406 Matching Results

Search Results

Advanced search parameters have been applied.

REVIEW OF REACTOR SAFETY ANALYSES OF FAST AND LIQUID METAL COOLED REACTORS

Description: Safety analysis reports on United States fast and liquid metal cooled reactors were reviewed to gain a better understanding of the safety philosophy applied to the design of these facilities. This information was compiled to help guide the design and safety analysis of the Fast Flux Test Facility. No attempt was made to draw conclusions concerning the relative merit of different approaches and philosophies used by different reactor design teams. The facilities reviewed were; Enrico Fermi Atomic Power Plant (FERMI) Hallam Nuclear Power Facility (HALLAM) Southwest Experimental Fast Oxide Reactor (SEFOR) Fast Reactor Test Facility (FARET) Experimental Breeder Reactor No. 1 (EBR-I) Experimental Breeder Reactor No. 2 (EBR-II) Fast Reactor Zero Power Experiment (ZPR - III). The information gathered from the safety analysis reports is tabulated under these headings: Control and Safety Systems; Reactor Protection Systems; Backup Systems; Containment or Confinement Systems; Inherent Reactivity Effects and Important Physics Parameters; Fuel and Fuel Handling; Accidents Considered and Chemical Problems; Site; Exhaust Ventilation System; and Waste Effluents.
Date: November 1, 1967
Creator: Shaver, R. E. & Wittenbrock, N. G.
Partner: UNT Libraries Government Documents Department

Deep Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors- Monthly Highlights October 2010

Description: The DB Program monthly highlights report for September 2010, ORNL/TM-2010/252, was distributed to program participants by email on October 26. This report discusses: (1) Core and Fuel Analysis; (2) Spent Fuel Management; (3) Fuel Cycle Integration of the HTR (high temperature helium-cooled reactor); (4) TRU (transuranic elements) HTR Fuel Qualification; (5) HTR Spent Fuel Recycle - (a) TRU Kernel Development (ORNL), (b) Coating Development (ORNL), (c) Characterization Development and Support, (d) ZrC Properties and Handbook; and (6) HTR Fuel Recycle.
Date: November 1, 2010
Creator: Snead, Lance Lewis; Besmann, Theodore M; Collins, Emory D & Bell, Gary L
Partner: UNT Libraries Government Documents Department

Design Study for a Low-enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2007

Description: This report documents progress made during fiscal year 2007 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low enriched uranium fuel (LEU). Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. A high volume fraction U/Mo-in-Al fuel could attain the same neutron flux performance as with the current, HEU fuel but materials considerations appear to preclude production and irradiation of such a fuel. A diffusion barrier would be required if Al is to be retained as the interstitial medium and the additional volume required for this barrier would degrade performance. Attaining the high volume fraction (55 wt. %) of U/Mo assumed in the computational study while maintaining the current fuel plate acceptance level at the fuel manufacturer is unlikely, i.e. no increase in the percentage of plates rejected for non-compliance with the fuel specification. Substitution of a zirconium alloy for Al would significantly increase the weight of the fuel element, the cost of the fuel element, and introduce an as-yet untried manufacturing process. A monolithic U-10Mo foil is the choice of LEU fuel for HFIR. Preliminary calculations indicate that with a modest increase in reactor power, the flux performance of the reactor can be maintained at the current level. A linearly-graded, radial fuel thickness profile is preferred to the arched profile currently used in HEU fuel because the LEU fuel media is a metal alloy foil rather than a powder. Developments in analysis capability and nuclear data processing techniques are underway with the goal of verifying the preliminary calculations of LEU flux performance. A conceptual study of the operational cost of an LEU fuel fabrication facility yielded the conclusion that the annual fuel cost to the HFIR would increase significantly from ...
Date: November 1, 2007
Creator: Primm, Trent; Ellis, Ronald James; Gehin, Jess C; Ilas, Germina; Miller, James Henry & Sease, John D
Partner: UNT Libraries Government Documents Department

LONG-TERM PERFORMANCE OF SOLID OXIDE STACKS WITH ELECTRODE-SUPPORTED CELLS OPERATING IN THE STEAM ELECTROLYSIS MODE

Description: Performance characterization and durability testing have been completed on two five-cell high-temperature electrolysis stacks constructed with advanced cell and stack technologies. The solid oxide cells incorporate a negative-electrode-supported multi-layer design with nickel-zirconia cermet negative electrodes, thin-film yttria-stabilized zirconia electrolytes, and multi-layer lanthanum ferrite-based positive electrodes. The per-cell active area is 100 cm2. The stack is internally manifolded with compliant mica-glass seals. Treated metallic interconnects with integral flow channels separate the cells. Stack compression is accomplished by means of a custom spring-loaded test fixture. Initial stack performance characterization was determined through a series of DC potential sweeps in both fuel cell and electrolysis modes of operation. Results of these sweeps indicated very good initial performance, with area-specific resistance values less than 0.5 ?.cm2. Long-term durability testing was performed with A test duration of 1000 hours. Overall performance degradation was less than 10% over the 1000-hour period. Final stack performance characterization was again determined by a series of DC potential sweeps at the same flow conditions as the initial sweeps in both electrolysis and fuel cell modes of operation. A final sweep in the fuel cell mode indicated a power density of 0.356 W/cm2, with average per-cell voltage of 0.71 V at a current of 50 A.
Date: November 1, 2011
Creator: O'Brien, J. E.; O'Brien, R. C.; Zhang, X.; Tao, G. & Butler, B. J.
Partner: UNT Libraries Government Documents Department

THE INTEGRATION OF PROCESS HEAT APPLICATIONS TO HIGH TEMPERATURE GAS REACTORS

Description: A high temperature gas reactor, HTGR, can produce industrial process steam, high-temperature heat-transfer gases, and/or electricity. In conventional industrial processes, these products are generated by the combustion of fossil fuels such as coal and natural gas, resulting in significant emissions of greenhouse gases such as carbon dioxide. Heat or electricity produced in an HTGR could be used to supply process heat or electricity to conventional processes without generating any greenhouse gases. Process heat from a reactor needs to be transported by a gas to the industrial process. Two such gases were considered in this study: helium and steam. For this analysis, it was assumed that steam was delivered at 17 MPa and 540 C and helium was delivered at 7 MPa and at a variety of temperatures. The temperature of the gas returning from the industrial process and going to the HTGR must be within certain temperature ranges to maintain the correct reactor inlet temperature for a particular reactor outlet temperature. The returning gas may be below the reactor inlet temperature, ROT, but not above. The optimal return temperature produces the maximum process heat gas flow rate. For steam, the delivered pressure sets an optimal reactor outlet temperature based on the condensation temperature of the steam. ROTs greater than 769.7 C produce no additional advantage for the production of steam.
Date: November 1, 2011
Creator: McKellar, Michael G.
Partner: UNT Libraries Government Documents Department

Waste Heat Recovery from the Advanced Test Reactor Secondary Coolant Loop

Description: This study investigated the feasibility of using a waste heat recovery system (WHRS) to recover heat from the Advanced Test Reactor (ATR) secondary coolant system (SCS). This heat would be used to preheat air for space heating of the reactor building, thus reducing energy consumption, carbon footprint, and energy costs. Currently, the waste heat from the reactor is rejected to the atmosphere via a four-cell, induced-draft cooling tower. Potential energy and cost savings are 929 kW and $285K/yr. The WHRS would extract a tertiary coolant stream from the SCS loop and pump it to a new plate and frame heat exchanger, from which the heat would be transferred to a glycol loop for preheating outdoor air supplied to the heating and ventilation system. The use of glycol was proposed to avoid the freezing issues that plagued and ultimately caused the failure of a WHRS installed at the ATR in the 1980s. This study assessed the potential installation of a new WHRS for technical, logistical, and economic feasibility.
Date: November 1, 2012
Creator: Guillen, Donna Post
Partner: UNT Libraries Government Documents Department

Pebble-bed pebble motion: Simulation and Applications

Description: Pebble bed reactors (PBR) have moving graphite fuel pebbles. This unique feature provides advantages, but also means that simulation of the reactor requires understanding the typical motion and location of the granular flow of pebbles. This report presents a method for simulation of motion of the pebbles in a PBR. A new mechanical motion simulator, PEBBLES, efficiently simulates the key elements of motion of the pebbles in a PBR. This model simulates gravitational force and contact forces including kinetic and true static friction. It's used for a variety of tasks including simulation of the effect of earthquakes on a PBR, calculation of packing fractions, Dancoff factors, pebble wear and the pebble force on the walls. The simulator includes a new differential static friction model for the varied geometries of PBRs. A new static friction benchmark was devised via analytically solving the mechanics equations to determine the minimum pebble-to-pebble friction and pebble-to-surface friction for a five pebble pyramid. This pyramid check as well as a comparison to the Janssen formula was used to test the new static friction equations. Because larger pebble bed simulations involve hundreds of thousands of pebbles and long periods of time, the PEBBLES code has been parallelized. PEBBLES runs on shared memory architectures and distributed memory architectures. For the shared memory architecture, the code uses a new O(n) lock-less parallel collision detection algorithm to determine which pebbles are likely to be in contact. The new collision detection algorithm improves on the traditional non-parallel O(n log(n)) collision detection algorithm. These features combine to form a fast parallel pebble motion simulation. The PEBBLES code provides new capabilities for understanding and optimizing PBRs. The PEBBLES code has provided the pebble motion data required to calculate the motion of pebbles during a simulated earthquake. The PEBBLES code provides the ability to ...
Date: November 1, 2011
Creator: Cogliati, Joshua J. & Ougouag, Abderrafi M.
Partner: UNT Libraries Government Documents Department

Performance Assessment of Single Electrode-Supported Solid Oxide Cells Operating in the Steam Electrolysis Mode

Description: An experimental study is under way to assess the performance of electrode-supported solid-oxide cells operating in the steam electrolysis mode for hydrogen production. Results presented in this paper were obtained from single cells, with an active area of 16 cm{sup 2} per cell. The electrolysis cells are electrode-supported, with yttria-stabilized zirconia (YSZ) electrolytes ({approx}10 {mu}m thick), nickel-YSZ steam/hydrogen electrodes ({approx}1400 {mu}m thick), and modified LSM or LSCF air-side electrodes ({approx}90 {mu}m thick). The purpose of the present study is to document and compare the performance and degradation rates of these cells in the fuel cell mode and in the electrolysis mode under various operating conditions. Initial performance was documented through a series of voltage-current (VI) sweeps and AC impedance spectroscopy measurements. Degradation was determined through long-term testing, first in the fuel cell mode, then in the electrolysis mode. Results generally indicate accelerated degradation rates in the electrolysis mode compared to the fuel cell mode, possibly due to electrode delamination. The paper also includes details of an improved single-cell test apparatus developed specifically for these experiments.
Date: November 1, 2011
Creator: Zhang, X.; O'Brien, J. E.; O'Brien, R. C. & Petigny, N.
Partner: UNT Libraries Government Documents Department

The Fukushima Daiichi Accident Study Information Portal

Description: This paper presents a description of The Fukushima Daiichi Accident Study Information Portal. The Information Portal was created by the Idaho National Laboratory as part of joint NRC and DOE project to assess the severe accident modeling capability of the MELCOR analysis code. The Fukushima Daiichi Accident Study Information Portal was created to collect, store, retrieve and validate information and data for use in reconstructing the Fukushima Daiichi accident. In addition to supporting the MELCOR simulations, the Portal will be the main DOE repository for all data, studies and reports related to the accident at the Fukushima Daiichi nuclear power station. The data is stored in a secured (password protected and encrypted) repository that is searchable and accessible to researchers at diverse locations.
Date: November 1, 2012
Creator: Germain, Shawn St.; Smith, Curtis; Schwieder, David & Phelan, Cherie
Partner: UNT Libraries Government Documents Department

TESTING AND PERFORMANCE ANALYSIS OF NASA 5 CM BY 5 CM BI-SUPPORTED SOLID OXIDE ELECTROLYSIS CELLS OPERATED IN BOTH FUEL CELL AND STEAM ELECTROLYSIS MODES

Description: A series of 5 cm by 5 cm bi-supported Solid Oxide Electrolysis Cells (SOEC) were produced by NASA for the Idaho National Laboratory (INL) and tested under the INL High Temperature Steam Electrolysis program. The results from the experimental demonstration of cell operation for both hydrogen production and operation as fuel cells is presented. An overview of the cell technology, test apparatus and performance analysis is also provided. The INL High Temperature Steam Electrolysis laboratory has developed significant test infrastructure in support of single cell and stack performance analyses. An overview of the single cell test apparatus is presented. The test data presented in this paper is representative of a first batch of NASA's prototypic 5 cm by 5 cm SOEC single cells. Clearly a significant relationship between the operational current density and cell degradation rate is evident. While the performance of these cells was lower than anticipated, in-house testing at NASA Glenn has yielded significantly higher performance and lower degradation rates with subsequent production batches of cells. Current post-test microstructure analyses of the cells tested at INL will be published in a future paper. Modification to cell compositions and cell reduction techniques will be altered in the next series of cells to be delivered to INL with the aim to decrease the cell degradation rate while allowing for higher operational current densities to be sustained. Results from the testing of new batches of single cells will be presented in a future paper.
Date: November 1, 2011
Creator: O'Brien, R. C.; O'Brien, J. E.; Stoots, C. M.; Zhang, X.; Farmer, S. C.; Cable, T. L. et al.
Partner: UNT Libraries Government Documents Department

Pre-test CFD Calculations for a Bypass Flow Standard Problem

Description: The bypass flow in a prismatic high temperature gas-cooled reactor (HTGR) is the flow that occurs between adjacent graphite blocks. Gaps exist between blocks due to variances in their manufacture and installation and because of the expansion and shrinkage of the blocks from heating and irradiation. Although the temperature of fuel compacts and graphite is sensitive to the presence of bypass flow, there is great uncertainty in the level and effects of the bypass flow. The Next Generation Nuclear Plant (NGNP) program at the Idaho National Laboratory has undertaken to produce experimental data of isothermal bypass flow between three adjacent graphite blocks. These data are intended to provide validation for computational fluid dynamic (CFD) analyses of the bypass flow. Such validation data sets are called Standard Problems in the nuclear safety analysis field. Details of the experimental apparatus as well as several pre-test calculations of the bypass flow are provided. Pre-test calculations are useful in examining the nature of the flow and to see if there are any problems associated with the flow and its measurement. The apparatus is designed to be able to provide three different gap widths in the vertical direction (the direction of the normal coolant flow) and two gap widths in the horizontal direction. It is expected that the vertical bypass flow will range from laminar to transitional to turbulent flow for the different gap widths that will be available.
Date: November 1, 2011
Creator: Johnson, Rich
Partner: UNT Libraries Government Documents Department

MODELING STRATEGIES TO COMPUTE NATURAL CIRCULATION USING CFD IN A VHTR AFTER A LOFA

Description: A prismatic gas-cooled very high temperature reactor (VHTR) is being developed under the next generation nuclear plant program (NGNP) of the U.S. Department of Energy, Office of Nuclear Energy. In the design of the prismatic VHTR, hexagonal shaped graphite blocks are drilled to allow insertion of fuel pins, made of compacted TRISO fuel particles, and coolant channels for the helium coolant. One of the concerns for the reactor design is the effects of a loss of flow accident (LOFA) where the coolant circulators are lost for some reason, causing a loss of forced coolant flow through the core. In such an event, it is desired to know what happens to the (reduced) heat still being generated in the core and if it represents a problem for the fuel compacts, the graphite core or the reactor vessel (RV) walls. One of the mechanisms for the transport of heat out of the core is by the natural circulation of the coolant, which is still present. That is, how much heat may be transported by natural circulation through the core and upwards to the top of the upper plenum? It is beyond current capability for a computational fluid dynamic (CFD) analysis to perform a calculation on the whole RV with a sufficiently refined mesh to examine the full potential of natural circulation in the vessel. The present paper reports the investigation of several strategies to model the flow and heat transfer in the RV. It is found that it is necessary to employ representative geometries of the core to estimate the heat transfer. However, by taking advantage of global and local symmetries, a detailed estimate of the strength of the resulting natural circulation and the level of heat transfer to the top of the upper plenum is obtained.
Date: November 1, 2012
Creator: Tung, Yu-Hsin; Johnson, Richard W.; Chieng, Ching-Chang & Ferng, Yuh-Ming
Partner: UNT Libraries Government Documents Department

New Reactor Physics Benchmark Data in the March 2012 Edition of the IRPhEP Handbook

Description: The International Reactor Physics Experiment Evaluation Project (IRPhEP) was established to preserve integral reactor physics experimental data, including separate or special effects data for nuclear energy and technology applications. Numerous experiments that have been performed worldwide, represent a large investment of infrastructure, expertise, and cost, and are valuable resources of data for present and future research. These valuable assets provide the basis for recording, development, and validation of methods. If the experimental data are lost, the high cost to repeat many of these measurements may be prohibitive. The purpose of the IRPhEP is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. Contributors from around the world collaborate in the evaluation and review of selected benchmark experiments for inclusion in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) [1]. Several new evaluations have been prepared for inclusion in the March 2012 edition of the IRPhEP Handbook.
Date: November 1, 2012
Creator: Bess, John D.; Briggs, J. Blair & Gulliford, Jim
Partner: UNT Libraries Government Documents Department

The Ongoing Impact of the U.S. Fast Reactor Integral Experiments Program

Description: The creation of a large database of integral fast reactor physics experiments advanced nuclear science and technology in ways that were unachievable by less capital intensive and operationally challenging approaches. They enabled the compilation of integral physics benchmark data, validated (or not) analytical methods, and provided assurance of future rector designs The integral experiments performed at Argonne National Laboratory (ANL) represent decades of research performed to support fast reactor design and our understanding of neutronics behavior and reactor physics measurements. Experiments began in 1955 with the Zero Power Reactor No. 3 (ZPR-3) and terminated with the Zero Power Physics Reactor (ZPPR, originally the Zero Power Plutonium Reactor) in 1990 at the former ANL-West site in Idaho, which is now part of the Idaho National Laboratory (INL). Two additional critical assemblies, ZPR-6 and ZPR-9, operated at the ANL-East site in Illinois. A total of 128 fast reactor assemblies were constructed with these facilities [1]. The infrastructure and measurement capabilities are too expensive to be replicated in the modern era, making the integral database invaluable as the world pushes ahead with development of liquid metal cooled reactors.
Date: November 1, 2012
Creator: Bess, John D.; Pope, Michael A. & McFarlane, Harold F.
Partner: UNT Libraries Government Documents Department

Basis for NGNP Reactor Design Down-Selection

Description: The purpose of this paper is to identify the extent of technology development, design and licensing maturity anticipated to be required to credibly identify differences that could make a technical choice practical between the prismatic and pebble bed reactor designs. This paper does not address a business decision based on the economics, business model and resulting business case since these will vary based on the reactor application. The selection of the type of reactor, the module ratings, the number of modules, the configuration of the balance of plant and other design selections will be made on the basis of optimizing the Business Case for the application. These are not decisions that can be made on a generic basis.
Date: November 1, 2011
Creator: Demick, L.E.
Partner: UNT Libraries Government Documents Department

Current Reactor Physics Benchmark Activities at the Idaho National Laboratory

Description: The International Reactor Physics Experiment Evaluation Project (IRPhEP) [1] and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) [2] 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.
Date: November 1, 2011
Creator: Bess, John D.; Marshall, Margaret A.; Gorham, Mackenzie L.; Christensen, Joseph; Turnbull, James C. & Clark, Kim
Partner: UNT Libraries Government Documents Department

Criticality Benchmark Analysis of the HTTR Annular Startup Core Configurations

Description: One of the high priority benchmarking activities for corroborating the Next Generation Nuclear Plant (NGNP) Project and Very High Temperature Reactor (VHTR) Program is evaluation of Japan's existing High Temperature Engineering Test Reactor (HTTR). The HTTR is a 30 MWt engineering test reactor utilizing graphite moderation, helium coolant, and prismatic TRISO fuel. A large amount of critical reactor physics data is available for validation efforts of High Temperature Gas-cooled Reactors (HTGRs). Previous international reactor physics benchmarking activities provided a collation of mixed results that inaccurately predicted actual experimental performance.1 Reevaluations were performed by the Japanese to reduce the discrepancy between actual and computationally-determined critical configurations.2-3 Current efforts at the Idaho National Laboratory (INL) involve development of reactor physics benchmark models in conjunction with the International Reactor Physics Experiment Evaluation Project (IRPhEP) for use with verification and validation methods in the VHTR Program. Annular cores demonstrate inherent safety characteristics that are of interest in developing future HTGRs.
Date: November 1, 2009
Creator: Bess, John D.
Partner: UNT Libraries Government Documents Department

An Independent Assessment of Evacuation Time Estimates for A Peak Population Scenario in the Emergency Planning Zone of the Seabrook Nuclear Power Station

Description: This study comprises two major tasks. First, it includes an independent assessment of the methods and assumptions used in calculating evacuation time estimates (ETEs) applicable to the general population for a peak population scenario in the emergency planning zone {EPZ) of the Seabrook Nuclear Power Station. This consists of a review and analysis of previous work by Public Service of New Hampshire {PSNH) and the Federal Emergency Management Agency (FEMA), as well as an independent calculation of evacuation times using the CLEAR model for the demographic data reported by PSNH. Secondly, this study includes independent estimations of evacuation time for the peak population scenario developed using demographic data prepared by the U.S. Nuclear Regulatory Commission {NRC). These evacuation time estimates are approximately 60% and 84% greater, respectively, than the estimate provided by PSNH for a simulataneous evacuation of the entire EPZ under peak conditions. The CLEAR model, which was developed by Pacific Northwest Laboratory {PNL) under the sponsorship of the. NRC, was also used for these latter calculations. The results of this study reveal the importance of the assumptions used for calculating evacuation times. Because traffic routings and management plans have not been prepared for the area, the CLEAR calculations utilized indepdently prepared traffic routings and assumptions. A detailed analysis of the results suggests that the ETEs submitted by PSNH are consistent with the methods and assumptions which provide the bases for PSNH•s evacuation time estimates. Differences among evacuation time estimates stem largely from differences jn the assumed size of the evacuating population and the estimated effectiveness of traffic controls.
Date: November 1, 1982
Creator: Moeller, M. P.; Urbanik, II, T.; Mclean, M. A. & Desrosiers, A. E.
Partner: UNT Libraries Government Documents Department

The Effect of Degraded Digital Instrumentation and Control systems on Human-system Interfaces and Operator Performance

Description: Integrated digital instrumentation and control (I&C) systems in new and advanced nuclear power plants (NPPs) will support operators in monitoring and controlling the plants. Even though digital systems typically are expected to be reliable, their potential for degradation or failure significantly could affect the operators performance and, consequently, jeopardize plant safety. This U.S. Nuclear Regulatory Commission (NRC) research investigated the effects of degraded I&C systems on human performance and on plant operations. The objective was to develop technical basis and guidance for human factors engineering (HFE) reviews addressing the operator's ability to detect and manage degraded digital I&C conditions. We reviewed pertinent standards and guidelines, empirical studies, and plant operating experience. In addition, we evaluated the potential effects of selected failure modes of the digital feedwater control system of a currently operating pressurized water reactor (PWR) on human-system interfaces (HSIs) and the operators performance. Our findings indicated that I&C degradations are prevalent in plants employing digital systems, and the overall effects on the plant's behavior can be significant, such as causing a reactor trip or equipment to operate unexpectedly. I&C degradations may affect the HSIs used by operators to monitor and control the plant. For example, deterioration of the sensors can complicate the operators interpretation of displays, and sometimes may mislead them by making it appear that a process disturbance has occurred. We used the findings as the technical basis upon which to develop HFE review guidance.
Date: November 7, 2010
Creator: OHara, J.M.; Gunther, B.; Martinez-Guridi, G. (BNL); Xing, J. & Barnes, V. (NRC)
Partner: UNT Libraries Government Documents Department

Markov Model of Accident Progression at Fukushima Daiichi

Description: On March 11, 2011, a magnitude 9.0 earthquake followed by a tsunami caused loss of offsite power and disabled the emergency diesel generators, leading to a prolonged station blackout at the Fukushima Daiichi site. After successful reactor trip for all operating reactors, the inability to remove decay heat over an extended period led to boil-off of the water inventory and fuel uncovery in Units 1-3. A significant amount of metal-water reaction occurred, as evidenced by the quantities of hydrogen generated that led to hydrogen explosions in the auxiliary buildings of the Units 1 & 3, and in the de-fuelled Unit 4. Although it was assumed that extensive fuel damage, including fuel melting, slumping, and relocation was likely to have occurred in the core of the affected reactors, the status of the fuel, vessel, and drywell was uncertain. To understand the possible evolution of the accident conditions at Fukushima Daiichi, a Markov model of the likely state of one of the reactors was constructed and executed under different assumptions regarding system performance and reliability. The Markov approach was selected for several reasons: It is a probabilistic model that provides flexibility in scenario construction and incorporates time dependence of different model states. It also readily allows for sensitivity and uncertainty analyses of different failure and repair rates of cooling systems. While the analysis was motivated by a need to gain insight on the course of events for the damaged units at Fukushima Daiichi, the work reported here provides a more general analytical basis for studying and evaluating severe accident evolution over extended periods of time. This work was performed at the request of the U.S. Department of Energy to explore 'what-if' scenarios in the immediate aftermath of the accidents.
Date: November 11, 2012
Creator: A., Cuadra; R., Bari; Cheng, L-Y; Ginsberg, T.; Lehner, J.; Martinez-Guridi, G. et al.
Partner: UNT Libraries Government Documents Department

Applying Human-performance Models to Designing and Evaluating Nuclear Power Plants: Review Guidance and Technical Basis

Description: Human performance models (HPMs) are simulations of human behavior with which we can predict human performance. Designers use them to support their human factors engineering (HFE) programs for a wide range of complex systems, including commercial nuclear power plants. Applicants to U.S. Nuclear Regulatory Commission (NRC) can use HPMs for design certifications, operating licenses, and license amendments. In the context of nuclear-plant safety, it is important to assure that HPMs are verified and validated, and their usage is consistent with their intended purpose. Using HPMs improperly may generate misleading or incorrect information, entailing safety concerns. The objective of this research was to develop guidance to support the NRC staff's reviews of an applicant's use of HPMs in an HFE program. The guidance is divided into three topical areas: (1) HPM Verification, (2) HPM Validation, and (3) User Interface Verification. Following this guidance will help ensure the benefits of HPMs are achieved in a technically sound, defensible manner. During the course of developing this guidance, I identified several issues that could not be addressed; they also are discussed.
Date: November 30, 2009
Creator: O'Hara, J.M.
Partner: UNT Libraries Government Documents Department

Condition Monitoring of Cables Task 3 Report: Condition Monitoring Techniques for Electric Cables

Description: For more than 20 years the NRC has sponsored research studying electric cable aging degradation, condition monitoring, and environmental qualification testing practices for electric cables used in nuclear power plants. This report summarizes several of the most effective and commonly used condition monitoring techniques available to detect damage and measure the extent of degradation in electric cable insulation. The technical basis for each technique is summarized, along with its application, trendability of test data, ease of performing the technique, advantages and limitations, and the usefulness of the test results to characterize and assess the condition of electric cables.
Date: November 30, 2009
Creator: Villaran, M.; Lofaro, R. & na
Partner: UNT Libraries Government Documents Department

Novel Processing of Unique Ceramic-Based Nuclear Materials and Fuels

Description: Advances in nuclear reactor technology and the use of gas-cooled fast reactors require the development of new materials that can operate at the higher temperatures expected in these systems. These include refractory alloys base on Nb, Zr, Ta, Mo, W, and Re; ceramics and composites such as those based on silicon carbide (SiCf-SiC); carbon-carbon composites; and advanced coatings. Besides the ability to handle higher expected temperatures, effective heat transfer between reactor componets is necessary for improved efficiency. Improving thermal conductivity of the materials used in nuclear fuels and other temperature critical components can lower the center-line fuel temperature and thereby enhance durability and reduce the risk of premature failure.
Date: November 30, 2008
Creator: Zhang, Hui & Singh, Raman P.
Partner: UNT Libraries Government Documents Department