Benchmarking of the 3-D CAD-based Discrete Ordinates code “ATTILA” for dose rate calculations against experiments and Monte Carlo calculations

Youssef, M.Z.; Feder, R.; Batistoni, P.; Fischer, U.; Jakhar, S.; Konno, Chikara; Loughlin, Michael; Villari, R.; Wu, Yican

doi:10.1016/j.fusengdes.2013.07.010

Cited by 8 publications

(3 citation statements)

References 20 publications

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“…The 3-D, FEM, Discrete Ordinates code, ATTILA [8][9], has been utilized to undertake the required nuclear analysis. The serial version and the distributed memory parallel (DMP) version of ATTILA-7.1.0 and ATTILA-7.2.0 have been used along with the FENDL2.1/FORNAX and the ICRP-74 flux-todose conversion factor [10] data bases.…”

Section: Calculation Procedures and Computational Tool And Parametersmentioning

confidence: 99%

Assessment and optimization of the interspace dose rate of the diagnostics equatorial port plug #3 in ITER with Attila

Youssef¹,

Feder²

2013

2013 IEEE 25th Symposium on Fusion Engineering (SOFE)

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According to ITER integration procurement arrangements, the installment of diagnostics in a port should not increase the shutdown dose rate (SDDR) in the port interspace area by no more than ~50 microSv/hr above the baseline, assuming another 50 microSv/hr is attributed to contribution from the port structure and other ITER in-vessel components, such that the upper SDDR limit of 100 microSv/hr is not exceeded 10 6 s after shutdown. It was found that placing the initial design of the Motional Stark Effect, MSE, and the Charge Exchange Recombination Spectroscopy, CXRS, in the equatorial port #3 (EPP3) resulted in an increase in the SDDR that far exceeded the limit. When we follow the optimization process discussed in this paper, substantial reduction in the port interspace SDDR was achieved. The results of the present work show that even when we combine the optimized CXRS and MSE diagnostics with a third GDC diagnostic, the excess of the SDDR over the baseline value did not exceed the allowed upper limit. The present work is based on utilizing the 3-D CAD-based Attila code for assessing the SDDR.

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Section: Calculation Procedures and Computational Tool And Parametersmentioning

confidence: 99%

Assessment and optimization of the interspace dose rate of the diagnostics equatorial port plug #3 in ITER with Attila

Youssef¹,

Feder²

2013

2013 IEEE 25th Symposium on Fusion Engineering (SOFE)

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“…Later, Oak Ridge National Laboratory following the steps of the Los Alamos National Laboratory developed ANISN, DORT and TORT [Rhoades and Simpson, 1997]. More recently, with the continuous improvement of the computing and parallelization technologies, other parallelized codes have appeared like PARTISN [Alcouffe et al, 2005], DRAGON [Marleau et al, 2008], ARES [Chen et al, 2017], DENOVO [Evans et al, 2010], ATTILA [Youssef et al, 2013] and TITAN [Yi, 2009].…”

Section: Discrete Ordinates Methodsmentioning

confidence: 99%

“…There are not several codes that make use of the P N formulation, some examples of them are EVENT [Ziver et al, 2005], FESH [Blomquist, 1979], TEPFEM [Cao and Wu, 2004] or the code developed by [Capilla et al, 2008]. S N method is more commonly used and it can be found implemented in codes such as, DORT/TORT [Rhoades and Childs, 1993], DANTSYS [Alcouffe et al, 1995], PEN-TRAN [Sjoden and Haghighat, 1997] or more recently using masive parallelization, PARTISN [Alcouffe et al, 2005], DRAGON [Marleau et al, 2008], ARES [Chen et al, 2017], DENOVO [Evans et al, 2010], ATTILA [Youssef et al, 2013] and TITAN [Yi, 2009].…”

Section: The Transport Equationmentioning

confidence: 99%

Contributions to solve the Multi-group Neutron Transport equation with different Angular Approaches

Rafet¹

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The most accurate way to know the movement of the neutrons through matter is achieved by solving the Neutron Transport Equation. Three different approaches to solve this equation have been investigated in this thesis: Discrete Ordinates Neutron Transport Equation, Neutron Diffusion Equation and Simplified Spherical Harmonics Equations.In order to solve the equations, different schemes of the Finite Differences Method were studied. The solution of these equations describes the population of neutrons and the occurred reactions inside a nuclear system. These variables are related with the flux and power, fundamental parameters for the Nuclear Safety Analysis.Por último, pero no menos importante, agradezco el apoyo de mi familia, mis hermanos y mi padre; y a los que me hacen sentir como tal, Jesús y Nati. Muy especialmente quiero dar las gracias a María, por estar siempre a mi lado e invitar a superarme. Finalmente, sería injusto no acordarme de mi madre a quien debo todo lo que soy.

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Neutron Transport Theory and Simulation

2017

Fusion Neutronics

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Benchmarking of the 3-D CAD-based Discrete Ordinates code “ATTILA” for dose rate calculations against experiments and Monte Carlo calculations

Cited by 8 publications

References 20 publications

Assessment and optimization of the interspace dose rate of the diagnostics equatorial port plug #3 in ITER with Attila

Assessment and optimization of the interspace dose rate of the diagnostics equatorial port plug #3 in ITER with Attila

Contributions to solve the Multi-group Neutron Transport equation with different Angular Approaches

Neutron Transport Theory and Simulation

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