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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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All sourcesNASA NTRSDOE OSTI
DOE OSTI2011

Materials and Fuels Complex Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables

Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Materials and Fuels Complex facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facility-specific documents. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool for developing the radioactive waste management basis.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Central Facilities Area Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables

Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Central Facilities Area facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facilityspecific documents. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool for developing the radioactive waste management basis.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

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

The Nuclear Energy Computational Fluid Dynamics Advanced Modeling and Simulation (NE-CAMS) system is being developed at the Idaho National Laboratory (INL) in collaboration with Bettis Laboratory, Sandia National Laboratory (SNL), Argonne National Laboratory (ANL), Utah State University (USU), and other interested parties with the objective of developing and implementing a comprehensive and readily accessible data and information management system for computational fluid dynamics (CFD) verification and validation (V&V) in support of nuclear energy systems design and safety analysis. The two key objectives of the NE-CAMS effort are to identify, collect, assess, store and maintain high resolution and high quality experimental data and related expert knowledge (metadata) for use in CFD V&V assessments specific to the nuclear energy field and to establish a working relationship with the U.S. Nuclear Regulatory Commission (NRC) to develop a CFD V&V database, including benchmark cases, that addresses and supports the associated NRC regulations and policies on the use of CFD analysis. In particular, the NE-CAMS system will support the Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program, which aims to develop and deploy advanced modeling and simulation methods and computational tools for reliable numerical simulation of nuclear reactor systems for design and safety analysis. Primary NE-CAMS Elements There are four primary elements of the NE-CAMS knowledge base designed to support computer modeling and simulation in the nuclear energy arena as listed below. Element 1. The database will contain experimental data that can be used for CFD validation that is relevant to nuclear reactor and plant processes, particularly those important to the nuclear industry and the NRC. Element 2. Qualification standards for data evaluation and classification will be incorporated and applied such that validation data sets will result in well-defined, well-characterized data. Element 3. Standards will be established for the design and operation of experiments for the generation of new validation data sets that are to be submitted to NE-CAMS that addresses the completeness and characterization of the dataset. Element 4. Standards will be developed for performing verification and validation (V&V) to establish confidence levels in CFD analyses of nuclear reactor processes; such processes will be acceptable and recognized by both CFD experts and the NRC.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2011

A Fruit of Yucca Mountain: The Remote Waste Package Closure System

Was the death of the Yucca Mountain repository the fate of a technical lemon or a political lemon? Without caution, this debate could lure us away from capitalizing on the fruits of the project. In March 2009, Idaho National Laboratory (INL) successfully demonstrated the Waste Package Closure System, a full-scale prototype system for closing waste packages that were to be entombed in the now abandoned Yucca Mountain repository. This article describes the system, which INL designed and built, to weld the closure lids on the waste packages, nondestructively examine the welds using four different techniques, repair the welds if necessary, mitigate crack initiating stresses in the surfaces of the welds, evacuate and backfill the packages with an inert gas, and perform all of these tasks remotely. As a nation, we now have a proven method for securely sealing nuclear waste packages for long term storage—regardless of whether or not the future destination for these packages will be an underground repository. Additionally, many of the system’s features and concepts may benefit other remote nuclear applications.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Systematic Approach to the Development, Evolution, and Effectiveness of Integrated Product Development Teams (IPDTs)

Integrated Product Development Teams (IPDT) are a key component of any systems engineering (SE) application, but since they are formed primarily from technical considerations, many IPDTs are far less productive than they otherwise could be. By recognizing specific personality types and skill sets, a random group of 'technical' individuals can be structured to become a highly effective team capable of delivering much more than the sum of its members.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2011

System Verification Through Reliability, Availability, Maintainability (RAM) Analysis & Technology Readiness Levels (TRLs)

The Next Generation Nuclear Plant (NGNP) Project, managed by the Idaho National Laboratory (INL), is authored by the Energy Policy Act of 2005, to research, develop, design, construct, and operate a prototype fourth generation nuclear reactor to meet the needs of the 21st Century. A section in this document proposes that the NGNP will provide heat for process heat applications. As with all large projects developing and deploying new technologies, the NGNP is expected to meet high performance and availability targets relative to current state of the art systems and technology. One requirement for the NGNP is to provide heat for the generation of hydrogen for large scale productions and this process heat application is required to be at least 90% or more available relative to other technologies currently on the market. To reach this goal, a RAM Roadmap was developed highlighting the actions to be taken to ensure that various milestones in system development and maturation concurrently meet required availability requirements. Integral to the RAM Roadmap was the use of a RAM analytical/simulation tool which was used to estimate the availability of the system when deployed based on current design configuration and the maturation level of the system.

08 HYDROGEN↗
DOE OSTI2011

The Need for Next Generation of Radiochemists in the USA

In 2009, the nuclear industry employed approximately 120,000 people. Nearly 38 percent of the nuclear industry force will be eligible to retire within the next five years. To maintain the current work force, the industry will need to hire approximately 25,000 more workers by 2015.1 The federal government will also need nuclear workers in the future in its laboratories, the military and government programs. There is a need not only for the entire nuclear community to work with the academia to recruit and train students in a standardized way for employment at nuclear facilities. Several strategies are taking place in the USA, as an example, an initiative developed at the Idaho National Laboratory (INL) is the Institute of Nuclear Science and Technology (INEST) with four Centers of Research and Education (COREs) selected to address some of the most challenging issues facing nuclear energy today: (1) Fuels and Materials, (2) Space Nuclear Research, (3) Fuel Cycle, and (4) Safety and Licensing. Another example is the development of a radiochemistry program at two universities: the University of Nevada Las Vegas (UNLV) and Washington State University (WSU) to attract the next generation work force. This paper will solely focus on the next generation of radiochemists needed in the US and will give examples illustrating the needs as well as the current activities in the academia and in the national laboratories to fulfill national needs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Evaluation Metrics for Intermediate Heat Exchangers for Next Generation Nuclear Reactors

The Department of Energy (DOE) is working with industry to develop a next generation, high-temperature gas-cooled reactor (HTGR) as a part of the effort to supply the United States with abundant, clean, and secure energy as initiated by the Energy Policy Act of 2005 (EPAct; Public Law 109-58,2005). The NGNP Project, led by the Idaho National Laboratory (INL), will demonstrate the ability of the HTGR to generate hydrogen, electricity, and/or high-quality process heat for a wide range of industrial applications.

08 HYDROGEN↗
DOE OSTI2011

A Reconstructed Discontinuous Galerkin Method for the Compressible Flows on Unstructured Tetrahedral Grids

A reconstruction-based discontinuous Galerkin (RDG) method is presented for the solution of the compressible Navier-Stokes equations on unstructured tetrahedral grids. The RDG method, originally developed for the compressible Euler equations, is extended to discretize viscous and heat fluxes in the Navier-Stokes equations using a so-called inter-cell reconstruction, where a smooth solution is locally reconstructed using a least-squares method from the underlying discontinuous DG solution. Similar to the recovery-based DG (rDG) methods, this reconstructed DG method eliminates the introduction of ad hoc penalty or coupling terms commonly found in traditional DG methods. Unlike rDG methods, this RDG method does not need to judiciously choose a proper form of a recovered polynomial, thus is simple, flexible, and robust, and can be used on unstructured grids. The preliminary results indicate that this RDG method is stable on unstructured tetrahedral grids, and provides a viable and attractive alternative for the discretization of the viscous and heat fluxes in the Navier-Stokes equations.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2011

A Framework for Analysis of Energy-Water Interdependency Problems

The overall objective of this work is to improve the holistic value of energy development strategies by integrating management criteria for water availability, water quality, and ecosystem health into the energy system planning process. The Snake River Basin (SRB) in southern Idaho is used as a case study to show options for improving full economic utilization of aquatic resources given multiple scenarios such as changing climate, additional regulations, and increasing population. Through the incorporation of multiple management criteria, potential crosscutting solutions to energy and water issues in the SRB can be developed. The final result of this work will be a multi-criteria decision support tool - usable by policy makers and researchers alike - that will give insight into the behavior of the management criteria over time and will allow the user to experiment with a range of potential solutions. Because several basins in the arid west are dealing with similar water, energy, and ecosystem issues, the tool and conclusions will be transferrable to a wide range of locations and applications. This is a very large project to be completed in phases. This paper deals with interactions between the hydrologic system and water use at a basin level. Future work will include the interdependency between energy use and water use in these systems.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2011

Nondestructive Evaluation of Nuclear-Grade Graphite

Nondestructive Evaluation of Nuclear Grade Graphite Dennis C. Kunerth and Timothy R. McJunkin Idaho National Laboratory Idaho Falls, ID, 83415 This paper discusses the nondestructive evaluation of nuclear grade graphite performed at the Idaho National Laboratory. Graphite is a composite material highly dependent on the base material and manufacturing methods. As a result, material variations are expected within individual billets as well billet to billet and lot to lot. Several methods of evaluating the material have been explored. Particular technologies each provide a subset of information about the material. This paper focuses on techniques that are applicable to in-service inspection of nuclear energy plant components. Eddy current examination of the available surfaces provides information on potential near surface structural defects and although limited, ultrasonics can be utilized in conventional volumetric inspection. Material condition (e.g. micro-cracking and porosity induced by radiation and stress) can be derived from backscatter or acousto-ultrasound (AU) methods. Novel approaches utilizing phased array ultrasonics have been attempted to expand the abilities of AU techniques. By combining variable placement of apertures, angle and depth of focus, the techniques provide the potential to obtain parameters at various depths in the material. Initial results of the study and possible procedures for application of the techniques are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2011

Benchmark Evaluation of the NRAD Reactor LEU Core Startup Measurements

The Neutron Radiography (NRAD) reactor is a 250-kW TRIGA-(Training, Research, Isotope Production, General Atomics)-conversion-type reactor at the Idaho National Laboratory; it is primarily used for neutron radiography analysis of irradiated and unirradiated fuels and materials. The NRAD reactor was converted from HEU to LEU fuel with 60 fuel elements and brought critical on March 31, 2010. This configuration of the NRAD reactor has been evaluated as an acceptable benchmark experiment and is available in the 2011 editions of the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) and the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Significant effort went into precisely characterizing all aspects of the reactor core dimensions and material properties; detailed analyses of reactor parameters minimized experimental uncertainties. The largest contributors to the total benchmark uncertainty were the 234 U, 236 U, Er, and Hf content in the fuel; the manganese content in the stainless-steel cladding; and the unknown level of water saturation in the graphite reflector blocks. A simplified benchmark model of the NRAD reactor was prepared with a keff of 1.0012 ± 0.0029 (1σ). Monte Carlo calculations with MCNP5 and KENO-VI and various neutron cross section libraries were performed and compared with the benchmark eigenvalue for the 60-fuel-element core configuration; all calculated eigenvalues are between 0.3 and 0.8% greater than the benchmark value. Benchmark evaluations of the NRAD reactor are beneficial in understanding biases and uncertainties affecting criticality safety analyses of storage, handling, or transportation applications with LEU-Er-Zr-H fuel.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Educating Next Generation Nuclear Criticality Safety Engineers at the Idaho National Laboratory

One of the challenges in educating our next generation of nuclear safety engineers is the limitation of opportunities to receive significant experience or hands-on training prior to graduation. Such training is generally restricted to on-the-job-training before this new engineering workforce can adequately provide assessment of nuclear systems and establish safety guidelines. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) can provide students and young professionals the opportunity to gain experience and enhance critical engineering skills. The ICSBEP and IRPhEP publish annual handbooks that contain evaluations of experiments along with summarized experimental data and peer-reviewed benchmark specifications to support the validation of neutronics codes, nuclear cross-section data, and the validation of reactor designs. Participation in the benchmark process not only benefits those who use these Handbooks within the international community, but provides the individual with opportunities for professional development, networking with an international community of experts, and valuable experience to be used in future employment. Traditionally students have participated in benchmarking activities via internships at national laboratories, universities, or companies involved with the ICSBEP and IRPhEP programs. Additional programs have been developed to facilitate the nuclear education of students while participating in the benchmark projects. These programs include coordination with the Center for Space Nuclear Research (CSNR) Next Degree Program, the Collaboration with the Department of Energy Idaho Operations Office to train nuclear and criticality safety engineers, and student evaluations as the basis for their master's thesis in nuclear engineering.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2011

THERMAL PREDICTIONS OF NEW COMPOSITE MATERIAL DURING INPILE TESTING

An inpile experiment is currently underway wherein specimens comprised of a newly developed material are being irradiated at Idaho National Laboratory's Advanced Test Reactor (ATR) in conjunction with Utah State University under the auspices of the ATR National Scientific User Facility. This paper provides the thermophysical properties of this new material measured prior to irradiation. After the irradiation campaign is complete, the thermophysical properties of the specimens will be measured and compared to the preirradiation values. A finite-element model was constructed to predict bounding specimen temperatures during irradiation. Results from the thermal hydraulic modeling, including the steady-state temperatures of the specimens within sealed capsules, are presented. After the irradiation campaign is completed, best-estimate thermal predictions will be performed for the individual specimens using the actual as-run irradiation power levels.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

THE MATHEMATICAL ANALYSIS OF A NOVEL APPROACH TO MAXIMIZE WASTE RECOVERY IN A LIFE SUPPORT SYSTEM

NASA has been evaluating closed-loop atmosphere revitalization architectures that include carbon dioxide (CO2) reduction technologies. The CO2 and steam (H2O) co-electrolysis process is one of the reduction options that NASA has investigated. Utilizing recent advances in the fuel cell technology sector, the Idaho National Laboratory, INL, has developed a CO2 and H2O co-electrolysis process to produce oxygen and syngas (carbon monoxide (CO) and hydrogen (H2) mixture) for terrestrial (energy production) application. The technology is a combined process that involves steam electrolysis, CO2 electrolysis, and the reverse water gas shift (RWGS) reaction. Two process models were developed to evaluate novel approaches for energy storage and resource recovery in a life support system. In the first model, products from the INL co-electrolysis process are combined to produce methanol fuel. In the second co-electrolysis, products are separated with a pressure swing adsorption (PSA) process. In both models the fuels are burned with added oxygen to produce H2O and CO2, the original reactants. For both processes, the overall power increases as the syngas ratio, H2/CO, increases because more water is needed to produce more hydrogen at a set CO2 incoming flow rate. The power for the methanol cases is less than pressure swing adsorption, PSA, because heat is available from the methanol reactor to preheat the water and carbon dioxide entering the co-electrolysis process.

08 HYDROGEN↗
DOE OSTI2011

Advanced Test Reactor National Scientific User Facility

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

Science-Based Simulation Model of Human Performance for Human Reliability Analysis

Human reliability analysis (HRA), a component of an integrated probabilistic risk assessment (PRA), is the means by which the human contribution to risk is assessed, both qualitatively and quantitatively. However, among the literally dozens of HRA methods that have been developed, most cannot fully model and quantify the types of errors that occurred at Three Mile Island. Furthermore, all of the methods lack a solid empirical basis, relying heavily on expert judgment or empirical results derived in non-reactor domains. Finally, all of the methods are essentially static, and are thus unable to capture the dynamics of an accident in progress. The objective of this work is to begin exploring a dynamic simulation approach to HRA, one whose models have a basis in psychological theories of human performance, and whose quantitative estimates have an empirical basis. This paper highlights a plan to formalize collaboration among the Idaho National Laboratory (INL), the University of Maryland, and The Ohio State University (OSU) to continue development of a simulation model initially formulated at the University of Maryland. Initial work will focus on enhancing the underlying human performance models with the most recent psychological research, and on planning follow-on studies to establish an empirical basis for the model, based on simulator experiments to be carried out at the INL and at the OSU.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

Current Reactor Physics Benchmark Activities at the Idaho National Laboratory

The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) were established to preserve integral reactor physics and criticality experiment data for present and future research. These valuable assets provide the basis for recording, developing, and validating our integral nuclear data, and experimental and computational methods. These projects are managed through the Idaho National Laboratory (INL) and the Organisation for Economic Co-operation and Development Nuclear Energy Agency (OECD-NEA). Staff and students at the Department of Energy - Idaho (DOE-ID) and INL are engaged in the development of benchmarks to support ongoing research activities. These benchmarks include reactors or assemblies that support Next Generation Nuclear Plant (NGNP) research, space nuclear Fission Surface Power System (FSPS) design validation, and currently operational facilities in Southeastern Idaho.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗