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Division Spotlight
Reactor Physics
The division's objectives are to promote the advancement of knowledge and understanding of the fundamental physical phenomena characterizing nuclear reactors and other nuclear systems. The division encourages research and disseminates information through meetings and publications. Areas of technical interest include nuclear data, particle interactions and transport, reactor and nuclear systems analysis, methods, design, validation and operating experience and standards. The Wigner Award heads the awards program.
Meeting Spotlight
International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2025)
April 27–30, 2025
Denver, CO|The Westin Denver Downtown
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
Argonne’s METL gears up to test more sodium fast reactor components
Argonne National Laboratory has successfully swapped out an aging cold trap in the sodium test loop called METL (Mechanisms Engineering Test Loop), the Department of Energy announced April 23. The upgrade is the first of its kind in the United States in more than 30 years, according to the DOE, and will help test components and operations for the sodium-cooled fast reactors being developed now.
Joseph W. Nielsen, Michael A. Reicheberger, Bryon J. Curnutt, Dong O. Choe, Irina Glagolenko, Jody Henley
Nuclear Technology | Volume 208 | Number 11 | November 2022 | Pages 1704-1720
Technical Paper | doi.org/10.1080/00295450.2022.2067448
Articles are hosted by Taylor and Francis Online.
One of the Advanced Test Reactor’s (ATR’s) functions is to irradiate and qualify nuclear fuels and materials. Due to the large number of experiment or test positions, the cost, and the limited number of vessel penetrations for instrumentation, in-core instrumentation for most experiments is not feasible. In such instances, modeling of experiment conditions using high-fidelity neutron transport codes can quantify such conditions as fission power density and fissile material burnup during irradiation. Validation of fissile material burnup can only be performed during post-irradiation examination, which typically occurs months—or even years—following irradiation. In most experiments, fission power density and fissile material burnup are directly proportional to the thermal neutron flux in the ATR. Additionally, fast neutrons are born from fission in the ATR core, affording a validation of power distribution within the reactor’s experiment locations. During each irradiation cycle, flux wires installed throughout the ATR can be used to validate computational models and determine an adjusted neutron flux for many of the experiment positions. The flux wires are installed as requested by the experiment sponsors in several of the ATR flux traps and consist of cobalt-aluminum alloy and nickel wires. Both kinds of wire enable measurements of the thermal and fast neutron flux in each experiment position. This paper presents the protocol for validating computational models for experiments using flux wires installed in the experiment positions, as well as the results for flux wires placed in the ATR safety rod guide tubes. The best estimate is typically referred to as the adjusted neutron flux. The calculated unadjusted neutron flux is referred to as the a priori neutron flux. The methods presented here provide the adjusted neutron flux, given both the measured and a prior fluxes. The adjusted flux is compared to the a priori flux to provide a bias in the calculated results and the adjusted results. Two model types are evaluated; an eigenvalue case and fixed-source case. Both conditions demonstrate relatively good agreement. The uncertainty for the adjusted flux ranges from 5% to 6% for all three energy ranges. For the eigenvalue case, the bias between the a priori and the adjusted neutron flux is within the statistical uncertainty in all but two wire pairs. For the fixed-source model, four wire pairs are outside of the uncertainty of the adjusted flux. The bias between the a priori and adjusted fast neutron flux is outside of the statistical range for four wires in the eigenvalue case and nine wires in the fixed-source model. As the differences are not contained to one flux trap, it can be assumed that the biases in the calculated models are attributed to localized effects in modeling. An additional evaluation was performed for the ATF-1 experiment in the ATR “I” positions. The differences between the adjusted and a priori are more pronounced in two of the test positions, indicating that additional model evaluation is needed, in particular in the region near the boundary of the ATR model. It is also noted that the eigenvalue model provides slightly better results in the flux trap positions. The fixed-source model is more computationally efficient though produces less accurate results; the differences in some cases are negligible. The work documented in this paper provides a methodology that extends the validation protocol established at the ATR for flux measurements to validate computational models with limited measurement capability during a cycle.