bed (Broughton et al 1989;Kuan eta!. 1989).…”

Section: Core Fragmentation and Hydrogen Generationmentioning

confidence: 99%

Identification and evaluation of PWR in-vessel severe accident management strategies

Dukelow

Harrison²,

Morgenstern

1992

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This reports documents work performed for the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents.

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Current Status of Core Degradation and Melt Progression in Severe LWR Accidents

Wright¹

2002

Advances in Nuclear Science and Technology

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Thermal Interactions during the Three Mile Island Unit 2 2-B Coolant Pump Transient

Cited by 6 publications

References 2 publications

Summary of important results and SCDAP/RELAP5 analysis for OECD LOFT experiment LP-FP-2

Summary of important results and SCDAP/RELAP5 analysis for OECD LOFT experiment LP-FP-2

Identification and evaluation of PWR in-vessel severe accident management strategies

Current Status of Core Degradation and Melt Progression in Severe LWR Accidents

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