Evaluation of the Safety Systems in the Next Generation Boiling Water Reactor

The thesis evaluates the safety systems in the next generation boiling water reactor by analyzing the main steam line break loss of coolant accident performed in the Purdue university multi-dimensional test assembly (PUMA). RELAP5 code simulations, both for the PUMA main steam line break (MSLB) case and for the simplified boiling water reactor (SBWR) MSLB case have been utilized to compare with the experiment data. The comparison shows that RELAP5 is capable to perform the safety analysis for SBWR. The comparison also validates the three-level scaling methodology applied to the design of the PUMA facility.

Name thesis:degree_name

Doctor of Philosophy (PhD)

Level thesis:degree_level

Dissertation

Discipline thesis:degree_discipline

Nuclear Engineering

Year

2015

Author dc:creator

• Cheng, Ling

Contributors dc:contributor

• Mamoru Ishii
• Takashi Hibiki
• Martin Lope-de- Bertodano
• Steven Wereley

• main steam line break analysis
• next generation boiling water reactor
• numerical simulation
• one dimensional integral model
• safety systems
• suppression pool mixing

Repository record dc:identifier

https://docs.lib.purdue.edu/open_access_dissertations/1176

OAI identifier oai:identifier

oai:docs.lib.purdue.edu:open_access_dissertations-2392