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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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Proving DRACO will deliver
The United States is now closer than it has been in over five decades to launching the first nuclear thermal rocket into space, thanks to DRACO—the Demonstration Rocket for Agile Cislunar Orbit.
Bryan L. Broadhead, Jabo S. Tang, Robert L. Childs, Cecil V. Parks, Hiroaki Taniuchi
Nuclear Technology | Volume 117 | Number 2 | February 1997 | Pages 206-222
Technical Paper | Radioactive Waste Management | doi.org/10.13182/NT97-A35326
Articles are hosted by Taylor and Francis Online.
The three-dimensional Monte Carlo code MORSE-SGC, as implemented in the SCALE system calcula-tional sequence SAS4, is applied to the analysis of a series of simple geometry benchmark experiments and prototypic spent-fuel storage cask measurements. The simple geometry experiments were performed in Japan and at the General Electric-Morris Operation facility; the cask measurements were performed at the Idaho National Engineering Laboratory. The quantification of uncertainties in a typical shielding analysis process for transport /storage casks can be accomplished by comparison of consistent trends between calculated and measured dose rate quantities in both benchmark and prototypic environments. Benchmark results typically measure the validity of cross-section data and computer code adequacy; prototypic environments, however, generally measure the overall validity of the calcula-tional procedure. A total of five storage cask problems and two simple geometry problems were analyzed to determine the expected accuracies of computational analyses using well-established source-generation and Monte Carlo codes. The general trends seen in this work are in agreement within 30% or better with the measurements for neutron dose rates along the cask side, lid, and bottom. The gamma-ray dose rates with substantial contributions from the top endfitting, plenum, and bottom end-fitting regions also are in good agreement. Based on the latest results, gamma-ray dose rate calculations with major contributions due to the active fuel region show a consistent factor of 1.6 overprediction of the measured quantities for casks with iron and concrete shields. Major uncertainties exist in the quantification of 59Co concentrations in endfitting hardware materials. The results presented support the accuracy of source generation methods and dose estimation methods in these regions given accurate impurity characterizations. Thus, it is felt that the practice of using upper bounds for 59Co initial concentrations should ensure conservative cask designs. Fortunately, the gamma-ray dose discrepancies seen along the sides of both the iron and concrete cask surfaces are overpredictions. The reason for overprediction is not fully known. Even though these overpredictions are not clearly understood, the trends observed, combined with some degree of code and data testing using these or similar benchmark measurements, should inspire confidence in the shielding results for a shipping/ storage package.