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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.
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Nuclear Energy Conference & Expo (NECX)
September 8–11, 2025
Atlanta, GA|Atlanta Marriott Marquis
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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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WIPP’s SSCVS: A breath of fresh air
This spring, the Department of Energy’s Office of Environmental Management announced that it had achieved a major milestone by completing commissioning of the Safety Significant Confinement Ventilation System (SSCVS) facility—a new, state-of-the-art, large-scale ventilation system at the Waste Isolation Pilot Plant, the DOE’s geologic repository for defense-related transuranic (TRU) waste in New Mexico.
Hisao Yamakoshi
Nuclear Science and Engineering | Volume 87 | Number 2 | June 1984 | Pages 152-180
Technical Paper | doi.org/10.13182/NSE84-A17709
Articles are hosted by Taylor and Francis Online.
By introducing a concept of shielding characteristics, a new method is proposed for shielding calculations of spent fuel shipping casks. The method separates ordinary shielding calculation into two steps, one calculates the radiation current leaking from the unshielded cavity region. The other method synthesizes the radiation dose rate outside the cask arising from the leaked current, the response functions for the radiation dose rate at the outer cask surface, and the functions for the radiation current reflected from the inner surface of the cask wall. In the synthesis, the effect of the coupling of the currents reflected between the cask wall and the cavity region is taken into account. The validity of the proposed method is confirmed by applying the method to an analysis of the measured data obtained for a CRIEPI cask. Response functions, the established characteristic functions for radiation shielding capabilities, are calculated for several typical actual casks. Calculated results are summarized for the convenience of applying the proposed method to actual cases. The merits of the present study are (a) the calculational code of the proposed method deals with only matrix calculations in short-step programming and is suitable for a microcomputer for which input data of characteristic functions are supplied from floppy disks, (b) with large and high-speed computers, one can evaluate radiation dose rates on the outer surface of a given cask in very short machine time and with good accuracy, (c) by application of the characteristic functions, one can extract information that will improve the design of the cask walls to provide more effective shielding by intercomparison of characteristic functions for several types of casks, and (d) one can foresee the influence of changes in the energy spectrum of source radiations on the neutron and the gamma-ray dose rates at the outer cask surface by the rule-of-thumb of superimposing the characteristic functions of the dose rate because they are functions of the incident energies.