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SOLIDWORKS Based MCNP6 Reactor Criti...

Tumuluri, Madhuranath.

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SOLIDWORKS Based MCNP6 Reactor Criticality Calculation.

Record Type:	Electronic resources : Monograph/item
Title/Author:	SOLIDWORKS Based MCNP6 Reactor Criticality Calculation./
Author:	Tumuluri, Madhuranath.
Published:	Ann Arbor : ProQuest Dissertations & Theses, : 2017,
Description:	84 p.
Notes:	Source: Masters Abstracts International, Volume: 57-05.
Contained By:	Masters Abstracts International57-05(E).
Subject:	Mechanical engineering. -
Online resource:	http://pqdd.sinica.edu.tw/twdaoapp/servlet/advanced?query=10788766
ISBN:	9780355849028

SOLIDWORKS Based MCNP6 Reactor Criticality Calculation.
Tumuluri, Madhuranath.

SOLIDWORKS Based MCNP6 Reactor Criticality Calculation. - Ann Arbor : ProQuest Dissertations & Theses, 2017 - 84 p.

Source: Masters Abstracts International, Volume: 57-05.

Thesis (M.S.)--Texas A&M University - Kingsville, 2017.

Reactor criticality calculation is one of the key steps in nuclear reactor design and development. Los Alamos National Laboratory has developed some codes to perform these calculations. One of the codes is MCNP6 (Monte Carlo N-Particle). The main objective of this research is to use SOLIDWORKS to design two reactor assemblies based on C5G7 benchmark model and calculate the criticality and pin power of the assemblies using MCNP6 and compare the results to C5G7 benchmark results. The C5G7 benchmark has very accurate Monte Carlo solutions for both two dimensional and three-dimensional configurations. In this work we consider two-dimensional assembly. MCNP6 is used in this work to perform reactor criticality modeling using Computer Aided Design (CAD) models designed in SOLIDWORKS. The CAD models are converted to ".DXF" files which are used by visual editing tool in MCNP6 to convert solid modeling data to MCNP6 understandable input format. MCNP6 package contains visual editor executable file called VISED which is used to view and edit the MCNP6 input file. In this research two assemblies are designed with different fuels. The models contain fuel-clad mix and moderator. Criticality factor "k-eff" and normalized pin power is calculated by executing the final input file. The results are compared to C5G7 geometry based MOCUM results and each pin power data is plotted using MATLAB data plotting feature.

ISBN: 9780355849028Subjects--Topical Terms:

649730
Mechanical engineering.

SOLIDWORKS Based MCNP6 Reactor Criticality Calculation.
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500 $a Source: Masters Abstracts International, Volume: 57-05.
500 $a Adviser: Xue Yang.
502 $a Thesis (M.S.)--Texas A&M University - Kingsville, 2017.
520 $a Reactor criticality calculation is one of the key steps in nuclear reactor design and development. Los Alamos National Laboratory has developed some codes to perform these calculations. One of the codes is MCNP6 (Monte Carlo N-Particle). The main objective of this research is to use SOLIDWORKS to design two reactor assemblies based on C5G7 benchmark model and calculate the criticality and pin power of the assemblies using MCNP6 and compare the results to C5G7 benchmark results. The C5G7 benchmark has very accurate Monte Carlo solutions for both two dimensional and three-dimensional configurations. In this work we consider two-dimensional assembly. MCNP6 is used in this work to perform reactor criticality modeling using Computer Aided Design (CAD) models designed in SOLIDWORKS. The CAD models are converted to ".DXF" files which are used by visual editing tool in MCNP6 to convert solid modeling data to MCNP6 understandable input format. MCNP6 package contains visual editor executable file called VISED which is used to view and edit the MCNP6 input file. In this research two assemblies are designed with different fuels. The models contain fuel-clad mix and moderator. Criticality factor "k-eff" and normalized pin power is calculated by executing the final input file. The results are compared to C5G7 geometry based MOCUM results and each pin power data is plotted using MATLAB data plotting feature.
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