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Nuclear Criticality Safety
NCSD provides communication among nuclear criticality safety professionals through the development of standards, the evolution of training methods and materials, the presentation of technical data and procedures, and the creation of specialty publications. In these ways, the division furthers the exchange of technical information on nuclear criticality safety with the ultimate goal of promoting the safe handling of fissionable materials outside reactors.
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Utility Working Conference and Vendor Technology Expo (UWC 2024)
August 4–7, 2024
Marco Island, FL|JW Marriott Marco Island
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
ARPA-E announces $40 million to develop transmutation technologies for UNF
The Department of Energy’s Advanced Research Projects Agency–Energy (ARPA-E) announced $40 million in funding to develop cutting-edge technologies to enable the transmutation of used nuclear fuel into less-radioactive substances. According to ARPA-E, the new initiative addresses one of the agency’s core goals as outlined by Congress: to provide transformative solutions to improve the management, cleanup, and disposal of radioactive waste and spent nuclear fuel.
Hyung Jin Shim, Sung Hoon Choi, Chang Hyo Kim
Nuclear Science and Engineering | Volume 176 | Number 1 | January 2014 | Pages 58-68
Technical Paper | doi.org/10.13182/NSE12-87
Articles are hosted by Taylor and Francis Online.
It is well known that the sample variance of a tally mean in Monte Carlo (MC) eigenvalue calculations is biased because of the intercycle correlations of the fission source distribution (FSD). This paper proposes the history-based batch method as a new method that can eliminate the dependency between samples and thereby estimate the real variance of the mean of the MC tally directly from routine cycle-by-cycle MC eigenvalue calculations. The new method estimates the real variance of the MC tally by the sample variance from tally estimates of the history-based batch defined as a set of histories with the same ancestor fission neutrons determined at the first active cycle MC run. The batch averages of the MC tally necessary for this estimate are obtained by correcting the individual tallies with the batch specific weight factors that are derived from independent FSD normalization of each individual batch. Diagnostic methods are also devised for small-batch-size problems, which one may encounter in applying the history-based batch method. The effectiveness of the history-based batch method is examined as a function of the dominance ratio and the batch size for the weakly coupled fissile array problems in comparison with those of bias estimation methods currently available. Its validity is also investigated in terms of the fuel storage facility problem exhibiting a slow source convergence.