SILENE and TDT: A code for collision probability calculations in XY geometries

Conference · Fri Jan 01 00:00:00 EST 1993 · Transactions of the American Nuclear Society; (United States)

OSTI ID:5874543

Sanchez, R; Stankovski, Z ^[1]

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1. Commissariat a l'Energie Atomique, Gif sur Yvette (France)

Collision probability methods are routinely used for cell and assembly multigroup transport calculations in core design tasks. Collision probability methods use a specialized tracking routine to compute neutron trajectories within a given geometric object. These trajectories are then used to generate the appropriate collision matrices in as many groups as required. Traditional tracking routines are based on [open quotes]global[close quotes] geometric descriptions (such as regular meshes) and are not able to cope with the geometric detail required in actual core calculations. Therefore, users have to modify their geometry in order to match the geometric model accepted by the tracking routine, introducing thus a modeling error whose evaluation requires the use of a [open quotes]reference[close quotes] method. Recently, an effort has been made to develop more flexible tracking routines either by directly adopting tracking Monte Carlo techniques or by coding of complicated geometries. Among these, the SILENE and TDT package is being developed at the Commissariat a l' Energie Atomique to provide routine as well as reference calculations in arbitrarily shaped XY geometries. This package combines a direct graphical acquisition system (SILENE) together with a node-based collision probability code for XY geometries (TDT).

• OSTI ID: 5874543
• Report Number(s): CONF-930601-; CODEN: TANSAO
• Journal Information: Transactions of the American Nuclear Society; (United States), Vol. 68; Conference: American Nuclear Society (ANS) annual meeting, San Diego, CA (United States), 20-24 Jun 1993; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

73 NUCLEAR PHYSICS AND RADIATION PHYSICS
22 GENERAL STUDIES OF NUCLEAR REACTORS
NEUTRON DIFFUSION EQUATION
NUMERICAL SOLUTION
NEUTRON TRANSPORT
COMPUTERIZED SIMULATION
NODAL EXPANSION METHOD
PARALLEL PROCESSING
REACTOR CORES
TRANSPORT THEORY
CALCULATION METHODS
DIFFERENTIAL EQUATIONS
EQUATIONS
NEUTRAL-PARTICLE TRANSPORT
PROGRAMMING
RADIATION TRANSPORT
REACTOR COMPONENTS
SIMULATION
663610* - Neutron Physics- (1992-)
220100 - Nuclear Reactor Technology- Theory & Calculation