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Division Spotlight
Nuclear Criticality Safety
NCSD provides communication among nuclear criticality safety professionals through the development of standards, the evolution of training methods and materials, the presentation of technical data and procedures, and the creation of specialty publications. In these ways, the division furthers the exchange of technical information on nuclear criticality safety with the ultimate goal of promoting the safe handling of fissionable materials outside reactors.
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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February 2025
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Latest News
Reboot: Nuclear needs a success . . . anywhere
The media have gleefully resurrected the language of a past nuclear renaissance. Beyond the hype and PR, many people in the nuclear community are taking a more measured view of conditions that could lead to new construction: data center demand, the proliferation of new reactor designs and start-ups, and the sudden ascendance of nuclear energy as the power source everyone wants—or wants to talk about.
Once built, large nuclear reactors can provide clean power for at least 80 years—outlasting 10 to 20 presidential administrations. Smaller reactors can provide heat and power outputs tailored to an end user’s needs. With all the new attention, are we any closer to getting past persistent supply chain and workforce issues and building these new plants? And what will the election of Donald Trump to a second term as president mean for nuclear?
As usual, there are more questions than answers, and most come down to money. Several developers are engaging with the Nuclear Regulatory Commission or have already applied for a license, certification, or permit. But designs without paying customers won’t get built. So where are the customers, and what will it take for them to commit?
Hsiang-Shou Cheng, David J. Diamond
Nuclear Technology | Volume 56 | Number 1 | January 1982 | Pages 40-54
Technical Paper | Fission Reactor | doi.org/10.13182/NT82-A32879
Articles are hosted by Taylor and Francis Online.
The center rod drop accident was calculated for a boiling water reactor using the two-dimensional (R,Z) core dynamics code BNL-TWIGL. Analysts frequently neglect moderator feedback under the assumption that it leads to conservative results. The present study shows that the peak of the power burst and peak fuel enthalpy can indeed be reduced by a factor of 2 or more by including this effect. The magnitude of the effect depends on reactor conditions. Moderator feedback is particularly important when there are voids in the core initially (i.e., at power conditions) or when the core is near saturation condition. When the reactor is initially at zero power and considerably subcooled, moderator feedback will influence the power peak by <10% but will have a much larger effect on the peak fuel enthalpy, which occurs later in time. The moderator feedback is the result of heat conducted from the fuel rod and direct energy deposition. At power conditions, the time constant for heat conduction is small and this is the primary mechanism for changing the steam void content during the accident. At zero power, the initial thermal constant is very large and, hence, any generation of voids at short times is due to direct energy deposition in the moderator. The effect of a different initial power level, flow rate, and inlet sub cooling, as well as the effect of delayed neutron fraction, rod drop speed, and accident rod worth, was calculated. In all cases, with moderator feedback accounted for, the maximum fuel enthalpy during the accident is well below presently established limits. Accident consequences are insensitive to the delayed neutron fraction and rod drop velocity. The parameters of most significance are inlet subcooling and accident rod worth. Most of the analysis used a fixed inlet flow and core pressure. A plant transient calculation was run to see how these parameters varied. The result was fed back into a bounding core calculation, which then showed that the change in pressure and flow increases the peak fuel enthalpy but not to an appreciable extent.