Calculation of radiation dose rates from a spent nuclear fuel shipping cask

Conference · Fri Jan 01 00:00:00 EST 1988

OSTI ID:6001131

Chen, S Y; Yuan, Y C

Radiation doses from a spent nuclear fuel cask are usually from various phases of operations during handling, shipping, and storage of the casks. Assessment of such doses requires knowledge of external radiation dose rates at various locations surrounding a cask. Under current practices, dose rates from gamma photons are usually estimated by means of point- or line-source approaches incorporating the conventional buildup factors. Although such simplified approaches may at times be easy to use, their accuracy has not been verified. For example, those simplified methods have not taken into account influencing factors such as the geometry of the cask and the presence of the ground surface, and the effects of these factors on the calculated dose rates are largely unknown. Moreover, similar empirical equations for buildup factors currently do not exist for neutrons. The objective of this study is to use a more accurate approach in calculating radiation dose rates for both neutrons and gamma photons from a spent fuel cask. The calculation utilizes the more sophisticated transport method and takes into account the geometry of the cask and the presence of the ground surface. The results of a detailed study of dose rates in the near field (within 20 meters) are presented and, for easy application, the cask centerline dose rates are fitted into empirical equations at cask centerline distances up to 2000 meters from the surface of the cask.

• Research Organization: Argonne National Lab., IL (United States)
• Sponsoring Organization: USDOE; USDOE, Washington, DC (United States)
• DOE Contract Number: W-31109-ENG-38
• OSTI ID: 6001131
• Report Number(s): ANL/CP-63976; CONF-880601-57; ON: DE92006999
• Resource Relation: Conference: American Nuclear Society (ANS) annual meeting, San Diego, CA (United States), 12-16 Jun 1988
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
42 ENGINEERING
SPENT FUEL CASKS
SPATIAL DOSE DISTRIBUTIONS
CALCULATION METHODS
DESIGN
DISTANCE
DOSE RATES
GAMMA SPECTRA
NEUTRON SPECTRA
RADIATION DOSES
TRANSPORT
CASKS
CONTAINERS
DOSES
RADIATION DOSE DISTRIBUTIONS
SPECTRA
050900* - Nuclear Fuels- Transport
Handling
& Storage
420204 - Engineering- Shipping Containers