SEU43 fuel bundle shielding analysis during spent fuel transport

Conference · Sat Jul 01 00:00:00 EDT 2006

OSTI ID:22039633

Margeanu, C A; Ilie, P; Olteanu, G ^[1]

---

1. Inst. for Nuclear Research Pitesti, No. 1 Campului Street, Mioveni 115400, Arges County (Romania)

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)

• Research Organization: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 22039633
• 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

Related Subjects

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