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The Path to Sustainable Nuclear Energy. Basic and Applied Research Opportunities for Advanced Fuel Cycles

Description: The objective of this report is to identify new basic science that will be the foundation for advances in nuclear fuel-cycle technology in the near term, and for changing the nature of fuel cycles and of the nuclear energy industry in the long term. The goals are to enhance the development of nuclear energy, to maximize energy production in nuclear reactor parks, and to minimize radioactive wastes, other environmental impacts, and proliferation risks. The limitations of the once-through fuel cycle can be overcome by adopting a closed fuel cycle, in which the irradiated fuel is reprocessed and its components are separated into streams that are recycled into a reactor or disposed of in appropriate waste forms. The recycled fuel is irradiated in a reactor, where certain constituents are partially transmuted into heavier isotopes via neutron capture or into lighter isotopes via fission. Fast reactors are required to complete the transmutation of long-lived isotopes. Closed fuel cycles are encompassed by the Department of Energy?s Advanced Fuel Cycle Initiative (AFCI), to which basic scientific research can contribute. Two nuclear reactor system architectures can meet the AFCI objectives: a ?single-tier? system or a ?dual-tier? system. Both begin with light water reactors and incorporate fast reactors. The ?dual-tier? systems transmute some plutonium and neptunium in light water reactors and all remaining transuranic elements (TRUs) in a closed-cycle fast reactor. Basic science initiatives are needed in two broad areas: ? Near-term impacts that can enhance the development of either ?single-tier? or ?dual-tier? AFCI systems, primarily within the next 20 years, through basic research. Examples: Dissolution of spent fuel, separations of elements for TRU recycling and transmutation Design, synthesis, and testing of inert matrix nuclear fuels and non-oxide fuels Invention and development of accurate on-line monitoring systems for chemical and nuclear species in the nuclear fuel ...
Date: September 1, 2005
Creator: Finck, P.; Edelstein, N.; Allen, T.; Burns, C.; Chadwick, M.; Corradini, M. et al.
Partner: UNT Libraries Government Documents Department

Flooding Experiments and Modeling for Improved Reactor Safety

Description: Countercurrent two-phase flow and “flooding” phenomena in light water reactor systems are being investigated experimentally and analytically to improve reactor safety of current and future reactors. The aspects that will be better clarified are the effects of condensation and tube inclination on flooding in large diameter tubes. The current project aims to improve the level of understanding of flooding mechanisms and to develop an analysis model for more accurate evaluations of flooding in the pressurizer surge line of a Pressurized Water Reactor (PWR). Interest in flooding has recently increased because Countercurrent Flow Limitation (CCFL) in the AP600 pressurizer surge line can affect the vessel refill rate following a small break LOCA and because analysis of hypothetical severe accidents with the current flooding models in reactor safety codes shows that these models represent the largest uncertainty in analysis of steam generator tube creep rupture. During a hypothetical station blackout without auxiliary feedwater recovery, should the hot leg become voided, the pressurizer liquid will drain to the hot leg and flooding may occur in the surge line. The flooding model heavily influences the pressurizer emptying rate and the potential for surge line structural failure due to overheating and creep rupture. The air-water test results in vertical tubes are presented in this paper along with a semi-empirical correlation for the onset of flooding. The unique aspects of the study include careful experimentation on large-diameter tubes and an integrated program in which air-water testing provides benchmark knowledge and visualization data from which to conduct steam-water testing.
Date: September 14, 2008
Creator: Solmos, M., Hogan, K.J., VIerow, K.
Partner: UNT Libraries Government Documents Department

COUNTERCURRENT FLOW LIMITATION EXPERIMENTS AND MODELING FOR IMPROVED REACTOR SAFETY

Description: This project is investigating countercurrent flow and “flooding” phenomena in light water reactor systems to improve reactor safety of current and future reactors. To better understand the occurrence of flooding in the surge line geometry of a PWR, two experimental programs were performed. In the first, a test facility with an acrylic test section provided visual data on flooding for air-water systems in large diameter tubes. This test section also allowed for development of techniques to form an annular liquid film along the inner surface of the “surge line” and other techniques which would be difficult to verify in an opaque test section. Based on experiences in the air-water testing and the improved understanding of flooding phenomena, two series of tests were conducted in a large-diameter, stainless steel test section. Air-water test results and steam-water test results were directly compared to note the effect of condensation. Results indicate that, as for smaller diameter tubes, the flooding phenomena is predominantly driven by the hydrodynamics. Tests with the test sections inclined were attempted but the annular film was easily disrupted. A theoretical model for steam venting from inclined tubes is proposed herein and validated against air-water data. Empirical correlations were proposed for air-water and steam-water data. Methods for developing analytical models of the air-water and steam-water systems are discussed, as is the applicability of the current data to the surge line conditions. This report documents the project results from July 1, 2005 through June 30, 2008.
Date: September 26, 2008
Creator: Vierow, Karen
Partner: UNT Libraries Government Documents Department

Integration Strategy for DB-MHR TRISO Fuel production in conjunction with MOX Fuel production

Description: One of the nuclear power options for the future involves the evolution of gas cooled reactors to support the likely high temperature operations needed for commercial scale hydrogen production. One such proposed option is to use a Gas Turbine Modular Helium Reactor fueled with uranium based TRISO (coated particle) fuel. It has also been suggested that such a MHR could be operated in a ''Deep Burn'' manner fueled with TRISO fuel produced from recycle spent nuclear fuel. This concept known as a DBMHR must withstand significant development and fuel fabrication cost to be economically viable. The purpose of this report is to consider and propose a strategy where synergy with a parallel MOX fuel to LWR program provides economic or other advantage for either or both programs. A strategy involving three phases has been envisioned with potential for economic benefit relative to a stand-alone TRISO/DBMHR program. Such a strategy and related timing will ultimately be driven by economics, but is offered here for consideration of value to the total AFCI program. Phase I Near-term. Conventional spent fuel aqueous processing, MOX fuel fabrication, and use of present and future LWR/ALWR's with objective of a ''Continuous Recycle'' mode of fuel cycle management. Phase II Intermediate. Augmentation of LWR/ALWR industry with MHR deployment as justified by hydrogen economy and/or electrical demand. Phase III Long-term. Introduction of DBMHR's to offer alternative method for transuranic destruction and associated repository benefits, in addition to Phase II benefits. The basic philosophy of this strategy appears sound. However, the details of the technology plans and economic evaluations should receive additional detail and evaluation in the next fiscal year as funding can support.
Date: September 30, 2005
Creator: MCGUIRE, DAVID
Partner: UNT Libraries Government Documents Department

University Reactor Conversion Lessons Learned Workshop for Purdue University Reactor

Description: The Department of Energy’s Idaho National Laboratory, under its programmatic responsibility for managing the University Research Reactor Conversions, has completed the conversion of the reactor at Purdue University Reactor. With this work completed and in anticipation of other impending conversion projects, the INL convened and engaged the project participants in a structured discussion to capture the lessons learned. The lessons learned process has allowed us to capture gaps, opportunities, and good practices, drawing from the project team’s experiences. These lessons will be used to raise the standard of excellence, effectiveness, and efficiency in all future conversion projects.
Date: September 1, 2008
Creator: Woolstenhulme, Eric C. & Hewit, Dana M.
Partner: UNT Libraries Government Documents Department

Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide

Description: This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactors’ spent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.
Date: September 1, 2012
Creator: Castle, Brian K.; Hoiland, Shauna A.; Rankin, Richard A. & Sterbentz, James W.
Partner: UNT Libraries Government Documents Department

Material Performance of Fully-Ceramic Micro-Encapsulated Fuel under Selected LWR Design Basis Scenarios: Final Report

Description: The extension to LWRs of the use of Deep-Burn coated particle fuel envisaged for HTRs has been investigated. TRISO coated fuel particles are used in Fully-Ceramic Microencapsulated (FCM) fuel within a SiC matrix rather than the graphite of HTRs. TRISO particles are well characterized for uranium-fueled HTRs. However, operating conditions of LWRs are different from those of HTRs (temperature, neutron energy spectrum, fast fluence levels, power density). Furthermore, the time scales of transient core behavior during accidents are usually much shorter and thus more severe in LWRs. The PASTA code was updated for analysis of stresses in coated particle FCM fuel. The code extensions enable the automatic use of neutronic data (burnup, fast fluence as a function of irradiation time) obtained using the DRAGON neutronics code. An input option for automatic evaluation of temperature rise during anticipated transients was also added. A new thermal model for FCM was incorporated into the code; so-were updated correlations (for pyrocarbon coating layers) suitable to estimating dimensional changes at the high fluence levels attained in LWR DB fuel. Analyses of the FCM fuel using the updated PASTA code under nominal and accident conditions show: (1) Stress levels in SiC-coatings are low for low fission gas release (FGR) fractions of several percent, as based on data of fission gas diffusion in UO{sub 2} kernels. However, the high burnup level of LWR-DB fuel implies that the FGR fraction is more likely to be in the range of 50-100%, similar to Inert Matrix Fuels (IMFs). For this range the predicted stresses and failure fractions of the SiC coating are high for the reference particle design (500 {micro}mm kernel diameter, 100 {micro}mm buffer, 35 {micro}mm IPyC, 35 {micro}mm SiC, 40 {micro}mm OPyC). A conservative case, assuming 100% FGR, 900K fuel temperature and 705 MWd/kg (77% FIMA) fuel burnup, ...
Date: September 1, 2011
Creator: Boer, B.; Sen, R. S.; Pope, M. A. & Ougouag, A. M.
Partner: UNT Libraries Government Documents Department

Small Break Air Ingress Experiment

Description: The small break air-ingress experiment, described in this report, is designed to investigate air-ingress phenomena postulated to occur in pipes in a very high temperature gas-cooled reactor (VHTRs). During this experiment, air-ingress rates were measured for various flow and break conditions through small holes drilled into a pipe of the experimental apparatus. The holes were drilled at right angles to the pipe wall such that a direction vector drawn from the pipe centerline to the center of each hole was at right angles with respect to the pipe centerline. Thus the orientation of each hole was obtained by measuring the included angle between the direction vector of each hole with respect to a reference line anchored on the pipe centerline and pointing in the direction of the gravitational force. Using this reference system, the influence of several important parameters on the air ingress flow rate were measured including break orientation, break size, and flow velocity . The approach used to study the influence of these parameters on air ingress is based on measuring the changes in oxygen concentrations at various locations in the helium flow circulation system as a function of time using oxygen sensors (or detectors) to estimate the air-ingress rates through the holes. The test-section is constructed of a stainless steel pipe which had small holes drilled at the desired locations.
Date: September 1, 2011
Creator: Oh, Chang & Kim, Eung Soo
Partner: UNT Libraries Government Documents Department

Summary of Planned Implementation for the HTGR Lessons Learned Applicable to the NGNP

Description: This document presents a reconciliation of the lessons learned during a 2010 comprehensive evaluation of pertinent lessons learned from past and present high temperature gas-cooled reactors that apply to the Next Generation Nuclear Plant Project along with current and planned activities. The data used are from the latest Idaho National Laboratory research and development plans, the conceptual design report from General Atomics, and the pebble bed reactor technology readiness study from AREVA. Only those lessons related to the structures, systems, and components of the Next Generation Nuclear Plant (NGNP), as documented in the recently updated lessons learned report are addressed. These reconciliations are ordered according to plant area, followed by the affected system, subsystem, or component; lesson learned; and finally an NGNP implementation statement. This report (1) provides cross references to the original lessons learned document, (2) describes the lesson learned, (3) provides the current NGNP implementation status with design data needs associated with the lesson learned, (4) identifies the research and development being performed related to the lesson learned, and (5) summarizes with a status of how the lesson learned has been addressed by the NGNP Project.
Date: September 1, 2011
Creator: Mckirdy, Ian
Partner: UNT Libraries Government Documents Department

Strategic Plan for Nuclear Energy -- Knowledge Base for Advanced Modeling and Simulation (NE-KAMS)

Description: NE-KAMS knowledge base will assist computational analysts, physics model developers, experimentalists, nuclear reactor designers, and federal regulators by: (1) Establishing accepted standards, requirements and best practices for V&V and UQ of computational models and simulations, (2) Establishing accepted standards and procedures for qualifying and classifying experimental and numerical benchmark data, (3) Providing readily accessible databases for nuclear energy related experimental and numerical benchmark data that can be used in V&V assessments and computational methods development, (4) Providing a searchable knowledge base of information, documents and data on V&V and UQ, and (5) Providing web-enabled applications, tools and utilities for V&V and UQ activities, data assessment and processing, and information and data searches. From its inception, NE-KAMS will directly support nuclear energy research, development and demonstration programs within the U.S. Department of Energy (DOE), including the Consortium for Advanced Simulation of Light Water Reactors (CASL), the Nuclear Energy Advanced Modeling and Simulation (NEAMS), the Light Water Reactor Sustainability (LWRS), the Small Modular Reactors (SMR), and the Next Generation Nuclear Power Plant (NGNP) programs. These programs all involve computational modeling and simulation (M&S) of nuclear reactor systems, components and processes, and it is envisioned that NE-KAMS will help to coordinate and facilitate collaboration and sharing of resources and expertise for V&V and UQ across these programs. In addition, from the outset, NE-KAMS will support the use of computational M&S in the nuclear industry by developing guidelines and recommended practices aimed at quantifying the uncertainty and assessing the applicability of existing analysis models and methods. The NE-KAMS effort will initially focus on supporting the use of computational fluid dynamics (CFD) and thermal hydraulics (T/H) analysis for M&S of nuclear reactor systems, components and processes, and will later expand to include materials, fuel system performance and other areas of M&S as time and funding ...
Date: September 1, 2011
Creator: Johnson, Rich; Mousseau, Kimberlyn C. & Lee, Hyung
Partner: UNT Libraries Government Documents Department

Light Water Reactor Sustainability Program Power Uprate Research and Development Strategy

Description: The economic incentives for low-cost electricity generation will continue to drive more plant owners to identify safe and reliable methods to increase the electrical power output of the current nuclear power plant fleet. A power uprate enables a nuclear power plant to increase its electrical output with low cost. However, power uprates brought new challenges to plant owners and operators. These include equipment damage or degraded performance, and unanticipated responses to plant conditions, etc. These problems have arisen mainly from using dated design and safety analysis tools and insufficient understanding of the full implications of the proposed power uprate or from insufficient attention to detail during the design and implementation phase. It is essential to demonstrate that all required safety margins have been properly retained and the existing safety level has been maintained or even increased, with consideration of all the conditions and parameters that have an influence on plant safety. The impact of the power uprate on plant life management for long term operation is also an important issue. Significant capital investments are required to extend the lifetime of an aging nuclear power plant. Power uprates can help the plant owner to recover the investment costs. However, plant aging issues may be aggravated by the power uprate due to plant conditions. More rigorous analyses, inspections and monitoring systems are required.
Date: September 1, 2011
Creator: Zhang, Hongbin
Partner: UNT Libraries Government Documents Department

Diffusion Welding of Alloys for Molten Salt Service - Status Report

Description: The present work is concerned with heat exchanger development for molten salt service, including the proposed molten salt reactor (MSR), a homogeneous reactor in which the fuel is dissolved in a circulating fluid of molten salt. It is an outgrowth of recent work done under the Next Generation Nuclear Plant (NGNP) program; what the two reactor systems have in common is an inherently safe nuclear plant with a high outlet temperature that is useful for process heat as well as more conventional generation The NGNP program was tasked with investigating the application of a new generation of nuclear power plants to a variety of energy needs. One baseline reactor design for this program is a high temperature, gas-cooled reactor (HTGR), which provides many options for energy use. These might include the conventional Rankine cycle (steam turbine) generation of electricity, but also other methods: for example, Brayton cycle (gas turbine) electrical generation, and the direct use of the high temperatures characteristic of HTGR output for process heat in the chemical industry. Such process heat is currently generated by burning fossil fuels, and is a major contributor to the carbon footprint of the chemical and petrochemical industries. The HTGR, based on graphite fuel elements, can produce very high output temperatures; ideally, temperatures of 900 °C or even greater, which has significant energy advantages. Such temperatures are, of course, at the frontiers of materials limitations, at the upper end of the performance envelope of the metallic materials for which robust construction codes exist, and within the realm of ceramic materials, the fabrication and joining of which, on the scale of large energy systems, are at an earlier stage of development. A considerable amount of work was done in the diffusion welding of materials of interest for HTGR service with alloys such as 617 ...
Date: September 1, 2012
Creator: Clark, Denis; Mizia, Ronald & Sabharwall, Piyush
Partner: UNT Libraries Government Documents Department

Development Plan for the Fuel Cycle Simulator

Description: The Fuel Cycle Simulator (FCS) project was initiated late in FY-10 as the activity to develop a next generation fuel cycle dynamic analysis tool for achieving the Systems Analysis Campaign 'Grand Challenge.' This challenge, as documented in the Campaign Implementation Plan, is to: 'Develop a fuel cycle simulator as part of a suite of tools to support decision-making, communication, and education, that synthesizes and visually explains the multiple attributes of potential fuel cycles.'
Date: September 1, 2011
Creator: Dixon, Brent
Partner: UNT Libraries Government Documents Department

Site Selection & Characterization Status Report for Next Generation Nuclear Plant (NGNP)

Description: In the near future, the US Department of Energy (DOE) will need to make important decisions regarding design and construction of the Next Generation Nuclear Plant (NGNP). One part of making these decisions is considering the potential environmental impacts that this facility may have, if constructed here at the Idaho National Laboratory (INL). The National Environmental Policy Act (NEPA) of 1969 provides DOE decision makers with a process to systematically consider potential environmental consequences of agency decisions. In addition, the Energy Policy Act of 2005 (Title VI, Subtitel C, Section 644) states that the 'Nuclear Regulatory Commission (NRC) shall have licensing and regulatory authority for any reactor authorized under this subtitle.' This stipulates that the NRC will license the NGNP for operation. The NRC NEPA Regulations (10 CFR Part 51) require tha thte NRC prepare an Environmental Impact Statement (EIS) for a permit to construct a nuclear power plant. The applicant is required to submit an Environmental report (ER) to aid the NRC in complying with NEPA.
Date: September 1, 2007
Creator: Holbrook, Mark
Partner: UNT Libraries Government Documents Department

Verification and Validation Strategy for LWRS Tools

Description: One intension of the Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) program is to create advanced computational tools for safety assessment that enable more accurate representation of a nuclear power plant safety margin. These tools are to be used to study the unique issues posed by lifetime extension and relicensing of the existing operating fleet of nuclear power plants well beyond their first license extension period. The extent to which new computational models / codes such as RELAP-7 can be used for reactor licensing / relicensing activities depends mainly upon the thoroughness with which they have been verified and validated (V&V). This document outlines the LWRS program strategy by which RELAP-7 code V&V planning is to be accomplished. From the perspective of developing and applying thermal-hydraulic and reactivity-specific models to reactor systems, the US Nuclear Regulatory Commission (NRC) Regulatory Guide 1.203 gives key guidance to numeric model developers and those tasked with the validation of numeric models. By creating Regulatory Guide 1.203 the NRC defined a framework for development, assessment, and approval of transient and accident analysis methods. As a result, this methodology is very relevant and is recommended as the path forward for RELAP-7 V&V. However, the unique issues posed by lifetime extension will require considerations in addition to those addressed in Regulatory Guide 1.203. Some of these include prioritization of which plants / designs should be studied first, coupling modern supporting experiments to the stringent needs of new high fidelity models / codes, and scaling of aging effects.
Date: September 1, 2012
Creator: Stoots, Carl M.; Schultz, Richard R.; Gougar, Hans D.; Larson, Thomas K; Corradini, Michael; Swiler, Laura et al.
Partner: UNT Libraries Government Documents Department

ASME Code Efforts Supporting HTGRs

Description: In 1999, an international collaborative initiative for the development of advanced (Generation IV) reactors was started. The idea behind this effort was to bring nuclear energy closer to the needs of sustainability, to increase proliferation resistance, and to support concepts able to produce energy (both electricity and process heat) at competitive costs. The U.S. Department of Energy has supported this effort by pursuing the development of the Next Generation Nuclear Plant, a high temperature gas-cooled reactor. This support has included research and development of pertinent data, initial regulatory discussions, and engineering support of various codes and standards development. This report discusses the various applicable American Society of Mechanical Engineers (ASME) codes and standards that are being developed to support these high temperature gascooled reactors during construction and operation. ASME is aggressively pursuing these codes and standards to support an international effort to build the next generation of advanced reactors so that all can benefit.
Date: September 1, 2011
Creator: Morton, D.K.
Partner: UNT Libraries Government Documents Department

Work Breakdown Structure and Plant/Equipment Designation System Numbering Scheme for the High Temperature Gas- Cooled Reactor (HTGR) Component Test Capability (CTC)

Description: This white paper investigates the potential integration of the CTC work breakdown structure numbering scheme with a plant/equipment numbering system (PNS), or alternatively referred to in industry as a reference designation system (RDS). Ideally, the goal of such integration would be a single, common referencing system for the life cycle of the CTC that supports all the various processes (e.g., information, execution, and control) that necessitate plant and equipment numbers be assigned. This white paper focuses on discovering the full scope of Idaho National Laboratory (INL) processes to which this goal might be applied as well as the factors likely to affect decisions about implementation. Later, a procedure for assigning these numbers will be developed using this white paper as a starting point and that reflects the resolved scope and outcome of associated decisions.
Date: September 1, 2009
Creator: Bryan, Jeffrey D
Partner: UNT Libraries Government Documents Department

Information Foraging in Nuclear Power Plant Control Rooms

Description: nformation foraging theory articulates the role of the human as an 'informavore' that seeks information and follows optimal foraging strategies (i.e., the 'information scent') to find meaningful information. This paper briefly reviews the findings from information foraging theory outside the nuclear domain and then discusses the types of information foraging strategies operators employ for normal and off-normal operations in the control room. For example, operators may employ a predatory 'wolf' strategy of hunting for information in the face of a plant upset. However, during routine operations, the operators may employ a trapping 'spider' strategy of waiting for relevant indicators to appear. This delineation corresponds to information pull and push strategies, respectively. No studies have been conducted to determine explicitly the characteristics of a control room interface that is optimized for both push and pull information foraging strategies, nor has there been empirical work to validate operator performance when transitioning between push and pull strategies. This paper explores examples of control room operators as wolves vs. spiders and con- cludes by proposing a set of research questions to investigate information foraging in control room settings.
Date: September 1, 2011
Creator: Boring, R.L.
Partner: UNT Libraries Government Documents Department

Influence of Hold Time on Creep-Fatigue Behavior of an Advanced Austenitic Alloy

Description: An advanced austenitic alloy, HT-UPS (high temperature-ultrafine precipitate strengthened), is a candidate material for the structural components of fast reactors and energy-conversion systems. HT-UPS provides improved creep resistance through a composition based on 316 stainless steel (SS) with additions of Ti and Nb to form nano-scale MC precipitates in the austenitic matrix. The low cycle fatigue and creep-fatigue behavior of a HT-UPS alloy has been investigated at 650 C, 1.0% total strain, and an R ratio of -1 with hold times as long as 9000 sec at peak tensile strain. The cyclic deformation response of HT-UPS is compared to that of 316 SS. The cycles to failure are similar, despite differences in peak stress profiles and the deformed microstructures. Cracking in both alloys is transgranular (initiation and propagation) in the case of continuous cycle fatigue, while the primary cracks also propagate transgranularly during creep-fatigue cycling. Internal grain boundary damage as a result of the tensile hold is present in the form of fine cracks for hold times of 3600 sec and longer and substantially more internal cracks are visible in 316 SS than HT-UPS. The dislocation substructures observed in the deformed material are different. An equiaxed cellular structure is observed in 316 SS, whereas tangles of dislocations are present at the nanoscale MC precipitates in HT-UPS and no cellular substructure is observed.
Date: September 1, 2011
Creator: Carroll, Mark & Carroll, Laura
Partner: UNT Libraries Government Documents Department

Integration of High-Temperature Gas-Cooled Reactors into Industrial Process Applications

Description: This report is a summary of analyses performed by the NGNP project to determine whether it is technically and economically feasible to integrate high temperature gas cooled reactor (HTGR) technology into industrial processes. To avoid an overly optimistic environmental and economic baseline for comparing nuclear integrated and conventional processes, a conservative approach was used for the assumptions and calculations.
Date: September 1, 2011
Creator: Nelson, Lee
Partner: UNT Libraries Government Documents Department

Plan for Demonstration of Online Monitoring for the Light Water Reactor Sustainability Online Monitoring Project

Description: Condition based online monitoring technologies and development of diagnostic and prognostic methodologies have drawn tremendous interest in the nuclear industry. It has become important to identify and resolve problems with structures, systems, and components (SSCs) to ensure plant safety, efficiency, and immunity to accidents in the aging fleet of reactors. The Machine Condition Monitoring (MCM) test bed at INL will be used to demonstrate the effectiveness to advancement in online monitoring, sensors, diagnostic and prognostic technologies on a pilot-scale plant that mimics the hydraulics of a nuclear plant. As part of this research project, INL will research available prognostics architectures and their suitability for deployment in a nuclear power plant. In addition, INL will provide recommendation to improve the existing diagnostic and prognostic architectures based on the experimental analysis performed on the MCM test bed.
Date: September 1, 2011
Creator: Tawfik, Magdy S.; Agarwal, Vivek & Lybeck, Nancy J.
Partner: UNT Libraries Government Documents Department

Advanced Fuels Campaign Execution Plan

Description: The purpose of the Advanced Fuels Campaign (AFC) Execution Plan is to communicate the structure and management of research, development, and demonstration (RD&D) activities within the Fuel Cycle Research and Development (FCRD) program. Included in this document is an overview of the FCRD program, a description of the difference between revolutionary and evolutionary approaches to nuclear fuel development, the meaning of science-based development of nuclear fuels, and the 'Grand Challenge' for the AFC that would, if achieved, provide a transformational technology to the nuclear industry in the form of a high performance, high reliability nuclear fuel system. The activities that will be conducted by the AFC to achieve success towards this grand challenge are described and the goals and milestones over the next 20 to 40 year period of research and development are established.
Date: September 1, 2011
Creator: Pasamehmetoglu, Kemal
Partner: UNT Libraries Government Documents Department