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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 Winter Conference and Expo
November 17–21, 2024
Orlando, FL|Renaissance Orlando at SeaWorld
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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Nuclear new build procurement considerations
It may seem counterintuitive, but the best time to enhance the ability to support operations and maintenance for a new plant is before construction starts. This is one of many lessons learned by the currently operating nuclear fleet. As construction and startup of many nuclear facilities was completed, it quickly became evident that the ability to efficiently support operations and maintenance was limited. Most of the information necessary to establish and manage procurement of spare and replacement items, maintenance, and configuration of the facilities was unavailable and had to be gathered on a case-by-case, “on-demand” basis. Absence of necessary information and the associated challenges resulted in the need for staff augmentation and multiyear-long projects to develop equipment bills of material and maintenance programs and to perform technical evaluations for the huge quantities of spare and replacement items being requested.
S. Al Issa, M. Murase, A. Tomiyama, K. Hayashi, R. Macián-Juan
Nuclear Science and Engineering | Volume 193 | Number 1 | January-February 2019 | Pages 147-159
Technical Paper – Selected papers from NURETH 2017 | doi.org/10.1080/00295639.2018.1489627
Articles are hosted by Taylor and Francis Online.
Countercurrent flow limitation (CCFL) in a pressurized water reactor hot-leg pipe geometry with a 190-mm pipe diameter was investigated experimentally and numerically at the COLLIDER test facility of the Technical University Munich in the past 3 years. This paper summarizes the most important CCFL findings learned from the COLLIDER test facility and tries to explain the reasons for obtaining different descriptions, results, and conclusions at different CCFL experimental investigations. The factors that can affect CCFL experimental results are explained in detail including some scale effects. The necessary preconditions to compare two sets of data from different CCFL experimental investigations are discussed in detail. The difference among CCFL-related limits/curves is clarified taking data at the COLLIDER as an example. The limits included the limit of the transition from a supercritical into a subcritical flow (SSTL); the onset of CCFL limit (iCCFL) inside the hot-leg pipe; the onset of CCFL limit (eCCFL) at the entrance of the steam generator; the deflooding limit (CCFLd); the CCFL characteristics curve (CCFLch), which predicts the water delivery rate after the onset of iCCFL; and the onset of hysteresis limit. It will be shown that among these limits only SSTL, CCFLch, and eCCFL are original limits while the rest are derivatives of them. In particular, it will be shown that the iCCFL limit is a combination of the SSTL and CCFLch limits. The effect of scale upon the eCCFL’s mechanism (whether a water accumulation or droplet entrainment at the entrance to the steam generator) is clarified via a comparison to a 50-mm CCFL facility at Kobe University. This paper tests the scalability of interface distribution at quasi-stationary conditions (i.e., points along the CCFLch curve) via a comparison of time-averaged interface distributions obtained at similar inlet conditions ( at the COLLIDER 190-mm and Kobe 50-mm channels. The comparison will show that interface distributions (which are directly linked to the pressure drop and interfacial momentum transfer) cannot be scaled at the bend/riser/entrance region because of the influence of the channel diameter upon occurring CCFL mechanism. Meanwhile, the water level gradient can be similar at the horizontal part, but not the relative water depth.