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Review of the margins for ASME code fatigue design curve - effects of surface roughness and material variability.

Description: The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. The Code specifies fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Existing fatigue strain-vs.-life ({var_epsilon}-N) data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. This report provides an overview of the existing fatigue {var_epsilon}-N data for carbon and low-alloy steels and wrought and cast austenitic SSs to define the effects of key material, loading, and environmental parameters on the fatigue lives of the steels. Experimental data are presented on the effects of surface roughness on the fatigue life of these steels in air and LWR environments. Statistical models are presented for estimating the fatigue {var_epsilon}-N curves as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are discussed. Data available in the literature have been reviewed to evaluate the conservatism in the existing ASME Code fatigue evaluations. A critical review of the margins for ASME Code fatigue design curves is presented.
Date: October 3, 2003
Creator: Chopra, O. K.; Shack, W. J. & Technology, Energy
Partner: UNT Libraries Government Documents Department

Global Nuclear Energy Partnership Programmatic Environmental Impact Statement

Description: Abstract: The proposed Global Nuclear Energy Partnership (GNEP) Program, which is part of the President’s Advanced Energy Initiative, is intended to support a safe, secure, and sustainable expansion of nuclear energy, both domestically and internationally. Domestically, the GNEP Program would promote technologies that support economic, sustained production of nuclear-generated electricity, while reducing the impacts associated with spent nuclear fuel disposal and reducing proliferation risks. The Department of Energy (DOE) proposed action envisions changing the United States nuclear energy fuel cycle from an open (or once-through) fuel cycle—in which nuclear fuel is used in a power plant one time and the resulting spent nuclear fuel is stored for eventual disposal in a geologic repository—to a closed fuel cycle in which spent nuclear fuel would be recycled to recover energy-bearing components for use in new nuclear fuel. At this time, DOE has no specific proposed actions for the international component of the GNEP Program. Rather, the United States, through the GNEP Program, is considering various initiatives to work cooperatively with other nations. Such initiatives include the development of grid-appropriate reactors and the development of reliable fuel services (to provide an assured supply of fresh nuclear fuel and assist with the management of the used fuel) for nations who agree to employ nuclear energy only for peaceful purposes, such as electricity generation.
Date: October 1, 2008
Creator: Wigeland, R.A.
Partner: UNT Libraries Government Documents Department

Alternative Waste Forms for Electro-Chemical Salt Waste

Description: This study was undertaken to examine alternate crystalline (ceramic/mineral) and glass waste forms for immobilizing spent salt from the Advanced Fuel Cycle Initiative (AFCI) electrochemical separations process. The AFCI is a program sponsored by U.S. Department of Energy (DOE) to develop and demonstrate a process for recycling spent nuclear fuel (SNF). The electrochemical process is a molten salt process for the reprocessing of spent nuclear fuel in an electrorefiner and generates spent salt that is contaminated with alkali, alkaline earths, and lanthanide fission products (FP) that must either be cleaned of fission products or eventually replaced with new salt to maintain separations efficiency. Currently, these spent salts are mixed with zeolite to form sodalite in a glass-bonded waste form. The focus of this study was to investigate alternate waste forms to immobilize spent salt. On a mole basis, the spent salt is dominated by alkali and Cl with minor amounts of alkaline earth and lanthanides. In the study reported here, we made an effort to explore glass systems that are more compatible with Cl and have not been previously considered for use as waste forms. In addition, alternate methods were explored with the hope of finding a way to produce a sodalite that is more accepting of as many FP present in the spent salt as possible. This study was done to investigate two different options: (1) alternate glass families that incorporate increased concentrations of Cl; and (2) alternate methods to produce a mineral waste form.
Date: October 28, 2009
Creator: Crum, Jarrod V.; Sundaram, S. K.; Riley, Brian J.; Matyas, Josef; Arreguin, Shelly A. & Vienna, John D.
Partner: UNT Libraries Government Documents Department

Integrated safeguards & security for material protection, accounting, and control.

Description: Traditional safeguards and security design for fuel cycle facilities is done separately and after the facility design is near completion. This can result in higher costs due to retrofits and redundant use of data. Future facilities will incorporate safeguards and security early in the design process and integrate the systems to make better use of plant data and strengthen both systems. The purpose of this project was to evaluate the integration of materials control and accounting (MC&A) measurements with physical security design for a nuclear reprocessing plant. Locations throughout the plant where data overlap occurs or where MC&A data could be a benefit were identified. This mapping is presented along with the methodology for including the additional data in existing probabilistic assessments to evaluate safeguards and security systems designs.
Date: October 1, 2009
Creator: Duran, Felicia Angelica & Cipiti, Benjamin B.
Partner: UNT Libraries Government Documents Department

Initial Laboratory-Scale Melter Test Results for Combined Fission Product Waste

Description: This report describes the methods and results used to vitrify a baseline glass, CSLNTM-C-2.5 in support of the AFCI (Advanced Fuel Cycle Initiative) using a Quartz Crucible Scale Melter at the Pacific Northwest National Laboratory. Document number AFCI-WAST-PMO-MI-DV-2009-000184.
Date: October 1, 2009
Creator: Riley, Brian J.; Crum, Jarrod V.; Buchmiller, William C.; Rieck, Bennett T.; Schweiger, Michael J. & Vienna, John D.
Partner: UNT Libraries Government Documents Department

Chemical effects head-loss research in support of generic safety issue 191.

Description: This summary report describes studies conducted at Argonne National Laboratory on the potential for chemical effects on head loss across sump screens. Three different buffering solutions were used for these tests: trisodium phosphate (TSP), sodium hydroxide, and sodium tetraborate. These pH control agents used following a LOCA at a nuclear power plant show various degrees of interaction with the insulating materials Cal-Sil and NUKON. Results for Cal-Sil dissolution tests in TSP solutions, settling rate tests of calcium phosphate precipitates, and benchmark tests in chemically inactive environments are also presented. The dissolution tests were intended to identify important environmental variables governing both calcium dissolution and subsequent calcium phosphate formation over a range of simulated sump pool conditions. The results from the dissolution testing were used to inform both the head loss and settling test series. The objective of the head loss tests was to assess the head loss produced by debris beds created by Cal-Sil, fibrous debris, and calcium phosphate precipitates. The effects of both the relative arrival time of the precipitates and insulation debris and the calcium phosphate formation process were specifically evaluated. The debris loadings, test loop flow rates, and test temperature were chosen to be reasonably representative of those expected in plants with updated sump screen configurations, although the approach velocity of 0.1 ft/s used for most of the tests is 3-10 times that expected in plants with large screens . Other variables were selected with the intent to reasonably bound the head loss variability due to arrival time and calcium phosphate formation uncertainty. Settling tests were conducted to measure the settling rates of calcium phosphate precipitates (formed by adding dissolved Ca to boric acid and TSP solutions) in water columns having no bulk directional flow. For PWRs where NaOH and sodium tetraborate are used to control sump ...
Date: October 31, 2006
Creator: Park, J. H.; Kasza, K.; Fisher, B.; Oras, J.; Natesan, K.; Shack, W. J. et al.
Partner: UNT Libraries Government Documents Department

Autonomous Control of Nuclear Power Plants

Description: A nuclear reactor is a complex system that requires highly sophisticated controllers to ensure that desired performance and safety can be achieved and maintained during its operations. Higher-demanding operational requirements such as reliability, lower environmental impacts, and improved performance under adverse conditions in nuclear power plants, coupled with the complexity and uncertainty of the models, necessitate the use of an increased level of autonomy in the control methods. In the opinion of many researchers, the tasks involved during nuclear reactor design and operation (e.g., design optimization, transient diagnosis, and core reload optimization) involve important human cognition and decisions that may be more easily achieved with intelligent methods such as expert systems, fuzzy logic, neural networks, and genetic algorithms. Many experts in the field of control systems share the idea that a higher degree of autonomy in control of complex systems such as nuclear plants is more easily achievable through the integration of conventional control systems and the intelligent components. Researchers have investigated the feasibility of the integration of fuzzy logic, neural networks, genetic algorithms, and expert systems with the conventional control methods to achieve higher degrees of autonomy in different aspects of reactor operations such as reactor startup, shutdown in emergency situations, fault detection and diagnosis, nuclear reactor alarm processing and diagnosis, and reactor load-following operations, to name a few. With the advancement of new technologies and computing power, it is feasible to automate most of the nuclear reactor control and operation, which will result in increased safety and economical benefits. This study surveys current status, practices, and recent advances made towards developing autonomous control systems for nuclear reactors.
Date: October 20, 2003
Creator: Basher, H.
Partner: UNT Libraries Government Documents Department

ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology

Description: We describe the next generation general purpose Evaluated Nuclear Data File, ENDF/B-VII.0, of recommended nuclear data for advanced nuclear science and technology applications. The library, released by the U.S. Cross Section Evaluation Working Group (CSEWG) in December 2006, contains data primarily for reactions with incident neutrons, protons, and photons on almost 400 isotopes. The new evaluations are based on both experimental data and nuclear reaction theory predictions. The principal advances over the previous ENDF/B-VI library are the following: (1) New cross sections for U, Pu, Th, Np and Am actinide isotopes, with improved performance in integral validation criticality and neutron transmission benchmark tests; (2) More precise standard cross sections for neutron reactions on H, {sup 6}Li, {sup 10}B, Au and for {sup 235,238}U fission, developed by a collaboration with the IAEA and the OECD/NEA Working Party on Evaluation Cooperation (WPEC); (3) Improved thermal neutron scattering; (4) An extensive set of neutron cross sections on fission products developed through a WPEC collaboration; (5) A large suite of photonuclear reactions; (6) Extension of many neutron- and proton-induced reactions up to an energy of 150 MeV; (7) Many new light nucleus neutron and proton reactions; (8) Post-fission beta-delayed photon decay spectra; (9) New radioactive decay data; and (10) New methods developed to provide uncertainties and covariances, together with covariance evaluations for some sample cases. The paper provides an overview of this library, consisting of 14 sublibraries in the same, ENDF-6 format, as the earlier ENDF/B-VI library. We describe each of the 14 sublibraries, focusing on neutron reactions. Extensive validation, using radiation transport codes to simulate measured critical assemblies, show major improvements: (a) The long-standing underprediction of low enriched U thermal assemblies is removed; (b) The {sup 238}U, {sup 208}Pb, and {sup 9}Be reflector biases in fast systems are largely removed; (c) ENDF/B-VI.8 good ...
Date: October 2, 2006
Creator: Chadwick, M B; Oblozinsky, P; Herman, M; Greene, N M; McKnight, R D; Smith, D L et al.
Partner: UNT Libraries Government Documents Department

Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses

Description: This report presents studies performed to support the development of a technically justifiable approach for addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality safety analyses. The effect of the axial-burnup distribution on reactivity and proposed approaches for addressing the axial-burnup distribution are briefly reviewed. A publicly available database of profiles is examined in detail to identify profiles that maximize the neutron multiplication factor, k{sub eff}, assess its adequacy for PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. A statistical evaluation of the k{sub eff} values associated with the profiles in the axial-burnup-profile database was performed, and the most reactive (bounding) profiles were identified as statistical outliers. The impact of these bounding profiles on k{sub eff} is quantified for a high-density burnup credit cask. Analyses are also presented to quantify the potential reactivity consequence of loading assemblies with axial-burnup profiles that are not bounded by the database. The report concludes with a discussion on the issues for consideration and recommendations for addressing axial burnup in criticality safety analyses using burnup credit for dry cask storage and transportation.
Date: October 23, 2002
Creator: Wagner, J.C.
Partner: UNT Libraries Government Documents Department

Tabulation of Fundamental Assembly Heat and Radiation Source Files

Description: The purpose of this calculation is to tabulate a set of computer files for use as input to the WPLOAD thermal loading software. These files contain details regarding heat and radiation from pressurized water reactor (PWR) assemblies and boiling water reactor (BWR) assemblies. The scope of this calculation is limited to rearranging and reducing the existing file information into a more streamlined set of tables for use as input to WPLOAD. The electronic source term files used as input to this calculation were generated from the output files of the SAS2H/ORIGIN-S sequence of the SCALE Version 4.3 modular code system, as documented in References 2.1.1 and 2.1.2, and are included in Attachment II.
Date: October 25, 2006
Creator: deBues, T. & Ryman, J.C.
Partner: UNT Libraries Government Documents Department

Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs

Description: This report documents the work performed during the first phase of the National Aeronautics and Space Administration (NASA), National Research Announcement (NRA) Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs. The document includes an optimization of both 100-kW{sub e} and 250-kW{sub e} (at the propulsion unit) Rankine cycle power conversion systems. In order to perform the mass optimization of these systems, several parametric evaluations of different design options were investigated. These options included feed and reheat, vapor superheat levels entering the turbine, three different material types, and multiple heat rejection system designs. The overall masses of these Nb-1%Zr systems are approximately 3100 kg and 6300 kg for the 100- kW{sub e} and 250-kW{sub e} systems, respectively, each with two totally redundant power conversion units, including the mass of the single reactor and shield. Initial conceptual designs for each of the components were developed in order to estimate component masses. In addition, an overall system concept was presented that was designed to fit within the launch envelope of a heavy lift vehicle. A technology development plan is presented in the report that describes the major efforts that are required to reach a technology readiness level of 6. A 10-year development plan was proposed.
Date: October 3, 2005
Creator: Yoder, G.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

Advanced Test Reactor Capabilities and Future Irradiation Plans

Description: The Advanced Test Reactor (ATR), located at the Idaho National Laboratory (INL), is one of the most versatile operating research reactors in the Untied States. The ATR has a long history of supporting reactor fuel and material research for the US government and other test sponsors. The INL is owned by the US Department of Energy (DOE) and currently operated by Battelle Energy Alliance (BEA). The ATR is the third generation of test reactors built at the Test Reactor Area, now named the Reactor Technology Complex (RTC), whose mission is to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The current experiments in the ATR are for a variety of customers--US DOE, foreign governments and private researchers, and commercial companies that need neutrons. The ATR has several unique features that enable the reactor to perform diverse simultaneous tests for multiple test sponsors. The ATR has been operating since 1967, and is expected to continue operating for several more decades. The remainder of this paper discusses the ATR design features, testing options, previous experiment programs, future plans for the ATR capabilities and experiments, and some introduction to the INL and DOE's expectations for nuclear research in the future.
Date: October 1, 2006
Creator: Marshall, Frances M.
Partner: UNT Libraries Government Documents Department

Engineering and Physics Optimization of Breed and Burn Fast Reactor Systems: Annual and Final Report

Description: The Idaho National Laboratory (INL) contribution to the Nuclear Energy Research Initiative (NERI) project number 2002-005 was divided into reactor physics, and thermal-hydraulics and plant design. The research targeted credible physics and thermal-hydraulics models for a gas-cooled fast reactor, analyzing various fuel and in-core fuel cycle options to achieve a true breed and burn core, and performing a design basis Loss of Coolant Accident (LOCA) analysis on that design. For the physics analysis, a 1/8 core model was created using different enrichments and simulated equilibrium fuel loadings. The model was used to locate the hot spot of the reactor, and the peak to average energy deposition at that location. The model was also used to create contour plots of the flux and energy deposition over the volume of the reactor. The eigenvalue over time was evaluated using three different fuel configurations with the same core geometry. The breeding capabilities of this configuration were excellent for a 7% U-235 model and good in both a plutonium model and a 14% U-235 model. Changing the fuel composition from the Pu fuel which provided about 78% U-238 for breeding to the 14% U-235 fuel with about 86% U-238 slowed the rate of decrease in the eigenvalue a noticeable amount. Switching to the 7% U-235 fuel with about 93% U-238 showed an increase in the eigenvalue over time. For the thermal-hydraulic analysis, the reactor design used was the one forwarded by the MIT team. This reactor design uses helium coolant, a Brayton cycle, and has a thermal power of 600 MW. The core design parameters were supplied by MIT; however, the other key reactor components that were necessary for a plausible simulation of a LOCA were not defined. The thermal-hydraulic and plant design research concentrated on determining reasonable values for those undefined components. The ...
Date: October 1, 2005
Creator: Weaver, Kevan D.; Marshall, Theron & Parry, James
Partner: UNT Libraries Government Documents Department

REACTOR PRESSURE VESSEL TEMPERATURE ANALYSIS OF CANDIDATE VERY HIGH TEMPERATURE REACTOR DESIGNS

Description: Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated 600 MW of thermal power. Calculations were performed for fluid outlet temperatures of 900 and 950 °C, corresponding to the expected range for the VHTR. The analyses were performed using the RELAP5-3D and PEBBED-THERMIX computer codes. Results of the calculations were compared with preliminary temperature limits derived from the ASME pressure vessel code. Because PEBBED-THERMIX has not been extensively validated, confirmatory calculations were also performed with RELAP5-3D for the pebble-bed design. During normal operation, the predicted axial profiles in reactor vessel temperature were similar with both codes and the predicted maximum values were within 2 °C. The trends of the calculated vessel temperatures were similar during the depressurized conduction cooldown accident. The maximum value predicted with RELAP5-3D during the depressurized conduction cooldown accident was about 40 °C higher than that predicted with PEBBED. This agreement is considered reasonable based on the expected uncertainty in either calculation. The differences between the PEBBED and RELAP5-3D calculations were not large enough to affect conclusions concerning comparisons between calculated and allowed maximum temperatures during normal operation and the depressurized conduction cooldown accident.
Date: October 1, 2006
Creator: Gougar, Hans D.; Davis, Cliff B.; Hayner, George & Weaver, Kevan
Partner: UNT Libraries Government Documents Department

Design of the Advanced Gas Reactor Fuel Experiments for Irradiation in the Advanced Test Reactor

Description: The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight particle fuel tests in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL) to support development of the next generation Very High Temperature Reactor (VHTR) in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments will be irradiated in an inert sweep gas atmosphere with on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The final design phase has just been completed on the first experiment (AGR-1) in this series and the support systems and fission product monitoring system that will monitor and control the experiment during irradiation. This paper discusses the development of the experimental hardware and support system designs and the status of the experiment.
Date: October 1, 2005
Creator: Grover, S. Blaine
Partner: UNT Libraries Government Documents Department

Capabilities and Facilities Available at the Advanced Test Reactor to Support Development of the Next Generation Reactors

Description: The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. It is a very versatile facility with a wide variety of experimental test capabilities for providing the environment needed in an irradiation experiment. These different capabilities include passive sealed capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. The Irradiation Test Vehicle (ITV) installed in 1999 enhanced these capabilities by providing a built in experiment monitoring and control system for instrumented and/or temperature controlled experiments. This built in control system significantly reduces the cost for an actively monitored/temperature controlled experiments by providing the thermocouple connections, temperature control system, and temperature control gas supply and exhaust systems already in place at the irradiation position. Although the ITV in-core hardware was removed from the ATR during the last core replacement completed in early 2005, it (or a similar facility) could be re-installed for an irradiation program when the need arises. The proposed Gas Test Loop currently being designed for installation in the ATR will provide additional capability for testing of not only gas reactor materials and fuels but will also include enhanced fast flux rates for testing of materials and fuels for other next generation reactors including preliminary testing for fast reactor fuels and materials. This paper discusses the different irradiation capabilities available and the cost benefit issues related to each capability.
Date: October 1, 2005
Creator: Grover, S. Blaine & Furstenau, Raymond V.
Partner: UNT Libraries Government Documents Department

Safety Assurance for ATR Irradiations

Description: The Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL) is the world’s premiere test reactor for performing high fluence, large volume, irradiation test programs. The ATR has many capabilities and a wide variety of tests are performed in this truly one of a kind reactor, including isotope production, simple self-contained static capsule experiments, instrumented/controlled experiments, and loop testing under pressurized water conditions. Along with the five pressurized water loops, ATR may also have gas (temperature controlled) lead experiments, fuel boosted fast flux experiments, and static sealed capsules all in the core at the same time. In addition, any or all of these tests may contain fuel or moderating materials that can affect reactivity levels in the ATR core. Therefore the safety analyses required to ensure safe operation of each experiment as well as the reactor itself are complex. Each test has to be evaluated against stringent reactor control safety criteria, as well as the effects it could have on adjacent tests and the reactor as well as the consequences of those effects. The safety analyses of each experiment are summarized in a document entitled the Experiment Safety Assurance Package (ESAP). The ESAP references and employs the results of the reactor physics, thermal, hydraulic, stress, seismic, vibration, and all other analyses necessary to ensure the experiment can be irradiated safely in the ATR. The requirements for reactivity worth, chemistry compatibilities, pressure limitations, material issues, etc. are all specified in the Technical Safety Requirements and the Upgraded Final Safety Analysis Report (UFSAR) for the ATR. This paper discusses the ESAP process, types of analyses, types of safety requirements and the approvals necessary to ensure an experiment can be safely irradiated in the ATR.
Date: October 1, 2006
Creator: Grover, S. Blaine
Partner: UNT Libraries Government Documents Department

The Modular Helium Reactor for Hydrogen Production

Description: For electricity and hydrogen production, an advanced reactor technology receiving considerable international interest is a modular, passively-safe version of the high-temperature, gas-cooled reactor (HTGR), known in the U.S. as the Modular Helium Reactor (MHR), which operates at a power level of 600 MW(t). For hydrogen production, the concept is referred to as the H2-MHR. Two concepts that make direct use of the MHR high-temperature process heat are being investigated in order to improve the efficiency and economics of hydrogen production. The first concept involves coupling the MHR to the Sulfur-Iodine (SI) thermochemical water splitting process and is referred to as the SI-Based H2-MHR. The second concept involves coupling the MHR to high-temperature electrolysis (HTE) and is referred to as the HTE-Based H2-MHR.
Date: October 1, 2006
Creator: Harvego, E.; Richards, M.; Shenoy, A.; Schultz, K.; Brown, L. & Fukuie, M.
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

Testing of Gas Reactor Fuel and Materials in the Advanced Test Reactor

Description: The recent growth in interest for high temperature gas reactors has resulted in an increased need for materials and fuel testing for this type of reactor. The Advanced Test Reactor (ATR), located at the US Department of Energy’s Idaho National Laboratory, has long been involved in testing gas reactor fuel and materials, and has facilities and capabilities to provide the right environment for gas reactor irradiation experiments. These capabilities include both passive sealed capsule experiments, and instrumented/actively controlled experiments. The instrumented/actively controlled experiments typically contain thermocouples and control the irradiation temperature, but on-line measurements and controls for pressure and gas environment have also been performed in past irradiations. The ATR has an existing automated gas temperature control system that can maintain temperature in an irradiation experiment within very tight bounds, and has developed an on-line fission product monitoring system that is especially well suited for testing gas reactor particle fuel. The ATR’s control system, which consists primarily of vertical cylinders used to rotate neutron poisons/reflectors toward or away from the reactor core, provides a constant vertical flux profile over the duration of each operating cycle. This constant chopped cosine shaped axial flux profile, with a relatively flat peak at the vertical centre of the core, is more desirable for experiments than a constantly moving axial flux peak resulting from a control system of axially positioned control components which are vertically withdrawn from the core.
Date: October 1, 2006
Creator: Grover, S. Blaine
Partner: UNT Libraries Government Documents Department

Theoretical Design of a Thermosyphon for Efficient Process Heat Removal from Next Generation Nuclear Plant (NGNP) for Production of Hydrogen

Description: The work reported here is the preliminary analysis of two-phase Thermosyphon heat transfer performance with various alkali metals. Thermosyphon is a device for transporting heat from one point to another with quite extraordinary properties. Heat transport occurs via evaporation and condensation, and the heat transport fluid is re-circulated by gravitational force. With this mode of heat transfer, the thermosyphon has the capability to transport heat at high rates over appreciable distances, virtually isothermally and without any requirement for external pumping devices. For process heat, intermediate heat exchangers (IHX) are required to transfer heat from the NGNP to the hydrogen plant in the most efficient way possible. The production of power at higher efficiency using Brayton Cycle, and hydrogen production requires both heat at higher temperatures (up to 1000oC) and high effectiveness compact heat exchangers to transfer heat to either the power or process cycle. The purpose for selecting a compact heat exchanger is to maximize the heat transfer surface area per volume of heat exchanger; this has the benefit of reducing heat exchanger size and heat losses. The IHX design requirements are governed by the allowable temperature drop between the outlet of the NGNP (900oC, based on the current capabilities of NGNP), and the temperatures in the hydrogen production plant. Spiral Heat Exchangers (SHE’s) have superior heat transfer characteristics, and are less susceptible to fouling. Further, heat losses to surroundings are minimized because of its compact configuration. SHEs have never been examined for phase-change heat transfer applications. The research presented provides useful information for thermosyphon design and Spiral Heat Exchanger.
Date: October 1, 2007
Creator: Sabharwall, Piyush; Gunnerson, Fred; Tokuhiro, Akira; Utgiker, Vivek; Weaver, Kevan & Sherman, Steven
Partner: UNT Libraries Government Documents Department

PEBBLES: A COMPUTER CODE FOR MODELING PACKING, FLOW AND RECIRCULATIONOF PEBBLES IN A PEBBLE BED REACTOR

Description: A comprehensive, high fidelity model for pebble flow has been developed and embodied in the PEBBLES computer code. In this paper, a description of the physical artifacts included in the model is presented and some results from using the computer code for predicting the features of pebble flow and packing in a realistic pebble bed reactor design are shown. The sensitivity of models to various physical parameters is also discussed.
Date: October 1, 2006
Creator: Cogliati, Joshua J. & Ougouag, Abderrafi M.
Partner: UNT Libraries Government Documents Department

The role of Z-pinch fusion transmutation of waste in the nuclear fuel cycle.

Description: The resurgence of interest in reprocessing in the United States with the Global Nuclear Energy Partnership has led to a renewed look at technologies for transmuting nuclear waste. Sandia National Laboratories has been investigating the use of a Z-Pinch fusion driver to burn actinide waste in a sub-critical reactor. The baseline design has been modified to solve some of the engineering issues that were identified in the first year of work, including neutron damage and fuel heating. An on-line control feature was added to the reactor to maintain a constant neutron multiplication with time. The transmutation modeling effort has been optimized to produce more accurate results. In addition, more attention was focused on the integration of this burner option within the fuel cycle including an investigation of overall costs. This report presents the updated reactor design, which is able to burn 1320 kg of actinides per year while producing 3,000 MWth.
Date: October 1, 2007
Creator: Smith, James Dean; Drennen, Thomas E. (Hobart & William Smith College, Geneva, NY); Rochau, Gary Eugene; Martin, William Joseph; Kamery, William (Hobart & William Smith College, Geneva, NY); Phruksarojanakun, Phiphat (University of Wisconsin, Madison, WI) et al.
Partner: UNT Libraries Government Documents Department