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The EBR-II spent fuel treatment program

Description: Argonne National Laboratory has refurbished and equipped an existing hot cell facility for demonstrating a high-temperature electrometallurgical process for treating spent nuclear fuel from the Experimental Breeder Reactor-11. Two waste forms will be produced and qualified for geologic disposal of the fission and activation products. Relatively pure uranium will be separated for storage. Following additional development, transuranium elements will be blended into one of the high-level waste streams. The spent fuel treatment program will help assess the viability of electrometallurgical technology as a spent fuel management option.
Date: December 1, 1995
Creator: Lineberry, M.J. & McFarlane, H.F.
Partner: UNT Libraries Government Documents Department

Progress in development of low-enriched U-Mo dispersion fuels.

Description: Results from postirradiation examinations and analyses of U-Mo/Al dispersion miniplates are presented. Irradiation test RERTR-5 contained mini-fuel plates with fuel loadings of 6 and 8 gU cm{sup -3}. The fuel material consisted of 6, 7 and 10 wt.% Mo-uranium-alloy powders in atomized and machined form. The swelling behavior of the various fuel types is analyzed, indicating athermal swelling of the U-Mo alloy and temperature-dependent swelling owing to U-Mo/Al interdiffusion.
Date: March 4, 2002
Creator: Hofman, G. L.; Snelgrove, J. L.; Hayes, S. L. & Meyer, M. K.
Partner: UNT Libraries Government Documents Department

Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2010

Description: Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or “Core Modeling Update”) Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).
Date: September 1, 2010
Creator: Aryaeinejad, Rahmat; Crawford, Douglas S.; DeHart, Mark D.; Griffith, George W.; Lucas, D. Scott; Nielsen, Joseph W. et al.
Partner: UNT Libraries Government Documents Department

Managing the Nuclear Fuel Cycle, The Big Picture

Description: The nuclear industry, at least in the United States, has failed to deliver on its promise of cheap, abundant energy. After pioneering the science and application and becoming a primary exporter of nuclear technologies, domestic use of nuclear power fell out-of-favor with the public and has been relatively stagnant for several decades. Recently, renewed interest has generated optimism and talk of a nuclear renaissance characterized by a new generation of safe, clean nuclear plants in this country. But, as illustrated by recent policy shifts regarding closure of the fuel cycle and geologic disposal of high-level radioactive wastes, significant hurdles have yet to be overcome. Using the principles of system dynamics, this paper will take a holistic look at the nuclear industry and the interactions between the key players to explore both the intended and unintended consequences of efforts to address the issues that have impeded the growth of the industry and also to illustrate aspects which must be effectively addressed if the renaissance of our industry is to be achieved and sustained.
Date: July 1, 2010
Creator: Carlsen, Brett W
Partner: UNT Libraries Government Documents Department

Selection of LWR cycle length and fuel reload fraction

Description: The continuing evolution of fuel having ever higher burnup capability and the increased emphasis on high plant capacity factor to keep nuclear power cost-competitive, motivates re-examination of some basic fuel management strategies. Specifically, what are the economic optimum goals for the fraction of core to be refueled, 1/n, and the length of the intra-refueling cycle, T{sub c}. The authors present a simple model to study these questions. They conclude that unless substantial improvements in technology are forthcoming, or economic circumstances change significantly, departure from 2- to 4-batch management, or longer than 2- to 3-year cycles in LWRs is not supported by their analysis.
Date: October 1, 1997
Creator: Driscoll, M.J.; Handwerk, C.S. & McMahon, M.V.
Partner: UNT Libraries Government Documents Department

A comparison of equilibrium and non-equilibrium cycle methods for Na-cooled ATW system.

Description: An equilibrium cycle method, embodied in the REBUS-3[1] code system, has generally been used in conventional fast reactor design activities. The equilibrium cycle method provides an efficient approach for modeling reactor system, compared to the more traditional non-equilibrium cycle fuel management calculation approach. Recently, the equilibrium analysis method has been utilized for designing Accelerator Transmutation of Waste (ATW)[2,3,4] cores, in which a scattered-reloading fuel management scheme is used. Compared with the conventional fast reactors, the ATW core is significantly different in several aspects since its main mission is to incinerate the transuranic (TRU) fuels. The high burnup non-fertile fuel has large variations in composition and reactivity during its lifetime. Furthermore, a relatively short cycle length is utilized in the ATW design to limit the potentially large reactivity swing over a cycle, and consequently 7 or 8-batch fuel management is usually assumed for a high fuel burnup. The validity of the equilibrium analysis method for the ATW core, therefore, needed to be verified. The main objective of this paper is to assess the validity of the equilibrium analysis method for a Na-cooled ATW core[4], which is an alternative core design of the ATW system under development.
Date: March 30, 2002
Creator: Kim, Y.; Hill, R. N. & Taiwo, T. A.
Partner: UNT Libraries Government Documents Department

Transition Core Properties during Conversion of the NBSR from HEU to LEU Fuel

Description: The transition of the NBSR from HEU to LEU fuel is challenging due to reactivity constraints and the need to maintain an uninterrupted science program, the mission of the NBSR. The transition cannot occur with a full change of HEU to LEU fuel elements since the excess reactivity would be large enough that the NBSR would violate the technical specification for shutdown margin. Manufacturing LEU fuel elements to represent irradiated fuel elements would be cost prohibitive since 26 one-of-a-kind fuel elements would need to be manufactured. For this report a gradual transition from the present HEU fuel to the proposed LEU fuel was studied. The gradual change approach would follow the present fuel management scheme and replace four HEU fuel elements with four LEU fuel elements each cycle. This manuscript reports the results of a series of calculations to predict the neutronic characteristics and how the neutronics will change during the transition from HEU to LEU in the NBSR.
Date: October 31, 2013
Creator: L., Hanson A. & D., Diamond
Partner: UNT Libraries Government Documents Department

The Application of the PEBBED Code Suite to the PBMR-400 Coupled Code Benchmark - FY 2006 Annual Report

Description: This document describes the recent developments of the PEBBED code suite and its application to the PBMR-400 Coupled Code Benchmark. This report addresses an FY2006 Level 2 milestone under the NGNP Design and Evaluation Methods Work Package. The milestone states "Complete a report describing the results of the application of the integrated PEBBED code package to the PBMR-400 coupled code benchmark". The report describes the current state of the PEBBED code suite, provides an overview of the Benchmark problems to which it was applied, discusses the code developments achieved in the past year, and states some of the results attained. Results of the steady state problems generated by the PEBBED fuel management code compare favorably to the preliminary results generated by codes from other participating institutions and to similar non-Benchmark analyses. Partial transient analysis capability has been achieved through the acquisition of the NEM-THERMIX code from Penn State University. Phase I of the task has been achieved through the development of a self-consistent set of tools for generating cross sections for design and transient analysis and in the successful execution of the steady state benchmark exercises.
Date: September 1, 2006
Partner: UNT Libraries Government Documents Department

CHARACTERIZATION OF MONOLITHIC FUEL FOIL PROPERTIES AND BOND STRENGTH

Description: Understanding fuel foil mechanical properties, and fuel / cladding bond quality and strength in monolithic plates is an important area of investigation and quantification. Specifically, what constitutes an acceptable monolithic fuel – cladding bond, how are the properties of the bond measured and determined, and what is the impact of fabrication process or change in parameters on the level of bonding? Currently, non-bond areas are quantified employing ultrasonic determinations that are challenging to interpret and understand in terms of irradiation impact. Thus, determining mechanical properties of the fuel foil and what constitutes fuel / cladding non-bonds is essential to successful qualification of monolithic fuel plates. Capabilities and tests related to determination of these properties have been implemented at the INL and are discussed, along with preliminary results.
Date: March 1, 2007
Creator: Burkes, D E; Keiser, D D; Wachs, D M; Larson, J S & Chapple, M D
Partner: UNT Libraries Government Documents Department

Microstructural Characterization of Burnable Absorber Materials Being Evaluated for Application in LEU U-Mo Fuel Plates

Description: The starting microstructure of a fuel plate will impact how it performs during irradiation. As a result, microstructural characterization has been performed on as-fabricated monolithic fuel plates to determine the changes in fuel plate microstructure that may result from changes in fabrication parameters. Particular focus has been given to the fuel plate U-10Mo/Zr and Zr/AA6061 cladding interfaces, since the integrity of these interfaces will play a big role in determining the overall performance of the fuel plate during irradiation. In addition, burnable absorber materials for potential incorporation into monolithic fuel plates have been characterized to identify their as-fabricated microstructures. This information will be important when trying to understand the PIE data from fuel plates with burnable absorbers that are irradiated in future irradiation experiments. This paper will focus on the microstructures observed using optical metallography, X-ray diffraction, and scanning and transmission electron microscopy for monolithic fuel plates exposed to different fabrication parameters and for as-fabricated burnable absorber materials.
Date: March 1, 2011
Creator: Jue, J. F.; Miller, B.; Yao, B.; Perez, E. & Sohn, Y. H.
Partner: UNT Libraries Government Documents Department

Neutronic assessment of stringer fuel assembly design for liquid-salt-cooledvery high temperature reactor (LS-VHTR).

Description: Neutronic studies of 18-pin and 36-pin stringer fuel assemblies have been performed to ascertain that core design requirements for the Liquid-Salt Cooled Very High Temperature Reactor (LS-VHTR) can be met. Parametric studies were performed to determine core characteristics required to achieve a target core cycle length of 18 months and fuel discharge burnup greater than 100 GWd/t under the constraint that the uranium enrichment be less than 20% in order to support non-proliferation goals. The studies were done using the WIMS9 lattice code and the linear reactivity model to estimate the core reactivity balance, fuel composition, and discharge burnup. The results show that the design goals can be met using a 1-batch fuel management scheme, uranium enrichment of 15% and a fuel packing fraction of 30% or greater for the 36-pin stringer fuel assembly design.
Date: September 15, 2006
Creator: Szakaly, F. J.; Kim, T. K. & Taiwo, T. A.
Partner: UNT Libraries Government Documents Department

AFCI Options Study

Description: This report describes the background and framework for both organizing the discussion and providing information on the potential for nuclear energy R&D to develop alternative nuclear fuel cycles that would address the issues with the current implementations of nuclear power, including nuclear waste disposal, proliferation risk, safety, security, economics, and sustainability. The disposition of used fuel is the cause of many of the concerns, and the possible approaches to used fuel management identify a number of basic technology areas that need to be considered. The basic science in each of the technology areas is discussed, emphasizing what science is currently available, where scientific knowledge may be insufficient, and especially to identify specific areas where transformational discoveries may allow achievement of performance goals not currently attainable. These discussions lead to the wide range of technical options that have been the basis for past and current research and development on advanced nuclear fuel cycles in the United States. The results of this work are then briefly reviewed to show the extent to which such approaches are capable of addressing the issues with nuclear power, the potential for moving further, and the inherent limitations.
Date: September 1, 2009
Creator: Wigeland, R.; Taiwo, T.; Todosow, M.; Halsey, W. & Gehin, J.
Partner: UNT Libraries Government Documents Department

Characterization of the Microstructure of Irradiated U-Mo Dispersion Fuel with a Matrix that Contains Si

Description: RERTR U-Mo dispersion fuel plates are being developed for application in research reactors throughout the world. Of particular interest is the irradiation performance of U-Mo dispersion fuels with Si added to the Al matrix. Si is added to improve the performance of U-Mo dispersion fuels. Microstructural examinations have been performed on fuel plates with Al-2Si matrix after irradiation to around 50% LEU burnup. Si-rich layers were observed in many areas around the various U-7Mo fuel particles. In one local area of one of the samples, where the Si-rich layer had developed into a layer devoid of Si, relatively large fission gas bubbles were observed in the interaction phase. There may be a connection between the growth of these bubbles and the amount of Si present in the interaction layer. Overall, it was found that having Si-rich layers around the fuel particles after fuel plate fabrication positively impacted the overall performance of the fuel plate.
Date: March 1, 2009
Creator: D. D. Keiser, Jr.; Robinson, A. B.; Jue, J. F.; Medvedev, P. & Finlay, M. R.
Partner: UNT Libraries Government Documents Department

Results of Recent Microstructural Characterization of Irradiated U-Mo Dispersion Fuels with Al Alloy Matrices that Contain Si

Description: RERTR U-Mo dispersion fuel plates are being developed for application in research reactors throughout the world. As part of this development, reactor experiments are being conducted in the Advanced Test Reactor to determine the irradiation performance of different dispersion fuels that contain U-Mo alloys with different Mo contents and Al alloy matrices with different Si contents. Of particular interest is the performance of the dispersion fuels depending on the Si content of the Al alloy matrix, since the addition of Si is being looked to for improving the performance of these dispersion fuels. This paper will describe the results of recent microstructural examinations that have been performed using optical metallography and scanning electron microscopy on as-fabricated and as-irradiated dispersion fuels with different amounts of Si added to the Al matrix. Differences in the microstructural development during irradiation as a function of the Si content in the Al matrix will be discussed, and comments will be made about the development and stability of the fuel/matrix interaction layers that are commonly present in irradiated dispersion fuels.
Date: March 1, 2008
Creator: D D. Keiser, Jr.; Robinson, A. B.; Janney, D. E. & Jue, J. F.
Partner: UNT Libraries Government Documents Department

Performance evaluation of the R6R018 fuel plate using PLATE code

Description: The paper presents results of performance evaluation of the R6R018 fuel plate using PLATE code. R6R018 is a U-7Mo dispersion type mini-plate with Al-3.5Si matrix irradiated in the RERTR-9B experiment. The design of this plate is prototypical of the planned LEONIDAS irradiation test. Therefore, a detailed performance analysis of this plate is important to confirm acceptable behavior in pile, and to provide baseline and justification for further analysis and testing. Specific results presented in the paper include fuel temperature history, growth of the interaction layer between the U-Mo and the matrix, swelling, growth of the corrosion layer, and degradation of thermal conductivity. The methodology of the analysis will be discussed including the newly developed capability to account for the formation of the interaction layer during fuel fabrication.
Date: March 1, 2010
Creator: Medvedev, Pavel G. & Ozaltun, Hakan
Partner: UNT Libraries Government Documents Department

Methods and codes for neutronic calculations of the MARIA research reactor.

Description: The core of the MARIA high flux multipurpose research reactor is highly heterogeneous. It consists of beryllium blocks arranged in 6 x 8 matrix, tubular fuel assemblies, control rods and irradiation channels. The reflector is also heterogeneous and consists of graphite blocks clad with aluminum. Its structure is perturbed by the experimental beam tubes. This paper presents methods and codes used to calculate the MARIA reactor neutronics characteristics and experience gained thus far at IAE and ANL. At ANL the methods of MARIA calculations were developed in connection with the RERTR program. At IAE the package of programs was developed to help its operator in optimization of fuel utilization.
Date: February 18, 2002
Creator: Andrzejewski, K.; Kulikowska, T.; Bretscher, M. M.; Hanan, N. A. & Matos, J. E.
Partner: UNT Libraries Government Documents Department

Management of spent nuclear fuel on the Oak Ridge Reservation, Oak Ridge, Tennessee: Environmental assessment

Description: On June 1, 1995, DOE issued a Record of Decision [60 Federal Register 28680] for the Department-wide management of spent nuclear fuel (SNF); regionalized storage of SNF by fuel type was selected as the preferred alternative. The proposed action evaluated in this environmental assessment is the management of SNF on the Oak Ridge Reservation (ORR) to implement this preferred alternative of regional storage. SNF would be retrieved from storage, transferred to a hot cell if segregation by fuel type and/or repackaging is required, loaded into casks, and shipped to off-site storage. The proposed action would also include construction and operation of a dry cask SNF storage facility on ORR, in case of inadequate SNF storage. Action is needed to enable DOE to continue operation of the High Flux Isotope Reactor, which generates SNF. This report addresses environmental impacts.
Date: February 1, 1996
Partner: UNT Libraries Government Documents Department

Preliminary scoping safety analyses of the limiting design basis protected accidents for the Fast Flux Test Facility tritium production core

Description: The SAS4A/SASSYS-l computer code is used to perform a series of analyses for the limiting protected design basis transient events given a representative tritium and medical isotope production core design proposed for the Fast Flux Test Facility. The FFTF tritium and isotope production mission will require a different core loading which features higher enrichment fuel, tritium targets, and medical isotope production assemblies. Changes in several key core parameters, such as the Doppler coefficient and delayed neutron fraction will affect the transient response of the reactor. Both reactivity insertion and reduction of heat removal events were analyzed. The analysis methods and modeling assumptions are described. Results of the analyses and comparison against fuel pin performance criteria are presented to provide quantification that the plant protection system is adequate to maintain the necessary safety margins and assure cladding integrity.
Date: November 19, 1997
Creator: Heard, F.J.
Partner: UNT Libraries Government Documents Department