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Title: SEU43 fuel bundle shielding analysis during spent fuel transport

Abstract

The basic task accomplished by the shielding calculations in a nuclear safety analysis consist in radiation doses calculation, in order to prevent any risks both for personnel protection and impact on the environment during the spent fuel manipulation, transport and storage. The paper investigates the effects induced by fuel bundle geometry modifications on the CANDU SEU spent fuel shielding analysis during transport. For this study, different CANDU-SEU43 fuel bundle projects, developed in INR Pitesti, have been considered. The spent fuel characteristics will be obtained by means of ORIGEN-S code. In order to estimate the corresponding radiation doses for different measuring points the Monte Carlo MORSE-SGC code will be used. Both codes are included in ORNL's SCALE 5 programs package. A comparison between the considered SEU43 fuel bundle projects will be also provided, with CANDU standard fuel bundle taken as reference. (authors)

Authors:
; ;  [1]
  1. Inst. for Nuclear Research Pitesti, No. 1 Campului Street, Mioveni 115400, Arges County (Romania)
Publication Date:
Research Org.:
American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
OSTI Identifier:
22039633
Resource Type:
Conference
Resource Relation:
Conference: PHYSOR-2006: American Nuclear Society's Topical Meeting on Reactor Physics - Advances in Nuclear Analysis and Simulation, Vancouver, BC (Canada), 10-14 Sep 2006; Other Information: Country of input: France; 12 refs.
Country of Publication:
United States
Language:
English
Subject:
61 RADIATION PROTECTION AND DOSIMETRY; 42 ENGINEERING; CANDU TYPE REACTORS; COMPARATIVE EVALUATIONS; ENVIRONMENTAL IMPACTS; FUEL ELEMENT CLUSTERS; MONTE CARLO METHOD; OCCUPATIONAL SAFETY; RADIATION DOSES; RADIATION PROTECTION; SAFETY ANALYSIS; SHIELDING; SPENT FUEL CASKS; SPENT FUEL ELEMENTS; SPENT FUEL STORAGE; WASTE TRANSPORTATION

Citation Formats

Margeanu, C A, Ilie, P, and Olteanu, G. SEU43 fuel bundle shielding analysis during spent fuel transport. United States: N. p., 2006. Web.
Margeanu, C A, Ilie, P, & Olteanu, G. SEU43 fuel bundle shielding analysis during spent fuel transport. United States.
Margeanu, C A, Ilie, P, and Olteanu, G. 2006. "SEU43 fuel bundle shielding analysis during spent fuel transport". United States.
@article{osti_22039633,
title = {SEU43 fuel bundle shielding analysis during spent fuel transport},
author = {Margeanu, C A and Ilie, P and Olteanu, G},
abstractNote = {The basic task accomplished by the shielding calculations in a nuclear safety analysis consist in radiation doses calculation, in order to prevent any risks both for personnel protection and impact on the environment during the spent fuel manipulation, transport and storage. The paper investigates the effects induced by fuel bundle geometry modifications on the CANDU SEU spent fuel shielding analysis during transport. For this study, different CANDU-SEU43 fuel bundle projects, developed in INR Pitesti, have been considered. The spent fuel characteristics will be obtained by means of ORIGEN-S code. In order to estimate the corresponding radiation doses for different measuring points the Monte Carlo MORSE-SGC code will be used. Both codes are included in ORNL's SCALE 5 programs package. A comparison between the considered SEU43 fuel bundle projects will be also provided, with CANDU standard fuel bundle taken as reference. (authors)},
doi = {},
url = {https://www.osti.gov/biblio/22039633}, journal = {},
number = ,
volume = ,
place = {United States},
year = {Sat Jul 01 00:00:00 EDT 2006},
month = {Sat Jul 01 00:00:00 EDT 2006}
}

Conference:
Other availability
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