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
Fuel Cycle & Waste Management
Devoted to all aspects of the nuclear fuel cycle including waste management, worldwide. Division specific areas of interest and involvement include uranium conversion and enrichment; fuel fabrication, management (in-core and ex-core) and recycle; transportation; safeguards; high-level, low-level and mixed waste management and disposal; public policy and program management; decontamination and decommissioning environmental restoration; and excess weapons materials disposition.
Meeting Spotlight
2024 ANS Winter Conference and Expo
November 17–21, 2024
Orlando, FL|Renaissance Orlando at SeaWorld
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
Aug 2024
Jan 2024
Latest Journal Issues
Nuclear Science and Engineering
October 2024
Nuclear Technology
Fusion Science and Technology
August 2024
Latest News
New laws offer nuclear industry incentives for existing power plant uprates
This year, the U.S. nuclear industry received a much-needed economic boost that could help preserve operating nuclear power plants and incentivize upgrades that extend their lifespan and power output.
Signed into law in 2022, the Inflation Reduction Act offers production tax credits (PTCs) for existing nuclear power plants and either PTCs or investment tax credits (ITCs) for new carbon-free generation. These credits could make power uprates—increasing the maximum power level at which a commercial plant may operate—a much more appealing option for utilities.
Donald Bogart
Nuclear Science and Engineering | Volume 123 | Number 2 | June 1996 | Pages 228-237
Technical Paper | doi.org/10.13182/NSE96-A24185
Articles are hosted by Taylor and Francis Online.
Although resonance neutron captures for 238U in water-moderated lattices are known to occur near moderator-fuel interfaces, the sharply attenuated spatial captures here have not been calculated by multigroup transport or Monte Carlo methods. Advances in computer speed and capacity have restored interest in applying Monte Carlo methods to evaluate spatial resonance captures in fueled lattices. Recently published studies have placed complete reliance on the ostensible precision of the Monte Carlo approach without auxiliary confirmation that resonance processes were followed adequately or that the Monte Carlo method was applied appropriately. Other methods of analysis that have evolved from early resonance integral theory have provided a basis for an alternative approach to determine radial resonance captures in fuel rods. A generalized method has been formulated and confirmed by comparison with published experiments of high spatial resolution for radial resonance captures in metallic uranium rods. The same analytical method has been applied to uranium-oxide fuels. The generalized method defined a spatial effective resonance cross section that is a continuous function of distance from the moderator-fuel interface and enables direct calculation of precise radial resonance capture distributions in fuel rods. This generalized method is used as a reference for comparison with two recent independent studies that have employed different Monte Carlo codes and cross-section libraries. The Monte Carlo studies have been found to undercount reference radial resonance captures in the moderator-fuel interface region. The steep radial capture gradients within 0.50 mm of the interface account for the majority of resonance captures and take place where Monte Carlo spatial resolution is poor and the effects of resonance peaks on neutron flux are large. Inconsistencies in the Monte Carlo application or in howpointwise cross-section libraries are sampled may exist. It is shown that refined Monte Carlo solutions with improved spatial resolution would not asymptotically approach the reference spatial capture distributions. It is suspected that the resolved resonance peak and off peak cross sections may not be represented or accounted for appropriately in the Monte Carlo calculations and should be reviewed. If these inconsistencies were cleared up, use of the generalized method might very well challenge the need to perform further Monte Carlo studies of radial resonance captures for isolated uranium-oxide fuel rods.