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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.
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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
Dragonfly, a Pu-fueled drone heading to Titan, gets key NASA approval
Curiosity landed on Mars sporting a radioisotope thermoelectric generator (RTG) in 2012, and a second NASA rover, Perseverance, landed in 2021. Both are still rolling across the red planet in the name of science. Another exploratory craft with a similar plutonium-238–fueled RTG but a very different mission—to fly between multiple test sites on Titan, Saturn’s largest moon—recently got one step closer to deployment.
On April 25, NASA and the Johns Hopkins University Applied Physics Laboratory (APL) announced that the Dragonfly mission to Saturn’s icy moon passed its critical design review. “Passing this mission milestone means that Dragonfly’s mission design, fabrication, integration, and test plans are all approved, and the mission can now turn its attention to the construction of the spacecraft itself,” according to NASA.
S. I. Bhuiyan, M. A. W. Mondal, M. M. Sarker, M. Rahman, M. S. Shahdatullah, M. Q. Huda, T. K. Chakrobortty, M. J. H. Khan
Nuclear Technology | Volume 130 | Number 2 | May 2000 | Pages 111-131
Technical Paper | Fission Reactors | doi.org/10.13182/NT00-A3081
Articles are hosted by Taylor and Francis Online.
This study deals with the analysis of some neutronics and safety parameters of the current core of a 3-MW TRIGA MARK-II research reactor and validation of the generated macroscopic cross-section library and calculational techniques by benchmarking with experimental, operational, and available Safety Analysis Report (SAR) values. The overall strategy is: (a) generation of the problem-dependent cross-section library from basic Evaluated Nuclear Data Files such as ENDF/B-VI and JENDL-3.2 with NJOY94.10+, (b) use of the WIMSD-5 package to generate a few-group neutron macroscopic cross section for all of the materials in the core and its immediate neighborhood, (c) use the three-dimensional CITATION code to perform the global analysis of the core, and (d) checking of the validity of the CITATION diffusion code with the MCNP4B2 Monte Carlo code. The ultimate objective is to establish methods for reshuffling the current core configuration to upgrade the thermal flux at irradiation locations for increased isotope production. The computational methods, tools and techniques, customization of cross-section libraries, various models for cells and supercells, and many associated utilities are standardized and established/validated for the overall neutronic analysis. The excess reactivity, neutron flux, power distribution, power peaking factors, determination of the hot spot, and fuel temperature reactivity coefficients f in the temperature range of 45 to 1000 °C are studied. All the analyses are performed using the 4- and 7-group libraries of the macroscopic cross sections generated from the 69-group WIMSD-5 library. The 7-group calculations yield comparatively better agreement with the experimental value of keff and the other core parameters. The CITATION test runs using different cross-section sets based on the different models applied in the WIMSD-5 calculations show a strong influence of those models on the final integral parameter. Some of the cells are specially treated with the Prize options available in WIMSD-5 to take into account the fine structure of the flux gradient in the fuel-reflector interface region. The hot spot is found physically at the fuel position C4 with a maximum power density of 1.044559 × 102 W/cm3. The calculated total peaking factor is 5.8867 compared to the original SAR value of 5.6325. The curve of f with the temperature at zero burnup shows that the curve deviates somewhat with that reported in the original SAR for low-enriched uranium fuel. The MCNP calculations establish that the CITATION calculations and the generated cross-section library are reasonably good for neutronic analysis of TRIGA reactors. The results obtained from the neutronic analysis will be used to analyze the thermal-hydraulic behavior and the safety margins of the core both for steady-state and pulse-mode operations.
