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Fusion Energy
This division promotes the development and timely introduction of fusion energy as a sustainable energy source with favorable economic, environmental, and safety attributes. The division cooperates with other organizations on common issues of multidisciplinary fusion science and technology, conducts professional meetings, and disseminates technical information in support of these goals. Members focus on the assessment and resolution of critical developmental issues for practical fusion energy applications.
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2024 ANS Winter Conference and Expo
November 17–21, 2024
Orlando, FL|Renaissance Orlando at SeaWorld
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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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Nuclear new build procurement considerations
It may seem counterintuitive, but the best time to enhance the ability to support operations and maintenance for a new plant is before construction starts. This is one of many lessons learned by the currently operating nuclear fleet. As construction and startup of many nuclear facilities was completed, it quickly became evident that the ability to efficiently support operations and maintenance was limited. Most of the information necessary to establish and manage procurement of spare and replacement items, maintenance, and configuration of the facilities was unavailable and had to be gathered on a case-by-case, “on-demand” basis. Absence of necessary information and the associated challenges resulted in the need for staff augmentation and multiyear-long projects to develop equipment bills of material and maintenance programs and to perform technical evaluations for the huge quantities of spare and replacement items being requested.
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.