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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
Conference on Nuclear Training and Education: A Biennial International Forum (CONTE 2025)
February 3–6, 2025
Amelia Island, FL|Omni Amelia Island Resort
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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How to talk about nuclear
In your career as a professional in the nuclear community, chances are you will, at some point, be asked (or volunteer) to talk to at least one layperson about the technology you know and love. You might even be asked to present to a whole group of nonnuclear folks, perhaps as a pitch to some company tangential to your company’s business. So, without further ado, let me give you some pointers on the best way to approach this important and surprisingly complicated task.
George J. Bohm, Amir N. Nahavandi
Nuclear Science and Engineering | Volume 47 | Number 4 | April 1972 | Pages 391-408
Technical Paper | doi.org/10.13182/NSE72-A22431
Articles are hosted by Taylor and Francis Online.
The dynamic analysis of the reactor internal structure in a typical pressurized-water reactor system, subjected to step, periodic, and seismic excitations, is presented. Employing the finite element approach of structural analysis, the governing differential equations describing the motion of the system are set up and integrated numerically in time. It is shown that the introduction of three types of structural elements, elastic, rigid and pin-joint members with nodes having three degrees of freedom, provides an adequate mathematical model for the solution of reactor structural dynamics problems. A main distinctive feature of this analysis is the application of “elements” global stiffness matrices in place of the standard structural global stiffness matrix. It is shown that this feature reduces the computer storage requirement and running time considerably. An examination of the system dynamic response characteristics indicates that when the clearance between the reactor internal components is relatively small, impact between various components could occur. The magnitude of the impact forces for periodic and seismic excitations is computed. Furthermore, a procedure for the calculation of the upper bound of integration time step is presented which ensures the numerical stability of the solution.