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Boiling Visualization and Critical Heat Flux Phenomena In Narrow Rectangular Gap

Description: An experimental study was performed to investifate the pool boling critical hear flux (CHF) on one-dimensional inclined rectangular channels with narrow gaps by changing the orientation of a copper test heater assembly. In a pool of saturated water at atmospheric pressure, the test parameters include the gap sizes of 1,2,5, and 10 mm, andthe surface orientation angles from the downward facing position (180 degrees) to the vertical position (90 degress) respectively.
Date: December 1, 2004
Creator: Kim, J. J.; Kim, Y. H.; Kim, S. J.; Noh, S. W.; Suh, K. Y.; Rempe, J. et al.
Partner: UNT Libraries Government Documents Department

Advanced Reactor Licensing: Experience with Digital I&C Technology in Evolutionary Plants

Description: This report presents the findings from a study of experience with digital instrumentation and controls (I&C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l&C systems and identified lessons learned. The report (1) gives an overview of the modern l&C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States.
Date: September 27, 2004
Creator: Wood, RT
Partner: UNT Libraries Government Documents Department

KINETICS OF SLURRY PHASE FISCHER-TROPSCH SYNTHESIS

Description: This report covers the second year of this three-year research grant under the University Coal Research program. The overall objective of this project is to develop a comprehensive kinetic model for slurry phase Fischer-Tropsch synthesis on iron catalysts. This model will be validated with experimental data obtained in a stirred tank slurry reactor (STSR) over a wide range of process conditions. The model will be able to predict concentrations of all reactants and major product species (H{sub 2}O, CO{sub 2}, linear 1- and 2-olefins, and linear paraffins) as a function of reaction conditions in the STSR. During the second year of the project we completed the STSR test SB-26203 (275-343 h on stream), which was initiated during the first year of the project, and another STSR test (SB-28603 lasting 341 h). Since the inception of the project we completed 3 STSR tests, and evaluated catalyst under 25 different sets of process conditions. A precipitated iron catalyst obtained from Ruhrchemie AG (Oberhausen-Holten, Germany) was used in all tests. This catalyst was used initially in commercial fixed bed reactors at Sasol in South Africa. Also, during the second year we performed a qualitative analysis of experimental data from all three STSR tests. Effects of process conditions (reaction temperature, pressure, feed composition and gas space velocity) on water-gas-shift (WGS) activity and hydrocarbon product distribution have been determined.
Date: September 29, 2004
Creator: Bukur, Dragomir B.
Partner: UNT Libraries Government Documents Department

Calculation of the radionuclides in PWR spent fuel samples for SFR experiment planning.

Description: This report documents the calculation of radionuclide content in the pressurized water reactor (PWR) spent fuel samples planned for use in the Spent Fuel Ratio (SPR) Experiments at Sandia National Laboratories, Albuquerque, New Mexico (SNL) to aid in experiment planning. The calculation methods using the ORIGEN2 and ORIGEN-ARP computer codes and the input modeling of the planned PWR spent fuel from the H. B. Robinson and the Surry nuclear power plants are discussed. The safety hazards for the calculated nuclide inventories in the spent fuel samples are characterized by the potential airborne dose and by the portion of the nuclear facility hazard category 2 and 3 thresholds that the experiment samples would present. In addition, the gamma ray photon energy source for the nuclide inventories is tabulated to facilitate subsequent calculation of the direct and shielded dose rates expected from the samples. The relative hazards of the high burnup 72 gigawatt-day per metric ton of uranium (GWd/MTU) spent fuel from H. B. Robinson and the medium burnup 36 GWd/MTU spent fuel from Surry are compared against a parametric calculation of various fuel burnups to assess the potential for higher hazard PWR fuel samples.
Date: June 1, 2004
Creator: Naegeli, Robert Earl
Partner: UNT Libraries Government Documents Department

Advanced nuclear energy analysis technology.

Description: A two-year effort focused on applying ASCI technology developed for the analysis of weapons systems to the state-of-the-art accident analysis of a nuclear reactor system was proposed. The Sandia SIERRA parallel computing platform for ASCI codes includes high-fidelity thermal, fluids, and structural codes whose coupling through SIERRA can be specifically tailored to the particular problem at hand to analyze complex multiphysics problems. Presently, however, the suite lacks several physics modules unique to the analysis of nuclear reactors. The NRC MELCOR code, not presently part of SIERRA, was developed to analyze severe accidents in present-technology reactor systems. We attempted to: (1) evaluate the SIERRA code suite for its current applicability to the analysis of next generation nuclear reactors, and the feasibility of implementing MELCOR models into the SIERRA suite, (2) examine the possibility of augmenting ASCI codes or alternatives by coupling to the MELCOR code, or portions thereof, to address physics particular to nuclear reactor issues, especially those facing next generation reactor designs, and (3) apply the coupled code set to a demonstration problem involving a nuclear reactor system. We were successful in completing the first two in sufficient detail to determine that an extensive demonstration problem was not feasible at this time. In the future, completion of this research would demonstrate the feasibility of performing high fidelity and rapid analyses of safety and design issues needed to support the development of next generation power reactor systems.
Date: May 1, 2004
Creator: Gauntt, Randall O.; Murata, Kenneth K.; Romero, Vicente JosÔe; Young, Michael Francis & Rochau, Gary Eugene
Partner: UNT Libraries Government Documents Department

DSNF AND OTHER WASTE FORM DEGRADATION ABSTRACTION

Description: Several hundred distinct types of DOE-owned spent nuclear fuel (DSNF) may potentially be disposed in the Yucca Mountain repository. These fuel types represent many more types than can be viably individually examined for their effect on the Total System Performance Assessment for the License Application (TSPA-LA). Additionally, for most of these fuel types, there is no known direct experimental test data for the degradation and dissolution of the waste form in repository groundwaters. The approach used in the TSPA-LA model is, therefore, to assess available information on each of 11 groups of DSNF, and to identify a model that can be used in the TSPA-LA model without differentiating between individual codisposal waste packages containing different DSNF types. The purpose of this report is to examine the available data and information concerning the dissolution kinetics of DSNF matrices for the purpose of abstracting a degradation model suitable for use in describing degradation of the DSNF inventory in the Total System Performance Assessment for the License Application. The data and information and associated degradation models were examined for the following types of DSNF: Group 1--Naval spent nuclear fuel; Group 2--Plutonium/uranium alloy (Fermi 1 SNF); Group 3--Plutonium/uranium carbide (Fast Flux Test Facility-Test Fuel Assembly SNF); Group 4--Mixed oxide and plutonium oxide (Fast Flux Test Facility-Demonstration Fuel Assembly/Fast Flux Test Facility-Test Demonstration Fuel Assembly SNF); Group 5--Thorium/uranium carbide (Fort St. Vrain SNF); Group 6--Thorium/uranium oxide (Shippingport light water breeder reactor SNF); Group 7--Uranium metal (N Reactor SNF); Group 8--Uranium oxide (Three Mile Island-2 core debris); Group 9--Aluminum-based SNF (Foreign Research Reactor SNF); Group 10--Miscellaneous Fuel; and Group 11--Uranium-zirconium hydride (Training Research Isotopes-General Atomics SNF). The analyses contained in this document provide an ''upper-limit'' (i.e., instantaneous degradation) model for use in the TSPA-LA model. ''Best-estimate'' models for the degradation of the fuels in each of ...
Date: November 19, 2004
Creator: CUNNANE, J.
Partner: UNT Libraries Government Documents Department

Production of Depleted UO<sub>2</sub>Kernels for the Advanced Gas-Cooled Reactor Program for Use in TRISO Coating Development

Description: The main objective of the Depleted UO{sub 2} Kernels Production Task at Oak Ridge National Laboratory (ORNL) was to conduct two small-scale production campaigns to produce 2 kg of UO{sub 2} kernels with diameters of 500 {+-} 20 {micro}m and 3.5 kg of UO{sub 2} kernels with diameters of 350 {+-} 10 {micro}m for the U.S. Department of Energy Advanced Fuel Cycle Initiative Program. The final acceptance requirements for the UO{sub 2} kernels are provided in the first section of this report. The kernels were prepared for use by the ORNL Metals and Ceramics Division in a development study to perfect the triisotropic (TRISO) coating process. It was important that the kernels be strong and near theoretical density, with excellent sphericity, minimal surface roughness, and no cracking. This report gives a detailed description of the production efforts and results as well as an in-depth description of the internal gelation process and its chemistry. It describes the laboratory-scale gel-forming apparatus, optimum broth formulation and operating conditions, preparation of the acid-deficient uranyl nitrate stock solution, the system used to provide uniform broth droplet formation and control, and the process of calcining and sintering UO{sub 3} {center_dot} 2H{sub 2}O microspheres to form dense UO{sub 2} kernels. The report also describes improvements and best past practices for uranium kernel formation via the internal gelation process, which utilizes hexamethylenetetramine and urea. Improvements were made in broth formulation and broth droplet formation and control that made it possible in many of the runs in the campaign to produce the desired 350 {+-} 10-{micro}m-diameter kernels, and to obtain very high yields.
Date: December 2, 2004
Creator: Collins, J. L.
Partner: UNT Libraries Government Documents Department

Advanced Core Design And Fuel Management For Pebble-Bed Reactors

Description: A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.
Date: October 1, 2004
Creator: Gougar, Hans D.; Ougouag, Abderrafi M. & Terry, William K.
Partner: UNT Libraries Government Documents Department

The Next Generation Nuclear Plant - Insights Gained from the INEEL Point Design Studies

Description: This paper provides the results of an assessment of two possible versions of the Next Generation Nuclear Plant (NGNP), a prismatic fuel type helium gas-cooled reactor and a pebble-bed fuel helium gas reactor. Insights gained regarding the strengths and weaknesses of the two designs are also discussed. Both designs will meet the three basic requirements that have been set for the NGNP: a coolant outlet temperature of 1000 C, passive safety, and a total power output consistent with that expected for commercial high-temperature gas-cooled reactors. Two major modifications of the current Gas Turbine- Modular Helium Reactor (GT-MHR) design were needed to obtain a prismatic block design with a 1000 C outlet temperature: reducing the bypass flow and better controlling the inlet coolant flow distribution to the core. The total power that could be obtained for different core heights without exceeding a peak transient fuel temperature of 1600 °C during a high or low-pressure conduction cooldown event was calculated. With a coolant inlet temperature of 490 °C and 10% nominal core bypass flow, it is estimated that the peak power for a 10-block high core is 686 MWt, for a 12-block high core is 786 MWt, and for a 14-block core is about 889 MWt. The core neutronics calculations showed that the NGNP will exhibit strongly negative Doppler and isothermal temperature coefficients of reactivity over the burnup cycle. In the event of rapid loss of the helium gas, there is negligible core reactivity change. However, water or steam ingress into the core coolant channels can produce a relatively large reactivity effect. Two versions of an annular pebble-bed NGNP have also been developed, a 300 and a 600 MWt module. From this work we learned how to design passively safe pebble bed reactors that produce more than 600 MWt. We also found ...
Date: August 1, 2004
Creator: MacDonald, Philip E.; Baxter, A. M.; Bayless, P. D.; Bolin, J. M.; Gougar, H. D.; Moore, R. L. et al.
Partner: UNT Libraries Government Documents Department

Gas-Cooled Fast Reactor (GFR) FY04 Annual Report

Description: The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.
Date: September 1, 2004
Creator: Weaver, K. D.; Totemeier, T. C.; Clark, D. E.; Feldman, E. E.; Hoffman, E. A.; Vilim, R. B. et al.
Partner: UNT Libraries Government Documents Department

Initial Requirements for Gas-Cooled Fast Reactor (GFR) System Design, Performance, and Safety Analysis Models

Description: The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.
Date: August 1, 2004
Creator: Weaver, Kevan D. & Wei, Thomas Y. C.
Partner: UNT Libraries Government Documents Department

Initial Scaling Studies and Conceptual Thermal Fluids Experiments for the Prismatic NGNP Point Design

Description: The objective of this report is to document the initial high temperature gas reactor scaling studies and conceptual experiment design for gas flow and heat transfer. The general approach of the project is to develop new benchmark experiments for assessment in parallel with CFD and coupled CFD/ATHENA/RELAP5-3D calculations for the same geometry. Two aspects of the complex flow in an NGNP are being addressed: (1) flow and thermal mixing in the lower plenum (&#34;hot streaking&#34; issue) and (2) turbulence and resulting temperature distributions in reactor cooling channels (&#34;hot channel&#34; issue). Current prismatic NGNP concepts are being examined to identify their proposed flow conditions and geometries over the range from normal operation to decay heat removal in a pressurized cooldown. Approximate analyses are being applied to determine key non-dimensional parameters and their magnitudes over this operating range. For normal operation, the flow in the coolant channels can be considered to be dominant forced convection with slight transverse property variation. The flow in the lower plenum can locally be considered to be a situation of multiple buoyant jets into a confined density-stratified crossflow -- with obstructions. Experiments are needed for the combined features of the lower plenum flows. Missing from the typical jet experiments are interactions with nearby circular posts and with vertical posts in the vicinity of vertical walls - with near stagnant surroundings at one extreme and significant crossflow at the other. Two heat transfer experiments are being considered. One addresses the &#34;hot channel&#34; problem, if necessary. The second experiment will treat heated jets entering a model plenum. Unheated MIR (Matched-Index-of-Refraction) experiments are first steps when the geometry is complicated. One does not want to use a computational technique which will not even handle constant properties properly. The MIR experiment will simulate flow features of the paths of jets as ...
Date: September 1, 2004
Creator: McEligot, D. M. & McCreery, G. E.
Partner: UNT Libraries Government Documents Department

Margin for In-Vessel Retention in the APR1400 - VESTA and SCDAP/RELAP5-3D Analyses

Description: If cooling is inadequate during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the lower head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with such plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe pressurized water reactor (PWR) (AP600), which relied upon external reactor vessel cooling (ERVC) for in-vessel retention (IVR), resulted in the U.S. Nuclear Regulatory Commission (USNRC) approving the design without requiring certain conventional features common to existing light water reactors (LWRs). IVR of core melt is therefore a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced LWRs. However, it is not clear that currently proposed ERVC without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a three-year, United States (U.S.) -Korean International Nuclear Energy Research Initiative (INERI) project was initiated in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korean Atomic Energy Research Institute (KAERI) explored options, such as enhanced ERVC performance and an enhanced in-vessel core catcher (IVCC), that have the potential to ensure that IVR is feasible for higher power reactors.
Date: December 1, 2004
Creator: Rempe, Joy & Knudson, D.
Partner: UNT Libraries Government Documents Department

Quality Assurance Program Plan for AGR Fuel Development and Qualification Program

Description: Quality Assurance Plan (QPP) is to document the Idaho National Engineering and Environmental Laboratory (INEEL) Management and Operating (M&O) Contractor’s quality assurance program for AGR Fuel Development and Qualification activities, which is under the control of the INEEL. The QPP is an integral part of the Gen IV Program Execution Plan (PEP) and establishes the set of management controls for those systems, structures and components (SSCs) and related quality affecting activities, necessary to provide adequate confidence that items will perform satisfactorily in service.
Date: February 1, 2004
Creator: Sowder, W. Ken
Partner: UNT Libraries Government Documents Department

A Systems Engineering Framework for Design, Construction and Operation of the Next Generation Nuclear Plant

Description: Not since the International Space Station has a project of such wide participation been proposed for the United States. Ten countries, the European Union, universities, Department of Energy (DOE) laboratories, and industry will participate in the research and development, design, construction and/or operation of the fourth generation of nuclear power plants with a demonstration reactor to be built at a DOE site and operational by the middle of the next decade. This reactor will be like no other. The Next Generation Nuclear Plant (NGNP) will be passively safe, economical, highly efficient, modular, proliferation resistant, and sustainable. In addition to electrical generation, the NGNP will demonstrate efficient and cost effective generation of hydrogen to support the President’s Hydrogen Initiative. To effectively manage this multi-organizational and technologically complex project, systems engineering techniques and processes will be used extensively to ensure delivery of the final product. The technological and organizational challenges are complex. Research and development activities are required, material standards require development, hydrogen production, storage and infrastructure requirements are not well developed, and the Nuclear Regulatory Commission may further define risk-informed/performance-based approach to licensing. Detailed design and development will be challenged by the vast cultural and institutional differences across the participants. Systems engineering processes must bring the technological and organizational complexity together to ensure successful product delivery. This paper will define the framework for application of systems engineering to this $1.5B - $1.9B project.
Date: June 1, 2004
Creator: Gorski, Edward J.; Park, Charles V. & Southworth, Finis H.
Partner: UNT Libraries Government Documents Department

Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor

Description: The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 900 degrees C or operational fuel temperatures above 1250 degrees C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR&#39;s higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Engineering and Environmental Laboratory (INEEL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world&#39;s computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertaninty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.
Date: November 1, 2004
Creator: Chang, H. Oh, PhD; Davis, Cliff & Moore, Richard
Partner: UNT Libraries Government Documents Department

Bounding Estimate for the &#39;Hot&#39; Channel Temperature and Preliminary Calculation of Mixing in the Lower Plenum for the NGNP Point Design Using CFD

Description: The power density in the core of the block next generation nuclear power plant (NGNP) will not be uniform due to geometry, core layout and fuel pin design. Recent calculations performed to optimize the core design indicate that the maximum radial variation will be 1.25 times the average. This significant radial variation in the local power density will create a variation in the temperature of the helium coolant as it cools the core. The coolant channel with the highest outlet temperature is referred to as the ‘hot’ channel. The concern is that the high temperature channels, which exit into the lower plenum as jets, called ‘hot streaking,’ will adversely affect materials in the lower plenum, the exit duct and the turbine, as well as affect the performance of the turbine. The objective of the present study is to determine or bound the maximum exit temperature of the ‘hot’ channel. The maximum hot channel temperature depends on the total coolant flow rate, which has not yet been fixed. Experiments need to be designed to capture the complex physics of the lower-plenum flow to allow assessment and validation of numerical simulations. While preliminary CFD simulations are not yet validated, they can be of use in the planning of the experiments, particularly in estimating where there are regions of high and low turbulence intensity. Mixing of the coolant is related to the turbulence intensity as well as to the overall nature of the mean flow. The purpose of the present task is to provide preliminary flow calculations of the coolant in the lower plenum to examine flow patterns and turbulence intensity.
Date: December 1, 2004
Creator: Johnson, Richard W. & Schultz, R. R.
Partner: UNT Libraries Government Documents Department

Development of a Supercritical Carbon Dioxide Brayton Cycle: Improving PBR Efficiency and Testing Material Compatibility - 2004 Annual Report

Description: The U.S. and other countries address major challenges related to energy security and the environmental impacts of fossil fuels. Solutions to these issues include carbon-free electricity generation and hydrogen production for fuel cell car, fertilizer synthesis, petroleum refining, and other applications. The Very High Temperature Gas Reactor (HTGR) has been recognized as a promising technology for high efficiency electricity generation and high temperature process heat applications. Therefore, the U.S. needs to make the HTGR intrinsically safe and proliferation-resistant. The U.S. and the world, however, must still overcome certain technical issues and the cost barrier before it can be built in the U.S. The establishment of a nuclear power cost goal of 3.3 cents/kWh is desirable in order to compete with fossil combined-cycle, gas turbine power generation. This goal requires approximately a 30% reduction in power cost for state-of-the-art nuclear plants. It has been demonstrated that this large cost differential can be overcome only by technology improvements that lead to a combination of better efficiency and more compatible reactor materials. The objectives of this research are (1) to develop a supercritical carbon dioxide Brayton cycle in the secondary power conversion side that can be applied to some Generation-IV reactors such as the HTGR and supercritical water reactor, (2) to improve the plant net efficiency by using the carbon dioxide Brayton cycle, and (3) to test material compatibility at high temperatures and pressures. The reduced volumetric flow rate of carbon dioxide due to higher density compared to helium will reduce compression work, which eventually increase turbine work enhancing the plant net efficiency.
Date: October 1, 2004
Creator: Oh, Chang; Lillo, Thomas; Windes, William; Totemeier, Terry & Moore, Richard
Partner: UNT Libraries Government Documents Department

Development of Improved Models and Designs for Coated-Particle Gas Reactor Fuels -- Final Report under the International Nuclear Energy Research Initiative (I-NERI)

Description: The objective of this INERI project was to develop improved fuel behavior models for gas reactor coated-particle fuels and to explore improved coated-particle fuel designs that could be used reliably at very high burnups and potentially in gas-cooled fast reactors. Project participants included the Idaho National Engineering Laboratory (INEEL), Centre Étude Atomique (CEA), and the Massachusetts Institute of Technology (MIT). To accomplish the project objectives, work was organized into five tasks.
Date: December 1, 2004
Creator: Petti, David; Martin, Philippe; Phélip, Mayeul; Ballinger, Ronald & account, Petti does not have NT
Partner: UNT Libraries Government Documents Department

Parametric Investigation of Brayton Cycle for High Temperature Gas-Cooled Reactor

Description: The Idaho National Engineering and Environmental Laboratory (INEEL) is investigating a Brayton cycle efficiency improvement on a high temperature gas-cooled reactor (HTGR) as part of Generation-IV nuclear engineering research initiative. In this project, we are investigating helium Brayton cycles for the secondary side of an indirect energy conversion system. Ultimately we will investigate the improvement of the Brayton cycle using other fluids, such as supercritical carbon dioxide. Prior to the cycle improvement study, we established a number of baseline cases for the helium indirect Brayton cycle. These cases look at both single-shaft and multiple-shaft turbomachinary. The baseline cases are based on a 250 MW thermal pebble bed HTGR. The results from this study are applicable to other reactor concepts such as a very high temperature gas-cooled reactor (VHTR), fast gas-cooled reactor (FGR), supercritical water reactor (SWR), and others. In this study, we are using the HYSYS computer code for optimization of the helium Brayton cycle. Besides the HYSYS process optimization, we performed parametric study to see the effect of important parameters on the cycle efficiency. For these parametric calculations, we use a cycle efficiency model that was developed based on the Visual Basic computer language. As a part of this study we are currently investigated single-shaft vs. multiple shaft arrangement for cycle efficiency and comparison, which will be published in the next paper.The ultimate goal of this study is to use supercritical carbon dioxide for the HTGR power conversion loop in order to improve the cycle efficiency to values great than that of the helium Brayton cycle. This paper includes preliminary calculations of the steady state overall Brayton cycle efficiency based on the pebble bed reactor reference design (helium used as the working fluid) and compares those results with an initial calculation of a CO2 Brayton cycle.
Date: July 1, 2004
Creator: Oh, Chang
Partner: UNT Libraries Government Documents Department

New Generation Nuclear Plant (NGNP) Project, Preliminary Point Design

Description: This paper provides a preliminary assessment of two possible versions of the Next Generation Nuclear Plant (NGNP), a prismatic fuel type helium gas-cooled reactor and a pebblebed fuel helium gas reactor. Both designs will meet the three basic requirements that have been set for the NGNP: a coolant outlet temperature of 1000 C, passive safety, and a total power output consistent with that expected for commercial high-temperature gas-cooled reactors.
Date: March 1, 2004
Creator: Southworth, F. H.; MacDonald, P. E.; Baxter, A. M.; Bayless, P. D.; Bolin, J. M.; Gougar, H. D. et al.
Partner: UNT Libraries Government Documents Department

Failure Forewarning in NPP Equipment NERI2000-109 Final Project Report

Description: The objective of this project is forewarning of machine failures in critical equipment at next-generation nuclear power plants (NPP). Test data were provided by two collaborating institutions: Duke Engineering and Services (first project year), and the Pennsylvania State University (Applied Research Laboratory) during the second and third project years. New nonlinear methods were developed and applied successfully to extract forewarning trends from process-indicative, time-serial data for timely, condition-based maintenance. Anticipation of failures in critical equipment at next-generation NPP will improve the scheduling of maintenance activities to minimize safety concerns, unscheduled non-productive downtime, and collateral damage due to unexpected failures. This approach provides significant economic benefit, and is expected to improve public acceptance of nuclear power. The approach is a multi-tiered, model-independent, and data-driven analysis that uses ORNL's novel nonlinear method to extract forewarning of machine failures from appropriate data. The first tier of the analysis provides a robust choice for the process-indicative data. The second tier rejects data of inadequate quality. The third tier removes signal artifacts that would otherwise confound the analysis, while retaining the relevant nonlinear dynamics. The fourth tier converts the artifact-filtered time-serial data into a geometric representation, that is then transformed to a discrete distribution function (DF). This method allows for noisy, finite-length datasets. The fifth tier obtains dissimilarity measures (DM) between the nominal-state DF and subsequent test-state DFs. Forewarning of a machine failure is indicated by several successive occurrences of the DM above a threshold, or by a statistically significant trend in the DM. This paradigm yields robust nonlinear signatures of degradation and its progression, allowing earlier and more accurate detection of the machine failure.
Date: March 26, 2004
Creator: Hively, LM
Partner: UNT Libraries Government Documents Department

Assessment of Food Chain Pathway Parameters in Biosphere Models: Annual Progress Report for Fiscal Year 2004

Description: This Annual Progress Report describes the work performed and summarizes some of the key observations to date on the U.S. Nuclear Regulatory Commission’s project Assessment of Food Chain Pathway Parameters in Biosphere Models, which was established to assess and evaluate a number of key parameters used in the food-chain models used in performance assessments of radioactive waste disposal facilities. Section 2 of this report describes activities undertaken to collect samples of soils from three regions of the United States, the Southeast, Northwest, and Southwest, and perform analyses to characterize their physical and chemical properties. Section 3 summarizes information gathered regarding agricultural practices and common and unusual crops grown in each of these three areas. Section 4 describes progress in studying radionuclide uptake in several representative crops from the three soil types in controlled laboratory conditions. Section 5 describes a range of international coordination activities undertaken by Project staff in order to support the underlying data needs of the Project. Section 6 provides a very brief summary of the status of the GENII Version 2 computer program, which is a “client” of the types of data being generated by the Project, and for which the Project will be providing training to the US NRC staff in the coming Fiscal Year. Several appendices provide additional supporting information.
Date: December 2, 2004
Creator: Napier, Bruce A.; Krupka, Kenneth M.; Fellows, Robert J.; Cataldo, Dominic A.; Valenta, Michelle M. & Gilmore, Tyler J.
Partner: UNT Libraries Government Documents Department