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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.
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2024 ANS Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
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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
Framatome signs contracts with Sizewell C
French nuclear developer Framatome is slated to deliver key equipment for Sizewell C Ltd.’s two large reactors planned for the United Kingdom’s Suffolk coast.
The agreement, reportedly worth multiple billions of euros, was announced this week and will involve Framatome from the design phase until commissioning. The company also agreed to a long-term fuel supply deal. Framatome is 80.5 percent owned by France’s EDF and 19.5 percent owned by Mitsubishi Heavy Industries.
D. Kotlyar, E. Fridman, E. Shwageraus
Nuclear Science and Engineering | Volume 179 | Number 3 | March 2015 | Pages 274-284
Technical Paper | doi.org/10.13182/NSE14-59
Articles are hosted by Taylor and Francis Online.
Allowing Monte Carlo (MC) codes to perform fuel cycle calculations requires coupling to a point depletion solver. In order to perform depletion calculations, one-group cross sections must be provided in advance. This paper focuses on generating accurate one-group cross-section values using Monte Carlo transport codes. The proposed method is an alternative to the conventional direct reaction rate tally approach, which requires substantial computational effort. The method presented here is based on the multigroup approach, in which pregenerated multigroup sets are collapsed with MC calculated flux. In our previous studies, we showed that generating accurate one-group cross sections requires their tabulation against the background cross section (σ0) to account for the self-shielding effect in the unresolved resonance energy range.However, in previous studies, the model used for the calculation of σ0 was simplified by relying on user-specified Bell and Dancoff factors. This work demonstrates that the one-group cross-section values calculated under the previous simplified model assumptions may not always agree with the directly tallied values. More specifically, the assumption is not universally applicable to the analysis of reactor systems with different neutron spectra and may be inaccurate when the number of energy groups is reduced (i.e., from tens of thousands to hundreds of groups). Therefore, the original background cross-section model was extended by implicitly accounting for the Dancoff and Bell factors. The method developed here reconstructs the correct value of σ0 by utilizing statistical data generated within the MC transport calculation by default. The proposed method was implemented in the BGCore code system. The one-group cross-section values generated by BGCore were compared with those tallied directly from the MCNP code. Very good agreement in the one-group cross-section values was observed. The method does not carry any additional computational burden, and it is universally applicable to the analysis of thermal as well as fast reactor systems. Adopting this multigroup methodology, which accounts for self-shielding, allows generation of highly accurate cross sections even if the number of energy groups is significantly reduced (to hundreds versus tens of thousands of groups). This reduction considerably improves the computational efficiency, which makes the analysis of large-scale reactor problems feasible.