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A ROBUST ARBITRARILY HIGH ORDER TRANSPORT METHOD OF THE CHARACTERISTIC TYPE FOR UNSTRUCTURED TETRAHEDRAL GRIDS

Description: We present a robust arbitrarily high order transport method of the characteristic type for unstructured tetrahedral grids. Previously encountered difficulties have been addressed through the reformulation of the method based on coordinate transformations, evaluation of the moments balance relation as a linear system of equations involving the expansion coefficients of the projected basis, and the asymptotic expansion of the integral kernels in the thin cell limit. The proper choice of basis functions for the high-order spatial expansion of the solution is discussed and its effect on problems involving scattering discussed. Numerical tests are presented to illustrate the beneficial effect of these improvements, and the improved robustness they yield.
Date: May 1, 2009
Creator: Ferrer, R. M. & Azmy, Y. Y.
Partner: UNT Libraries Government Documents Department

Nodal Diffusion Burnable Poison Treatment for Prismatic Reactor Cores

Description: The prismatic block version of the High Temperature Reactor (HTR) considered as a candidate Very High Temperature Reactor (VHTR)design may use burnable poison pins in locations at some corners of the fuel blocks (i.e., assembly equivalent structures). The presence of any highly absorbing materials, such as these burnable poisons, within fuel blocks for hexagonal geometry, graphite-moderated High Temperature Reactors (HTRs) causes a local inter-block flux depression that most nodal diffusion-based method have failed to properly model or otherwise represent. The location of these burnable poisons near vertices results in an asymmetry in the morphology of the assemblies (or blocks). Hence the resulting inadequacy of traditional homogenization methods, as these “spread” the actually local effect of the burnable poisons throughout the assembly. Furthermore, the actual effect of the burnable poison is primarily local with influence in its immediate vicinity, which happens to include a small region within the same assembly as well as similar regions in the adjacent assemblies. Traditional homogenization methods miss this artifact entirely. This paper presents a novel method for treating the local effect of the burnable poison explicitly in the context of a modern nodal method.
Date: October 1, 2010
Creator: Ougouag, A. M. & Ferrer, R. M.
Partner: UNT Libraries Government Documents Department

A Proposal for User-defined Reductions in OpenMP

Description: Reductions are commonly used in parallel programs to produce a global result from partial results computed in parallel. Currently, OpenMP only supports reductions for primitive data types and a limited set of base language operators. This is a significant limitation for those applications that employ user-defined data types (e. g., objects). Implementing manual reduction algorithms makes software development more complex and error-prone. Additionally, an OpenMP runtime system cannot optimize a manual reduction algorithm in ways typically applied to reductions on primitive types. In this paper, we propose new mechanisms to allow the use of most pre-existing binary functions on user-defined data types as User-Defined Reduction (UDR) operators. Our measurements show that our UDR prototype implementation provides consistently good performance across a range of thread counts without increasing general runtime overheads.
Date: March 22, 2010
Creator: Duran, A; Ferrer, R; Klemm, M; de Supinski, B R & Ayguade, E
Partner: UNT Libraries Government Documents Department

Evaluation of Heterogeneous Options: Effects of MgO versus UO2 Matrix Selection for Minor Actinide Targets in a Sodium Fast Reactor

Description: The primary focus of this work was to compare MgO with UO2 as target matrix material options for burning minor actinides in a transmutation target within a sodium fast reactor. This analysis compared the transmutation performance of target assemblies having UO2 matrix to those having specifically MgO inert matrix.
Date: March 1, 2008
Creator: Pope, M.; Bays, S. & Ferrer, R.
Partner: UNT Libraries Government Documents Department

Impact of Including Higher Actinides in Fast Reactor Transmutation Analyses

Description: Previous fast reactor transmutation studies generally disregarded higher mass minor actinides beyond Cm-246 due to various considerations including deficiencies in nuclear cross-section data. Although omission of these higher mass actinides does not significantly impact the neutronic calculations and fuel cycle performance parameters follow-on neutron dose calculations related to fuel recycling, transportation and handling are significantly impacted. This report shows that including the minor actinides in the equilibrium fast reactor calculations will increase the predicted neutron emission by about 30%. In addition a sensitivity study was initiated by comparing the impact of different cross-section evaluation file for representing these minor actinides.
Date: September 1, 2007
Creator: Forget, B.; Asgari, M.; Ferrer, R. & Bays, S.
Partner: UNT Libraries Government Documents Department

Fast Reactor Alternative Studies: Effects of Transuranic Groupings on Metal and Oxide Sodium Fast Reactor Designs

Description: A 1000 MWth commercial-scale Sodium Fast Reactor (SFR) design with a conversion ratio (CR) of 0.50 was selected in this study to perform perturbations on the external feed coming from Light Water Reactor Spent Nuclear Fuel (LWR SNF) and separation groupings in the reprocessing scheme. A secondary SFR design with a higher conversion ratio (CR=0.75) was also analyzed as a possible alternative, although no perturbations were applied to this model.
Date: September 1, 2007
Creator: Ferrer, R.; Asgari, M.; Bays, S. & Forget, B.
Partner: UNT Libraries Government Documents Department

Evaluation of Homogeneous Options: Effects of Minor Actinide Exclusion from Single and Double Tier Recycle in Sodium Fast Reactors

Description: The Systems Analysis Campaign under the Global Nuclear Energy Partnership (GNEP) has requested the fuel cycle analysis group at the Idaho National Laboratory (INL) to analyze and provide isotopic data for four scenarios in which different strategies for Minor Actinides (MA) management are investigated. A 1000 MWth commercial-scale Sodium Fast Reactor (SFR) design was selected as the baseline in this scenario study. Two transuranic (TRU) conversion ratios, defined as the ratio of the amount of TRU produced over the TRU destroyed in the reactor core, along with different fuel-types were investigated.
Date: March 1, 2008
Creator: Ferrer, R. M.; Bays, S. & Pope, M.
Partner: UNT Libraries Government Documents Department

Sensitivity Analysis of Reprocessing Cooling Times on Light Water Reactor and Sodium Fast Reactor Fuel Cycles

Description: The purpose of this study is to quantify the effects of variations of the Light Water Reactor (LWR) Spent Nuclear Fuel (SNF) and fast reactor reprocessing cooling time on a Sodium Fast Reactor (SFR) assuming a single-tier fuel cycle scenario. The results from this study show the effects of different cooling times on the SFR’s transuranic (TRU) conversion ratio (CR) and transuranic fuel enrichment. Also, the decay heat, gamma heat and neutron emission of the SFR’s fresh fuel charge were evaluated. A 1000 MWth commercial-scale SFR design was selected as the baseline in this study. Both metal and oxide CR=0.50 SFR designs are investigated.
Date: April 1, 2008
Creator: Ferrer, R. M.; Bays, S. & Pope, M.
Partner: UNT Libraries Government Documents Department

Computational Neutronics Methods and Transmutation Performance Analyses for Fast Reactors

Description: The once-through fuel cycle strategy in the United States for the past six decades has resulted in an accumulation of Light Water Reactor (LWR) Spent Nuclear Fuel (SNF). This SNF contains considerable amounts of transuranic (TRU) elements that limit the volumetric capacity of the current planned repository strategy. A possible way of maximizing the volumetric utilization of the repository is to separate the TRU from the LWR SNF through a process such as UREX+1a, and convert it into fuel for a fast-spectrum Advanced Burner Reactor (ABR). The key advantage in this scenario is the assumption that recycling of TRU in the ABR (through pyroprocessing or some other approach), along with a low capture-to-fission probability in the fast reactor’s high-energy neutron spectrum, can effectively decrease the decay heat and toxicity of the waste being sent to the repository. The decay heat and toxicity reduction can thus minimize the need for multiple repositories. This report summarizes the work performed by the fuel cycle analysis group at the Idaho National Laboratory (INL) to establish the specific technical capability for performing fast reactor fuel cycle analysis and its application to a high-priority ABR concept. The high-priority ABR conceptual design selected is a metallic-fueled, 1000 MWth SuperPRISM (S-PRISM)-based ABR with a conversion ratio of 0.5. Results from the analysis showed excellent agreement with reference values. The independent model was subsequently used to study the effects of excluding curium from the transuranic (TRU) external feed coming from the LWR SNF and recycling the curium produced by the fast reactor itself through pyroprocessing. Current studies to be published this year focus on analyzing the effects of different separation strategies as well as heterogeneous TRU target systems.
Date: March 1, 2007
Creator: Ferrer, R.; Asgari, M.; Bays, S. & Forget, B.
Partner: UNT Libraries Government Documents Department

Nodal Green’s Function Method Singular Source Term and Burnable Poison Treatment in Hexagonal Geometry

Description: An accurate and computationally efficient two or three-dimensional neutron diffusion model will be necessary for the development, safety parameters computation, and fuel cycle analysis of a prismatic Very High Temperature Reactor (VHTR) design under Next Generation Nuclear Plant Project (NGNP). For this purpose, an analytical nodal Green’s function solution for the transverse integrated neutron diffusion equation is developed in two and three-dimensional hexagonal geometry. This scheme is incorporated into HEXPEDITE, a code first developed by Fitzpatrick and Ougouag. HEXPEDITE neglects non-physical discontinuity terms that arise in the transverse leakage due to the transverse integration procedure application to hexagonal geometry and cannot account for the effects of burnable poisons across nodal boundaries. The test code being developed for this document accounts for these terms by maintaining an inventory of neutrons by using the nodal balance equation as a constraint of the neutron flux equation. The method developed in this report is intended to restore neutron conservation and increase the accuracy of the code by adding these terms to the transverse integrated flux solution and applying the nodal Green’s function solution to the resulting equation to derive a semi-analytical solution.
Date: September 1, 2009
Creator: Bingham, A.A.; Ferrer, R.M. & ougouag, A.M.
Partner: UNT Libraries Government Documents Department

Summary Report on New Transmutation Analysis for the Evaluation of Homogeneous and Heterogeneous Options in Fast Reactors

Description: A 1000 MWth commercial-scale Sodium Fast Reactor (SFR) design was selected as the baseline in this scenario study. Traditional approaches to Light Water Reactor (LWR) Spent Nuclear Fuel (SNF) transuranic waste (TRU) burning in a fast spectrum system have typically focused on the continual homogeneous recycling (reprocessing) of the discharge fast reactor fuel. The effective reduction of transuranic inventories has been quantified through the use of the transuranics conversion ratio (TRU CR). The implicit assumption in the use of this single parameter is a homogeneous fast reactor option where equal weight is given to the destruction of transuranics, either by fission or eventual fission via transmutation. This work explores the potential application of alternative fast reactor fuel cycles in which the minor actinide (MA) component of the TRU is considered ‘waste’, while the plutonium component is considered as fuel. Specifically, a set of potential designs that incorporate radial heterogeneous target assemblies is proposed and results relevant to transmutation and system analysis are presented. In this work we consider exclusively minor actinide-bearing radial targets in a continual reprocessing scenario (as opposed to deep-burn options). The potential use of targets in a deep burn mode is not necessarily ruled out as an option. However, due to work scope constraints and material limit considerations, it was preferred to leave the target assemblies reach either the assumed limit of 200 DPA at discharge or maximum allowable gas pressure caused by helium production from transmutation. The number and specific design of the target assemblies was chosen to satisfy the necessary core symmetry and physical dimensions (available space for a certain amount of mass in an assembly based on an iterated mass density).
Date: August 1, 2008
Creator: Ferrer, R. M.; Bays, S.; Pope, M.; Forget, B.; Skerjanc, W. & Asgari, M.
Partner: UNT Libraries Government Documents Department

Computational Neutronics Methods and Transmutation Performance Analyses for Light Water Reactors

Description: The urgency for addressing repository impacts has grown in the past few years as a result of Spent Nuclear Fuel (SNF) accumulation from commercial nuclear power plants. One obvious path that has been explored by many is to eliminate the transuranic (TRU) inventory from the SNF thus reducing the need for additional long term repository storage sites. One strategy for achieving this is to burn the separated TRU elements in the currently operating U.S. Light Water Reactor (LWR) fleet. Many studies have explored the viability of this strategy by loading a percentage of LWR cores with TRU in the form of either Mixed Oxide (MOX) fuels or Inert Matrix Fuels (IMF). A task was undertaken at INL to establish specific technical capabilities to perform neutronics analyses in order to further assess several key issues related to the viability of thermal recycling. The initial computational study reported here is focused on direct thermal recycling of IMF fuels in a heterogeneous Pressurized Water Reactor (PWR) bundle design containing Plutonium, Neptunium, Americium, and Curium (IMF-PuNpAmCm) in a multi-pass strategy using legacy 5 year cooled LWR SNF. In addition to this initial high-priority analysis, three other alternate analyses with different TRU vectors in IMF pins were performed. These analyses provide comparison of direct thermal recycling of PuNpAmCm, PuNpAm, PuNp, and Pu.
Date: March 1, 2007
Creator: Asgari, M.; Forget, B.; Piet, S.; Ferrer, R. & Bays, S.
Partner: UNT Libraries Government Documents Department

Graphene Layer Growth Chemistry: Five-Six-Ring Flip Reaction

Description: Reaction pathways are presented for hydrogen-mediated isomerization of a five and six member carbon ring complex on the zigzag edge of a graphene layer. A new reaction sequence that reverses orientation of the ring complex, or 'flips' it, was identified. Competition between the flip reaction and 'ring separation' was examined. Ring separation is the reverse of the five and six member ring complex formation reaction, previously reported as 'ring collision'. The elementary steps of the pathways were analyzed using density-functional theory (DFT). Rate coefficients were obtained by solution of the energy master equation and classical transition state theory utilizing the DFT energies, frequencies, and geometries. The results indicate that the flip reaction pathway dominates the separation reaction and should be competitive with other pathways important to the graphene zigzag edge growth in high temperature environments.
Date: December 1, 2007
Creator: Whitesides, R.; Domin, D.; Salomon-Ferrer, R.; Lester Jr., W.A. & Frenklach, M.
Partner: UNT Libraries Government Documents Department

Creation of a GUI for Zori, a Quantum Monte Carlo program, usingRappture

Description: In their research laboratories, academic institutions produce some of the most advanced software for scientific applications. However, this software is usually developed only for local application in the research laboratory or for method development. In spite of having the latest advances in the particular field of science, such software often lacks adequate documentation and therefore is difficult to use by anyone other than the code developers. As such codes become more complex, so typically do the input files and command statements necessary to operate them. Many programs offer the flexibility of performing calculations based on different methods that have their own set of variables and options to be specified. Moreover, situations can arise in which certain options are incompatible with each other. For this reason, users outside the development group can be unaware of how the program runs in detail. The opportunity can be lost to make the software readily available outside of the laboratory of origin. This is a long-standing problem in scientific programming. Rappture, Rapid Application Infrastructure [1], is a new GUI development kit that enables a developer to build an I/O interface for a specific application. This capability enables users to work only with the generated GUI and avoids the problem of the user needing to learn details of the code. Further, it reduces input errors by explicitly specifying the variables required. Zori, a quantum Monte Carlo (QMC) program, developed by the Lester group at the University of California, Berkeley [2], is one of the few free tools available for this field. Like many scientific computer packages, Zori suffers from the problems described above. Potential users outside the research group have acquired it, but some have found the code difficult to use. Furthermore, new members of the Lester group usually have to take considerable time learning all ...
Date: December 1, 2007
Creator: Olivares-Amaya, R.; Salomon Ferrer, R.; Lester Jr., W.A. & Amador-Bedolla, C.
Partner: UNT Libraries Government Documents Department

Transmutation Performance Analysis for Inert Matrix Fuels in Light Water Reactors and Computational Neutronics Methods Capabilities at INL

Description: The urgency for addressing repository impacts has grown in the past few years as a result of Spent Nuclear Fuel (SNF) accumulation from commercial nuclear power plants. One path that has been explored by many is to eliminate the transuranic (TRU) inventory from the SNF, thus reducing the need for additional long term repository storage sites. One strategy for achieving this is to burn the separated TRU elements in the currently operating U.S. Light Water Reactor (LWR) fleet. Many studies have explored the viability of this strategy by loading a percentage of LWR cores with TRU in the form of either Mixed Oxide (MOX) fuels or Inert Matrix Fuels (IMF). A task was undertaken at INL to establish specific technical capabilities to perform neutronics analyses in order to further assess several key issues related to the viability of thermal recycling. The initial computational study reported here is focused on direct thermal recycling of IMF fuels in a heterogeneous Pressurized Water Reactor (PWR) bundle design containing Plutonium, Neptunium, Americium, and Curium (IMF-PuNpAmCm) in a multi-pass strategy using legacy 5 year cooled LWR SNF. In addition to this initial high-priority analysis, three other alternate analyses with different TRU vectors in IMF pins were performed. These analyses provide comparison of direct thermal recycling of PuNpAmCmCf, PuNpAm, PuNp, and Pu. The results of this infinite lattice assembly-wise study using SCALE 5.1 indicate that it may be feasible to recycle TRU in this manner using an otherwise typical PWR assembly without violating peaking factor limits.
Date: May 1, 2009
Creator: Pope, Michael A.; Bays, Samuel E.; Piet, S.; Ferrer, R.; Asgari, Mehdi & Forget, Benoit
Partner: UNT Libraries Government Documents Department

Deterministic Modeling of the High Temperature Test Reactor

Description: Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine INL’s current prismatic reactor deterministic analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 column thin annular core, and the fully loaded core critical condition with 30 columns. Special emphasis is devoted to the annular core modeling, which shares more characteristics with the NGNP base design. The DRAGON code is used in this study because it offers significant ease and versatility in modeling prismatic designs. Despite some geometric limitations, the code performs quite well compared to other lattice physics codes. DRAGON can generate transport solutions via collision probability (CP), method of characteristics (MOC), and discrete ordinates (Sn). A fine group cross section library based on the SHEM 281 energy structure is used in the DRAGON calculations. HEXPEDITE is the hexagonal z full core solver used in this study and is based on the Green’s Function solution of the transverse integrated equations. In addition, two Monte Carlo (MC) based codes, MCNP5 and PSG2/SERPENT, provide benchmarking capability for the DRAGON and the nodal diffusion solver codes. The results from this study show a consistent bias of 2–3% for the core multiplication factor. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement stems from the fact that ...
Date: June 1, 2010
Creator: Ortensi, J.; Cogliati, J. J.; Pope, M. A.; Ferrer, R. M. & Ougouag, A. M.
Partner: UNT Libraries Government Documents Department

Deterministic Modeling of the High Temperature Test Reactor with DRAGON-HEXPEDITE

Description: The Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine the INL’s current prismatic reactor analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 fuel column thin annular core, and the fully loaded core critical condition with 30 fuel columns. Special emphasis is devoted to physical phenomena and artifacts in HTTR that are similar to phenomena and artifacts in the NGNP base design. The DRAGON code is used in this study since it offers significant ease and versatility in modeling prismatic designs. DRAGON can generate transport solutions via Collision Probability (CP), Method of Characteristics (MOC) and Discrete Ordinates (Sn). A fine group cross-section library based on the SHEM 281 energy structure is used in the DRAGON calculations. The results from this study show reasonable agreement in the calculation of the core multiplication factor with the MC methods, but a consistent bias of 2–3% with the experimental values is obtained. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement partially stems from the fact that during the experiments the control rods were adjusted to maintain criticality, whereas in the model, the rod positions were fixed. In addition, this work includes a brief study of a cross section generation approach that seeks to decouple the ...
Date: October 1, 2010
Creator: Ortensi, J.; Pope, M.A.; Ferrer, R.M.; Cogliati, J.J.; Bess, J.D. & Ougouag, A.M.
Partner: UNT Libraries Government Documents Department