Published 2017 | Version v1
Book Metadata-only

Elaborate SMART MCNP modelling using ANSYS and its applications

Creators

  • 1. Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

Description

An MCNP 3-dimensional model can be widely used to evaluate various design parameters such as a core design or shielding design. Conventionally, a simplified 3-dimensional MCNP model is applied to calculate these parameters because of the cumbersomeness of modelling by hand. ANSYS has a function for converting the CAD 'stp' format into an MCNP input in the geometry part. Using ANSYS and a 3- dimensional CAD file, a very detailed and sophisticated MCNP 3-dimensional model can be generated. The MCNP model is applied to evaluate the assembly weighting factor at the ex-core detector of SMART, and the result is compared with a simplified MCNP SMART model and assembly weighting factor calculated by DORT, which is a deterministic Sn code. (authors)

Availability note (English)

Available from doi: http://dx.doi.org/10.1051/epjconf/201715307040
Part of:
EPJ Web of Conferences, Proceedings of the 13. international conference on radiation shielding and 19. topical meeting of the radiation protection and shielding division of the American Nuclear Society - 2016

Additional details

Identifiers

DOI
10.1051/epjconf/201715307040;

Publishing Information

Publisher
EDP Sciences
Imprint Place
Les Ulis (France)
Imprint Title
EPJ Web of Conferences, Proceedings of the 13. international conference on radiation shielding and 19. topical meeting of the radiation protection and shielding division of the American Nuclear Society - 2016
Imprint Pagination
v. 153 [1590 p.]
Journal Page Range
p. 07040.p.1-07040.p.5

Conference

Title
13. international conference on radiation shielding; RPSD-2016: 19. topical meeting of the radiation protection and shielding division of the American Nuclear Society
Acronym
ICRS-13
Dates
3-6 Oct 2016
Place
Paris (France)

INIS

Country of Publication
France
Country of Input or Organization
France
INIS RN
51050056
Subject category
S22: GENERAL STUDIES OF NUCLEAR REACTORS;
Resource subtype / Literary indicator
Conference
Descriptors DEI
A CODES; COMPUTERIZED SIMULATION; GRAPHICAL USER INTERFACE; M CODES; MONTE CARLO METHOD; NEUTRON TRANSPORT; REACTOR CORES; VALIDATION
Descriptors DEC
CALCULATION METHODS; COMPUTER CODES; NEUTRAL-PARTICLE TRANSPORT; RADIATION TRANSPORT; REACTOR COMPONENTS; SIMULATION; TESTING

Optional Information

Notes
11 refs.; This record replaces 51039609