A condensed review of nuclear reactor thermal-hydraulic computer codes for two-phase flow analysis
A review is made of the computer codes developed in the U.S. for thermal-hydraulic analysis of nuclear reactors. The intention of this review is to compare these codes on the basis of their numerical method and physical models with particular attention to the two-phase flow and heat transfer...
| Main Authors: | , |
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| Format: | Technical Report |
| Language: | en_US |
| Published: |
MIT Energy Laboratory
2006
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| Subjects: | |
| Online Access: | http://hdl.handle.net/1721.1/35164 |
| _version_ | 1811095566275313664 |
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| author | Kazimi, Mujid S. Massoud, Mahmoud |
| author_facet | Kazimi, Mujid S. Massoud, Mahmoud |
| author_sort | Kazimi, Mujid S. |
| collection | MIT |
| description | A review is made of the computer codes developed in the
U.S. for thermal-hydraulic analysis of nuclear reactors. The
intention of this review is to compare these codes on the
basis of their numerical method and physical models with
particular attention to the two-phase flow and heat transfer
characteristics. A chronology of the most documented codes
such as COBRA and RELAP is given. The features of the recent
codes as RETRAN, TRAC and THERMIT are also reviewed. The
range of application as well as limitations of the various
codes are discussed. |
| first_indexed | 2024-09-23T16:19:31Z |
| format | Technical Report |
| id | mit-1721.1/35164 |
| institution | Massachusetts Institute of Technology |
| language | en_US |
| last_indexed | 2024-09-23T16:19:31Z |
| publishDate | 2006 |
| publisher | MIT Energy Laboratory |
| record_format | dspace |
| spelling | mit-1721.1/351642019-04-12T08:38:23Z A condensed review of nuclear reactor thermal-hydraulic computer codes for two-phase flow analysis Kazimi, Mujid S. Massoud, Mahmoud Nuclear fuel elements |x Computer programs. Two-phase flow. A review is made of the computer codes developed in the U.S. for thermal-hydraulic analysis of nuclear reactors. The intention of this review is to compare these codes on the basis of their numerical method and physical models with particular attention to the two-phase flow and heat transfer characteristics. A chronology of the most documented codes such as COBRA and RELAP is given. The features of the recent codes as RETRAN, TRAC and THERMIT are also reviewed. The range of application as well as limitations of the various codes are discussed. Sponsored by Boston Edison Company and others under MIT Energy Laboratory Electric Utility Program. 2006-12-19T16:00:57Z 2006-12-19T16:00:57Z 1980-02 Technical Report 06526448 http://hdl.handle.net/1721.1/35164 en_US MIT-EL 79-018 5295128 bytes application/pdf application/pdf MIT Energy Laboratory |
| spellingShingle | Nuclear fuel elements |x Computer programs. Two-phase flow. Kazimi, Mujid S. Massoud, Mahmoud A condensed review of nuclear reactor thermal-hydraulic computer codes for two-phase flow analysis |
| title | A condensed review of nuclear reactor thermal-hydraulic computer codes for two-phase flow analysis |
| title_full | A condensed review of nuclear reactor thermal-hydraulic computer codes for two-phase flow analysis |
| title_fullStr | A condensed review of nuclear reactor thermal-hydraulic computer codes for two-phase flow analysis |
| title_full_unstemmed | A condensed review of nuclear reactor thermal-hydraulic computer codes for two-phase flow analysis |
| title_short | A condensed review of nuclear reactor thermal-hydraulic computer codes for two-phase flow analysis |
| title_sort | condensed review of nuclear reactor thermal hydraulic computer codes for two phase flow analysis |
| topic | Nuclear fuel elements |x Computer programs. Two-phase flow. |
| url | http://hdl.handle.net/1721.1/35164 |
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