ANS is committed to advancing, fostering, and promoting the development and application of nuclear sciences and technologies to benefit society.
Explore the many uses for nuclear science and its impact on energy, the environment, healthcare, food, and more.
Division Spotlight
Nuclear Nonproliferation Policy
The mission of the Nuclear Nonproliferation Policy Division (NNPD) is to promote the peaceful use of nuclear technology while simultaneously preventing the diversion and misuse of nuclear material and technology through appropriate safeguards and security, and promotion of nuclear nonproliferation policies. To achieve this mission, the objectives of the NNPD are to: Promote policy that discourages the proliferation of nuclear technology and material to inappropriate entities. Provide information to ANS members, the technical community at large, opinion leaders, and decision makers to improve their understanding of nuclear nonproliferation issues. Become a recognized technical resource on nuclear nonproliferation, safeguards, and security issues. Serve as the integration and coordination body for nuclear nonproliferation activities for the ANS. Work cooperatively with other ANS divisions to achieve these objective nonproliferation policies.
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!
Latest Magazine Issues
Mar 2024
Jan 2024
Latest Journal Issues
Nuclear Science and Engineering
April 2024
Nuclear Technology
Fusion Science and Technology
February 2024
Latest News
Can hydrogen be the transportation fuel in an otherwise nuclear economy?
Let’s face it: The global economy should be powered primarily by nuclear power. And it probably will by the end of this century, with a still-significant assist from renewables and hydro. Once nuclear systems are dominant, the costs come down to where gas is now; and when carbon emissions are reduced to a small portion of their present state, it will become obvious that most other sources are only good in niche settings. I mean, why use small modular reactors to load-follow when they can just produce that power instead of buffering it?
David L. Aumiller, Jeffrey W. Lane
Nuclear Science and Engineering | Volume 184 | Number 3 | November 2016 | Pages 463-471
Technical Paper | doi.org/10.13182/NSE16-12
Articles are hosted by Taylor and Francis Online.
COBRA-IE is a three-field subchannel analysis code that was originally based on the COBRA-TF code series. The default interfacial drag model in COBRA-IE has been assessed against a wide range of pressure drop data taken in confined geometries and has been shown to perform very well. The difference in interfacial drag behavior for confined flow paths compared to large open regions where the bubbles are not constrained by the physical geometry of the flow path has been well documented in the open literature. Therefore, a dedicated interfacial drag model for large, open regions has been developed and implemented in COBRA-IE. This alternative interfacial drag model is based on the drift flux formulation and is activated by user input. A combination of the Kataoka-Ishii and the Zuber-Findley drift flux correlations has been implemented in COBRA-IE to calculate the weighted mean drift velocity and distribution parameter. The implementation of the model is described in this paper, and the interface functions to transition between the drift flux and two-fluid formulations are emphasized.
An assessment of the predictive capability of COBRA-IE for the transient level swell phenomena for the experiments performed by General Electric (GE) has been performed. Level swell is an important phenomenon for reactor safety analysis because it impacts water distribution within the reactor vessel during the blowdown phase of the transient as well as the residual inventory available to provide core cooling. The initial assessment of the code using the default interfacial drag modeling package showed an overprediction of the level swell and liquid carryover for the GE experiments, which is indicative of an overprediction of the interfacial drag for these situations. In addition to using the new code to reexamine the GE level swell experiment, assessments of the new model have been performed using the steady-state void fraction data collected in the Beattie-Sugawara and Smith experiments and are presented in this paper.