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Title: Comparison of the 3-D Deterministic Neutron Transport Code Attila® To Measure Data, MCNP And MCNPX For The Advanced Test Reactor

Abstract

An LDRD (Laboratory Directed Research and Development) project is underway at the Idaho National Laboratory (INL) to apply the three-dimensional multi-group deterministic neutron transport code (Attila®) to criticality, flux and depletion calculations of the Advanced Test Reactor (ATR). This paper discusses the development of Attila models for ATR, capabilities of Attila, the generation and use of different cross-section libraries, and comparisons to ATR data, MCNP, MCNPX and future applications.

Authors:
;
Publication Date:
Research Org.:
Idaho National Laboratory (INL)
Sponsoring Org.:
USDOE
OSTI Identifier:
911127
Report Number(s):
INL/CON-05-00662
TRN: US0704408
DOE Contract Number:
DE-AC07-99ID-13727
Resource Type:
Conference
Resource Relation:
Conference: M&C 2005 International Topical,Avignon, FR,09/12/2005,09/15/2005
Country of Publication:
United States
Language:
English
Subject:
22 - GENERAL STUDIES OF NUCLEAR REACTORS; CRITICALITY; NEUTRON TRANSPORT; TEST REACTORS; Advanced Test Reactor, 3D Transport, Attila

Citation Formats

D. Scott Lucas, and D. S. Lucas. Comparison of the 3-D Deterministic Neutron Transport Code Attila® To Measure Data, MCNP And MCNPX For The Advanced Test Reactor. United States: N. p., 2005. Web.
D. Scott Lucas, & D. S. Lucas. Comparison of the 3-D Deterministic Neutron Transport Code Attila® To Measure Data, MCNP And MCNPX For The Advanced Test Reactor. United States.
D. Scott Lucas, and D. S. Lucas. Thu . "Comparison of the 3-D Deterministic Neutron Transport Code Attila® To Measure Data, MCNP And MCNPX For The Advanced Test Reactor". United States. doi:. https://www.osti.gov/servlets/purl/911127.
@article{osti_911127,
title = {Comparison of the 3-D Deterministic Neutron Transport Code Attila® To Measure Data, MCNP And MCNPX For The Advanced Test Reactor},
author = {D. Scott Lucas and D. S. Lucas},
abstractNote = {An LDRD (Laboratory Directed Research and Development) project is underway at the Idaho National Laboratory (INL) to apply the three-dimensional multi-group deterministic neutron transport code (Attila®) to criticality, flux and depletion calculations of the Advanced Test Reactor (ATR). This paper discusses the development of Attila models for ATR, capabilities of Attila, the generation and use of different cross-section libraries, and comparisons to ATR data, MCNP, MCNPX and future applications.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Thu Sep 01 00:00:00 EDT 2005},
month = {Thu Sep 01 00:00:00 EDT 2005}
}

Conference:
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  • The 3D neutron transport code Attila{sup R} has been used in a Laboratory Directed Research and Development (LDRD) project for the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). The purpose is to examine the feasibility of replacing the current diffusion based Core Safety Analysis Methods with a neutron transport code. This is a discussion of the development of the Attila models, and their comparison to models from other codes and historical data from the ATR. Additional comparisons have been made to the ATR Critical Facility (ATRC), the low power version of ATR, used for physics testing. (authors)
  • An LDRD (Laboratory Directed Research and Development) project is ongoing at the Idaho National Engineering and Environmental Laboratory (INEEL) for applying the three-dimensional multi-group deterministic neutron transport code (Attila{reg_sign}) to criticality, flux and depletion calculations of the Advanced Test Reactor (ATR). This paper discusses the model development, capabilities of Attila, generation of the cross-section libraries, and comparisons to an ATR MCNP model and future.
  • An LDRD (Laboratory Directed Research and Development) project is ongoing at the Idaho National Engineering and Environmental Laboratory (INEEL) for applying the three-dimensional multi-group deterministic neutron transport code (Attila®) to criticality, flux and depletion calculations of the Advanced Test Reactor (ATR). This paper discusses the model development, capabilities of Attila, generation of the cross-section libraries, and comparisons to an ATR MCNP model and future.
  • This paper discusses several topics related to our efforts to accomplish research leveraging 3-D radiation transport models to ultimately yield a complete mosaic of the radiation spectrum from a fissile source. Here, we effectively characterize and construct SNM neutron source terms, and demonstrate that the BUGLE-96 multigroup library is applicable to our general deterministic transport neutron detection scenarios. We also investigated our initial design of a moderated He-3 detector array. In performing our computations, we demonstrate how the PENTRAN parallel 3-D Sn code is quite efficient at parallel computation with several domain decomposition strategies, achieving a parallel (Amdahl) fraction ofmore » 0.96 on up to 16 dedicated processors while yielding our adjoint Sn transport results. Finally, we have established a procedure for analyzing He-3 response in a graded detector-moderator array, and are moving closer in our efforts to attribute a neutron spectrum from the resulting neutron responses in our graded moderator He-3 detector array, simulated entirely via computational methods for SNM sources of high interest. (authors)« less
  • First-principles NaI and BGO detector response functions calculations made with the MCNP code are compared to measurements. Excellent agreement is achieved for the experiments analyzed. Such calculational methodology can be used to achieve a better understanding of the physics of detector response and to maximize the information content available from measured data.