Failure Assessment Methodologies for Pressure-Retaining Components under Severe Accident Loading
During postulated high-pressure core melt accident scenarios, temperature values of more than 800°C can be reached in the reactor coolant line and the surge line of a pressurised water reactor (PWR), before the bottom of the reactor pressure vessel experiences a significant temperature increase due...
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| Format: | Article |
| Language: | English |
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Wiley
2012-01-01
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| Series: | Science and Technology of Nuclear Installations |
| Online Access: | http://dx.doi.org/10.1155/2012/487371 |
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| author | J. Arndt H. Grebner J. Sievers |
| author_facet | J. Arndt H. Grebner J. Sievers |
| author_sort | J. Arndt |
| collection | DOAJ |
| description | During postulated high-pressure core melt accident scenarios, temperature values of more than 800°C can be reached in the reactor coolant line and the surge line of a pressurised water reactor (PWR), before the bottom of the reactor pressure vessel experiences a significant temperature increase due to core melting. For the assessment of components of the primary cooling circuit, two methods are used by GRS. One is the simplified method ASTOR (approximated structural time of rupture). This method employs the hypothesis of linear damage accumulation for modeling damage progression. A failure time surface which is generated by structural finite element (FE) analysis of varying pressure and temperature loads serves as a basis for estimations of failure times. The second method is to perform thermohydraulic and structure mechanic calculations for the accident scenario under consideration using complex calculation models. The paper shortly describes both assessment procedures. Validation of the ASTOR method concerning a large-scale test on a pipe section with geometric properties similar to a reactor coolant line is presented as well as severe accident scenarios investigated with both methods. |
| format | Article |
| id | doaj-art-7670b4f062cb446697e08825d661eb4e |
| institution | Kabale University |
| issn | 1687-6075 1687-6083 |
| language | English |
| publishDate | 2012-01-01 |
| publisher | Wiley |
| record_format | Article |
| series | Science and Technology of Nuclear Installations |
| spelling | doaj-art-7670b4f062cb446697e08825d661eb4e2025-02-03T01:32:17ZengWileyScience and Technology of Nuclear Installations1687-60751687-60832012-01-01201210.1155/2012/487371487371Failure Assessment Methodologies for Pressure-Retaining Components under Severe Accident LoadingJ. Arndt0H. Grebner1J. Sievers2Reactor Safety Research–Barrier Effectiveness, Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH, Schwertnergasse 1, 50667 Cologne, GermanyReactor Safety Research–Barrier Effectiveness, Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH, Schwertnergasse 1, 50667 Cologne, GermanyReactor Safety Research–Barrier Effectiveness, Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH, Schwertnergasse 1, 50667 Cologne, GermanyDuring postulated high-pressure core melt accident scenarios, temperature values of more than 800°C can be reached in the reactor coolant line and the surge line of a pressurised water reactor (PWR), before the bottom of the reactor pressure vessel experiences a significant temperature increase due to core melting. For the assessment of components of the primary cooling circuit, two methods are used by GRS. One is the simplified method ASTOR (approximated structural time of rupture). This method employs the hypothesis of linear damage accumulation for modeling damage progression. A failure time surface which is generated by structural finite element (FE) analysis of varying pressure and temperature loads serves as a basis for estimations of failure times. The second method is to perform thermohydraulic and structure mechanic calculations for the accident scenario under consideration using complex calculation models. The paper shortly describes both assessment procedures. Validation of the ASTOR method concerning a large-scale test on a pipe section with geometric properties similar to a reactor coolant line is presented as well as severe accident scenarios investigated with both methods.http://dx.doi.org/10.1155/2012/487371 |
| spellingShingle | J. Arndt H. Grebner J. Sievers Failure Assessment Methodologies for Pressure-Retaining Components under Severe Accident Loading Science and Technology of Nuclear Installations |
| title | Failure Assessment Methodologies for Pressure-Retaining Components under Severe Accident Loading |
| title_full | Failure Assessment Methodologies for Pressure-Retaining Components under Severe Accident Loading |
| title_fullStr | Failure Assessment Methodologies for Pressure-Retaining Components under Severe Accident Loading |
| title_full_unstemmed | Failure Assessment Methodologies for Pressure-Retaining Components under Severe Accident Loading |
| title_short | Failure Assessment Methodologies for Pressure-Retaining Components under Severe Accident Loading |
| title_sort | failure assessment methodologies for pressure retaining components under severe accident loading |
| url | http://dx.doi.org/10.1155/2012/487371 |
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