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Division Spotlight
Thermal Hydraulics
The division provides a forum for focused technical dialogue on thermal hydraulic technology in the nuclear industry. Specifically, this will include heat transfer and fluid mechanics involved in the utilization of nuclear energy. It is intended to attract the highest quality of theoretical and experimental work to ANS, including research on basic phenomena and application to nuclear system design.
Meeting Spotlight
2024 ANS Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
Standards Program
The Standards Committee is responsible for the development and maintenance of voluntary consensus standards that address the design, analysis, and operation of components, systems, and facilities related to the application of nuclear science and technology. Find out What’s New, check out the Standards Store, or Get Involved today!
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Latest News
Why should safeguards by design be a global effort?
Jeremy Whitlock
I can’t think of a more exciting time to be working in nuclear, with the diversity of advanced reactor development and increasing global support for nuclear in sustainable energy planning. But we can’t lose sight of the need to plan for efficient international safeguards at the same time.
Global nuclear deployment has been underpinned since 1970 by the Treaty on the Non-Proliferation of Nuclear Weapons (NPT), making it a key customer requirement for governments to demonstrate unequivocally that the technology is not being misused for weapons development.
The International Atomic Energy Agency (IAEA) has helped verify this commitment for more than 50 years, but it has never safeguarded many of the advanced reactors (and related fuel cycle processes) being developed today.
Wei-Nian Su, Shih-Jen Wang, I-Ming Huang, Show-Chyuan Chiang
Nuclear Technology | Volume 155 | Number 3 | September 2006 | Pages 253-264
Technical Paper | Fission Reactors | doi.org/10.13182/NT06-A3760
Articles are hosted by Taylor and Francis Online.
Containment flooding is an important strategy for severe accident management of a conventional boiling water reactor (BWR) system. The execution of containment flooding requires information about the water level in the primary containment. However, there is no instrument to measure the drywell water level for most Mark-III systems. Furthermore, because of the design feature of the Mark-III containment, the water level in the containment does not necessarily guarantee that there is an equivalent water level in the drywell. Therefore, the development of a drywell water level computational aid becomes very useful. The purpose of this work is to develop and validate the drywell water level computational aid and to investigate the implementation of the proposed computational aid on the containment flooding strategy of a Mark-III system. The Kuosheng nuclear power plant (NPP) is a typical BWR-6 NPP with Mark-III containment, and the Severe Accident Management Guideline (SAMG) of the Kuosheng NPP has been developed based on the BWR Owners' Group Emergency Procedure Guidelines and Severe Accident Guidelines, Revision 2. Therefore, the Kuosheng NPP is selected as the plant for study, and the MAAP4 code is chosen as the tool for analysis.