Results of EPRI/ANL DCH investigations and model development
The results of a series of five experiments are described addressing the severity and mitigation of direct containment heating. The tests were performed in a 1:30 linear scale mockup of the Zion PWR containment system using a reactor-material corium melt consisting of 60% UO/sub 2/, 16% ZrO/sub 2/,...
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Format: | Article in Journal/Newspaper |
Language: | English |
Published: |
Argonne National Laboratory
1988
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Online Access: | https://digital.library.unt.edu/ark:/67531/metadc1210917/ |
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author | Spencer, B. W. Sienicki, J. J. Sehgal, B. R. Merilo, M. |
author_facet | Spencer, B. W. Sienicki, J. J. Sehgal, B. R. Merilo, M. |
author_sort | Spencer, B. W. |
collection | University of North Texas: UNT Digital Library |
description | The results of a series of five experiments are described addressing the severity and mitigation of direct containment heating. The tests were performed in a 1:30 linear scale mockup of the Zion PWR containment system using a reactor-material corium melt consisting of 60% UO/sub 2/, 16% ZrO/sub 2/, 24% SSt at nominally 2800C initial temperature. A ''worst-case'' type test involving unimpeded corium dispersal through an air atmosphere in a closed vessel produced an atmosphere heatup of 323K, equivalent to a DCH efficiency of 62%. With the addition of structural features which impeded the corium dispersal, representative of dispersal pathway features at Zion, the DCH efficiency was reduced to 1--5%. (This important result is scale dependent and requires larger scale tests such as the SURTSEY program at SNL plus mechanistic modeling for application to the reactor system.) With the addition of water in the cavity region, there was no measurable heatup of the atmosphere. This was attributable to the vigorous codispersal of water with corium which prevented the temperature of the atmosphere from significantly exceeding T/sub sat/. In this case the DCH load was replaced by the more benign ''steam spike'' from corium quench. Significant oxidation of the corium constituents occurred in the tests, adding chemical energy to the system and producing hydrogen. Overall, the results suggest that with consideration of realistic, plant specific, mitigating features, DCH may be no worse and possibly far less severe than the previously examined steam spike. Implications for accident management are addressed. 17 refs., 7 figs., 4 tabs. |
format | Article in Journal/Newspaper |
genre | Surtsey |
genre_facet | Surtsey |
geographic | Surtsey |
geographic_facet | Surtsey |
id | ftunivnotexas:info:ark/67531/metadc1210917 |
institution | Open Polar |
language | English |
long_lat | ENVELOPE(-20.608,-20.608,63.301,63.301) |
op_collection_id | ftunivnotexas |
op_relation | other: DE89007372 rep-no: CONF-881014-20 grantno: W-31109-ENG-38 osti: 6519463 https://digital.library.unt.edu/ark:/67531/metadc1210917/ ark: ark:/67531/metadc1210917 |
op_source | International European Nuclear Society/American Nuclear Society meeting on thermal reactor safety, Avignon, France, 2 Oct 1988 |
publishDate | 1988 |
publisher | Argonne National Laboratory |
record_format | openpolar |
spelling | ftunivnotexas:info:ark/67531/metadc1210917 2025-01-17T01:01:26+00:00 Results of EPRI/ANL DCH investigations and model development Spencer, B. W. Sienicki, J. J. Sehgal, B. R. Merilo, M. 1988-01-01 14 pages Text https://digital.library.unt.edu/ark:/67531/metadc1210917/ English eng Argonne National Laboratory other: DE89007372 rep-no: CONF-881014-20 grantno: W-31109-ENG-38 osti: 6519463 https://digital.library.unt.edu/ark:/67531/metadc1210917/ ark: ark:/67531/metadc1210917 International European Nuclear Society/American Nuclear Society meeting on thermal reactor safety, Avignon, France, 2 Oct 1988 Heat Transfer Elements Chemical Reactions Enriched Uranium Reactors Oxidation Hydraulics 22 General Studies Of Nuclear Reactors Computer Codes Nonmetals 21 Specific Nuclear Reactors And Associated Plants Fluid Mechanics Heating Power Reactors S Codes Water Cooled Reactors Containment Structural Models Accidents Thermal Reactors Reactors Water Moderated Reactors 220900* -- Nuclear Reactor Technology-- Reactor Safety M Codes Mechanics Testing Reactor Safety 210200 -- Power Reactors Nonbreeding Light-Water Moderated Nonboiling Water Cooled Pwr Type Reactors Blowdown Energy Transfer H Codes Hydrogen Loss Of Coolant Oxygen Compounds Corium Scale Models Zion-1 Reactor Hydrogen Compounds Zion-2 Reactor Water Safety C Codes Reactor Accidents Article 1988 ftunivnotexas 2020-05-09T22:08:19Z The results of a series of five experiments are described addressing the severity and mitigation of direct containment heating. The tests were performed in a 1:30 linear scale mockup of the Zion PWR containment system using a reactor-material corium melt consisting of 60% UO/sub 2/, 16% ZrO/sub 2/, 24% SSt at nominally 2800C initial temperature. A ''worst-case'' type test involving unimpeded corium dispersal through an air atmosphere in a closed vessel produced an atmosphere heatup of 323K, equivalent to a DCH efficiency of 62%. With the addition of structural features which impeded the corium dispersal, representative of dispersal pathway features at Zion, the DCH efficiency was reduced to 1--5%. (This important result is scale dependent and requires larger scale tests such as the SURTSEY program at SNL plus mechanistic modeling for application to the reactor system.) With the addition of water in the cavity region, there was no measurable heatup of the atmosphere. This was attributable to the vigorous codispersal of water with corium which prevented the temperature of the atmosphere from significantly exceeding T/sub sat/. In this case the DCH load was replaced by the more benign ''steam spike'' from corium quench. Significant oxidation of the corium constituents occurred in the tests, adding chemical energy to the system and producing hydrogen. Overall, the results suggest that with consideration of realistic, plant specific, mitigating features, DCH may be no worse and possibly far less severe than the previously examined steam spike. Implications for accident management are addressed. 17 refs., 7 figs., 4 tabs. Article in Journal/Newspaper Surtsey University of North Texas: UNT Digital Library Surtsey ENVELOPE(-20.608,-20.608,63.301,63.301) |
spellingShingle | Heat Transfer Elements Chemical Reactions Enriched Uranium Reactors Oxidation Hydraulics 22 General Studies Of Nuclear Reactors Computer Codes Nonmetals 21 Specific Nuclear Reactors And Associated Plants Fluid Mechanics Heating Power Reactors S Codes Water Cooled Reactors Containment Structural Models Accidents Thermal Reactors Reactors Water Moderated Reactors 220900* -- Nuclear Reactor Technology-- Reactor Safety M Codes Mechanics Testing Reactor Safety 210200 -- Power Reactors Nonbreeding Light-Water Moderated Nonboiling Water Cooled Pwr Type Reactors Blowdown Energy Transfer H Codes Hydrogen Loss Of Coolant Oxygen Compounds Corium Scale Models Zion-1 Reactor Hydrogen Compounds Zion-2 Reactor Water Safety C Codes Reactor Accidents Spencer, B. W. Sienicki, J. J. Sehgal, B. R. Merilo, M. Results of EPRI/ANL DCH investigations and model development |
title | Results of EPRI/ANL DCH investigations and model development |
title_full | Results of EPRI/ANL DCH investigations and model development |
title_fullStr | Results of EPRI/ANL DCH investigations and model development |
title_full_unstemmed | Results of EPRI/ANL DCH investigations and model development |
title_short | Results of EPRI/ANL DCH investigations and model development |
title_sort | results of epri/anl dch investigations and model development |
topic | Heat Transfer Elements Chemical Reactions Enriched Uranium Reactors Oxidation Hydraulics 22 General Studies Of Nuclear Reactors Computer Codes Nonmetals 21 Specific Nuclear Reactors And Associated Plants Fluid Mechanics Heating Power Reactors S Codes Water Cooled Reactors Containment Structural Models Accidents Thermal Reactors Reactors Water Moderated Reactors 220900* -- Nuclear Reactor Technology-- Reactor Safety M Codes Mechanics Testing Reactor Safety 210200 -- Power Reactors Nonbreeding Light-Water Moderated Nonboiling Water Cooled Pwr Type Reactors Blowdown Energy Transfer H Codes Hydrogen Loss Of Coolant Oxygen Compounds Corium Scale Models Zion-1 Reactor Hydrogen Compounds Zion-2 Reactor Water Safety C Codes Reactor Accidents |
topic_facet | Heat Transfer Elements Chemical Reactions Enriched Uranium Reactors Oxidation Hydraulics 22 General Studies Of Nuclear Reactors Computer Codes Nonmetals 21 Specific Nuclear Reactors And Associated Plants Fluid Mechanics Heating Power Reactors S Codes Water Cooled Reactors Containment Structural Models Accidents Thermal Reactors Reactors Water Moderated Reactors 220900* -- Nuclear Reactor Technology-- Reactor Safety M Codes Mechanics Testing Reactor Safety 210200 -- Power Reactors Nonbreeding Light-Water Moderated Nonboiling Water Cooled Pwr Type Reactors Blowdown Energy Transfer H Codes Hydrogen Loss Of Coolant Oxygen Compounds Corium Scale Models Zion-1 Reactor Hydrogen Compounds Zion-2 Reactor Water Safety C Codes Reactor Accidents |
url | https://digital.library.unt.edu/ark:/67531/metadc1210917/ |