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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

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

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

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

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

RERTR progress in MO-99 production from LEU.

Description: The ANL RERTR program is performing R and D supporting conversion of {sup 99}Mo production from HEU to LEU targets. Irradiation and processing of LEU targets were demonstrated at the Argentine Ezeiza Atomic Center. Target irradiation and disassembly were flawless, but the processing is not fully developed. In addition to preparing for, assisting in, and analyzing results of the demonstration, they performed other R and D related to LEU conversion: (1) designing a prototype production dissolver for digesting irradiated LEU foils in alkaline solutions and developing means to simplify digestion, (2) modifying ion-exchange columns used in the CNEA recovery and purification of {sup 99}Mo to deal with the lower volumes generated from LEU-foil digestion, (3) measuring the performance of new inorganic sorbents that outperform alumina for recovering Mo(VI) from nitric acid solutions containing high concentrations of uranium nitrate, and (4) developing means to facilitate the concentration and calcination of waste nitric-acid/LEU-nitrate solutions from {sup 99}Mo production.
Date: February 13, 2002
Creator: Vandegrift, G. F.; Conner, C.; Aase, S.; Bakel, A.; Bowers, D.; Freiberg, E. et al.
Partner: UNT Libraries Government Documents Department

Progress of the RERTR program in 2001.

Description: This paper describes the 2001 progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners. Postirradiation examinations of microplates have continued to reveal excellent irradiation behavior of U-Mo dispersion fuels in a variety of compositions and irradiating conditions. Irradiation of two new batches of miniplates of greater sizes was completed in the ATR to investigate the swelling behavior of these fuels under prototypic conditions. These materials hold the promise of achieving the program goal of developing LEU research reactor fuels with uranium densities in the 8-9 g/cm{sup 3} range. Qualification of the U-Mo dispersion fuels has been delayed by a patent issue involving KAERI. Test fuel elements with uranium density of 6 g/cm{sup 3} are being fabricated by BWXT and are expected to begin undergoing irradiation in the HFR-Petten reactor around March 2003, with a goal of qualifying this fuel by mid-2005. U-Mo fuel with uranium density of 8-9 g/cm{sup 3} is expected to be qualified by mid-2007. Final irradiation tests of LEU {sup 99}Mo targets in the RAS-GAS reactor at BATAN, in Indonesia, had to be postponed because of the 9/11 attacks, but the results collected to date indicate that these targets will soon be ready for commercial production. Excellent cooperation is also in progress with the CNEA in Argentina, MDSN/AECL in Canada, and ANSTO in Australia. Irradiation testing of five WWR-M2 tube-type fuel assemblies fabricated by the NZChK and containing LEU UO{sub 2} dispersion fuel was successfully completed within the Russian RERTR program. A new LEU U-Mo pin-type fuel that could be used to convert most Russian-designed research reactors has been developed by VNIINM and is ready for testing. Four additional shipments containing 822 spent fuel assemblies from foreign research reactors were accepted by the U.S. by September ...
Date: March 7, 2002
Creator: Travelli, A.
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

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

Macroscopic cross sections for the management of weapons-grade Pu fuels in BWRs

Description: One possible method to reduce the enrichment of the surplus of weapons-grade plutonium is to irradiate mixed oxide fuels (MOX) in commercial nuclear reactors like the boiling water reactors built by General Electric. Contributions to evaluate this possibility are currently made by the Oak Ridge National Laboratory (ORNL) and its contractors. Because important decisions are to be made based on calculations, the calculational procedures, in particular the two-dimensional Scandpower proprietary code HELIOS, were benchmarked against available experience with near weapons-grade Pu fuel and against other codes. In this work the authors report their calculations of diffusion theory macroscopic cross section as a function of burnup, for different combinations of operational parameters. These results are to be input later to the code FORMOSA-B, developed at North Carolina State University (NCSU), to study fuel management strategies in the long range operation of BWR`s with MOX fuels. Seventeen cases for various conditions of the fuel assemblies were specified by NCSU. These correspond to different combinations of void fractions, fuel temperatures, control rods and history.
Date: November 1, 1998
Creator: Difilippo, F.C.
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

Alternate fluid to improve energy efficiency of supercritical water oxidation process

Description: This report discusses the replacement of water by carbon dioxide in both the quench stream and the supercritical water oxidation (SCWO) reactor feed in order to reduce the energy utilization in the process. FLUENT was used to generate the input requirements and ASPEN PLUS was used to model the SCWO process. Simulations were made for normal MODAR operating conditions (baseline case) and two other cases replacing water by carbon dioxide. The basis for and assumptions used in the simulation are given. Economic evaluations were made and costs were compared with the baseline case and a case with 60% replacement of water by carbon dioxide. The equipment cost is almost the same. However, the case with replacement of water by carbon dioxide reduces the energy requirement in the end process by a factor of three, which is a significant energy savings in the operation. Also, the injection of carbon dioxide into the SCWO reactor feed is expected to reduce corrosion and makes salt particles non-sticky. However, these advantages need to be confirmed by experiment.
Date: March 1, 1996
Creator: Oh, C.H.
Partner: UNT Libraries Government Documents Department

Cost optimization of long-cycle LWR operation

Description: The continuing emphasis on improvement of plant capacity factor, as a major means to make nuclear energy more cost competitive in the current deregulatory environment, motivates heightened interest in long intra-refueling intervals and high burnup in LWR units. This study examines the economic implications of these trends, to determine the envelope of profitable fuel management tactics. One batch management is found to be significantly more expensive than two-batch management. Parametric studies were carried out varying the most important input parameters. If ultra-high burnup can be achieved, then n = 3 or even n = 4 management may be preferable. For n = 1 or 2, economic performance declines at higher burnups, hence providing no great incentive for moving further in that direction. Values for n > 2 are also attractive because, for a given burnup target, required enrichment decreases as n increases. This study was limited to average batch burnups below 60,000 MWd/MT.
Date: October 1, 1997
Creator: Handwerk, C.S.; Driscoll, M.J.; McMahon, M.V. & Todreas, N.E.
Partner: UNT Libraries Government Documents Department

Mixed enrichment core design for the NC State University PULSTAR Reactor

Description: The North Carolina State University PULSTAR Reactor license was renewed for an additional 20 years of operation on April 30, 1997. The relicensing period added additional years to the facility operating time through the end of the second license period, increasing the excess reactivity needs as projected in 1988. In 1995, the Nuclear Reactor Program developed a strategic plan that addressed the future maintenance, development, and utilization of the facility. Goals resulting from this plan included increased academic utilization of the facility in accordance with its role as a university research facility, and increased industrial service use in accordance with the mission of a land grant university. The strategic plan was accepted, and it is the intent of the College of Engineering to operate the PULSTAR Reactor as a going concern through at least the end of the current license period. In order to reach the next relicensing review without prejudice due to low excess reactivity, it is desired to maintain sufficient excess reactivity so that, if relicensed again, the facility could continue to operate without affecting users until new fuel assistance was provided. During the NC State University license renewal, the operation of the PULSTAR Reactor at the State University of New York at Buffalo (SUNY Buffalo) was terminated. At that time, the SUNY Buffalo facility had about 240 unused PULSTAR Reactor fuel pins with 6% enrichment. The objective of the work reported here was to develop a mixed enrichment core design for the NC State University PULSTAR reactor which would: (1) demonstrate that 6% enriched SUNY buffalo fuel could be used in the NC State University PULSTAR Reactor within the existing technical specification safety limits for core physics parameters; (2) show that use of this fuel could permit operating the NC State University PULSTAR Reactor to 2017 with ...
Date: December 1, 1997
Creator: Mayo, C.W.; Verghese, K. & Huo, Y.G.
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

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