15 Matching Results

Search Results

Advanced search parameters have been applied.

Power distributions in fresh and depleted LEU and HEU cores of the MITR reactor.

Description: The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (UMo) is expected to allow the conversion of U.S. domestic high performance reactors like the MITR-II reactor. Toward this goal, core geometry and power distributions are presented. Distributions of power are calculated for LEU cores depleted with MCODE using an MCNP5 Monte Carlo model. The MCNP5 HEU and LEU MITR models were previously compared to experimental benchmark data for the MITR-II. This same model was used with a finer spatial depletion in order to generate power distributions for the LEU cores. The objective of this work is to generate and characterize a series of fresh and depleted core peak power distributions, and provide a thermal hydraulic evaluation of the geometry which should be considered for subsequent thermal hydraulic safety analyses.
Date: April 4, 2012
Creator: Wilson, E.H.; Horelik, N.E.; Dunn, F.E.; Newton, T.H., Jr.; Hu, L.; Stevens, J.G. (Nuclear Engineering Division) et al.
Partner: UNT Libraries Government Documents Department

Comparison and validation of HEU and LEU modeling results to HEU experimental benchmark data for the Massachusetts Institute of Technology MITR reactor.

Description: The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (UMo) is expected to allow the conversion of U.S. domestic high performance reactors like the MITR-II reactor. Towards this goal, comparisons of MCNP5 Monte Carlo neutronic modeling results for HEU and LEU cores have been performed. Validation of the model has been based upon comparison to HEU experimental benchmark data for the MITR-II. The objective of this work was to demonstrate a model which could represent the experimental HEU data, and therefore could provide a basis to demonstrate LEU core performance. This report presents an overview of MITR-II model geometry and material definitions which have been verified, and updated as required during the course of validation to represent the specifications of the MITR-II reactor. Results of calculations are presented for comparisons to historical HEU start-up data from 1975-1976, and to other experimental benchmark data available for the MITR-II Reactor through 2009. This report also presents results of steady state neutronic analysis of an all-fresh LEU fueled core. Where possible, HEU and LEU calculations were performed for conditions equivalent to HEU experiments, which serves as a starting point for safety analyses for conversion of ...
Date: March 2, 2011
Creator: Newton, T. H.; Wilson, E. H; Bergeron, A.; Horelik, N.; Stevens, J. (Nuclear Engineering Division) & Lab.), (MIT Nuclear Reactor
Partner: UNT Libraries Government Documents Department

Computational fluid dynamics analyses of lateral heat conduction, coolant azimuthal mixing and heat transfer predictions in a BR2 fuel assembly geometry.

Description: To support the analyses related to the conversion of the BR2 core from highly-enriched (HEU) to low-enriched (LEU) fuel, the thermal-hydraulics codes PLTEMP and RELAP-3D are used to evaluate the safety margins during steady-state operation (PLTEMP), as well as after a loss-of-flow, loss-of-pressure, or a loss of coolant event (RELAP). In the 1-D PLTEMP and RELAP simulations, conduction in the azimuthal and axial directions is not accounted. The very good thermal conductivity of the cladding and the fuel meat and significant temperature gradients in the lateral directions (axial and azimuthal directions) could lead to a heat flux distribution that is significantly different than the power distribution. To evaluate the significance of the lateral heat conduction, 3-D computational fluid dynamics (CFD) simulations, using the CFD code STAR-CD, were performed. Safety margin calculations are typically performed for a hot stripe, i.e., an azimuthal region of the fuel plates/coolant channel containing the power peak. In a RELAP model, for example, a channel between two plates could be divided into a number of RELAP channels (stripes) in the azimuthal direction. In a PLTEMP model, the effect of azimuthal power peaking could be taken into account by using engineering factors. However, if the thermal mixing in the azimuthal direction of a coolant channel is significant, a stripping approach could be overly conservative by not taking into account this mixing. STAR-CD simulations were also performed to study the thermal mixing in the coolant. Section II of this document presents the results of the analyses of the lateral heat conduction and azimuthal thermal mixing in a coolant channel. Finally, PLTEMP and RELAP simulations rely on the use of correlations to determine heat transfer coefficients. Previous analyses showed that the Dittus-Boelter correlation gives significantly more conservative (lower) predictions than the correlations of Sieder-Tate and Petukhov. STAR-CD 3-D simulations ...
Date: May 23, 2011
Creator: Tzanos, C. P. & Dionne, B. (Nuclear Engineering Division)
Partner: UNT Libraries Government Documents Department

FY2012 summary of tasks completed on PROTEUS-thermal work.

Description: PROTEUS is a suite of the neutronics codes, both old and new, that can be used within the SHARP codes being developed under the NEAMS program. Discussion here is focused on updates and verification and validation activities of the SHARP neutronics code, DeCART, for application to thermal reactor analysis. As part of the development of SHARP tools, the different versions of the DeCART code created for PWR, BWR, and VHTR analysis were integrated. Verification and validation tests for the integrated version were started, and the generation of cross section libraries based on the subgroup method was revisited for the targeted reactor types. The DeCART code has been reorganized in preparation for an efficient integration of the different versions for PWR, BWR, and VHTR analysis. In DeCART, the old-fashioned common blocks and header files have been replaced by advanced memory structures. However, the changing of variable names was minimized in order to limit problems with the code integration. Since the remaining stability problems of DeCART were mostly caused by the CMFD methodology and modules, significant work was performed to determine whether they could be replaced by more stable methods and routines. The cross section library is a key element to obtain accurate solutions. Thus, the procedure for generating cross section libraries was revisited to provide libraries tailored for the targeted reactor types. To improve accuracy in the cross section library, an attempt was made to replace the CENTRM code by the MCNP Monte Carlo code as a tool obtaining reference resonance integrals. The use of the Monte Carlo code allows us to minimize problems or approximations that CENTRM introduces since the accuracy of the subgroup data is limited by that of the reference solutions. The use of MCNP requires an additional set of libraries without resonance cross sections so that reference ...
Date: June 6, 2012
Creator: Lee, C.H. & Smith, M.A. (Nuclear Engineering Division)
Partner: UNT Libraries Government Documents Department

A user's guide to the PLTEMP/ANL code.

Description: PLTEMP/ANL V4.1 is a FORTRAN program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of ''PLTEMP'' codes in use at ANL for the past 20 years. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each with its own axial power shape. The temperature solution is effectively 2-dimensional. It begins with a one-dimensional solution across all coolant channels and fuel plates/tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available with which to determine safety margins such as Onset-of-Nucleate boiling (ONB), departure from nucleate boiling (DNB), and onset of flow instability (FI). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst's time.
Date: July 5, 2011
Creator: Kalimullah, M. (Nuclear Engineering Division)
Partner: UNT Libraries Government Documents Department

Status report on high fidelity reactor simulation.

Description: This report presents the effort under way at Argonne National Laboratory toward a comprehensive, integrated computational tool intended mainly for the high-fidelity simulation of sodium-cooled fast reactors. The main activities carried out involved neutronics, thermal hydraulics, coupling strategies, software architecture, and high-performance computing. A new neutronics code, UNIC, is being developed. The first phase involves the application of a spherical harmonics method to a general, unstructured three-dimensional mesh. The method also has been interfaced with a method of characteristics. The spherical harmonics equations were implemented in a stand-alone code that was then used to solve several benchmark problems. For thermal hydraulics, a computational fluid dynamics code called Nek5000, developed in the Mathematics and Computer Science Division for coupled hydrodynamics and heat transfer, has been applied to a single-pin, periodic cell in the wire-wrap geometry typical of advanced burner reactors. Numerical strategies for multiphysics coupling have been considered and higher-accuracy efficient methods proposed to finely simulate coupled neutronic/thermal-hydraulic reactor transients. Initial steps have been taken in order to couple UNIC and Nek5000, and simplified problems have been defined and solved for testing. Furthermore, we have begun developing a lightweight computational framework, based in part on carefully selected open source tools, to nonobtrusively and efficiently integrate the individual physics modules into a unified simulation tool.
Date: December 11, 2006
Creator: Palmiotti, G.; Smith, M.; Rabiti, C.; Lewis, E.; Yang, W.; Leclere,M. et al.
Partner: UNT Libraries Government Documents Department

Multilevel method for modeling large-scale networks.

Description: Understanding the behavior of real complex networks is of great theoretical and practical significance. It includes developing accurate artificial models whose topological properties are similar to the real networks, generating the artificial networks at different scales under special conditions, investigating a network dynamics, reconstructing missing data, predicting network response, detecting anomalies and other tasks. Network generation, reconstruction, and prediction of its future topology are central issues of this field. In this project, we address the questions related to the understanding of the network modeling, investigating its structure and properties, and generating artificial networks. Most of the modern network generation methods are based either on various random graph models (reinforced by a set of properties such as power law distribution of node degrees, graph diameter, and number of triangles) or on the principle of replicating an existing model with elements of randomization such as R-MAT generator and Kronecker product modeling. Hierarchical models operate at different levels of network hierarchy but with the same finest elements of the network. However, in many cases the methods that include randomization and replication elements on the finest relationships between network nodes and modeling that addresses the problem of preserving a set of simplified properties do not fit accurately enough the real networks. Among the unsatisfactory features are numerically inadequate results, non-stability of algorithms on real (artificial) data, that have been tested on artificial (real) data, and incorrect behavior at different scales. One reason is that randomization and replication of existing structures can create conflicts between fine and coarse scales of the real network geometry. Moreover, the randomization and satisfying of some attribute at the same time can abolish those topological attributes that have been undefined or hidden from researchers. We propose to develop multilevel methods to model complex networks. The key point of the proposed ...
Date: February 24, 2012
Creator: Safro, I. M. (Mathematics and Computer Science)
Partner: UNT Libraries Government Documents Department

Interim report on fuel cycle neutronics code development.

Description: As part of the Global Nuclear Energy Partnership (GNEP), a fast reactor simulation program was launched in April 2007 to develop a suite of modern simulation tools specifically for the analysis and design of sodium cooled fast reactors. The general goal of the new suite of codes is to reduce the uncertainties and biases in the various areas of reactor design activities by enhanced prediction capabilities. Under this fast reactor simulation program, a high-fidelity deterministic neutron transport code named UNIC is being developed. The final objective is to produce an integrated, advanced neutronics code that allows the high fidelity description of a nuclear reactor and simplifies the multi-step design process by direct coupling with thermal-hydraulics and structural mechanics calculations. Currently there are three solvers for the neutron transport code incorporated in UNIC: PN2ND, SN2ND, and MOCFE. PN2ND is based on a second-order even-parity spherical harmonics discretization of the transport equation and its primary target area of use is the existing homogenization approaches that are prevalent in reactor physics. MOCFE is based upon the method of characteristics applied to an unstructured finite element mesh and its primary target area of use is the fine grained nature of the explicit geometrical problems which is the long term goal of this project. SN2ND is based on a second-order, even-parity discrete ordinates discretization of the transport equation and its primary target area is the modeling transition region between the PN2ND and MOCFE solvers. The major development goal in fiscal year 2008 for the MOCFE solver was to include a two-dimensional capability that is scalable to hundreds of processors. The short term goal of this solver is to solve two-dimensional representations of reactor systems such that the energy and spatial self-shielding are accounted for and reliable cross sections can be generated for the homogeneous calculations. ...
Date: May 13, 2008
Creator: Rabiti, C; Smith, M. A.; Kaushik, D. & Yang, W. S.
Partner: UNT Libraries Government Documents Department

Scoping analysis of the Advanced Test Reactor using SN2ND

Description: A detailed set of calculations was carried out for the Advanced Test Reactor (ATR) using the SN2ND solver of the UNIC code which is part of the SHARP multi-physics code being developed under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in DOE-NE. The primary motivation of this work is to assess whether high fidelity deterministic transport codes can tackle coupled dynamics simulations of the ATR. The successful use of such codes in a coupled dynamics simulation can impact what experiments are performed and what power levels are permitted during those experiments at the ATR. The advantages of the SN2ND solver over comparable neutronics tools are its superior parallel performance and demonstrated accuracy on large scale homogeneous and heterogeneous reactor geometries. However, it should be noted that virtually no effort from this project was spent constructing a proper cross section generation methodology for the ATR usable in the SN2ND solver. While attempts were made to use cross section data derived from SCALE, the minimal number of compositional cross section sets were generated to be consistent with the reference Monte Carlo input specification. The accuracy of any deterministic transport solver is impacted by such an approach and clearly it causes substantial errors in this work. The reasoning behind this decision is justified given the overall funding dedicated to the task (two months) and the real focus of the work: can modern deterministic tools actually treat complex facilities like the ATR with heterogeneous geometry modeling. SN2ND has been demonstrated to solve problems with upwards of one trillion degrees of freedom which translates to tens of millions of finite elements, hundreds of angles, and hundreds of energy groups, resulting in a very high-fidelity model of the system unachievable by most deterministic transport codes today. A space-angle convergence study was conducted to determine ...
Date: July 26, 2012
Creator: Wolters, E.; Smith, M. (NE NEAMS PROGRAM) & SC), (
Partner: UNT Libraries Government Documents Department

ZPR-6 assembly 7 high {sup 240} PU core : a cylindrical assemby with mixed (PU, U)-oxide fuel and a central high {sup 240} PU zone.

Description: Over a period of 30 years more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited to form the basis for criticality safety benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. The term 'benchmark' in a ZPR program connotes a particularly simple loading aimed at gaining basic reactor physics insight, as opposed to studying a reactor design. In fact, the ZPR-6/7 Benchmark Assembly (Reference 1) had a very simple core unit cell assembled from plates of depleted uranium, sodium, iron oxide, U3O8, and plutonium. The ZPR-6/7 core cell-average composition is typical of the interior region of liquid-metal fast breeder reactors (LMFBRs) of the era. It was one part of the Demonstration Reactor Benchmark Program,a which provided integral experiments characterizing ...
Date: October 1, 2007
Creator: Lell, R. M.; Schaefer, R. W.; McKnight, R. D.; Tsiboulia, A.; Rozhikhin, Y.; Division, Nuclear Engineering et al.
Partner: UNT Libraries Government Documents Department

Numerical Homogenization on Approach for Stokesian Suspensions.

Description: In this technical report we investigate efficient methods for numerical simulation of active suspensions. The prototypical system is a suspension of swimming bacteria in a Newtonian fluid. Rheological and other macroscopic properties of such suspensions can differ dramatically from the same properties of the suspending fluid alone or of suspensions of similar but inactive particles. Elongated bacteria, such as E. coli or B. subtilis, swim along their principal axis, propelling themselves with the help of flagella, attached at the anterior of the organism and pushing it forward in the manner of a propeller. They interact hydrodynamically with the surrounding fluid and, because of their asymmetrical shape, have the propensity to align with the local flow. This, along with the dipolar nature of bacteria (the two forces a bacterium exerts on a fluid - one due to self-propulsion and the other opposing drag - have equal magnitude and point in opposite directions), causes nearby bacteria to tend to align, resulting in a intermittent local ordering on the mesoscopic scale, which is between the microscopic scale of an individual bacterium and the macroscopic scale of the suspension (e.g., its container). The local ordering is sometimes called a collective mode or collective swimming. Thanks to self-propulsion, collective modes inject momentum into the fluid in a coherent way. This enhances the local strain rate without changing the macroscopic stress applied at the boundary of the container. The macroscopic effective viscosity of the suspension is defined roughly as the ratio of the applied stress to the bulk strain rate. If local alignment and therefore local strain-rate enhancement, are significant, the effective viscosity can be appreciably lower than that of the corresponding passive suspension or even of the surrounding fluid alone. Indeed, a sevenfold decrease in the effective viscosity was observed in experiments with B. subtilis. ...
Date: January 20, 2012
Creator: Haines, Brian M.; Berlyand, Leonid V. & Karpeev, Dmitry A.
Partner: UNT Libraries Government Documents Department

Benchmark specifications and data requirements for initial modeling of the China experimental fast reactor.

Description: A specification is proposed for an initial transient benchmark analysis of the China Experimental Fast Reactor design based on the analysis capabilities of the SAS4A/SASSYS-1 code. For the initial benchmark, a single-channel protected transient overpower accident is defined. Reactivity feedback coefficients will not be required and simplified material properties are recommended. This report also describes the data required for developing the modeling input. This data includes assembly geometry, reactor power distributions, kinetics and decay heat data, and material properties. Comparisons of benchmark results will take place at a future SAS4A/SASSYS-1 training meeting planned to occur at Argonne National Laboratory. Future benchmark specifications will be planned to expand upon this initial model to include more complex reactivity feedback models, material properties, additional assembly geometry, and primary and intermediate coolant systems.
Date: June 4, 2010
Creator: Fanning, T. H. & Division, Nuclear Engineering
Partner: UNT Libraries Government Documents Department

Progress report on development of intermediate fidelity full assembly analysis methods.

Description: While high fidelity modeling capabilities for various physics phenomena are being pursued under advanced modeling and simulation initiatives under the DOE Office of Nuclear Energy, they generally rely on high-performance computation facilities and are too expensive to be used for parameter-space exploration or design analysis. One-dimensional system codes have been used for a long time and have reached a degree of maturity, but limit their validity to specific applications. Thus, an intermediate fidelity (IF) modeling method is being pursued in this work for a fast-running, modest-fidelity, whole-core transient analyses capability. The new approach is essential for design scoping and engineering analyses and could lead to improvements in the design of the new generations of reactors and to the reduction of uncertainties in safety analysis. This report summarizes the initial effort on the development of the intermediate-fidelity full assembly modeling method. The requirements and the desired merits of the IF approach have been defined. A three-dimensional momentum source model has been developed to model the anisotropic flow in the wire-wrapped rod bundle without the need to resolve the geometric details. It has been confirmed that the momentum source model works well if its affecting region is accurately imposed. The validity of the model is further verified by mesh and parameter sensitivity studies. The developed momentum source model, in principle, can be applied to any wire-wrapped bundle geometries and any flow regimes; while the modeling strategy can be applied to other conditions with complex or distorted geometry, such as flow in blocked channels.
Date: September 30, 2011
Creator: Hu, R. & Fanning, T. H. (Nuclear Engineering Division)
Partner: UNT Libraries Government Documents Department

ZPR-6 assembly 7 high {sup 240}Pu core experiments : a fast reactor core with mixed (Pu,U)-oxide fuel and a centeral high{sup 240}Pu zone.

Description: ZPR-6 Assembly 7 (ZPR-6/7) encompasses a series of experiments performed at the ZPR-6 facility at Argonne National Laboratory in 1970 and 1971 as part of the Demonstration Reactor Benchmark Program (Reference 1). Assembly 7 simulated a large sodium-cooled LMFBR with mixed oxide fuel, depleted uranium radial and axial blankets, and a core H/D near unity. ZPR-6/7 was designed to test fast reactor physics data and methods, so configurations in the Assembly 7 program were as simple as possible in terms of geometry and composition. ZPR-6/7 had a very uniform core assembled from small plates of depleted uranium, sodium, iron oxide, U{sub 3}O{sub 8} and Pu-U-Mo alloy loaded into stainless steel drawers. The steel drawers were placed in square stainless steel tubes in the two halves of a split table machine. ZPR-6/7 had a simple, symmetric core unit cell whose neutronic characteristics were dominated by plutonium and {sup 238}U. The core was surrounded by thick radial and axial regions of depleted uranium to simulate radial and axial blankets and to isolate the core from the surrounding room. The ZPR-6/7 program encompassed 139 separate core loadings which include the initial approach to critical and all subsequent core loading changes required to perform specific experiments and measurements. In this context a loading refers to a particular configuration of fueled drawers, radial blanket drawers and experimental equipment (if present) in the matrix of steel tubes. Two principal core configurations were established. The uniform core (Loadings 1-84) had a relatively uniform core composition. The high {sup 240}Pu core (Loadings 85-139) was a variant on the uniform core. The plutonium in the Pu-U-Mo fuel plates in the uniform core contains 11% {sup 240}Pu. In the high {sup 240}Pu core, all Pu-U-Mo plates in the inner core region (central 61 matrix locations per half of the split ...
Date: February 23, 2009
Creator: Lell, R. M.; Morman, J. A.; Schaefer, R.W. & McKnight, R.D.
Partner: UNT Libraries Government Documents Department

Thermal-fluid and electrochemical modeling and performance study of a planar solid oxide electrolysis cell : analysis on SOEC resistances, size, and inlet flow conditions.

Description: Argonne National Laboratory and Idaho National Laboratory researchers are analyzing the electrochemical and thermal-fluid behavior of solid oxide electrolysis cells (SOECs) for high temperature steam electrolysis using computational fluid dynamics (CFD) techniques. The major challenges facing commercialization of steam electrolysis technology are related to efficiency, cost, and durability of the SOECs. The goal of this effort is to guide the design and optimization of performance for high temperature electrolysis (HTE) systems. An SOEC module developed by FLUENT Inc. as part of their general CFD code was used for the SOEC analysis by INL. ANL has developed an independent SOEC model that combines the governing electrochemical mechanisms based on first principals to the heat transfer and fluid dynamics in the operation of SOECs. The ANL model was embedded into the commercial STAR-CD CFD software, and is being used for the analysis of SOECs by ANL. The FY06 analysis performed by ANL and reported here covered the influence of electrochemical properties, SOEC component resistances and their contributing factors, SOEC size and inlet flow conditions, and SOEC flow configurations on the efficiency and expected durability of these systems. Some of the important findings from the ANL analysis are: (1) Increasing the inlet mass flux while going to larger cells can be a compromise to overcome increasing thermal and current density gradients while increasing the cell size. This approach could be beneficial for the economics of the SOECs; (2) The presence of excess hydrogen at the SOEC inlet to avoid Ni degradation can result in a sizeable decrease in the process efficiency; (3) A parallel-flow geometry for SOEC operation (if such a thing be achieved without sealing problems) yields smaller temperature gradients and current density gradients across the cell, which is favorable for the durability of the cells; (4) Contact resistances can significantly influence ...
Date: June 25, 2008
Creator: Yildiz, B.; Smith, J. & Sofu, T.
Partner: UNT Libraries Government Documents Department