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
Decommissioning & Environmental Sciences
The mission of the Decommissioning and Environmental Sciences (DES) Division is to promote the development and use of those skills and technologies associated with the use of nuclear energy and the optimal management and stewardship of the environment, sustainable development, decommissioning, remediation, reutilization, and long-term surveillance and maintenance of nuclear-related installations, and sites. The target audience for this effort is the membership of the Division, the Society, and the public at large.
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!
Latest Magazine Issues
Dec 2024
Jul 2024
Latest Journal Issues
Nuclear Science and Engineering
December 2024
Nuclear Technology
Fusion Science and Technology
November 2024
Latest News
Bipartisan Nuclear REFUEL Act introduced in the U.S. House
To streamline the licensing requirements for nuclear fuel recycling facilities and help increase investment in nuclear energy in the United States, U.S. Reps. Bob Latta (R., Ohio) and Scott Peters (D., Calif.) have introduced the bipartisan Nuclear REFUEL Act in the House of Representatives.
The bill, introduced on December 6, would amend the definition of “production facility” in the Atomic Energy Act, clarifying that a reprocessing facility producing uranium-transuranic mixed fuel would be licensed only under 10 CFR Part 70. According to the lawmakers, this single-step licensing process would significantly streamline the licensing requirements for fuel recycling facilities.
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.