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Evaluation of the Safety Systems in ...

Cheng, Ling.

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

紀錄類型:	書目-電子資源 : Monograph/item
正題名/作者:	Evaluation of the Safety Systems in the Next Generation Boiling Water Reactor./
作者:	Cheng, Ling.
面頁冊數:	150 p.
附註:	Source: Dissertation Abstracts International, Volume: 77-04(E), Section: B.
Contained By:	Dissertation Abstracts International77-04B(E).
標題:	Nuclear engineering. -
電子資源:	http://pqdd.sinica.edu.tw/twdaoapp/servlet/advanced?query=3736756
ISBN:	9781339256917

Evaluation of the Safety Systems in the Next Generation Boiling Water Reactor.
Cheng, Ling.

Evaluation of the Safety Systems in the Next Generation Boiling Water Reactor. - 150 p.

Source: Dissertation Abstracts International, Volume: 77-04(E), Section: B.

Thesis (Ph.D.)--Purdue University, 2015.

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.

ISBN: 9781339256917Subjects--Topical Terms:

595435
Nuclear engineering.

Evaluation of the Safety Systems in the Next Generation Boiling Water Reactor.
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500 $a Source: Dissertation Abstracts International, Volume: 77-04(E), Section: B.
500 $a Adviser: Mamoru Ishii.
502 $a Thesis (Ph.D.)--Purdue University, 2015.
520 $a 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.
520 $a The PUMA suppression pool mixing and condensation test data have been studied to give the detailed understanding on this important local phenomenon. A simple one dimensional integral model, which can reasonably simulate the mixing process inside suppression pool have been developed and the comparison between the model prediction and the experiment data demonstrates the model can be utilized for analyzing the suppression pool mixing process.
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