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Identification and Assessment of Material Models for Age-Related Degradation of Structures and Passive Components in Nuclear Power Plants

Description: When performing seismic safety assessments of nuclear power plants (NPPs), the potential effects of age-related degradation on structures, systems, and components (SSCs) should be considered. To address the issue of aging degradation, the Korea Atomic Energy Research Institute (KAERI) has embarked on a five-year research project to develop a realistic seismic risk evaluation system which will include the consideration of aging of structures and components in NPPs. Three specific areas that are included in the KAERI research project, related to seismic probabilistic risk assessment (PRA), are probabilistic seismic hazard analysis, seismic fragility analysis including the effects of aging, and a plant seismic risk analysis. To support the development of seismic capability evaluation technology for degraded structures and components, KAERI entered into a collaboration agreement with Brookhaven National Laboratory (BNL) in 2007. The collaborative research effort is intended to continue over a five year period with the goal of developing seismic fragility analysis methods that consider the potential effects of age-related degradation of SSCs, and using these results as input to seismic PRAs. In the Year 1 scope of work BNL collected and reviewed degradation occurrences in US NPPs and identified important aging characteristics needed for the seismic capability evaluations that will be performed in the subsequent evaluations in the years that follow. This information is presented in the Annual Report for the Year 1 Task, identified as BNL Report-81741-2008 and also designated as KAERI/RR-2931/2008. The report presents results of the statistical and trending analysis of this data and compares the results to prior aging studies. In addition, the report provides a description of U.S. current regulatory requirements, regulatory guidance documents, generic communications, industry standards and guidance, and past research related to aging degradation of SSCs. This report describes the research effort performed by BNL for the Year 2 scope of ...
Date: April 27, 2009
Creator: Nie,J.; Braverman, J.; Hofmayer, C.; Kim, M. K. & Choi, I-K.
Partner: UNT Libraries Government Documents Department

Westinghouse Reactor Protection System Unavailability, 1984-1995

Description: An analysis was performed of the safety-related performance of the reactor protection system (RPS) at U.S. Westinghouse commercial reactors during the period 1984 through 1995. RPS operational data were collected from the Nuclear Plant Reliability Data System and Licensee Event Reports. A risk-based analysis was performed on the data to estimate the observed unavailability of the RPS, based on a fault tree model of the system. Results were compared with existing unavailability estimates from Individual Plant Examinations and other reports.
Date: August 1, 1999
Creator: Gentillon, C. D.; Marksberry, D.; Rasmuson, D.; Calley, M. B.; Eide, S. A. & Wierman, T.
Partner: UNT Libraries Government Documents Department

General Electric Reactor Protection System Unavailability, 1984-1995

Description: An analysis was performed of the safety-related performance of the reactor protection system (RPS) at U.S. General Electric commercial reactors during the period 1984 through 1995. RPS operational data were collected from the Nuclear Plant Reliability Data System and Licensee Event Reports. A risk-based analysis was performed on the data to estimate the observed unavailability of the RPS, based on a fault tree model of the system. Results were compared with existing unavailability estimates from Individual Plant Examinations and other reports.
Date: August 1, 1999
Creator: Gentillon, C. D.; Rasmuson, D.; Hamzehee, H.; Calley, M. B.; Eide, S. A. & Wierman, T.
Partner: UNT Libraries Government Documents Department

Design of an Actinide Burning, Lead-Bismuth Cooled Reactor That Produces Low Cost Electricity

Description: The purpose of this project is to investigate the suitability of lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. The choice of lead-bismuth for the reactor coolant is an actinide burning fast reactor offers enhanced safety and reliability. The advantages of lead-bismuth over sodium as a coolant are related to the following material characteristics: chemical inertness with air and water; higher atomic number; lower vapor pressure at operating temperatures; and higher boiling temperature. Given the status of the field, it was agreed that the focus of this investigation in the first two years will be on the assessment of approaches to optimize core and plant arrangements in order to provide maximum safety and economic potential in this type of reactor.
Date: July 1, 1999
Creator: Buongiorno, Jacopo; Ballinger, Ron; Davis, Cliff; Doyoung, Kim; Driscoll, Mikael; Hejzler, Pavel et al.
Partner: UNT Libraries Government Documents Department

Chemical and Radiochemical Constituents in Water from Wells in the Vicinity of the Naval Reactors Facility, Idaho National Engineering and Environmental Laboratory, Idaho, 1996

Description: The U.S. Geological Survey, in response to a request from the U.S. Department of Energy's Pittsburgh Naval Reactors Office, Idaho Branch Office (IBO), samples water from 13 wells during 1996 as part of a long-term project to monitor water quality to the Snake River Plain aquifer in the vicinity of the Naval Reactors Facility (NRF), Idaho National Engineering and Environmental Laboratory, Idaho. The IBO requires information about the mobility of radionuclide- and chemical-waste constituents in the Snake River Plain aquifer. Waste-constituent mobility is determined principally by (1) the rate and direction of ground-water flow; (2) the locations, quantities, and methods of waste disposal; (3) waste-constituents chemistry; and (4) the geochemical processes taking place in the aquifer. The purpose of the data-collection program is to provide IBO with water-chemistry data to evaluate the effect of NRF activities on the water quality of the Snake River Plain aquifer. Water samples were analyzed for naturally occurring constituents and man-made contaminants.
Date: October 1, 1999
Creator: Knobel, L. L.; Bartholomay, R. C.; Tucker, B. J. & Williams, L. M.
Partner: UNT Libraries Government Documents Department

A Program for Risk Assessment Associated with IGSCC of BWR Vessel Internals

Description: A program is being carried out for the US Nuclear Regulatory Commission (NRC) by the Idaho National Engineering and Environmental Laboratory (INEEL), to conduct an independent risk assessment of the consequences of failures initiated by intergranular stress corrosion cracking (IGSCC) of the reactor vessel internals of boiling water reactor (BWR) plants. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, both singly and in combination with the failures of others, with specific consideration given to potential cascading and common mode effects on system performance. This paper presents a description of the overall program that is underway to modify an existing probabilistic risk assessment (PRA) of the BWR/4 plant to include IGSCC-initiated failures, subsequently to complete a quantitative PRA.
Date: August 1, 1999
Creator: Ware, Arthur G.; Morton, D. Keith; Page, Joel D.; Nitzel, Michael E.; Eide, Steven A. & Chang, T. -Y.
Partner: UNT Libraries Government Documents Department

Preliminary Risk Assessment Associated with IGSCC of BWR Vessel Internals

Description: BWR core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission (NRC) has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory (INEEL) to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. The paper presents an overview of the program, discusses the results of a preliminary qualitative assessment, and summarizes a simplified risk assessment that was conducted on sequences resulting from failures of jet pump components of a BWR/4 plant.
Date: August 1, 1999
Creator: Ware, Arthur; Morton, Keith; Nitzel, Michael; Chokshi, Nilesh & Chang, Tsun-Yung
Partner: UNT Libraries Government Documents Department

Final Technical Report for the Period September 2002 through September 2005; H2-MHR Pre-Conceptual Design Report: SI-Based Plant; H2-MHR Pre-Conceptual Design Report: HTE-Based Plant

Description: For electricity and hydrogen production, an advanced reactor technology receiving considerable international interest is a modular, passively-safe version of the high-temperature, gas-cooled reactor, known in the U.S. as the Modular Helium Reactor (MHR), which operates at a power level of 600 MW(t). For electricity production, the MHR operates with an outlet helium temperature of 850 C to drive a direct, Brayton-cycle power-conversion system with a thermal-to-electrical conversion efficiency of 48 percent. This concept is referred to as the Gas Turbine MHR (GT-MHR). For hydrogen production, both electricity and process heat from the MHR are used to produce hydrogen. This concept is referred to as the H2-MHR. This report provides pre-conceptual design descriptions of full-scale, nth-of-a-kind H2 MHR plants based on thermochemical water splitting using the Sulfur-Iodine process and High-Temperature Electrolysis.
Date: April 19, 2006
Creator: Richards, M.; Shenoy, A.; Brown, L.; Buckingham, R.; Harvego, E.; Peddicord, K. et al.
Partner: UNT Libraries Government Documents Department

Advanced Power Ultra-Uprates of Existing Plants (APPU) Final Scientific/Technical Report

Description: This project assessed the feasibility of a Power Ultra-Uprate on an existing nuclear plant. The study determined the technical and design limitations of the current components, both inside and outside the containment. Based on the identified plant bottlenecks, the design changes for major pieces of equipment required to meet the Power Ultra-Uprate throughput were determined. Costs for modified pieces of equipment and for change-out and disposal of the replaced equipment were evaluated. These costs were then used to develop capital, fuel and operating and maintenance cost estimates for the Power Ultra-Uprate plant. The cost evaluation indicates that the largest cost components are the replacement of power (during the outage required for the uprate) and the new fuel loading. Based on these results, the study concluded that, for a ?standard? 4-loop plant, the proposed Power Ultra-Uprate is technically feasible. However, the power uprate is likely to be more expensive than the cost (per Kw electric installed) of a new plant when large capacity uprates are considered (>25%). Nevertheless, the concept of the Power Ultra-Uprate may be an attractive option for specific nuclear power plants where a large margin exists in the steam and power conversion system or where medium power increases (~600 MWe) are needed. The results of the study suggest that development efforts on fuel technologies for current nuclear power plants should be oriented towards improving the fuel performance (fretting-wear, corrosion, uranium load, manufacturing, safety) required to achieve higher burnup rather focusing on potential increases in the fuel thermal output.
Date: March 31, 2006
Creator: Rubiolo, Pablo R.; Conway, Lawarence E.; Oriani, Luca; Lahoda, Edward J.; DeSilva, Greg (Westinghouse Science and Technology Department); Hu, Min H. et al.
Partner: UNT Libraries Government Documents Department

Isolation of Metals from Liquid Wastes: Reactive Scavenging in Turbulent Thermal Reactors

Description: Sorption of cesium and strontium on kaolinite powders was investigated as a means to minimize the emissions of these metals during certain high temperature processes currently being developed to isolate and dispose of radiological and mixed wastes. In this work, non-radioactive aqueous cesium acetate or strontium acetate was atomized down the center of a natural gas flame supported on a variable-swirl burner in a refractory-lined laboratory-scale combustion facility. Kaolinite powder was injected at a post-flame location in the combustor. Cesium readily vaporizes in the high temperature regions of the combustor, but was reactively scavenged onto dispersed kaolinite. Global sorption mechanisms of cesium vapor on kaolinite were quantified, and are related to those available in the literature for sodium and lead. Both metal adsorption and substrate deactivation steps are important, and so there is an optimum temperature, between 1400 and 1500 K, at which maximum sorption occurs. The presence of chlorine inhibits cesium sorption. In contrast to cesium, and in the absence of chlorine, strontium was only partially vaporized and was, therefore, only partially scavengeable. The strontium data did not allow quantification of global kinetic mechanisms of interaction, although equilibrium arguments provided insight into the effects of chlorine on strontium sorption. These results have implications for the use of sorbents to control cesium and strontium emissions during high temperature waste processing including incineration and vitrification.
Date: December 16, 2004
Creator: Linak, William
Partner: UNT Libraries Government Documents Department

Full fuel-cycle comparison of forklift propulsion systems.

Description: Hydrogen has received considerable attention as an alternative to fossil fuels. The U.S. Department of Energy (DOE) investigates the technical and economic feasibility of promising new technologies, such as hydrogen fuel cells. A recent report for DOE identified three near-term markets for fuel cells: (1) Emergency power for state and local emergency response agencies, (2) Forklifts in warehousing and distribution centers, and (3) Airport ground support equipment markets. This report examines forklift propulsion systems and addresses the potential energy and environmental implications of substituting fuel-cell propulsion for existing technologies based on batteries and fossil fuels. Industry data and the Argonne Greenhouse Gases, Regulated Emissions, and Energy Use in Transportation (GREET) model are used to estimate full fuel-cycle emissions and use of primary energy sources, back to the primary feedstocks for fuel production. Also considered are other environmental concerns at work locations. The benefits derived from using fuel-cell propulsion are determined by the sources of electricity and hydrogen. In particular, fuel-cell forklifts using hydrogen made from the reforming of natural gas had lower impacts than those using hydrogen from electrolysis.
Date: November 5, 2008
Creator: Gaines, L. L.; Elgowainy, A.; Wang, M. Q. & Systems, Energy
Partner: UNT Libraries Government Documents Department

Fuel cycle comparison of distributed power generation technologies.

Description: The fuel-cycle energy use and greenhouse gas (GHG) emissions associated with the application of fuel cells to distributed power generation were evaluated and compared with the combustion technologies of microturbines and internal combustion engines, as well as the various technologies associated with grid-electricity generation in the United States and California. The results were primarily impacted by the net electrical efficiency of the power generation technologies and the type of employed fuels. The energy use and GHG emissions associated with the electric power generation represented the majority of the total energy use of the fuel cycle and emissions for all generation pathways. Fuel cell technologies exhibited lower GHG emissions than those associated with the U.S. grid electricity and other combustion technologies. The higher-efficiency fuel cells, such as the solid oxide fuel cell (SOFC) and molten carbonate fuel cell (MCFC), exhibited lower energy requirements than those for combustion generators. The dependence of all natural-gas-based technologies on petroleum oil was lower than that of internal combustion engines using petroleum fuels. Most fuel cell technologies approaching or exceeding the DOE target efficiency of 40% offered significant reduction in energy use and GHG emissions.
Date: December 8, 2008
Creator: Elgowainy, A.; Wang, M. Q. & Systems, Energy
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

Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

Description: This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high-flow operation to natural circulation. Low-flow coolant events are the most difficult to design for because they involve the most complex thermal-hydraulic behavior induced by the dominance of thermal-buoyancy forces acting on the ...
Date: June 30, 2007
Creator: Kasza, K.; Grandy, C.; Chang, Y. & Khalil, H.
Partner: UNT Libraries Government Documents Department

Preliminary report on blending strategies for inert-matrix fuel recycling in LWRs.

Description: Various recycle strategies have been proposed to manage the inventory of transuranics in commercial spent nuclear fuel (CSNF), with a particular goal of increasing the loading capacity of spent fuel and reprocessing wastes in the Yucca Mountain repository. Transuranic recycling in commercial LWRs can be seen as a viable means of slowing the accumulation of transuranics in the nationwide CSNF stockpile. Furthermore, this type of approach is an important first step in demonstrating the benefits of a nuclear fuel cycle which incorporates recycling, such as envisioned for Generation-IV reactor systems under development. Recycling strategies of this sort are not proposed as an attempt to eliminate the need of a geologic nuclear waste repository, but as a means to enhance the usefulness of the repository currently under construction in the U.S., perhaps circumventing the need for a second facility. A US-DOE Secretarial recommendation on the need for the construction of a second geologic repository is required by 2010. The Advanced Fuel Cycle Initiative (AFCI) has supported a breadth of work to evaluate the ideal transuranic separation and recycle strategy. Previous AFCI studies of LWR-based transmutation have considered the benefits of homogeneously recycling plutonium, plutonium and neptunium, and all transuranic (TRU) species. A study of a wide range of hypothetical separation schemes (Pu, Pu+Np, Pu+Np+Am, etc.) with multi-recycling has also been performed, focusing on the proliferation resistance of the various fuel cycles and fuel handling issues. The direct recycle of the recovered TRU from spent inert-matrix fuel (IMF) into new IMF was found to be quite limited due to the rapid burndown of the fissile plutonium. The IMF is very effective at destroying the fissile fraction of the TRU with destruction rates in excess of 80% of the fissile material without recycling the IMF. Blending strategies have been proposed to mitigate the ...
Date: April 29, 2005
Creator: Hoffman, E. A.
Partner: UNT Libraries Government Documents Department

Summary report on safety and licensing strategy support for the ABR prototype.

Description: Argonne National Laboratory is providing support to the US Department of Energy in the Global Nuclear Energy Partnership (GNEP) in certification of an advanced, sodium-cooled fast reactor. The reactor is to be constructed as a prototype for future commercial power reactors that will produce electricity while consuming actinides recovered from light water reactor spent fuel. This prototype reactor has been called the Advanced Burner Reactor, or ABR, and is now often referred to as the advanced recycle reactor. As part of its activities, Argonne is providing technical services to assist definition of a safety and licensing strategy for the ABR prototype, and to further implementation of the strategy. In FY06, an organizational meeting was held for DOE and its laboratory contractors to discuss licensing alternatives and review previous licensing experience for the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor Plant (CRBRP). Near the end of FY06, a report summarizing the discussions and conclusions was written. One of the top-level conclusions recorded in the report was a recommendation to follow a licensing strategy that included the US Nuclear Regulatory Commission (NRC) as the regulatory review and licensing authority. In FY07, activities ar Argonne to support safety and licensing progress have continued. These activities have focused on further evaluation of licensing alternatives; assessment of design, analysis, and documentation implications of licensing paths; and initial technical interactions with the Nuclear Regulatory Commission. This report summarizes FY07 activities.
Date: September 18, 2007
Creator: Cahalan, J. E.
Partner: UNT Libraries Government Documents Department

Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

Description: Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of the structural materials needed to ensure their safe ...
Date: August 1, 2008
Creator: Corwin, William R; Burchell, Timothy D; Katoh, Yutai; McGreevy, Timothy E; Nanstad, Randy K; Ren, Weiju et al.
Partner: UNT Libraries Government Documents Department

Roadmap to an Engineering-Scale Nuclear Fuel Performance & Safety Code

Description: Developing new fuels and qualifying them for large-scale deployment in power reactors is a lengthy and expensive process, typically spanning a period of two decades from concept to licensing. Nuclear fuel designers serve an indispensable role in the process, at the initial exploratory phase as well as in analysis of the testing results. In recent years fuel performance capabilities based on first principles have been playing more of a role in what has traditionally been an empirically dominated process. Nonetheless, nuclear fuel behavior is based on the interaction of multiple complex phenomena, and recent evolutionary approaches are being applied more on a phenomenon-by-phenomenon basis, targeting localized problems, as opposed to a systematic approach based on a fundamental understanding of all interacting parameters. Advanced nuclear fuels are generally more complex, and less understood, than the traditional fuels used in existing reactors (ceramic UO{sub 2} with burnable poisons and other minor additives). The added challenges are primarily caused by a less complete empirical database and, in the case of recycled fuel, the inherent variability in fuel compositions. It is clear that using the traditional approach to develop and qualify fuels over the entire range of variables pertinent to the U.S. Department of Energy (DOE) Office of Nuclear Energy on a timely basis with available funds would be very challenging, if not impossible. As a result the DOE Office of Nuclear Energy has launched the Nuclear Energy Advanced Modeling and Simulation (NEAMS) approach to revolutionize fuel development. This new approach is predicated upon transferring the recent advances in computational sciences and computer technologies into the fuel development program. The effort will couple computational science with recent advances in the fundamental understanding of physical phenomena through ab initio modeling and targeted phenomenological testing to leapfrog many fuel-development activities. Realizing the full benefits of this ...
Date: September 1, 2009
Creator: Turner, John A; Clarno, Kevin T & Hansen, Glen A
Partner: UNT Libraries Government Documents Department

ADVANCED CERAMIC MATERIALS FOR NEXT-GENERATION NUCLEAR APPLICATIONS

Description: Rising global energy demands coupled with increased environmental concerns point to one solution; they must reduce their dependence on fossil fuels that emit greenhouse gases. As the global community faces the challenge of maintaining sovereign nation security, reducing greenhouse gases, and addressing climate change nuclear power will play a significant and likely growing role. In the US, nuclear energy already provides approximately one-fifth of the electricity used to power factories, offices, homes, and schools with 104 operating nuclear power plants, located at 65 sites in 31 states. Additionally, 19 utilities have applied to the US Nuclear Regulatory Commission (NRC) for construction and operating licenses for 26 new reactors at 17 sites. This planned growth of nuclear power is occurring worldwide and has been termed the 'nuclear renaissance.' As major industrial nations craft their energy future, there are several important factors that must be considered about nuclear energy: (1) it has been proven over the last 40 years to be safe, reliable and affordable (good for Economic Security); (2) its technology and fuel can be domestically produced or obtained from allied nations (good for Energy Security); and (3) it is nearly free of greenhouse gas emissions (good for Environmental Security). Already an important part of worldwide energy security via electricity generation, nuclear energy can also potentially play an important role in industrial processes and supporting the nation's transportation sector. Coal-to-liquid processes, the generation of hydrogen and supporting the growing potential for a greatly increased electric transportation system (i.e. cars and trains) mean that nuclear energy could see dramatic growth in the near future as we seek to meet our growing demand for energy in cleaner, more secure ways. In order to address some of the prominent issues associated with nuclear power generation (i.e., high capital costs, waste management, and proliferation), the ...
Date: September 29, 2010
Creator: Marra, J.
Partner: UNT Libraries Government Documents Department

PASSIVATION LAYER STABILITY OF A METALLIC ALLOY WASTE FORM

Description: Alloy waste form development under the Waste Forms Campaign of the DOE-NE Fuel Cycle Research & Development program includes the process development and characterization of an alloy system to incorporate metal species from the waste streams generated during nuclear fuel recycling. This report describes the tests and results from the FY10 activities to further investigate an Fe-based waste form that uses 300-series stainless steel as the base alloy in an induction furnace melt process to incorporate the waste species from a closed nuclear fuel recycle separations scheme. This report is focused on the initial activities to investigate the formation of oxyhydroxide layer(s) that would be expected to develop on the Fe-based waste form as it corrodes under aqueous repository conditions. Corrosion tests were used to evaluate the stability of the layer(s) that can act as a passivation layer against further corrosion and would affect waste form durability in a disposal environment.
Date: August 16, 2010
Creator: Williamson, M.; Mickalonis, J.; Fisher, D. & Sindelar, R.
Partner: UNT Libraries Government Documents Department

FUEL CYCLE POTENTIAL WASTE FOR DISPOSITION

Description: The United States (U.S.) currently utilizes a once-through fuel cycle where used nuclear fuel (UNF) is stored on-site in either wet pools or in dry storage systems with ultimate disposal in a deep mined geologic repository envisioned. Within the Department of Energy's (DOE) Office of Nuclear Energy (DOE-NE), the Fuel Cycle Research and Development Program (FCR&D) develops options to the current commercial fuel cycle management strategy to enable the safe, secure, economic, and sustainable expansion of nuclear energy while minimizing proliferation risks by conducting research and development of advanced fuel cycles, including modified open and closed cycles. The safe management and disposition of used nuclear fuel and/or nuclear waste is a fundamental aspect of any nuclear fuel cycle. Yet, the routine disposal of used nuclear fuel and radioactive waste remains problematic. Advanced fuel cycles will generate different quantities and forms of waste than the current LWR fleet. This study analyzes the quantities and characteristics of potential waste forms including differing waste matrices, as a function of a variety of potential fuel cycle alternatives including: (1) Commercial UNF generated by uranium fuel light water reactors (LWR). Four once through fuel cycles analyzed in this study differ by varying the assumed expansion/contraction of nuclear power in the U.S; (2) Four alternative LWR used fuel recycling processes analyzed differ in the reprocessing method (aqueous vs. electro-chemical), complexity (Pu only or full transuranic (TRU) recovery) and waste forms generated; (3) Used Mixed Oxide (MOX) fuel derived from the recovered Pu utilizing a single reactor pass; and (4) Potential waste forms generated by the reprocessing of fuels derived from recovered TRU utilizing multiple reactor passes.
Date: October 13, 2010
Creator: Jones, R. & Carter, J.
Partner: UNT Libraries Government Documents Department

ZPR-3 Assembly 11 : A cylindrical sssembly of highly enriched uranium and depleted uranium with an average {sup 235}U enrichment of 12 atom % and a depleted uranium reflector.

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 for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP 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. ZPR-3 Assembly 11 (ZPR-3/11) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 12 at.% and a depleted uranium reflector. Approximately 79.7% of the total fissions in this assembly occur above 100 keV, approximately 20.3% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor ...
Date: September 30, 2010
Creator: Lell, R. M.; McKnight, R. D.; Tsiboulia, A.; Rozhikhin, Y.; Security, National & Engineering, Inst. of Physics and Power
Partner: UNT Libraries Government Documents Department

ZPR-3 Assembly 12 : A cylindrical assembly of highly enriched uranium, depleted uranium and graphite with an average {sup 235}U enrichment of 21 atom %.

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 for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP 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. ZPR-3 Assembly 12 (ZPR-3/12) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 21 at.%. Approximately 68.9% of the total fissions in this assembly occur above 100 keV, approximately 31.1% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 9 in the ...
Date: September 30, 2010
Creator: Lell, R. M.; McKnight, R. D.; Perel, R. L.; Wagschal, J. J.; Division, Nuclear Engineering & Physics, Racah Inst. of
Partner: UNT Libraries Government Documents Department

ZPR-3 Assembly 6F : A spherical assembly of highly enriched uranium, depleted uranium, aluminum and steel with an average {sup 235}U enrichment of 47 atom %.

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 for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP 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. ZPR-3 Assembly 6 consisted of six phases, A through F. In each phase a critical configuration was constructed to simulate a very simple shape such as a slab, cylinder or sphere that could be analyzed with the limited analytical tools available in the 1950s. In each case the configuration consisted of a core region of metal plates surrounded by a thick depleted uranium metal reflector. The average ...
Date: September 30, 2010
Creator: Lell, R. M.; McKnight, R. D; Schaefer, R. W. & Division, Nuclear Engineering
Partner: UNT Libraries Government Documents Department