MODELING LOFW IN A PWR USING MELCOR
The Probabilistic Safety Assessment (PSA) is part of a Nuclear Power Plant (NPP) licensing process. It considers the elaboration and updating of probabilistic models that estimate the risk associated to the operation, allowing the risk monitoring from the design to the plant decommissioning, for bot...
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| Format: | Article |
| Language: | English |
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Brazilian Radiation Protection Society (Sociedade Brasileira de Proteção Radiológica, SBPR)
2021-02-01
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| Series: | Brazilian Journal of Radiation Sciences |
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| Online Access: | https://bjrs.org.br/revista/index.php/REVISTA/article/view/1355 |
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| author | Maritza Rodríguez Gual Marcos C. Maturana Nathália N. Araújo Marcelo R. Martins |
| author_facet | Maritza Rodríguez Gual Marcos C. Maturana Nathália N. Araújo Marcelo R. Martins |
| author_sort | Maritza Rodríguez Gual |
| collection | DOAJ |
| description | The Probabilistic Safety Assessment (PSA) is part of a Nuclear Power Plant (NPP) licensing process. It considers the elaboration and updating of probabilistic models that estimate the risk associated to the operation, allowing the risk monitoring from the design to the plant decommissioning, for both operational as regulatory activities. The PSA identifies those components or plant systems whose unavailability contributes significantly to the Core Damage Frequency (CDF) and to the Large Early Release Frequency (LERF) of radioactive material. Based on the PSA Level 1 results for a reference plant under design, the Analysis, Evaluating and Risk Management Laboratory (LabRisco), located in the University of São Paulo (USP), Brazil, started the analytical investigation of severe accident phenomena using the US Nuclear Regulatory Commission (NRC) MELCOR2.2 code – focusing on the qualification of a group of specialists who will subsidize a PSA Level 2 for the same plant. This PSA Level 1 shows that the accident with large CDF contribution is the Loss of Feed Water Accident (LOFW). Therefore, the initial objective of the investigation was to model the progression of severe accidents during a LOFW for the reference Pressurized Water Reactor (PWR) and to analyze the response of the plant under these accident scenarios. During the course of the hypothetical LOFW in the reference plant, hydrogen was generated – by a reaction between the high temperature steam water and the fuel-cladding inside the reactor pressure vessel (RPV) but not representing a serious threat to the RPV integrity. |
| format | Article |
| id | doaj-art-6398d79875154a6fa596f21ffe2d2139 |
| institution | Kabale University |
| issn | 2319-0612 |
| language | English |
| publishDate | 2021-02-01 |
| publisher | Brazilian Radiation Protection Society (Sociedade Brasileira de Proteção Radiológica, SBPR) |
| record_format | Article |
| series | Brazilian Journal of Radiation Sciences |
| spelling | doaj-art-6398d79875154a6fa596f21ffe2d21392025-08-20T03:50:53ZengBrazilian Radiation Protection Society (Sociedade Brasileira de Proteção Radiológica, SBPR)Brazilian Journal of Radiation Sciences2319-06122021-02-0183A (Suppl.)10.15392/bjrs.v8i3A.13551057MODELING LOFW IN A PWR USING MELCORMaritza Rodríguez Gual0Marcos C. Maturana1Nathália N. Araújo2Marcelo R. Martins3University of São Paulo (USP)University of São Paulo (USP)University of São Paulo (USP)University of São Paulo (USP)The Probabilistic Safety Assessment (PSA) is part of a Nuclear Power Plant (NPP) licensing process. It considers the elaboration and updating of probabilistic models that estimate the risk associated to the operation, allowing the risk monitoring from the design to the plant decommissioning, for both operational as regulatory activities. The PSA identifies those components or plant systems whose unavailability contributes significantly to the Core Damage Frequency (CDF) and to the Large Early Release Frequency (LERF) of radioactive material. Based on the PSA Level 1 results for a reference plant under design, the Analysis, Evaluating and Risk Management Laboratory (LabRisco), located in the University of São Paulo (USP), Brazil, started the analytical investigation of severe accident phenomena using the US Nuclear Regulatory Commission (NRC) MELCOR2.2 code – focusing on the qualification of a group of specialists who will subsidize a PSA Level 2 for the same plant. This PSA Level 1 shows that the accident with large CDF contribution is the Loss of Feed Water Accident (LOFW). Therefore, the initial objective of the investigation was to model the progression of severe accidents during a LOFW for the reference Pressurized Water Reactor (PWR) and to analyze the response of the plant under these accident scenarios. During the course of the hypothetical LOFW in the reference plant, hydrogen was generated – by a reaction between the high temperature steam water and the fuel-cladding inside the reactor pressure vessel (RPV) but not representing a serious threat to the RPV integrity.https://bjrs.org.br/revista/index.php/REVISTA/article/view/1355severe accidentmelcorpsa. |
| spellingShingle | Maritza Rodríguez Gual Marcos C. Maturana Nathália N. Araújo Marcelo R. Martins MODELING LOFW IN A PWR USING MELCOR Brazilian Journal of Radiation Sciences severe accident melcor psa. |
| title | MODELING LOFW IN A PWR USING MELCOR |
| title_full | MODELING LOFW IN A PWR USING MELCOR |
| title_fullStr | MODELING LOFW IN A PWR USING MELCOR |
| title_full_unstemmed | MODELING LOFW IN A PWR USING MELCOR |
| title_short | MODELING LOFW IN A PWR USING MELCOR |
| title_sort | modeling lofw in a pwr using melcor |
| topic | severe accident melcor psa. |
| url | https://bjrs.org.br/revista/index.php/REVISTA/article/view/1355 |
| work_keys_str_mv | AT maritzarodriguezgual modelinglofwinapwrusingmelcor AT marcoscmaturana modelinglofwinapwrusingmelcor AT nathalianaraujo modelinglofwinapwrusingmelcor AT marcelormartins modelinglofwinapwrusingmelcor |