Evaluation of shielding analysis methods for spent-fuel casks

Conference · Wed Nov 01 04:00:00 EST 1989 · Transactions of the American Nuclear Society; (United States)

OSTI ID:5346144

Parks, C V ^[1]

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1. Oak Ridge National Laboratory, TN (USA)

Accurate results from shielding analyses of spent-fuel casks are increasingly important as the desire for optimized designs increases. As-low-as-reasonably-achievable dose-level concerns also contribute to the need for accurate dose evaluations for casks. Three areas require the attention of cask shielding analysts: radiation source generation, utilization of cross-section data, and the radiation transport and dose evaluation. This paper reviews recent efforts carried out at Oak Ridge National Laboratory to evaluate the impact of various codes, data, and analysis assumptions on the calculation of radiation doses from spent-fuel casks.

• OSTI ID: 5346144
• Report Number(s): CONF-891103--
• Conference Information: Journal Name: Transactions of the American Nuclear Society; (United States) Journal Volume: 60
• Country of Publication: United States
• Language: English

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11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
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ACCURACY
ALARA
BOSONS
BURNUP
CALCULATION METHODS
CASKS
CONTAINERS
CROSS SECTIONS
DESIGN
DOSES
ELEMENTARY PARTICLES
MASSLESS PARTICLES
MONTE CARLO METHOD
NATIONAL ORGANIZATIONS
NEUTRAL-PARTICLE TRANSPORT
NEUTRON TRANSPORT
OPTIMIZATION
ORNL
PHOTONS
PWR TYPE REACTORS
RADIATION DOSES
RADIATION PROTECTION
RADIATION TRANSPORT
REACTORS
SELF-SHIELDING
SENSITIVITY
SHIELDING
SPENT FUEL CASKS
SPENT FUEL STORAGE
STORAGE
TRANSPORT
US AEC
US DOE
US ERDA
US ORGANIZATIONS
WATER COOLED REACTORS
WATER MODERATED REACTORS