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Optimization of multi-group cross sections for fast reactor analysis

Optimization of multi-group cross sections for fast reactor ...

International Conference on Mathematics and Computational Methods Applied to nuclear Science and engineering, M and C 2013

作者： Chin, M.R. Manalo, K.L. Edgar, C.A. Paul, J.N. Molinar, M.P. Redd, E.M. Yi, C. Sjoden, G.E. Nuclear and Radiological Engineering Program George W. Woodruff School of Mechanical Engineering Georgia Institute of Technology 770 State St Atlanta GA 30332-0745 United States

ISBN: (纸本)9781627486439

The selection of the number of broad energy groups, collapsed broad energy group boundaries, and their associated evaluation into collapsed macroscopic cross sections from a general 238-group ENDF/B-VII library dramatically impacted the k eigenvalue for fast reactor analysis. An analysis was undertaken to assess the minimum number of energy groups that would preserve problem physics;this involved studies using the 3D deterministic transport parallel code PENTRAN, the 2D deterministic transport code SCALE6.1, the Monte Carlo based MCNP5 code, and the YGROUP cross section collapsing tool on a spatially discretized MOX fuel pin comprised of 21% PuO2-UO2 with sodium coolant. The various cases resulted in a few hundred pcm difference between cross section libraries that included the 238 multi-group reference, and cross sections rendered using various reaction and adjoint weighted cross sections rendered by the YGROUP tool, and a reference continuous energy MCNP case. Particular emphasis was placed on the higher energies characteristic of fission neutrons in a fast spectrum;adjoint computations were performed to determine the average per-group adjoint fission importance for the MOX fuel pin. This study concluded that at least 10 energy groups for neutron transport calculations are required to accurately predict the eigenvalue for a fast reactor system to within 250 pcm of the 238 group case. In addition, the cross section collapsing/weighting schemes within YGROUP that provided a collapsed library rendering eigenvalues closest to the reference were the contributon collapsed, reaction rate weighted scheme. A brief analysis on homogenization of the MOX fuel pin is also provided, although more work is in progress in this area.

关键词： Fast reactors

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Advanced quadratures and periodic boundary conditions in parallel 3D S n transport

Advanced quadratures and periodic boundary conditions in par...

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International Conference on Mathematics and Computational Methods Applied to nuclear Science and engineering, M and C 2013

作者： Manalo, K. Yi, C. Huang, M. Sjoden, G. Nuclear and Radiological Engineering Program G.W. Woodruff School of Mechanical Engineering Georgia Institute of Technology 770 State Street Atlanta GA 30332-0745 United States

ISBN: (纸本)9781627486439

Significant updates in numerical quadratures have warranted investigation with 3D Sn discrete ordinates transport. We show new applications of quadrature departing from level symmetric (2o) and Pn-Tn (>S 2o). investigating 3 recently developed quadratures: Even-Odd (EO), Linear-Discontinuous Finite Element - Surface Area (LDFE-SA), and the non-symmetric Icosahedral Quadrature (IC). We discuss implementation changes to 3D Sn codes (applied to Hybrid MOC-Sn TITAN and 3D parallel PENTRAN) that can be performed to accommodate Icosahedral Quadrature, as this quadrature is not 90-degree rotation invariant. In particular, as demonstrated using PENTRAN, the properties of Icosahedral Quadrature are suitable for trivial application using periodic BCs versus that of reflective BCs. In addition to implementing periodic BCs for 3D Sn PENTRAN, we implemented a technique termed 'angular resweep' which properly conditions periodic BCs for outer eigenvalue iterative loop convergence. As demonstrated by two simple transport problems (3group fixed source and 3group reflected/periodic eigenvalue pin cell), we remark that all of the quadratures we investigated are generally superior to level symmetric quadrature, with Icosahedral Quadrature performing the most efficiently for problems tested.

关键词： Eigenvalues and eigenfunctions

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PULSE HEIGHT TALLY RESPONSE EXPANSION METHOD FOR APPLICATION IN DETECTOR PROBLEMS

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International Journal of Modern Physics: Conference Series 2014年第27期ijmpcs卷

作者： TRAVIS ZIPPERER FARZAD RAHNEMA DINGKANG ZHANG Nuclear and Radiological Engineering Program Georgia Institute of Technology 770 State Street Atlanta Georgia 30332-0745 USA

A new i ncident fl ux r e sponse e x pansion method (IFLEX) is developed to perform on-the-fly detector pulse height spectra calculations with Monte Carlo accuracy. Given the flux incident on the detector window, the method uses pre-computed continuous energy Monte Carlo based response functions to generate the pulse height tallies. B-spline functions are selected for the expansion of the incident flux in the energy phase space. Response functions are generated for a CsI(Na) crystal using an energy dependent B-spline form of the IFLEX method. The method is verified for two incident flux spectra on the crystal. Results of the energy dependent B-spline form of the method are compared to the solutions generated using the direct Monte Carlo code MCNP5. The method is shown to be several orders faster than MCNP5 while maintaining paralleled accuracy.

关键词： Detector modeling pulse height tally hybrid response method

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Application of subgroup decomposition in diffusion theory to gas cooled thermal reactor problem

Application of subgroup decomposition in diffusion theory to...

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International Conference on Mathematics and Computational Methods Applied to nuclear Science and engineering, M and C 2013

作者： Yasseri, Saam Rahnema, Farzad Nuclear and Radiological Engineering and Medical Physics Program George W. Woodruff School Georgia Institute of Technology Atlanta GA 30332-0405 United States

ISBN: (纸本)9781627486439

In this paper, the accuracy and computational efficiency of the subgroup decomposition (SGD) method in diffusion theory is assessed in a ID benchmark problem characteristic of gas cooled thermal systems. This method can be viewed as a significant improvement in accuracy of standard coarse-group calculations used for VHTR whole core analysis in which core environmental effect and energy angle coupling are pronounced. It is shown that a 2-group SGD calculation reproduces fine-group (47) results with 1.5 to 6 times faster computational speed depending on the stabilizing schemes while it is as efficient as single standard 6-group diffusion calculation.

关键词： Computational efficiency

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Development of active interrogation for monitoring special-nuclear- materials (AIMS) hybrid tool

Development of active interrogation for monitoring special-n...

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2013 ANS Annual Meeting - Next Generation nuclear Energy: Prospects and Challenges

作者： Royston, K. Walters, W. Haghighat, A. Yi, C. Sjoden, G. Nuclear Engineering Program VTRC Virginia Tech 900 N Glebe Road Arlington VA 22203 United States Nuclear and Radiological Engineering and Medical Physics Programs Georgia Tech. 770 State St. NW Atlanta GA 30332 United States

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A new integrated framework to advance fire probabilistic risk analysis of nuclear power plants

A new integrated framework to advance fire probabilistic ris...

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2013 Winter Meeting on Transactions and Embedded Topical Meetings: Risk Management for Complex Socio-Technical Systems, 2nd ANS SMR 2013 Conference, nuclear Nonproliferation - 1st Fission to the Future

作者： Sakurahara, Tatsuya Mohaghegh, Zahra Kee, Ernie Rodgers, Shawn Brandyberry, Mark Kazemi, Reza Reihani, Seyed A. Dept. of Nuclear Plasma and Radiological Eng. University of Illinois Urbana-Champaign United States South Texas Project Nuclear Operating Company United States Computational Science and Engineering Program University of Illinois Urbana-Champaign United States Soteria Consultants LLC United States

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ANTINEUTRINOS FOR REACTOR SAFEGUARDS: EFFECT OF FUEL LOADING AND BURNUP ON THE SIGNAL

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International Journal of Modern Physics: Conference Series 2014年第27期ijmpcs卷

作者： ANNA ERICKSON ADAM BERNSTEIN NATHANIEL BOWDEN Nuclear and Radiological Engineering Program Georgia Institute of Technology 770 State St. NW Atlanta GA 30332 Atlanta Advanced Detectors Group Lawrence Livermore National Laboratory USA

Various types of nuclear reactor related information, including relative power level and fuel evolution parameters, can be inferred remotely using antineutrino detectors. We show that it is possible to verify assembly-level burnup using information derived from an antineutrino detector if the nominal reactor fuel loading is known. Alternatively, if the core power is measured using an independent method, for example, a thermal hydraulic element, and the nominal core behavior is known, the antineutrino detector has a capability to determine previously unknown MOX loading in the core.

关键词： Antineutrinos remote reactor monitoring nuclear fuel cycle

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Risk-informed resolution of Generic Safety Issue 191

Risk-informed resolution of Generic Safety Issue 191

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International Topical Meeting on Probabilistic Safety Assessment and Analysis 2013, PSA 2013

作者： Mohaghegh, Zahra Kee, Ernie Reihani, Seyed A. Kazemi, Reza Johnson, David Grantom, Rick Fleming, Karl Sande, Tim Letellier, Bruce Zigler, Gilbert Morton, David Tejada, Jeremy Howe, Kerry Leavitt, Janet Hassan, Yassin A. Vaghetto, Rodolfo Lee, Saya Blossom, Steve Department of Nuclear Plasma and Radiological Engineering University of Illinois at Urbana-Champaign United States South Texas Project Nuclear Operating Company United States Soteria Consultants LLC United States ABS Consulting United States Knf Consulting Services LLC United States Enercon Services Inc. Germany Alion Science and Technology United States Graduate Program in Operations Research and Industrial Engineering University of Texas at Austin United States Civil Engineering Department University of New Mexico United States Department of Nuclear Engineering Texas A and M University United States

ISBN: (纸本)9781629938172

The United States nuclear Regulatory Commission (USNRC) relies on Probabilistic Risk Assessment (PRA) as one of the main pillars of its risk-informed regulatory and oversight functions. In 2011, the South Texas Project nuclear Operating Company (STPNOC) initiated a risk-informed project to resolve Generic Safety Issue 191 (GSI-191), which is related to the performance of the emergency core cooling system (ECCS) following a loss of coolant accident (LOCA). This paper summarizes the results of collaborative work on this project by teams of experts from industry and academia. The progress of the research and its implementation have also significantly benefitted from the discussions and feedback received from regulatory representatives. An integrative framework was developed to explicitly provide failure probabilities for the post-LOCA PRA basic events associated with the concerns raised in GSI-191. These basic event probabilities are estimated using the CASA Grande program, which was developed as part of the project. This program encompasses the time-dependent modeling of the underlying physical phenomena of the basic events and the propagation of the uncertainties in the physical models. This paper summarizes the elements of the integrative framework including PRA and the CASA Grande program, as well as the input parameters, assumptions, methodology, and results of the STPNOC analysis. The results show that the risk of core damage or large early release related to the concerns raised in GSI-191 in the as-built, as-operated design for STPNOC is very small (as defined in Regulatory Guide 1.174).

关键词： Loss of coolant accidents

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Atomistic Ordering in Body Centered Cubic Uranium-Zirconium Alloy

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MRS Online Proceedings Library 2013年第1期1514卷 27-35页

作者： Alex P Moore Ben Beeler Michael Baskes Maria Okuniewski Chaitanya S Deo Nuclear and Radiological Engineering Program George W Woodruff School of Mechanical Engineering Georgia Institute of Technology Atlanta USA University of California San Diego La Jolla USA Los Alamos National Laboratory Los Alamos USA Idaho National Laboratory Idaho Falls USA

The metallic binary-alloy fuel Uranium-Zirconium is important for the use of the new generation of advanced fast reactors. Uranium-Zirconium goes through a phase transition at higher temperatures to a (gamma) Body Centered Cubic (BCC) phase. The BCC high temperature phase is particularly important, since the BCC phase corresponds to the temperature range in which the fast reactors will operate. A semi-empirical MEAM (Modified Embedded Atom Method) potential is presented for Uranium-Zirconium. The physical properties of the Uranium-Zirconium binary alloy were reproduced using Molecular Dynamics (MD) simulations and Monte Carlo (MC) simulations with the MEAM potential. This is a large step in making a computationally acceptable fuel performance code.

关键词： Zr U nuclear materials

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Atomistic ordering in body centered cubic Uranium-Zirconium alloy

Atomistic ordering in body centered cubic Uranium-Zirconium ...

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2012 MRS Fall Meeting

作者： Moore, Alex P. Beeler, Ben Baskes, Michael Okuniewski, Maria Deo, Chaitanya S. Nuclear and Radiological Engineering Program George W Woodruff School of Mechanical Engineering Georgia Institute of Technology 770 State Street Atlanta GA 30332 United States University of California San Diego 9500 Gilman Drive San Diego CA 92093 United States Los Alamos National Laboratory PO Box 1663 Los Alamos NM 87545 United States Idaho National Laboratory PO Box 1625 Idaho Falls ID 83415 United States

ISBN: (纸本)9781605114910

关键词： Molecular dynamics

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