Testimony of Fred R. Mynatt before the Energy Research and Development Subcommittee of the Committee on Science, Space, and Technology, US House of Representatives. [Advanced fuel technology, gas-cooled reactor technology, and liquid metal-cooled reactor technology programs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mynatt, F.R.

1987-03-18

This report provides a description of the statements submitted for the record to the committee on Science, Space, and Technology of the United States House of Representatives. These statements describe three principal areas of activity of the Advanced Reactor Technology Program of the Department of Energy (DOE). These areas are advanced fuel cycle technology, modular high-temperature gas-cooled reactor technology, and liquid metal-cooled reactor. The areas of automated reactor control systems, robotics, materials and structural design shielding and international cooperation were included in these statements describing the Oak Ridge National Laboratory's efforts in these areas. (FI)

Review of Rover fuel element protective coating development at Los Alamos

NASA Technical Reports Server (NTRS)

Wallace, Terry C.

1991-01-01

The Los Alamos Scientific Laboratory (LASL) entered the nuclear propulsion field in 1955 and began work on all aspects of a nuclear propulsion program with a target exhaust temperature of about 2750 K. A very extensive chemical vapor deposition coating technology for preventing catastrophic corrosion of reactor core components by the high temperature, high pressure hydrogen propellant gas was developed. Over the 17-year term of the program, more than 50,000 fuel elements were coated and evaluated. Advances in performance were achieved only through closely coupled interaction between the developing fuel element fabrication and protective coating technologies. The endurance of fuel elements in high temperature, high pressure hydrogen environment increased from several minutes at 2000 K exit gas temperature to 2 hours at 2440 K exit gas temperature in a reactor test and 10 hours at 2350 K exit gas temperature in a hot gas test. The purpose of this paper is to highlight the rationale for selection of coating materials used (NbC and ZrC), identify critical fuel element-coat interactions that had to be modified to increase system performance, and review the evolution of protective coating technology.

Molten Salts for High Temperature Reactors: University of Wisconsin Molten Salt Corrosion and Flow Loop Experiments -- Issues Identified and Path Forward

DOE Office of Scientific and Technical Information (OSTI.GOV)

Piyush Sabharwall; Matt Ebner; Manohar Sohal

2010-03-01

Considerable amount of work is going on regarding the development of high temperature liquid salts technology to meet future process needs of Next Generation Nuclear Plant. This report identifies the important characteristics and concerns of high temperature molten salts (with lesson learned at University of Wisconsin-Madison, Molten Salt Program) and provides some possible recommendation for future work

NASA-EPA automotive thermal reactor technology program

NASA Technical Reports Server (NTRS)

Blankenship, C. P.; Hibbard, R. R.

1972-01-01

The status of the NASA-EPA automotive thermal reactor technology program is summarized. This program is concerned primarily with materials evaluation, reactor design, and combustion kinetics. From engine dynamometer tests of candidate metals and coatings, two ferritic iron alloys (GE 1541 and Armco 18-SR) and a nickel-base alloy (Inconel 601) offer promise for reactor use. None of the coatings evaluated warrant further consideration. Development studies on a ceramic thermal reactor appear promising based on initial vehicle road tests. A chemical kinetic study has shown that gas temperatures of at least 900 K to 1000 K are required for the effective cleanup of carbon monoxide and hydrocarbons, but that higher temperatures require shorter combustion times and thus may permit smaller reactors.

Design of conduction cooling system for a high current HTS DC reactor

NASA Astrophysics Data System (ADS)

Dao, Van Quan; Kim, Taekue; Le Tat, Thang; Sung, Haejin; Choi, Jongho; Kim, Kwangmin; Hwang, Chul-Sang; Park, Minwon; Yu, In-Keun

2017-07-01

A DC reactor using a high temperature superconducting (HTS) magnet reduces the reactor’s size, weight, flux leakage, and electrical losses. An HTS magnet needs cryogenic cooling to achieve and maintain its superconducting state. There are two methods for doing this: one is pool boiling and the other is conduction cooling. The conduction cooling method is more effective than the pool boiling method in terms of smaller size and lighter weight. This paper discusses a design of conduction cooling system for a high current, high temperature superconducting DC reactor. Dimensions of the conduction cooling system parts including HTS magnets, bobbin structures, current leads, support bars, and thermal exchangers were calculated and drawn using a 3D CAD program. A finite element method model was built for determining the optimal design parameters and analyzing the thermo-mechanical characteristics. The operating current and inductance of the reactor magnet were 1,500 A, 400 mH, respectively. The thermal load of the HTS DC reactor was analyzed for determining the cooling capacity of the cryo-cooler. The study results can be effectively utilized for the design and fabrication of a commercial HTS DC reactor.

Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

NASA Astrophysics Data System (ADS)

Cisneros, Anselmo Tomas, Jr.

The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP and PEBBED for a high temperature gas cooled pebble bed reactor. Three parametric studies were performed for exploring the design space of the PB-FHR---to select a fuel design for the PB-FHR] to select a core configuration; and to optimize the PB-FHR design. These parametric studies investigated trends in the dependence of important reactor performance parameters such as burnup, temperature reactivity feedback, radiation damage, etc on the reactor design variables and attempted to understand the underlying reactor physics responsible for these trends. A pebble fuel parametric study determined that pebble fuel should be designed with a carbon to heavy metal ratio (C/HM) less than 400 to maintain negative coolant temperature reactivity coefficients. Seed and thorium blanket-, seed and inert pebble reflector- and seed only core configurations were investigated for annular FHR PBRs---the C/HM of the blanket pebbles and discharge burnup of the thorium blanket pebbles were additional design variable for core configurations with thorium blankets. Either a thorium blanket or graphite pebble reflector is required to shield the outer graphite reflector enough to extend its service lifetime to 60 EFPY. The fuel fabrication costs and long cycle lengths of the thorium blanket fuel limit the potential economic advantages of using a thorium blanket. Therefore, the seed and pebble reflector core configuration was adopted as the baseline core configuration. Multi-objective optimization with respect to economics was performed for the PB-FHR accounting for safety and other physical design constraints derived from the high-level safety regulatory criteria. These physical constraints were applied along in a design tool, Nuclear Application Value Estimator, that evaluated a simplified cash flow economics model based on estimates of reactor performance parameters calculated using correlations based on the results of parametric design studies for a specific PB-FHR design and a set of economic assumptions about the electricity market to evaluate the economic implications of design decisions. The optimal PB-FHR design---Mark 1 PB-FHR---is described along with a detailed summary of its performance characteristics including: the burnup, the burnup evolution, temperature reactivity coefficients, the power distribution, radiation damage distributions, control element worths, decay heat curves and tritium production rates. The Mk1 PB-FHR satisfies the PB-FHR safety criteria. The fuel, moderator (pebble core, pebble shell, graphite matrix, TRISO layers) and coolant have global negative temperature reactivity coefficients and the fuel temperatures are well within their limits.

High-Temperature Fluid-Wall Reactor Technology Research, Test and Evaluation Performed at Naval Construction Battalion Center, Gulfport, Mississippi, for the United States Air Force Installation/Restoration Program

DTIC Science & Technology

1988-01-01

the reactor Duties: The Process Engineers rotate with the Lead Operator to monitor the process at the top of the reactor through the site glass...pant cuffs and coverhoods of coveralls, will be attached to gloves, boots and coveralls, using duct tape. * IF AMBIENT WORK STATIONS TEMPERATURE IS...L of the sample fortification solution (Section ýý8) containing 1C 12-2,3,7,8-TCDD at a concentration of 0.5 ng/1,Land C14-2,3,7,8-TCDD at a

Atom probe tomography analysis of high dose MA957 at selected irradiation temperatures

NASA Astrophysics Data System (ADS)

Bailey, Nathan A.; Stergar, Erich; Toloczko, Mychailo; Hosemann, Peter

2015-04-01

Oxide dispersion strengthened (ODS) alloys are meritable structural materials for nuclear reactor systems due to the exemplary resistance to radiation damage and high temperature creep. Summarized in this work are atom probe tomography (APT) investigations on a heat of MA957 that underwent irradiation in the form of in-reactor creep specimens in the Fast Flux Test Facility-Materials Open Test Assembly (FFTF-MOTA) for the Liquid Metal Fast Breeder Reactor (LMFBR) program. The oxide precipitates appear stable under irradiation at elevated temperature over extended periods of time. Nominally, the precipitate chemistry is unchanged by the accumulated dose; although, evidence suggests that ballistic dissolution and reformation processes are occurring at all irradiation temperatures. At 412 °C-109 dpa, chromium enrichments - consistent with the α‧ phase - appear between the oxide precipitates, indicating radiation induced segregation. Grain boundaries, enriched with several elements including nickel and titanium, are observed at all irradiation conditions. At 412 °C-109 dpa, the grain boundaries are also enriched in molecular titanium oxide (TiO).

Aerodynamic drag characterization and deposition studies of irregular particles. Part 3: Analysis of flow and temperature field inside the Combustion Deposition Entrained Reactor (CDER)

NASA Astrophysics Data System (ADS)

Celik, I.; Katragadda, S.; Nagarajan, R.

1990-01-01

An experimental and numerical analysis was performed of the temperature and flow field involved in co-axial, confined, non-reacting heated jets in a drop tube reactor. An electrically heated 2-inch (50.8 mm) diameter drop tube reactor was utilized to study the jet characteristics. Profiles of gas temperature, typically in the range of 800 to 1600 K were measured in the mixing zone of the jet with a K-Type thermocouple. Measured temperatures were corrected for conduction, convection, and radiation heat losses. Because of limited access to the mixing zone, characterization of the flow field at high temperatures with laser Doppler or hot wire anemometry were impractical. A computer program which solves the full equations of motion and energy was employed to simulate the temperature and flow fields. The location of the recirculation region, the flow regimes, and the mixing phenomena were studied. The wall heating, laminar and turbulent flow regimes were considered in the simulations. The predictions are in fairly good agreement with the corrected temperature measurements provided that the flow is turbulent. The results of this study demonstrate how a numerical method and measurement can be used together to analyze the flow conditions inside a reactor which has limited access because of very high temperatures.

Determination of the Arrhenius Activation Energy Using a Temperature-Programmed Flow Reactor.

ERIC Educational Resources Information Center

Chan, Kit-ha C.; Tse, R. S.

1984-01-01

Describes a novel method for the determination of the Arrhenius activation energy, without prejudging the validity of the Arrhenius equation or the concept of activation energy. The method involves use of a temperature-programed flow reactor connected to a concentration detector. (JN)

Space reactor power 1986 - A year of choices and transition

NASA Technical Reports Server (NTRS)

Wiley, R. L.; Verga, R. L.; Schnyer, A. D.; Sholtis, J. A., Jr.; Wahlquist, E. J.

1986-01-01

Both the SP-100 and Multimegawatt programs have made significant progress over the last year and that progress is the focus of this paper. In the SP-100 program the thermoelectric energy conversion concept powered by a compact, high-temperature, lithium-cooled, uranium-nitride-fueled fast spectrum reactor was selected for engineering development and ground demonstration testing at an electrical power level of 300 kilowatts. In the Multimegawatt program, activities moved from the planning phase into one of technology development and assessment with attendant preliminary definition and evaluation of power concepts against requirements of the Strategic Defense Initiative.

The use of a very high temperature nuclear reactor in the manufacture of synthetic fuels

NASA Technical Reports Server (NTRS)

Farbman, G. H.; Brecher, L. E.

1976-01-01

The three parts of a program directed toward creating a cost-effective nuclear hydrogen production system are described. The discussion covers the development of a very high temperature nuclear reactor (VHTR) as a nuclear heat and power source capable of producing the high temperature needed for hydrogen production and other processes; the development of a hydrogen generation process based on water decomposition, which can utilize the outputs of the VHTR and be integrated with many different ultimate hydrogen consuming processes; and the evaluation of the process applications of the nuclear hydrogen systems to assess the merits and potential payoffs. It is shown that the use of VHTR for the manufacture of synthetic fuels appears to have a very high probability of making a positive contribution to meeting the nation's energy needs in the future.

High pressure/high temperature thermogravimetric apparatus. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Calo, J.M.; Suuberg, E.M.

1999-12-01

The purpose of this instrumentation grant was to acquire a state-of-the-art, high pressure, high temperature thermogravimetric apparatus (HP/HT TGA) system for the study of the interactions between gases and carbonaceous solids for the purpose of solving problems related to coal utilization and applications of carbon materials. The instrument that we identified for this purpose was manufactured by DMT (Deutsche Montan Technologies)--Institute of Cokemaking and Coal Chemistry of Essen, Germany. Particular features of note include: Two reactors: a standard TGA reactor, capable of 1100 C at 100 bar; and a high temperature (HT) reactor, capable of operation at 1600 C andmore » 100 bar; A steam generator capable of generating steam to 100 bar; Flow controllers and gas mixing system for up to three reaction gases, plus a separate circuit for steam, and another for purge gas; and An automated software system for data acquisition and control. The HP/TP DMT-TGA apparatus was purchased in 1996 and installed and commissioned during the summer of 1996. The apparatus was located in Room 128 of the Prince Engineering Building at Brown University. A hydrogen alarm and vent system were added for safety considerations. The system has been interfaced to an Ametek quadruple mass spectrometer (MA 100), pumped by a Varian V250 turbomolecular pump, as provided for in the original proposed. With this capability, a number of gas phase species of interest can be monitored in a near-simultaneous fashion. The MS can be used in a few different modes. During high pressure, steady-state gasification experiments, it is used to sample, measure, and monitor the reactant/product gases. It can also be used to monitor gas phase species during nonisothermal temperature programmed reaction (TPR) or temperature programmed desorption (TPD) experiments.« less

Gaseous fuel reactors for power systems

NASA Technical Reports Server (NTRS)

Kendall, J. S.; Rodgers, R. J.

1977-01-01

Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

Johnson Noise Thermometry for Advanced Small Modular Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Britton, C.L.,Jr.; Roberts, M.; Bull, N.D.

Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor’s physical condition. In and near the core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurementmore » due to their fundamental natures. Small Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of the current ORNL-led project, conducted under the Instrumentation, Controls, and Human-Machine Interface (ICHMI) research pathway of the U.S. Department of Energy (DOE) Advanced SMR Research and Development (R&D) program, is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.« less

Testing piezoelectric sensors in a nuclear reactor environment

NASA Astrophysics Data System (ADS)

Reinhardt, Brian T.; Suprock, Andy; Tittmann, Bernhard

2017-02-01

Several Department of Energy Office of Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development (FCRD), Advanced Reactor Concepts (ARC), Light Water Reactor Sustainability, and Next Generation Nuclear Power Plants (NGNP), are investigating new fuels, materials, and inspection paradigms for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials during irradiation. In DOE-NE's FCRD program, ultrasonic based technology was identified as a key approach that should be pursued to obtain the high-fidelity, high-accuracy data required to characterize the behavior and performance of new candidate fuels and structural materials during irradiation testing. The radiation, high temperatures, and pressure can limit the available tools and characterization methods. In this work piezoelectric transducers capable of making these measurements are developed. Specifically, three piezoelectric sensors (Bismuth Titanate, Aluminum Nitride, and Zinc Oxide) are tested in the Massachusetts Institute of Technology Research reactor to a fast neutron fluence of 8.65×1020 nf/cm2. It is demonstrated that Bismuth Titanate is capable of transduction up to 5 × 1020 nf/cm2, Zinc Oxide is capable of transduction up to at least 6.27 × 1020 nf/cm2, and Aluminum Nitride is capable of transduction up to at least 8.65 × 1020 nf/cm2.

NEET Enhanced Micro-Pocket Fission Detector for High Temperature Reactors - FY16 Status Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Unruh, Troy; Reichenberger, Michael; Stevenson, Sarah

2016-09-01

A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core fission chambers and thermocouple. As discussed within this report,more » the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year two of this three year project. Highlights from research accomplishments include: • Continuation of a joint collaboration between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. • An updated parallel wire HT MPFD design was developed. • Program support for HT MPFD deployments was given to Accident Tolerant Fuels (ATF) and Advanced Gas-cooled Reactor (AGR) irradiation test programs. • Quality approved materials for HT MPFD construction were procured by irradiation test programs for upcoming deployments. • KSU improved and performed electrical contact and fissile material plating. • KSU delivered fissile HT MPFD parts to INL for final construction of HT MPFD prototype. • A prototype HT MPFD was constructed and analyzed at INL. • The HT MPFD has been modeled in MCNP to optimize the amount of fissile material deposition. • The HT MPFD has been modeled in MCNP to optimize the sensor location in the irradiation test. • The fissile material deposition is undergoing independent verifications. • Detector amplifier electronics have been revised and tested by KSU. • Several project meetings were held at INL and KSU to discuss the roles and responsibilities between INL, KSU, and CEA for development and deployment of the HT MPFDs. As documented in this report, FY16 funding has allowed the project to meet year two planned accomplishments to develop a HT MPFD. In addition, the accomplishments of this project have attracted independent funding from other Department of Energy Office of Nuclear Energy (DOE-NE) programs for MTR irradiations of the MPFD technology. These are significant opportunities for this NEET Enhanced Micro-Pocket Fission Detector for High Temperature Reactors project because the irradiation expense of these experiments could not be included in the original project scope.« less

Development of toroid-type HTS DC reactor series for HVDC system

NASA Astrophysics Data System (ADS)

Kim, Kwangmin; Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho; Lee, Sangjin; Oh, Yunsang; Park, Minwon; Yu, In-Keun

2015-11-01

This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

Consolidated fuel reprocessing program

NASA Astrophysics Data System (ADS)

1985-04-01

A survey of electrochemical methods applications in fuel reprocessing was completed. A dummy fuel assembly shroud was cut using the remotely operated laser disassembly equipment. Operations and engineering efforts have continued to correct equipment operating, software, and procedural problems experienced during the previous uranium compaigns. Fuel cycle options were examined for the liquid metal reactor fuel cycle. In high temperature gas cooled reactor spent fuel studies, preconceptual designs were completed for the concrete storage cask and open field drywell storage concept. These and other tasks operating under the consolidated fuel reprocessing program are examined.

NEET Enhanced Micro Pocket Fission Detector for High Temperature Reactors - FY15 Status Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Unruh, Troy; McGregor, Douglas; Ugorowski, Phil

2015-09-01

A new project, that is a collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core parallel plate fission chambermore » and thermocouple. As discussed within this report, the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year one of this three year project. Highlights from research accomplishments include: A joint collaboration was initiated between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. An updated HT MPFD design was developed. New high temperature-compatible materials for HT MPFD construction were procured. Construction methods to support the new design were evaluated at INL. Laboratory evaluations of HT MPFD were initiated. Electrical contact and fissile material plating has been performed at KSU. Updated detector electronics are undergoing evaluations at KSU. A project meeting was held at KSU to discuss the roles and responsibilities between INL and KSU for development of the HT MPFDs. Provide input to various irradiation programs for installation of the MPFD technology in irradiation tests. As documented in this report, FY15 funding has allowed the project to meet year one planned accomplishments to develop a HT MPFD that offers US MTR users enhanced capabilities for real-time measurement of flux and temperature with a single detector. In addition, the accomplishments of this project have attracted funding from other Department of Energy Office of Nuclear Energy (DOE-NE) programs for additional applications. The work in those programs will build on current activities completed in this NEETASI HT MPFD project, but the MPFD will be specifically tailored to meet their program needs.« less

AGC 2 Irradiated Material Properties Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohrbaugh, David Thomas

2017-05-01

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. , Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core componentsmore » within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.« less

AGC 2 Irradiation Creep Strain Data Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Windes, William E.; Rohrbaugh, David T.; Swank, W. David

2016-08-01

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. Nuclear graphite H-451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within amore » commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.« less

Present limits and improvements of structural materials for fusion reactors - a review

NASA Astrophysics Data System (ADS)

Tavassoli, A.-A. F.

2002-04-01

Since the transition from ITER or DEMO to a commercial power reactor would involve a significant change in system and materials options, a parallel R&D path has been put in place in Europe to address these issues. This paper assesses the structural materials part of this program along with the latest R&D results from the main programs. It is shown that stainless steels and ferritic/martensitic steels, retained for ITER and DEMO, will also remain the principal contenders for the future FPR, despite uncertainties over irradiation induced embrittlement at low temperatures and consequences of high He/dpa ratio. Neither one of the present advanced high temperature materials has to this date the structural integrity reliability needed for application in critical components. This situation is unlikely to change with the materials R&D alone and has to be mitigated in close collaboration with blanket system design.

Role of nuclear grade graphite in controlling oxidation in modular HTGRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Windes, Willaim; Strydom, G.; Kane, J.

2014-11-01

The passively safe High Temperature Gas-cooled Reactor (HTGR) design is one of the primary concepts considered for Generation IV and Small Modular Reactor (SMR) programs. The helium cooled, nuclear grade graphite moderated core achieves extremely high operating temperatures allowing either industrial process heat or electricity generation at high efficiencies. In addition to their neutron moderating properties, nuclear grade graphite core components provide excellent high temperature stability, thermal conductivity, and chemical compatibility with the high temperature nuclear fuel form. Graphite has been continuously used in nuclear reactors since the 1940’s and has performed remarkably well over a wide range of coremore » environments and operating conditions. Graphite moderated, gas-cooled reactor designs have been safely used for research and power production purposes in multiple countries since the inception of nuclear energy development. However, graphite is a carbonaceous material, and this has generated a persistent concern that the graphite components could actually burn during either normal or accident conditions [ , ]. The common assumption is that graphite, since it is ostensibly similar to charcoal and coal, will burn in a similar manner. While charcoal and coal may have the appearance of graphite, the internal microstructure and impurities within these carbonaceous materials are very different. Volatile species and trapped moisture provide a source of oxygen within coal and charcoal allowing them to burn. The fabrication process used to produce nuclear grade graphite eliminates these oxidation enhancing impurities, creating a dense, highly ordered form of carbon possessing high thermal diffusivity and strongly (covalently) bonded atoms.« less

Expansion of high-temperature; high-pressure data set for coal gasification. Fifth quarterly report, September 28-December 28, 1985

DOE Office of Scientific and Technical Information (OSTI.GOV)

Solomon, P.R.; Serio, M.A.; Hamblen, D.G.

1985-01-01

During the fifth quarter, the gas mixing station for the high pressure reactor (HPR) system was completed. This station allows us to make reproducible binary mixtures of any two gases. It will be used for pyrolysis experiments in helium/nitrogen or oxygen/nitrogen and gasification experiments in helium/nitrogen or oxygen/nitrogen and gasification experiments in carbon dioxide/nitrogen. In addition, work began on modifications of the HPR system for high pressure (600 psig) operation. A limited amount of data was taken with the HPR system due to the modifications for the mixing station. However, the test plan experiments for pyrolysis in mixtures of heliummore » and nitrogen were completed. In general, there is a slightly higher yield of volatiles and lower yield of char as the helium content (heating rate) increases. A new technique for measuring char reactivity resulted from an Army SBIR program and was further developed under our other METC Contract. It has also been used to characterize chars generated under the current program. It was evident that the severity of the thermal treatment had a direct effect on char reactivity. In this regard, rapid heating to a relatively low temperature was most favorable while slow heating to a high temperature was least favorable. With regard to pressure effects on reactivity, our preliminary data indicated that higher pressures produce chars lower initial reactivity. A total of four experiments were done in the heated tube reactor (HTR) at 60 psig, 800/sup 0/C maximum tube temperature. The trends are the same as observed in the atmospheric pressure experiments for the same tube temperature and cold gas velocity. During the past quarter, a particle temperature (PT) model was under development for the high pressure entrained flow reactor (HPR). 5 refs., 5 figs.« less

AGR-1 Compact 1-3-1 Post-Irradiation Examination Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Demkowicz, Paul Andrew

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport (INL 2015). A seriesmore » of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously (Grover, Petti, and Maki 2010, Maki 2009).« less

AGR-1 Compact 5-3-1 Post-Irradiation Examination Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Demkowicz, Paul; Harp, Jason; Winston, Phil

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance, and fission product transport (INL 2015). A series ofmore » fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously.« less

Synergies Between ' and Cavity Formation in HT-9 Following High Dose Neutron Irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G.; Parish, Chad M.; Saleh, Tarik A.

Candidate cladding materials for advanced nuclear power reactors including fast reactor designs require materials capable of withstanding high dose neutron irradiation at elevated temperatures. One candidate material, HT-9, through various research programs have demonstrated the ability to withstand significant swelling and other radiation-induced degradation mechanisms in the high dose regime (>50 displacements per atom, dpa) at elevated temperatures (>300 C). Here, high efficiency multi-dimensional scanning transmission electron microscopy (STEM) acquisition with the aid of a three-dimensional (3D) reconstruction and modeling technique is used to probe the microstructural features that contribute to the exceptional swelling resistance of HT-9. In particular, themore » synergies between ' and fine-scale and moderate-scale cavity formation is investigated.« less

Material Control and Accounting Design Considerations for High-Temperature Gas Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trond Bjornard; John Hockert

The subject of this report is domestic safeguards and security by design (2SBD) for high-temperature gas reactors, focusing on material control and accountability (MC&A). The motivation for the report is to provide 2SBD support to the Next Generation Nuclear Plant (NGNP) project, which was launched by Congress in 2005. This introductory section will provide some background on the NGNP project and an overview of the 2SBD concept. The remaining chapters focus specifically on design aspects of the candidate high-temperature gas reactors (HTGRs) relevant to MC&A, Nuclear Regulatory Commission (NRC) requirements, and proposed MC&A approaches for the two major HTGR reactormore » types: pebble bed and prismatic. Of the prismatic type, two candidates are under consideration: (1) GA's GT-MHR (Gas Turbine-Modular Helium Reactor), and (2) the Modular High-Temperature Reactor (M-HTR), a derivative of Areva's Antares reactor. The future of the pebble-bed modular reactor (PBMR) for NGNP is uncertain, as the PBMR consortium partners (Westinghouse, PBMR [Pty] and The Shaw Group) were unable to agree on the path forward for NGNP during 2010. However, during the technology assessment of the conceptual design phase (Phase 1) of the NGNP project, AREVA provided design information and technology assessment of their pebble bed fueled plant design called the HTR-Module concept. AREVA does not intend to pursue this design for NGNP, preferring instead a modular reactor based on the prismatic Antares concept. Since MC&A relevant design information is available for both pebble concepts, the pebble-bed HTGRs considered in this report are: (1) Westinghouse PBMR; and (2) AREVA HTR-Module. The DOE Office of Nuclear Energy (DOE-NE) sponsors the Fuel Cycle Research and Development program (FCR&D), which contains an element specifically focused on the domestic (or state) aspects of SBD. This Material Protection, Control and Accountancy Technology (MPACT) program supports the present work summarized in this report, namely the development of guidance to support the consideration of MC&A in the design of both pebble-bed and prismatic-fueled HTGRs. The objective is to identify and incorporate design features into the facility design that will cost effectively aid in making MC&A more effective and efficient, with minimum impact on operations. The theft of nuclear material is addressed through both MC&A and physical protection, while the threat of sabotage is addressed principally through physical protection.« less

Clean catalytic combustor program

NASA Technical Reports Server (NTRS)

Ekstedt, E. E.; Lyon, T. F.; Sabla, P. E.; Dodds, W. J.

1983-01-01

A combustor program was conducted to evolve and to identify the technology needed for, and to establish the credibility of, using combustors with catalytic reactors in modern high-pressure-ratio aircraft turbine engines. Two selected catalytic combustor concepts were designed, fabricated, and evaluated. The combustors were sized for use in the NASA/General Electric Energy Efficient Engine (E3). One of the combustor designs was a basic parallel-staged double-annular combustor. The second design was also a parallel-staged combustor but employed reverse flow cannular catalytic reactors. Subcomponent tests of fuel injection systems and of catalytic reactors for use in the combustion system were also conducted. Very low-level pollutant emissions and excellent combustor performance were achieved. However, it was obvious from these tests that extensive development of fuel/air preparation systems and considerable advancement in the steady-state operating temperature capability of catalytic reactor materials will be required prior to the consideration of catalytic combustion systems for use in high-pressure-ratio aircraft turbine engines.

NGNP Data Management and Analysis System Analysis and Web Delivery Capabilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cynthia D. Gentillon

2010-09-01

Projects for the Very High Temperature Reactor Technology Development Office provide data in support of Nuclear Regulatory Commission licensing of the very high temperature reactor. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high-temperature and high-fluence environments. In addition, thermal-hydraulic experiments are conducted to validate codes used to assess reactor safety. The Very High Temperature Reactor Technology Development Office has established the NGNP Data Management and Analysis System (NDMAS) at the Idaho National Laboratory to ensure that very high temperature reactor data are (1) qualified for use, (2) stored in amore » readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the third NDMAS objective. It describes capabilities for displaying the data in meaningful ways and for data analysis to identify useful relationships among the measured quantities.« less

Status of Fuel Development and Manufacturing for Space Nuclear Reactors at BWX Technologies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carmack, W.J.; Husser, D.L.; Mohr, T.C.

2004-02-04

New advanced nuclear space propulsion systems will soon seek a high temperature, stable fuel form. BWX Technologies Inc (BWXT) has a long history of fuel manufacturing. UO2, UCO, and UCx have been fabricated at BWXT for various US and international programs. Recent efforts at BWXT have focused on establishing the manufacturing techniques and analysis capabilities needed to provide a high quality, high power, compact nuclear reactor for use in space nuclear powered missions. To support the production of a space nuclear reactor, uranium nitride has recently been manufactured by BWXT. In addition, analytical chemistry and analysis techniques have been developedmore » to provide verification and qualification of the uranium nitride production process. The fabrication of a space nuclear reactor will require the ability to place an unclad fuel form into a clad structure for assembly into a reactor core configuration. To this end, BWX Technologies has reestablished its capability for machining, GTA welding, and EB welding of refractory metals. Specifically, BWX Technologies has demonstrated GTA welding of niobium flat plate and EB welding of niobium and Nb-1Zr tubing. In performing these demonstration activities, BWX Technologies has established the necessary infrastructure to manufacture UO2, UCx, or UNx fuel, components, and complete reactor assemblies in support of space nuclear programs.« less

Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

NASA Astrophysics Data System (ADS)

Fic, Adam; Składzień, Jan; Gabriel, Michał

2015-03-01

Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle), which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle). The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

Exploratory study of several advanced nuclear-MHD power plant systems.

NASA Technical Reports Server (NTRS)

Williams, J. R.; Clement, J. D.; Rosa, R. J.; Yang, Y. Y.

1973-01-01

In order for efficient multimegawatt closed cycle nuclear-MHD systems to become practical, long-life gas cooled reactors with exit temperatures of about 2500 K or higher must be developed. Four types of nuclear reactors which have the potential of achieving this goal are the NERVA-type solid core reactor, the colloid core (rotating fluidized bed) reactor, the 'light bulb' gas core reactor, and the 'coaxial flow' gas core reactor. Research programs aimed at developing these reactors have progressed rapidly in recent years so that prototype power reactors could be operating by 1980. Three types of power plant systems which use these reactors have been analyzed to determine the operating characteristics, critical parameters and performance of these power plants. Overall thermal efficiencies as high as 80% are projected, using an MHD turbine-compressor cycle with steam bottoming, and slightly lower efficiencies are projected for an MHD motor-compressor cycle.

Design of a Low Power, Fast-Spectrum, Liquid-Metal Cooled Surface Reactor System

NASA Astrophysics Data System (ADS)

Marcille, T. F.; Dixon, D. D.; Fischer, G. A.; Doherty, S. P.; Poston, D. I.; Kapernick, R. J.

2006-01-01

In the current 2005 US budget environment, competition for fiscal resources make funding for comprehensive space reactor development programs difficult to justify and accommodate. Simultaneously, the need to develop these systems to provide planetary and deep space-enabling power systems is increasing. Given that environment, designs intended to satisfy reasonable near-term surface missions, using affordable technology-ready materials and processes warrant serious consideration. An initial lunar application design incorporating a stainless structure, 880 K pumped NaK coolant system and a stainless/UO2 fuel system can be designed, fabricated and tested for a fraction of the cost of recent high-profile reactor programs (JIMO, SP-100). Along with the cost reductions associated with the use of qualified materials and processes, this design offers a low-risk, high-reliability implementation associated with mission specific low temperature, low burnup, five year operating lifetime requirements.

Historical perspectives - The role of the NASA Lewis Research Center in the national space nuclear power programs

NASA Technical Reports Server (NTRS)

Bloomfield, H. S.; Sovie, R. J.

1991-01-01

The history of the NASA Lewis Research Center's role in space nuclear power programs is reviewed. Lewis has provided leadership in research, development, and the advancement of space power and propulsion systems. Lewis' pioneering efforts in nuclear reactor technology, shielding, high temperature materials, fluid dynamics, heat transfer, mechanical and direct energy conversion, high-energy propellants, electric propulsion and high performance rocket fuels and nozzles have led to significant technical and management roles in many natural space nuclear power and propulsion programs.

Historical perspectives: The role of the NASA Lewis Research Center in the national space nuclear power programs

NASA Technical Reports Server (NTRS)

Bloomfield, H. S.; Sovie, R. J.

1991-01-01

The history of the NASA Lewis Research Center's role in space nuclear power programs is reviewed. Lewis has provided leadership in research, development, and the advancement of space power and propulsion systems. Lewis' pioneering efforts in nuclear reactor technology, shielding, high temperature materials, fluid dynamics, heat transfer, mechanical and direct energy conversion, high-energy propellants, electric propulsion and high performance rocket fuels and nozzles have led to significant technical and management roles in many national space nuclear power and propulsion programs.

Status of FeCrAl ODS Irradiations in the High Flux Isotope Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G.; Howard, Richard H.

2016-08-19

FeCrAl oxide-dispersion strengthened (ODS) alloys are an attractive sub-set alloy class of the more global FeCrAl material class for nuclear applications due to their high-temperature steam oxidation resistance and hypothesized enhanced radiation tolerance. A need currently exists to determine the radiation tolerance of these newly developed alloys. To address this need, a preliminary study was conducted using the High Flux Isotope Reactor (HFIR) to irradiate an early generation FeCrAl ODS alloy, 125YF. Preliminary post-irradiation examination (PIE) on these irradiated specimens have shown good radiation tolerance at elevated temperatures (≥330°C) but possible radiation-induced hardening and embrittlement at irradiations of 200°C tomore » a damage level of 1.9 displacement per atom (dpa). Building on this experience, a new series of irradiations are currently being conceptualized. This irradiation series called the FCAD irradiation program will irradiate the latest generation FeCrAl ODS and FeCr ODS alloys to significantly higher doses. These experiments will provide the necessary information to determine the mechanical performance of irradiated FeCrAl ODS alloys at light water reactor and fast reactor conditions.« less

Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tan, Lizhen; Yang, Ying; Sridharan, K.

2015-12-01

The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computationalmore » tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe 2+) irradiation.« less

Cermet coating tribological behavior in high temperature helium

DOE Office of Scientific and Technical Information (OSTI.GOV)

CACHON, Lionel; ALBALADEJO, Serge; TARAUD, Pascal

As the CEA is highly involved in the Generation IV Forum, a comprehensive research and development program has been conducted for several years, in order to establish the feasibility of Gas Cooled Reactor (GCR) technology projects using helium as a cooling fluid. Within this framework, a tribology program was launched in order to select and qualify coatings and materials, and to provide recommendations for the sliding components operating in GCRs. The purpose of this paper is to describe the CEA Helium tribology study on several GCR components (thermal barriers, control rod drive mechanisms, reactor internals, ..) requiring protection against wearmore » and bonding. Tests in helium atmosphere are necessary to be fully representative of tribological environments and to assess the material or coating candidates which can provide a reliable answer to these situations. This paper focuses on the tribology tests performed on CERMET (Cr{sub 3}C-2- NiCr) coatings within a temperature range of between 800 and 1000 deg C.« less

NEET Micro-Pocket Fission Detector. Final Project report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Unruh, T.; Rempe, Joy; McGregor, Douglas

2014-09-01

A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Alternative Energies and Atomic Energy Commission, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), is funded by the Nuclear Energy Enabling Technologies (NEET) program to develop and test Micro-Pocket Fission Detectors (MPFDs), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package. When deployed, these sensors will significantly advance flux detection capabilities for irradiation tests in US Material Test Reactors (MTRs). Ultimately, evaluations may lead to a more compact, more accurate, andmore » longer lifetime flux sensor for critical mock-ups, and high performance reactors, allowing several Department of Energy Office of Nuclear Energy (DOE-NE) programs to obtain higher accuracy/higher resolution data from irradiation tests of candidate new fuels and materials. Specifically, deployment of MPFDs will address several challenges faced in irradiations performed at MTRs: Current fission chamber technologies do not offer the ability to measure fast flux, thermal flux and temperature within a single compact probe; MPFDs offer this option. MPFD construction is very different than current fission chamber construction; the use of high temperature materials allow MPFDs to be specifically tailored to survive harsh conditions encountered in-core of high performance MTRs. The higher accuracy, high fidelity data available from the compact MPFD will significantly enhance efforts to validate new high-fidelity reactor physics codes and new multi-scale, multi-physics codes. MPFDs can be built with variable sensitivities to survive the lifetime of an experiment or fuel assembly in some MTRs, allowing for more efficient and cost effective power monitoring. The small size of the MPFDs allows multiple sensors to be deployed, offering the potential to accurately measure the flux and temperature profiles in the reactor. This report summarizes the status at the end of year two of this three year project. As documented in this report, all planned accomplishments for developing this unique new, compact, multipurpose sensor have been completed.« less

In situ TEM and synchrotron characterization of U–10Mo thin specimen annealed at the fast reactor temperature regime

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yun, Di, E-mail: diyun1979@xjtu.edu.cn; Xi'an Jiao Tong University, 28 Xian Ning West Road, Xi'an 710049; Mo, Kun

2015-12-15

U–Mo metallic alloys have been extensively used for the Reduced Enrichment for Research and Test Reactors (RERTR) program, which is now known as the Office of Material Management and Minimization under the Conversion Program. This fuel form has also recently been proposed as fast reactor metallic fuels in the recent DOE Ultra-high Burnup Fast Reactor project. In order to better understand the behavior of U–10Mo fuels within the fast reactor temperature regime, a series of annealing and characterization experiments have been performed. Annealing experiments were performed in situ at the Intermediate Voltage Electron Microscope (IVEM-Tandem) facility at Argonne National Laboratorymore » (ANL). An electro-polished U–10Mo alloy fuel specimen was annealed in situ up to 700 °C. At an elevated temperature of about 540 °C, the U–10Mo specimen underwent a relatively slow microstructure transition. Nano-sized grains were observed to emerge near the surface. At the end temperature of 700 °C, the near-surface microstructure had evolved to a nano-crystalline state. In order to clarify the nature of the observed microstructure, Laue diffraction and powder diffraction experiments were carried out at beam line 34-ID of the Advanced Photon Source (APS) at ANL. Phases present in the as-annealed specimen were identified with both Laue diffraction and powder diffraction techniques. The U–10Mo was found to recrystallize due to thermally-induced recrystallization driven by a high density of pre-existing dislocations. A separate in situ annealing experiment was carried out with a Focused Ion Beam processed (FIB) specimen. A similar microstructure transition occurred at a lower temperature of about 460 °C with a much faster transition rate compared to the electro-polished specimen. - Highlights: • TEM annealing experiments were performed in situ at the IVEM facility up to fast reactor temperature. • At 540 °C, the U-10Mo specimen underwent a slow microstructure transition where nano-sized grains were observed to emerge. • UO{sub 2} phase exists at the thin area of the as-annealed specimen whereas U-10Mo γ phase dominated at the thicker part. • Bcc γ U-10Mo recrystallized to become nano-meter sized crystallites near the specimen surface. • A separateannealing experiment was conducted with a FIB processed specimen where similar transition occurred at a lower temperature of 460 °C with a faster rate.« less

USSR Report, Energy, No. 147.

DTIC Science & Technology

1983-05-18

based on low-temperature reactors ; atomic heat and electric power stations (ATETs); The restructuring of the energy balance for the 1980-2000 period...ASPT) based on low-temperature reactors ; atomic heat and electric power stations (TETs); industrial atomic power stations (AETS) based on high-temper...ature reactors ) and high-efficiency long-distance heat transport (in conjunc- tion with high-temperature nuclear power sources: ASDT). The

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.; Sterbentz, James W.; Snoj, Luka

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

Nonlinear Ultrasonic Measurements in Nuclear Reactor Environments

NASA Astrophysics Data System (ADS)

Reinhardt, Brian T.

Several Department of Energy Office of Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development (FCRD), Advanced Reactor Concepts (ARC), Light Water Reactor Sustainability, and Next Generation Nuclear Power Plants (NGNP), are investigating new fuels, materials, and inspection paradigms for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials during irradiation. In DOE-NE's FCRD program, ultrasonic based technology was identified as a key approach that should be pursued to obtain the high-fidelity, high-accuracy data required to characterize the behavior and performance of new candidate fuels and structural materials during irradiation testing. The radiation, high temperatures, and pressure can limit the available tools and characterization methods. In this thesis, two ultrasonic characterization techniques will be explored. The first, finite amplitude wave propagation has been demonstrated to be sensitive to microstructural material property changes. It is a strong candidate to determine fuel evolution; however, it has not been demonstrated for in-situ reactor applications. In this thesis, finite amplitude wave propagation will be used to measure the microstructural evolution in Al-6061. This is the first demonstration of finite amplitude wave propagation at temperatures in excess of 200 °C and during an irradiation test. Second, a method based on contact nonlinear acoustic theory will be developed to identify compressed cracks. Compressed cracks are typically transparent to ultrasonic wave propagation; however, by measuring harmonic content developed during finite amplitude wave propagation, it is shown that even compressed cracks can be characterized. Lastly, piezoelectric transducers capable of making these measurements are developed. Specifically, three piezoelectric sensors (Bismuth Titanate, Aluminum Nitride, and Zinc Oxide) are tested in the Massachusetts Institute of Technology Research reactor to a fast neutron fluence of 8.65x10 20 n/cm2. It is demonstrated that Bismuth Titanate is capable of transduction up to 5 x1020 n/cm2, Zinc Oxide is capable of transduction up to 6.27 x1020 n/cm 2, and Aluminum Nitride is capable of transduction up to 8.65x x10 20 n/cm2.

NEUTRONIC REACTOR CORE

DOEpatents

Thomson, W.B.; Corbin, A. Jr.

1961-07-18

An improved core for a gas-cooled power reactor which admits gas coolant at high temperatures while affording strong integral supporting structure and efficient moderation of neutrons is described. The multiplicities of fuel elements constituting the critical amassment of fissionable material are supported and confined by a matrix of metallic structure which is interspersed therebetween. Thermal insulation is interposed between substantially all of the metallic matrix and the fuel elements; the insulation then defines the principal conduit system for conducting the coolant gas in heat-transfer relationship with the fuel elements. The metallic matrix itseif comprises a system of ducts through which an externally-cooled hydrogeneous liquid, such as water, is circulated to serve as the principal neutron moderant for the core and conjointly as the principal coolant for the insulated metallic structure. In this way, use of substantially neutron transparent metals, such as aluminum, becomes possible for the supporting structure, despite the high temperatures of the proximate gas. The Aircraft Nuclear Propulsion program's "R-1" reactor design is a preferred embodiment.

HTR-PROTEUS pebble bed experimental program cores 9 & 10: columnar hexagonal point-on-point packing with a 1:1 moderator-to-fuel pebble ratio

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.

2014-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 5, 6, 7, & 8: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:2 MODERATOR-TO-FUEL PEBBLE RATIO

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess

2013-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 9 & 10: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess

2013-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

Results from the DOE Advanced Gas Reactor Fuel Development and Qualification Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

David Petti

2014-06-01

Modular HTGR designs were developed to provide natural safety, which prevents core damage under all design basis accidents and presently envisioned severe accidents. The principle that guides their design concepts is to passively maintain core temperatures below fission product release thresholds under all accident scenarios. This level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude and allows potential elimination of the need for evacuation and sheltering beyond a small exclusion area. This level, however, is predicated on exceptionally high fuel fabrication quality and performance under normal operation and accident conditions. Germanymore » produced and demonstrated high quality fuel for their pebble bed HTGRs in the 1980s, but no U.S. manufactured fuel had exhibited equivalent performance prior to the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The design goal of the modular HTGRs is to allow elimination of an exclusion zone and an emergency planning zone outside the plant boundary fence, typically interpreted as being about 400 meters from the reactor. To achieve this, the reactor design concepts require a level of fuel integrity that is better than that claimed for all prior US manufactured TRISO fuel, by a few orders of magnitude. The improved performance level is about a factor of three better than qualified for German TRISO fuel in the 1980’s. At the start of the AGR program, without a reactor design concept selected, the AGR fuel program selected to qualify fuel to an operating envelope that would bound both pebble bed and prismatic options. This resulted in needing a fuel form that could survive at peak fuel temperatures of 1250°C on a time-averaged basis and high burnups in the range of 150 to 200 GWd/MTHM (metric tons of heavy metal) or 16.4 to 21.8% fissions per initial metal atom (FIMA). Although Germany has demonstrated excellent performance of TRISO-coated UO2 particle fuel up to about 10% FIMA and 1150°C, UO2 fuel is known to have limitations because of CO formation and kernel migration at the high burnups, power densities, temperatures, and temperature gradients that may be encountered in the prismatic modular HTGRs. With uranium oxycarbide (UCO) fuel, the kernel composition is engineered to prevent CO formation and kernel migration, which are key threats to fuel integrity at higher burnups, temperatures, and temperature gradients. Furthermore, the recent poor fuel performance of UO2 TRISO fuel pebbles measured in Chinese irradiation testing in Russia and in German pebbles irradiated at 1250°C, and historic data on poorer fuel performance in safety testing of German pebbles that experienced burnups in excess of 10% FIMA [1] have each raised concern about the use of UO2 TRISO above 10% FIMA and 1150°C and the degree of margin available in the fuel system. This continues to be an active area of study internationally.« less

Thermal-Hydraulic Design of a Fluoride High-Temperature Demonstration Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carbajo, Juan J; Qualls, A L

2016-01-01

INTRODUCTION The Fluoride High-Temperature Reactor (FHR) named the Demonstration Reactor (DR) is a novel reactor concept using molten salt coolant and TRIstructural ISOtropic (TRISO) fuel that is being developed at Oak Ridge National Laboratory (ORNL). The objective of the FHR DR is to advance the technology readiness level of FHRs. The FHR DR will demonstrate technologies needed to close remaining gaps to commercial viability. The FHR DR has a thermal power of 100 MWt, very similar to the SmAHTR, another FHR ORNL concept (Refs. 1 and 2) with a power of 125 MWt. The FHR DR is also a smallmore » version of the Advanced High Temperature Reactor (AHTR), with a power of 3400 MWt, cooled by a molten salt and also being developed at ORNL (Ref. 3). The FHR DR combines three existing technologies: (1) high-temperature, low-pressure molten salt coolant, (2) high-temperature coated-particle TRISO fuel, (3) and passive decay heat cooling systems by using Direct Reactor Auxiliary Cooling Systems (DRACS). This paper presents FHR DR thermal-hydraulic design calculations.« less

3D thermal modeling of TRISO fuel coupled with neutronic simulation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Jianwei; Uddin, Rizwan

2010-01-01

The Very High Temperature Gas Reactor (VHTR) is widely considered as one of the top candidates identified in the Next Generation Nuclear Power-plant (NGNP) Technology Roadmap under the U.S . Depanment of Energy's Generation IV program. TRlSO particle is a common element among different VHTR designs and its performance is critical to the safety and reliability of the whole reactor. A TRISO particle experiences complex thermo-mechanical changes during reactor operation in high temperature and high burnup conditions. TRISO fuel performance analysis requires evaluation of these changes on micro scale. Since most of these changes are temperature dependent, 3D thermal modelingmore » of TRISO fuel is a crucial step of the whole analysis package. In this paper, a 3D numerical thermal model was developed to calculate temperature distribution inside TRISO and pebble under different scenarios. 3D simulation is required because pebbles or TRISOs are always subjected to asymmetric thermal conditions since they are randomly packed together. The numerical model was developed using finite difference method and it was benchmarked against ID analytical results and also results reported from literature. Monte-Carlo models were set up to calculate radial power density profile. Complex convective boundary condition was applied on the pebble outer surface. Three reactors were simulated using this model to calculate temperature distribution under different power levels. Two asymmetric boundary conditions were applied to the pebble to test the 3D capabilities. A gas bubble was hypothesized inside the TRISO kernel and 3D simulation was also carried out under this scenario. Intuition-coherent results were obtained and reported in this paper.« less

High temperature durable catalyst development

NASA Technical Reports Server (NTRS)

Snow, G. C.; Tong, H.

1981-01-01

A program has been carried out to develop a catalytic reactor capable of operation in environments representative of those anticipated for advanced automotive gas turbine engines. A reactor consisting of a graded cell honeycomb support with a combination of noble metal and metal oxide catalyst coatings was built and successfully operated for 1000 hr. At an air preheat temperature of 740 K and a propane/air ratio of 0.028 by mass, the adiabatic flame temperature was held at about 1700 K. The graded cell monolithic reaction measured 5 cm in diameter by 10.2 cm in length and was operated at a reference velocity of 14.0 m/s at 1 atm. Measured NOx levels remained below 5 ppm, while unburned hydrocarbon concentrations registered near zero and carbon monoxide levels were nominally below 20 ppm.

AGC-2 Graphite Pre-irradiation Data Package

DOE Office of Scientific and Technical Information (OSTI.GOV)

David Swank; Joseph Lord; David Rohrbaugh

2010-08-01

The NGNP Graphite R&D program is currently establishing the safe operating envelope of graphite core components for a Very High Temperature Reactor (VHTR) design. The program is generating quantitative data necessary for predicting the behavior and operating performance of the new nuclear graphite grades. To determine the in-service behavior of the graphite for pebble bed and prismatic designs, the Advanced Graphite Creep (AGC) experiment is underway. This experiment is examining the properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences and compressive loads. Each experiment consists of over 400 graphite specimens that are characterizedmore » prior to irradiation and following irradiation. Six experiments are planned with the first, AGC-1, currently being irradiated in the Advanced Test Reactor (ATR) and pre-irradiation characterization of the second, AGC-2, completed. This data package establishes the readiness of 512 specimens for assembly into the AGC-2 capsule.« less

Tory II-A: a nuclear ramjet test reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hadley, J.W.

Declassified 28 Nov 1973. The first test reactor in the Pluto program, leading to development of a nuclear ramjet engine, is called Tory II-A. While it is not an actual prototype engine, this reactor embodies a core design which is considered feasible for an engine, and operation of the reactor will provide a test of that core type as well as more generalized values in reactor design and testing. The design of Tory II-A and construction of the reactor and of its test facility are described. Operation of the Tory II-A core at a total power of 160 megawatts, withmore » 800 pounds of air per second passing through the core and emerging at a temperature of 2000 deg F, is the central objective of the test program. All other reactor and facility components exist to support operation of the core, and preliminary steps in the test program itself will be directed primarily toward ensuring attalnment of full-power operation and collection of meaningful data on core behavior during that operation. The core, 3 feet in diameter and 41/2 feet long, will be composed of bundled ceramic tubes whose central holes will provide continuous air passages from end to end of the reactor. These tubes are to be composed of a homogeneous mixture of UO/sub 2/ fuel and BeO moderator, compacted and sintered to achieve high strength and density. (30 references) (auth)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salko, Robert K; Sung, Yixing; Kucukboyaci, Vefa

The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time stepmore » of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.« less

Acquisition of Raman Spectrometer and High Temperature and Pressure Reactor for Synthesis and Characterization of Carbon Based Hybrid Nanoparticles from Waste Wood

DTIC Science & Technology

2015-04-27

from waste biomass using these two high temperature reactors. We have extensively used a Raman spectrometer to analyse as synthesized carbon materials...corporation). These tools were fully installed and operational. We have also synthesized carbon materials from waste biomass using these two high...materials from waste biomass using these two high temperature reactors. We have extensively used a Raman spectrometer to analyse as synthesized carbon

A summary of the results from the DOE advanced gas reactor (AGR) fuel development and qualification program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Petti, David Andrew

2017-04-01

Modular high temperature gas-cooled reactor (HTGR) designs were developed to provide natural safety, which prevents core damage under all licensing basis events. The principle that guides their design concepts is to passively maintain core temperatures below fission product release thresholds under all accident scenarios. The required level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude relative to source terms for other reactor types and allows a graded approach to emergency planning and the potential elimination of the need for evacuation and sheltering beyond a small exclusion area. Achieving this level, however,more » is predicated on exceptionally high coated-particle fuel fabrication quality and excellent performance under normal operation and accident conditions. The design goal of modular HTGRs is to meet the Environmental Protection Agency (EPA) Protective Action Guides (PAGs) for offsite dose at the Exclusion Area Boundary (EAB). To achieve this, the reactor design concepts require a level of fuel integrity that is far better than that achieved for all prior U.S.-manufactured tristructural isotropic (TRISO) coated particle fuel.« less

Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

PubMed

Hill, R N; Nutt, W M; Laidler, J J

2011-01-01

The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described. Copyright © 2010 Health Physics Society

Studies of the use of high-temperature nuclear heat from an HTGR for hydrogen production

NASA Technical Reports Server (NTRS)

Peterman, D. D.; Fontaine, R. W.; Quade, R. N.; Halvers, L. J.; Jahromi, A. M.

1975-01-01

The results of a study which surveyed various methods of hydrogen production using nuclear and fossil energy are presented. A description of these methods is provided, and efficiencies are calculated for each case. The process designs of systems that utilize the heat from a general atomic high temperature gas cooled reactor with a steam methane reformer and feed the reformer with substitute natural gas manufactured from coal, using reforming temperatures, are presented. The capital costs for these systems and the resultant hydrogen production price for these cases are discussed along with a research and development program.

PIE on Safety-Tested Loose Particles from Irradiated Compact 4-4-2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hunn, John D.; Gerczak, Tyler J.; Morris, Robert Noel

2016-04-01

Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High Temperature Gas-cooled Reactors (HTGRs). This work is sponsored by the Department of Energy Office of Nuclear Energy (DOE-NE) through the Advanced Reactor Technologies (ART) Office under the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program. The AGR-1 experiment was the first in a series of TRISO fuel irradiation tests initiated in 2006. The AGR-1 TRISO particles and fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 using laboratory-scale equipment and irradiated for 3 years in themore » Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. Post-irradiation examination was performed at INL and ORNL to study how the fuel behaved during irradiation, and to test fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing and post-safety testing PIE conducted at ORNL on loose particles extracted from irradiated AGR-1 Compact 4-4-2.« less

The Shock and Vibration Digest. Volume 15, Number 3

DTIC Science & Technology

1983-03-01

High Temperature Gas-Cooled Reactor Core with Block-type Fuel (2nd Report: An Analytical Method of Two-dmentmnal Vibration of Interacting CohunM) T...Computer-aided techniquei, Detign techniquei A wite of computer programs hat been developed which allow« advanced fatigue analyiit procedures to be...valuei with those developed by bearing analysis computer programs were used to formulate an understanding of the mechanisms that induce ball skidding

Evaluation of catalytic combustion of actual coal-derived gas

NASA Technical Reports Server (NTRS)

Blanton, J. C.; Shisler, R. A.

1982-01-01

The combustion characteristics of a Pt-Pl catalytic reactor burning coal-derived, low-Btu gas were investigated. A large matrix of test conditions was explored involving variations in fuel/air inlet temperature and velocity, reactor pressure, and combustor exit temperature. Other data recorded included fuel gas composition, reactor temperatures, and exhaust emissions. Operating experience with the reactor was satisfactory. Combustion efficiencies were quite high (over 95 percent) over most of the operating range. Emissions of NOx were quite high (up to 500 ppm V and greater), owing to the high ammonia content of the fuel gas.

Fuel development for gas-cooled fast reactors

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Fielding, R.; Gan, J.

2007-09-01

The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

Nondestructive evaluation of nuclear-grade graphite

NASA Astrophysics Data System (ADS)

Kunerth, D. C.; McJunkin, T. R.

2012-05-01

The material of choice for the core of the high-temperature gas-cooled reactors being developed by the U.S. Department of Energy's Next Generation Nuclear Plant Program is graphite. Graphite is a composite material whose properties are highly dependent on the base material and manufacturing methods. In addition to the material variations intrinsic to the manufacturing process, graphite will also undergo changes in material properties resulting from radiation damage and possible oxidation within the reactor. Idaho National Laboratory is presently evaluating the viability of conventional nondestructive evaluation techniques to characterize the material variations inherent to manufacturing and in-service degradation. Approaches of interest include x-ray radiography, eddy currents, and ultrasonics.

Demonstration of catalytic combustion with residual fuel

NASA Technical Reports Server (NTRS)

Dodds, W. J.; Ekstedt, E. E.

1981-01-01

An experimental program was conducted to demonstrate catalytic combustion of a residual fuel oil. Three catalytic reactors, including a baseline configuration and two backup configurations based on baseline test results, were operated on No. 6 fuel oil. All reactors were multielement configurations consisting of ceramic honeycomb catalyzed with palladium on stabilized alumina. Stable operation on residual oil was demonstrated with the baseline configuration at a reactor inlet temperature of about 825 K (1025 F). At low inlet temperature, operation was precluded by apparent plugging of the catalytic reactor with residual oil. Reduced plugging tendency was demonstrated in the backup reactors by increasing the size of the catalyst channels at the reactor inlet, but plugging still occurred at inlet temperature below 725 K (845 F). Operation at the original design inlet temperature of 589 K (600 F) could not be demonstrated. Combustion efficiency above 99.5% was obtained with less than 5% reactor pressure drop. Thermally formed NO sub x levels were very low (less than 0.5 g NO2/kg fuel) but nearly 100% conversion of fuel-bound nitrogen to NO sub x was observed.

Silicon carbide, an emerging high temperature semiconductor

NASA Technical Reports Server (NTRS)

Matus, Lawrence G.; Powell, J. Anthony

1991-01-01

In recent years, the aerospace propulsion and space power communities have expressed a growing need for electronic devices that are capable of sustained high temperature operation. Applications for high temperature electronic devices include development instrumentation within engines, engine control, and condition monitoring systems, and power conditioning and control systems for space platforms and satellites. Other earth-based applications include deep-well drilling instrumentation, nuclear reactor instrumentation and control, and automotive sensors. To meet the needs of these applications, the High Temperature Electronics Program at the Lewis Research Center is developing silicon carbide (SiC) as a high temperature semiconductor material. Research is focussed on developing the crystal growth, characterization, and device fabrication technologies necessary to produce a family of silicon carbide electronic devices and integrated sensors. The progress made in developing silicon carbide is presented, and the challenges that lie ahead are discussed.

Computer modeling of a hot filament diamond deposition reactor

NASA Technical Reports Server (NTRS)

Kuczmarski, Maria A.; Washlock, Paul A.; Angus, John C.

1991-01-01

A commercial fluid mechanics program, FLUENT, has been applied to the modeling of a hot-filament diamond deposition reactor. Streamlines and contours of constant temperature and species concentrations are obtained for practical reactor geometries and conditions. The modeling is presently restricted to two-dimensional simulations and to a chemical mechanism of ten independent homogeneous and surface reactions. Comparisons are made between predicted power consumption, substrate temperature, and concentrations of atomic hydrogen and methyl-radical with values taken from the literature. The results to date indicate that the modeling can aid in the rational design and analysis of practical reactor configurations.

UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Novitrian,; Pramuditya, Syeilendra

High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. Themore » result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.« less

Goals of thermionic program for space power

NASA Technical Reports Server (NTRS)

English, R. E.

1981-01-01

The thermionic and Brayton reactor concepts were compared for application to space power. For a turbine inlet temperature of 15000 K the Brayton powerplant weighted 5 to 40% less than the thermionic concept. The out of core concept separates the thermionic converters from their reactor. Technical risks are diminished by: (1) moving the insolator out of the reactor; (2) allowing a higher thermal flux for the thermionic converters than is required of the reactor fuel; and (3) eliminating fuel swelling's threat against lifetime of the thermionic converters. Overall performance can be improved by including power processing in system optimization for design and technology on more efficient, higher temperature power processors. The thermionic reactors will be larger than those for competitive systems with higher conversion efficiency and lower reactor operating temperatures. It is concluded that although the effect of reactor size on shield weight will be modest for unmanned spacecraft, the penalty in shield weight will be large for manned or man-tended spacecraft.

Coupling a Supercritical Carbon Dioxide Brayton Cycle to a Helium-Cooled Reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Middleton, Bobby; Pasch, James Jay; Kruizenga, Alan Michael

2016-01-01

This report outlines the thermodynamics of a supercritical carbon dioxide (sCO 2) recompression closed Brayton cycle (RCBC) coupled to a Helium-cooled nuclear reactor. The baseline reactor design for the study is the AREVA High Temperature Gas-Cooled Reactor (HTGR). Using the AREVA HTGR nominal operating parameters, an initial thermodynamic study was performed using Sandia's deterministic RCBC analysis program. Utilizing the output of the RCBC thermodynamic analysis, preliminary values of reactor power and of Helium flow rate through the reactor were calculated in Sandia's HelCO 2 code. Some research regarding materials requirements was then conducted to determine aspects of corrosion related tomore » both Helium and to sCO 2 , as well as some mechanical considerations for pressures and temperatures that will be seen by the piping and other components. This analysis resulted in a list of materials-related research items that need to be conducted in the future. A short assessment of dry heat rejection advantages of sCO 2> Brayton cycles was also included. This assessment lists some items that should be investigated in the future to better understand how sCO 2 Brayton cycles and nuclear can maximally contribute to optimizing the water efficiency of carbon free power generation« less

PIE on Safety-Tested AGR-1 Compact 5-1-1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hunn, John D.; Morris, Robert Noel; Baldwin, Charles A.

Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High-Temperature Gas-cooled Reactors (HTGRs). AGR-1 was the first in a series of TRISO fuel irradiation experiments initiated in 2006 under the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program; this work continues to be funded by the Department of Energy's Office of Nuclear Energy as part of the Advanced Reactor Technologies (ART) initiative. AGR-1 fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 and irradiated for three years in the Idaho National Laboratory (INL) Advanced Test Reactormore » (ATR) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. PIE is being performed at INL and ORNL to study how the fuel behaved during irradiation, and to examine fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing of irradiated AGR-1 Compact 5-1-1 in the ORNL Core Conduction Cooldown Test Facility (CCCTF) and post-safety testing PIE.« less

Station Blackout Analysis of HTGR-Type Experimental Power Reactor

NASA Astrophysics Data System (ADS)

Syarip; Zuhdi, Aliq; Falah, Sabilul

2018-01-01

The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

Progress towards developing neutron tolerant magnetostrictive and piezoelectric transducers

NASA Astrophysics Data System (ADS)

Reinhardt, Brian; Tittmann, Bernhard; Rempe, Joy; Daw, Joshua; Kohse, Gordon; Carpenter, David; Ames, Michael; Ostrovsky, Yakov; Ramuhalli, Pradeep; Montgomery, Robert; Chien, Hualte; Wernsman, Bernard

2015-03-01

Current generation light water reactors (LWRs), sodium cooled fast reactors (SFRs), small modular reactors (SMRs), and next generation nuclear plants (NGNPs) produce harsh environments in and near the reactor core that can severely tax material performance and limit component operational life. To address this issue, several Department of Energy Office of Nuclear Energy (DOE-NE) research programs are evaluating the long duration irradiation performance of fuel and structural materials used in existing and new reactors. In order to maximize the amount of information obtained from Material Testing Reactor (MTR) irradiations, DOE is also funding development of enhanced instrumentation that will be able to obtain in-situ, real-time data on key material characteristics and properties, with unprecedented accuracy and resolution. Such data are required to validate new multi-scale, multi-physics modeling tools under development as part of a science-based, engineering driven approach to reactor development. It is not feasible to obtain high resolution/microscale data with the current state of instrumentation technology. However, ultrasound-based sensors offer the ability to obtain such data if it is demonstrated that these sensors and their associated transducers are resistant to high neutron flux, high gamma radiation, and high temperature. To address this need, the Advanced Test Reactor National Scientific User Facility (ATR-NSUF) is funding an irradiation, led by PSU, at the Massachusetts Institute of Technology Research Reactor to test the survivability of ultrasound transducers. As part of this effort, PSU and collaborators have designed, fabricated, and provided piezoelectric and magnetostrictive transducers that are optimized to perform in harsh, high flux, environments. Four piezoelectric transducers were fabricated with either aluminum nitride, zinc oxide, or bismuth titanate as the active element that were coupled to either Kovar or aluminum waveguides and two magnetostrictive transducers were fabricated with Remendur or Galfenol as the active elements. Pulse-echo ultrasonic measurements of these transducers are made in-situ. This paper will present an overview of the test design including selection criteria for candidate materials and optimization of test assembly parameters, data obtained from both out-of-pile and in-pile testing at elevated temperatures, and an assessment based on initial data of the expected performance of ultrasonic devices in irradiation conditions.

Development of monolithic nuclear fuels for RERTR by hot isostatic pressing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jue, J.-F.; Park, Blair; Chapple, Michael

2008-07-15

The RERTR Program (Reduced Enrichment for Research and Test Reactors) is developing advanced nuclear fuels for high power test reactors. Monolithic fuel design provides a higher uranium loading than that of the traditional dispersion fuel design. In order to bond monolithic fuel meat to aluminum cladding, several bonding methods such as roll bonding, friction stir bonding and hot isostatic pressing, have been explored. Hot isostatic pressing is a promising process for low cost, batch fabrication of monolithic RERTR fuel plates. The progress on the development of this process at the Idaho National Laboratory will be presented. Due to the relativelymore » high processing temperature used, the reaction between fuel meat and aluminum cladding to form brittle intermetallic phases may be a concern. The effect of processing temperature and time on the fuel/cladding reaction will be addressed. The influence of chemical composition on the reaction will also be discussed. (author)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peterson, Per F.

A high-temperature containment-isolation system for transferring heat from a nuclear reactor containment to a high-pressure heat exchanger is presented. The system uses a high-temperature, low-volatility liquid coolant such as a molten salt or a liquid metal, where the coolant flow path provides liquid free surfaces a short distance from the containment penetrations for the reactor hot-leg and the cold-leg, where these liquid free surfaces have a cover gas maintained at a nearly constant pressure and thus prevent high-pressures from being transmitted into the reactor containment, and where the reactor vessel is suspended within a reactor cavity with a plurality ofmore » refractory insulator blocks disposed between an actively cooled inner cavity liner and the reactor vessel.« less

Comparisons of RELAP5-3D Analyses to Experimental Data from the Natural Convection Shutdown Heat Removal Test Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bucknor, Matthew; Hu, Rui; Lisowski, Darius

2016-04-17

The Reactor Cavity Cooling System (RCCS) is an important passive safety system being incorporated into the overall safety strategy for high temperature advanced reactor concepts such as the High Temperature Gas- Cooled Reactors (HTGR). The Natural Convection Shutdown Heat Removal Test Facility (NSTF) at Argonne National Laboratory (Argonne) reflects a 1/2-scale model of the primary features of one conceptual air-cooled RCCS design. The project conducts ex-vessel, passive heat removal experiments in support of Department of Energy Office of Nuclear Energy’s Advanced Reactor Technology (ART) program, while also generating data for code validation purposes. While experiments are being conducted at themore » NSTF to evaluate the feasibility of the passive RCCS, parallel modeling and simulation efforts are ongoing to support the design, fabrication, and operation of these natural convection systems. Both system-level and high fidelity computational fluid dynamics (CFD) analyses were performed to gain a complete understanding of the complex flow and heat transfer phenomena in natural convection systems. This paper provides a summary of the RELAP5-3D NSTF model development efforts and provides comparisons between simulation results and experimental data from the NSTF. Overall, the simulation results compared favorably to the experimental data, however, further analyses need to be conducted to investigate any identified differences.« less

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville

2016-01-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville

2016-03-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

DYNAMIC AND STATIC PARAMETERS OF THE AQUEOUS HOMOGENEOUS ARMOUR RESEARCH REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Terrell, C.W.; McElroy, W.N.

1959-06-01

A brief description of the aqueous homogeneous Armour Research Reactor is given. The negative reactivity coefficient resulting from a temperature increase was determined over a fuel temperature range of 37 to 150 deg F. Possession of an accurately calibrated rod and temperature coefficient permitted a direct measurement of the void coefficient. The reactor was taken to different power levels, and from the calibrated rod the total reduction in excess reactivity was obtained. During the power increase program additional U/sup 235/ and water were added to the core to determine the worth of U/sup 235/ and water. (W.D.M.)

Advance High Temperature Inspection Capabilities for Small Modular Reactors: Part 1 - Ultrasonics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bond, Leonard J.; Bowler, John R.

The project objective was to investigate the development non-destructive evaluation techniques for advanced small modular reactors (aSMR), where the research sought to provide key enabling inspection technologies needed to support the design and maintenance of reactor component performance. The project tasks for the development of inspection techniques to be applied to small modular reactor are being addressed through two related activities. The first is focused on high temperature ultrasonic transducers development (this report Part 1) and the second is focused on an advanced eddy current inspection capability (Part 2). For both inspection techniques the primary aim is to develop in-servicemore » inspection techniques that can be carried out under standby condition in a fast reactor at a temperature of approximately 250°C in the presence of liquid sodium. The piezoelectric material and the bonding between layers have been recognized as key factors fundamental for development of robust ultrasonic transducers. Dielectric constant characterization of bismuth scantanate-lead titanate ((1-x)BiScO 3-xPbTiO 3) (BS-PT) has shown a high Curie temperature in excess of 450°C , suitable for hot stand-by inspection in liquid metal reactors. High temperature pulse-echo contact measurements have been performed with BS-PT bonded to 12.5 mm thick 1018-low carbon steel plate from 20C up to 260 C. High temperature air-backed immersion transducers have been developed with BS-PT, high temperature epoxy and quarter wavlength nickel plate, needed for wetting ability in liquid sodium. Ultrasonic immersion measurements have been performed in water up to 92C and in silicone oil up to 140C. Physics based models have been validated with room temperature experimental data with benchmark artifical defects.« less

High Temperature Fusion Reactor Cooling Using Brayton Cycle Based Partial Energy Conversion

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.; Sawicki, Jerzy T.

2003-01-01

For some future space power systems using high temperature nuclear heat sources most of the output energy will be used in other than electrical form, and only a fraction of the total thermal energy generated will need to be converted to electrical work. The paper describes the conceptual design of such a partial energy conversion system, consisting of a high temperature fusion reactor operating in series with a high temperature radiator and in parallel with dual closed cycle gas turbine (CCGT) power systems, also referred to as closed Brayton cycle (CBC) systems, which are supplied with a fraction of the reactor thermal energy for conversion to electric power. Most of the fusion reactor's output is in the form of charged plasma which is expanded through a magnetic nozzle of the interplanetary propulsion system. Reactor heat energy is ducted to the high temperature series radiator utilizing the electric power generated to drive a helium gas circulation fan. In addition to discussing the thermodynamic aspects of the system design the authors include a brief overview of the gas turbine and fan rotor-dynamics and proposed bearing support technology along with performance characteristics of the three phase AC electric power generator and fan drive motor.

High Temperature Fusion Reactor Cooling Using Brayton Cycle Based Partial Energy Conversion

NASA Astrophysics Data System (ADS)

Juhasz, Albert J.; Sawicki, Jerzy T.

2004-02-01

For some future space power systems using high temperature nuclear heat sources most of the output energy will be used in other than electrical form, and only a fraction of the total thermal energy generated will need to be converted to electrical work. The paper describes the conceptual design of such a ``partial energy conversion'' system, consisting of a high temperature fusion reactor operating in series with a high temperature radiator and in parallel with dual closed cycle gas turbine (CCGT) power systems, also referred to as closed Brayton cycle (CBC) systems, which are supplied with a fraction of the reactor thermal energy for conversion to electric power. Most of the fusion reactor's output is in the form of charged plasma which is expanded through a magnetic nozzle of the interplanetary propulsion system. Reactor heat energy is ducted to the high temperature series radiator utilizing the electric power generated to drive a helium gas circulation fan. In addition to discussing the thermodynamic aspects of the system design the authors include a brief overview of the gas turbine and fan rotor-dynamics and proposed bearing support technology along with performance characteristics of the three phase AC electric power generator and fan drive motor.

First-wall structural analysis of the self-cooled water blanket concept

DOE Office of Scientific and Technical Information (OSTI.GOV)

O'Brien, D.A.; Steiner, D.; Embrechts, M.J.

1986-01-01

A novel blanket concept recently proposed utilizes water with small amounts of dissolved lithium compound as both coolant and breeder. The inherent simplicity of this idea should result in an attractive breeding blanket for fusion reactors. In addition, the available base of relevant information accumulated through water-cooled fission reactor programs should greatly facilitate the R and D effort required to validate this concept. First-wall and blanket designs have been developed first for the tandem mirror reactor (TMR) due to the obvious advantages of this geometry. First-wall and blanket designs will also be developed for toroidal reactors. A simple plate designmore » with coolant tubes welded on the back (side away from plasma) was chosen as the first wall for the TMR application. Dimensions and materials were chosen to minimize temperature differences and thermal stresses. A finite element code (STRAW), originally developed for the analysis of core components subjected to high-pressure transients in the fast breeder program, was utilized to evaluate stresses in the first wall.« less

DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-TEMPERATURE GAS-COOLED TEST REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James; Bayless, Paul; Strydom, Gerhard

2016-11-01

Uncertainty and sensitivity analysis is an indispensable element of any substantial attempt in reactor simulation validation. The quantification of uncertainties in nuclear engineering has grown more important and the IAEA Coordinated Research Program (CRP) on High-Temperature Gas Cooled Reactor (HTGR) initiated in 2012 aims to investigate the various uncertainty quantification methodologies for this type of reactors. The first phase of the CRP is dedicated to the estimation of cell and lattice model uncertainties due to the neutron cross sections co-variances. Phase II is oriented towards the investigation of propagated uncertainties from the lattice to the coupled neutronics/thermal hydraulics core calculations.more » Nominal results for the prismatic single block (Ex.I-2a) and super cell models (Ex.I-2c) have been obtained using the SCALE 6.1.3 two-dimensional lattice code NEWT coupled to the TRITON sequence for cross section generation. In this work, the TRITON/NEWT-flux-weighted cross sections obtained for Ex.I-2a and various models of Ex.I-2c is utilized to perform a sensitivity analysis of the MHTGR-350 core power densities and eigenvalues. The core solutions are obtained with the INL coupled code PHISICS/RELAP5-3D, utilizing a fixed-temperature feedback for Ex. II-1a.. It is observed that the core power density does not vary significantly in shape, but the magnitude of these variations increases as the moderator-to-fuel ratio increases in the super cell lattice models.« less

Preliminary design of high temperature ultrasonic transducers for liquid sodium environments

NASA Astrophysics Data System (ADS)

Prowant, M. S.; Dib, G.; Qiao, H.; Good, M. S.; Larche, M. R.; Sexton, S. S.; Ramuhalli, P.

2018-04-01

Advanced reactor concepts include fast reactors (including sodium-cooled fast reactors), gas-cooled reactors, and molten-salt reactors. Common to these concepts is a higher operating temperature (when compared to light-water-cooled reactors), and the proposed use of new alloys with which there is limited operational experience. Concerns about new degradation mechanisms, such as high-temperature creep and creep fatigue, that are not encountered in the light-water fleet and longer operating cycles between refueling intervals indicate the need for condition monitoring technology. Specific needs in this context include periodic in-service inspection technology for the detection and sizing of cracking, as well as technologies for continuous monitoring of components using in situ probes. This paper will discuss research on the development and evaluation of high temperature (>550°C; >1022°F) ultrasonic probes that can be used for continuous monitoring of components. The focus of this work is on probes that are compatible with a liquid sodium-cooled reactor environment, where the core outlet temperatures can reach 550°C (1022°F). Modeling to assess sensitivity of various sensor configurations and experimental evaluation have pointed to a preferred design and concept of operations for these probes. This paper will describe these studies and ongoing work to fabricate and fully evaluate survivability and sensor performance over extended periods at operational temperatures.

High yields of hydrogen production from methanol steam reforming with a cross-U type reactor

PubMed Central

Zhang, Shubin; Chen, Junyu; Zhang, Xuelin; Liu, Xiaowei

2017-01-01

This paper presents a numerical and experimental study on the performance of a methanol steam reformer integrated with a hydrogen/air combustion reactor for hydrogen production. A CFD-based 3D model with mass and momentum transport and temperature characteristics is established. The simulation results show that better performance is achieved in the cross-U type reactor compared to either a tubular reactor or a parallel-U type reactor because of more effective heat transfer characteristics. Furthermore, Cu-based micro reformers of both cross-U and parallel-U type reactors are designed, fabricated and tested for experimental validation. Under the same condition for reforming and combustion, the results demonstrate that higher methanol conversion is achievable in cross-U type reactor. However, it is also found in cross-U type reactor that methanol reforming selectivity is the lowest due to the decreased water gas shift reaction under high temperature, thereby carbon monoxide concentration is increased. Furthermore, the reformed gas generated from the reactors is fed into a high temperature proton exchange membrane fuel cell (PEMFC). In the test of discharging for 4 h, the fuel cell fed by cross-U type reactor exhibits the most stable performance. PMID:29121067

High yields of hydrogen production from methanol steam reforming with a cross-U type reactor.

PubMed

Zhang, Shubin; Zhang, Yufeng; Chen, Junyu; Zhang, Xuelin; Liu, Xiaowei

2017-01-01

This paper presents a numerical and experimental study on the performance of a methanol steam reformer integrated with a hydrogen/air combustion reactor for hydrogen production. A CFD-based 3D model with mass and momentum transport and temperature characteristics is established. The simulation results show that better performance is achieved in the cross-U type reactor compared to either a tubular reactor or a parallel-U type reactor because of more effective heat transfer characteristics. Furthermore, Cu-based micro reformers of both cross-U and parallel-U type reactors are designed, fabricated and tested for experimental validation. Under the same condition for reforming and combustion, the results demonstrate that higher methanol conversion is achievable in cross-U type reactor. However, it is also found in cross-U type reactor that methanol reforming selectivity is the lowest due to the decreased water gas shift reaction under high temperature, thereby carbon monoxide concentration is increased. Furthermore, the reformed gas generated from the reactors is fed into a high temperature proton exchange membrane fuel cell (PEMFC). In the test of discharging for 4 h, the fuel cell fed by cross-U type reactor exhibits the most stable performance.

NGNP High Temperature Materials White Paper

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lew Lommers; George Honma

2012-08-01

This white paper is one in a series of white papers that address key generic issues of the combined construction and operating license (COL) pre-application program key generic issues for the Next Generation Nuclear Plant reactor using the prismatic block fuel technology. The purpose of the pre-application program interactions with the NRC staff is to reduce the time required for COL application review by identifying and addressing key regulatory issues and, if possible, obtaining agreements for their resolution

Low exchange element for nuclear reactor

DOEpatents

Brogli, Rudolf H.; Shamasunder, Bangalore I.; Seth, Shivaji S.

1985-01-01

A flow exchange element is presented which lowers temperature gradients in fuel elements and reduces maximum local temperature within high temperature gas-cooled reactors. The flow exchange element is inserted within a column of fuel elements where it serves to redirect coolant flow. Coolant which has been flowing in a hotter region of the column is redirected to a cooler region, and coolant which has been flowing in the cooler region of the column is redirected to the hotter region. The safety, efficiency, and longevity of the high temperature gas-cooled reactor is thereby enhanced.

Syngas Production By Thermochemical Conversion Of H2o And Co2 Mixtures Using A Novel Reactor Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pearlman, Howard; Chen, Chien-Hua

The Department of Energy awarded Advanced Cooling Technologies, Inc. (ACT) an SBIR Phase II contract (#DE-SC0004729) to develop a high-temperature solar thermochemical reactor for syngas production using water and/or carbon dioxide as feedstocks. The technology aims to provide a renewable and sustainable alternative to fossil fuels, promote energy independence and mitigate adverse issues associated with climate change by essentially recycling carbon from carbon dioxide emitted by the combustion of hydrocarbon fuels. To commercialize the technology and drive down the cost of solar fuels, new advances are needed in materials development and reactor design, both of which are integral elements inmore » this program.« less

Development of a Polysilicon Process Based on Chemical Vapor Deposition of Dichlorosilane in an Advanced Siemen's Reactor

NASA Technical Reports Server (NTRS)

Arevidson, A. N.; Sawyer, D. H.; Muller, D. M.

1983-01-01

Dichlorosilane (DCS) was used as the feedstock for an advanced decomposition reactor for silicon production. The advanced reactor had a cool bell jar wall temperature, 300 C, when compared to Siemen's reactors previously used for DCS decomposition. Previous reactors had bell jar wall temperatures of approximately 750 C. The cooler wall temperature allows higher DCS flow rates and concentrations. A silicon deposition rate of 2.28 gm/hr-cm was achieved with power consumption of 59 kWh/kg. Interpretation of data suggests that a 2.8 gm/hr-cm deposition rate is possible. Screening of lower cost materials of construction was done as a separate program segment. Stainless Steel (304 and 316), Hastalloy B, Monel 400 and 1010-Carbon Steel were placed individually in an experimental scale reactor. Silicon was deposited from trichlorosilane feedstock. The resultant silicon was analyzed for electrically active and metallic impurities as well as carbon. No material contributed significant amounts of electrically active or metallic impurities, but all contributed carbon.

New Temperature Monitoring Devices for High-Temperature Irradiation Experiments in the High Flux Reactor Petten

NASA Astrophysics Data System (ADS)

Laurie, M.; Futterer, M. A.; Lapetite, J. M.; Fourrez, S.; Morice, R.

2011-10-01

Within the European High Temperature Reactor Technology Network (HTR-TN) and related projects a number of HTR fuel irradiations are planned in the High Flux Reactor Petten (HFR), The Netherlands, with the objective to explore the potential of recently produced fuel for even higher temperature and burn-up. Irradiating fuel under defined conditions to extremely high burn-ups will provide a better understanding of fission product release and failure mechanisms if particle failure occurs. After an overview of the irradiation rigs used in the HFR, this paper sums up data collected from previous irradiation tests in terms of thermocouple data. Some R&D for further improvement of thermocouples and other on-line instrumentation will be outlined.

Coupled field-structural analysis of HGTR fuel brick using ABAQUS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohanty, S.; Jain, R.; Majumdar, S.

2012-07-01

High-temperature, gas-cooled reactors (HTGRs) are usually helium-gas cooled, with a graphite core that can operate at reactor outlet temperatures much higher than can conventional light water reactors. In HTGRs, graphite components moderate and reflect neutrons. During reactor operation, high temperature and high irradiation cause damage to the graphite crystal and grains and create other defects. This cumulative structural damage during the reactor lifetime leads to changes in graphite properties, which can alter the ability to support the designed loads. The aim of the present research is to develop a finite-element code using commercially available ABAQUS software for the structural integritymore » analysis of graphite core components under extreme temperature and irradiation conditions. In addition, the Reactor Geometry Generator tool-kit, developed at Argonne National Laboratory, is used to generate finite-element mesh for complex geometries such as fuel bricks with multiple pin holes and coolant flow channels. This paper presents the proposed concept and discusses results of stress analysis simulations of a fuel block with H-451 grade material properties. (authors)« less

Dual-mode, high energy utilization system concept for mars missions

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.

2000-01-01

This paper describes a dual-mode, high energy utilization system concept based on the Pellet Bed Reactor (PeBR) to support future manned missions to Mars. The system uses proven Closed Brayton Cycle (CBC) engines to partially convert the reactor thermal power to electricity. The electric power generated is kept the same during the propulsion and the power modes, but the reactor thermal power in the former could be several times higher, while maintaining the reactor temperatures almost constant. During the propulsion mode, the electric power of the system, minus ~1-5 kWe for house keeping, is used to operate a Variable Specific Impulse Magnetoplasma Rocket (VASIMR). In addition, the reactor thermal power, plus more than 85% of the head load of the CBC engine radiators, are used to heat hydrogen. The hot hydrogen is mixed with the high temperature plasma in a VASIMR to provide both high thrust and Isp>35,000 N.s/kg, reducing the travel time to Mars to about 3 months. The electric power also supports surface exploration of Mars. The fuel temperature and the inlet temperatures of the He-Xe working fluid to the nuclear reactor core and the CBC turbine are maintained almost constant during both the propulsion and power modes to minimize thermal stresses. Also, the exit temperature of the He-Xe from the reactor core is kept at least 200 K below the maximum fuel design temperature. The present system has no single point failure and could be tested fully assembled in a ground facility using electric heaters in place of the nuclear reactor. Operation and design parameters of a 40-kWe prototype are presented and discussed to illustrate the operation and design principles of the proposed system. .

Nuclear fuel elements made from nanophase materials

DOEpatents

Heubeck, Norman B.

1998-01-01

A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000.degree. F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics.

Nuclear fuel elements made from nanophase materials

DOEpatents

Heubeck, N.B.

1998-09-08

A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000 F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics. 5 figs.

Next Generation Nuclear Plant Methods Research and Development Technical Program Plan -- PLN-2498

DOE Office of Scientific and Technical Information (OSTI.GOV)

Richard R. Schultz; Abderrafi M. Ougouag; David W. Nigg

2008-09-01

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope ofmore » the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.« less

Next Generation Nuclear Plant Methods Technical Program Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Richard R. Schultz; Abderrafi M. Ougouag; David W. Nigg

2010-12-01

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope ofmore » the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.« less

Next Generation Nuclear Plant Methods Technical Program Plan -- PLN-2498

DOE Office of Scientific and Technical Information (OSTI.GOV)

Richard R. Schultz; Abderrafi M. Ougouag; David W. Nigg

2010-09-01

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope ofmore » the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.« less

Top shield temperatures, C and K Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Agar, J.D.

1964-12-28

A modification program is now in progress at the C and K Reactors consisting of an extensive renovation of the graphite channels in the vertical safety rod ststems. The present VSR channels are being enlarged by a graphite coring operation and channel sleeves will be installed in the larger channels. One problem associated with the coring operation is the danger of damaging top thermal shield cooling tubes located close to the VSR channels to such an extent that these tubes will have to be removed from service. If such a condition should exist at one or a number of locationsmore » in the top shield of the reactors after reactor startup, the question remains -- what would the resulting temperatures be of the various components of the top shields? This study was initiated to determine temperature distributions in the top shield complex at the C and K Reactors for various top thermal shield coolant system conditions. Since the top thermal shield cooling system at C Reactor is different than those at the K Reactors, the study was conducted separately for the two different systems.« less

An numerical analysis of high-temperature helium reactor power plant for co-production of hydrogen and electricity

NASA Astrophysics Data System (ADS)

Dudek, M.; Podsadna, J.; Jaszczur, M.

2016-09-01

In the present work, the feasibility of using a high temperature gas cooled nuclear reactor (HTR) for electricity generation and hydrogen production are analysed. The HTR is combined with a steam and a gas turbine, as well as with the system for heat delivery for medium temperature hydrogen production. Industrial-scale hydrogen production using copper-chlorine (Cu-Cl) thermochemical cycle is considered and compared with high temperature electrolysis. Presented cycle shows a very promising route for continuous, efficient, large-scale and environmentally benign hydrogen production without CO2 emissions. The results show that the integration of a high temperature helium reactor, with a combined cycle for electric power generation and hydrogen production, may reach very high efficiency and could possibly lead to a significant decrease of hydrogen production costs.

Design and Analysis of Embedded I&C for a Fully Submerged Magnetically Suspended Impeller Pump

DOE PAGES

Melin, Alexander M.; Kisner, Roger A.

2018-04-03

Improving nuclear reactor power system designs and fuel-processing technologies for safer and more efficient operation requires the development of new component designs. In particular, many of the advanced reactor designs such as the molten salt reactors and high-temperature gas-cooled reactors have operating environments beyond the capability of most currently available commercial components. To address this gap, new cross-cutting technologies need to be developed that will enable design, fabrication, and reliable operation of new classes of reactor components. The Advanced Sensor Initiative of the Nuclear Energy Enabling Technologies initiative is investigating advanced sensor and control designs that are capable of operatingmore » in these extreme environments. Under this initiative, Oak Ridge National Laboratory (ORNL) has been developing embedded instrumentation and control (I&C) for extreme environments. To develop, test, and validate these new sensing and control techniques, ORNL is building a pump test bed that utilizes submerged magnetic bearings to levitate the shaft. The eventual goal is to apply these techniques to a high-temperature (700°C) canned rotor pump that utilizes active magnetic bearings to eliminate the need for mechanical bearings and seals. The technologies will benefit the Next Generation Power Plant, Advanced Reactor Concepts, and Small Modular Reactor programs. In this paper, we will detail the design and analysis of the embedded I&C test bed with submerged magnetic bearings, focusing on the interplay between the different major systems. Then we will analyze the forces on the shaft and their role in the magnetic bearing design. Next, we will develop the radial and thrust bearing geometries needed to meet the operational requirements of the test bed. In conclusion, we will present some initial system identification results to validate the theoretical models of the test bed dynamics.« less

Design and Analysis of Embedded I&C for a Fully Submerged Magnetically Suspended Impeller Pump

DOE Office of Scientific and Technical Information (OSTI.GOV)

Melin, Alexander M.; Kisner, Roger A.

Improving nuclear reactor power system designs and fuel-processing technologies for safer and more efficient operation requires the development of new component designs. In particular, many of the advanced reactor designs such as the molten salt reactors and high-temperature gas-cooled reactors have operating environments beyond the capability of most currently available commercial components. To address this gap, new cross-cutting technologies need to be developed that will enable design, fabrication, and reliable operation of new classes of reactor components. The Advanced Sensor Initiative of the Nuclear Energy Enabling Technologies initiative is investigating advanced sensor and control designs that are capable of operatingmore » in these extreme environments. Under this initiative, Oak Ridge National Laboratory (ORNL) has been developing embedded instrumentation and control (I&C) for extreme environments. To develop, test, and validate these new sensing and control techniques, ORNL is building a pump test bed that utilizes submerged magnetic bearings to levitate the shaft. The eventual goal is to apply these techniques to a high-temperature (700°C) canned rotor pump that utilizes active magnetic bearings to eliminate the need for mechanical bearings and seals. The technologies will benefit the Next Generation Power Plant, Advanced Reactor Concepts, and Small Modular Reactor programs. In this paper, we will detail the design and analysis of the embedded I&C test bed with submerged magnetic bearings, focusing on the interplay between the different major systems. Then we will analyze the forces on the shaft and their role in the magnetic bearing design. Next, we will develop the radial and thrust bearing geometries needed to meet the operational requirements of the test bed. In conclusion, we will present some initial system identification results to validate the theoretical models of the test bed dynamics.« less

SAM Theory Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Rui

The System Analysis Module (SAM) is an advanced and modern system analysis tool being developed at Argonne National Laboratory under the U.S. DOE Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. SAM development aims for advances in physical modeling, numerical methods, and software engineering to enhance its user experience and usability for reactor transient analyses. To facilitate the code development, SAM utilizes an object-oriented application framework (MOOSE), and its underlying meshing and finite-element library (libMesh) and linear and non-linear solvers (PETSc), to leverage modern advanced software environments and numerical methods. SAM focuses on modeling advanced reactormore » concepts such as SFRs (sodium fast reactors), LFRs (lead-cooled fast reactors), and FHRs (fluoride-salt-cooled high temperature reactors) or MSRs (molten salt reactors). These advanced concepts are distinguished from light-water reactors in their use of single-phase, low-pressure, high-temperature, and low Prandtl number (sodium and lead) coolants. As a new code development, the initial effort has been focused on modeling and simulation capabilities of heat transfer and single-phase fluid dynamics responses in Sodium-cooled Fast Reactor (SFR) systems. The system-level simulation capabilities of fluid flow and heat transfer in general engineering systems and typical SFRs have been verified and validated. This document provides the theoretical and technical basis of the code to help users understand the underlying physical models (such as governing equations, closure models, and component models), system modeling approaches, numerical discretization and solution methods, and the overall capabilities in SAM. As the code is still under ongoing development, this SAM Theory Manual will be updated periodically to keep it consistent with the state of the development.« less

Facile synthesis of graphene on dielectric surfaces using a two-temperature reactor CVD system

NASA Astrophysics Data System (ADS)

Zhang, C.; Man, B. Y.; Yang, C.; Jiang, S. Z.; Liu, M.; Chen, C. S.; Xu, S. C.; Sun, Z. C.; Gao, X. G.; Chen, X. J.

2013-10-01

Direct deposition of graphene on a dielectric substrate is demonstrated using a chemical vapor deposition system with a two-temperature reactor. The two-temperature reactor is utilized to offer sufficient, well-proportioned floating Cu atoms and to provide a temperature gradient for facile synthesis of graphene on dielectric surfaces. The evaporated Cu atoms catalyze the reaction in the presented method. C atoms and Cu atoms respectively act as the nuclei for forming graphene film in the low-temperature zone and the zones close to the high-temperature zones. A uniform and high-quality graphene film is formed in an atmosphere of sufficient and well-proportioned floating Cu atoms. Raman spectroscopy, scanning electron microscopy and atomic force microscopy confirm the presence of uniform and high-quality graphene.

Operational Philosophy for the Advanced Test Reactor National Scientific User Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Benson; J. Cole; J. Jackson

2013-02-01

In 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). At its core, the ATR NSUF Program combines access to a portion of the available ATR radiation capability, the associated required examination and analysis facilities at the Idaho National Laboratory (INL), and INL staff expertise with novel ideas provided by external contributors (universities, laboratories, and industry). These collaborations define the cutting edge of nuclear technology research in high-temperature and radiation environments, contribute to improved industry performance of current and future light-water reactors (LWRs), and stimulate cooperative research between user groupsmore » conducting basic and applied research. To make possible the broadest access to key national capability, the ATR NSUF formed a partnership program that also makes available access to critical facilities outside of the INL. Finally, the ATR NSUF has established a sample library that allows access to pre-irradiated samples as needed by national research teams.« less

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Belles, Randy; Poore, III, Willis P.; Brown, Nicholas R.

2017-03-01

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-basedmore » description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.« less

High-Temperature Gas-Cooled Test Reactor Point Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville

2016-04-01

A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

Problems and prospects connected with development of high-temperature filtration technology at nuclear power plants equipped with VVER-1000 reactors

NASA Astrophysics Data System (ADS)

Shchelik, S. V.; Pavlov, A. S.

2013-07-01

Results of work on restoring the service properties of filtering material used in the high-temperature reactor coolant purification system of a VVER-1000 reactor are presented. A quantitative assessment is given to the effect from subjecting a high-temperature sorbent to backwashing operations carried out with the use of regular capacities available in the design process circuit in the first years of operation of Unit 3 at the Kalinin nuclear power plant. Approaches to optimizing this process are suggested. A conceptual idea about comprehensively solving the problem of achieving more efficient and safe operation of the high-temperature active water treatment system (AWT-1) on a nuclear power industry-wide scale is outlined.

FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Davidson, J.K.

1963-11-19

A fuel element structure particularly useful in high temperature nuclear reactors is presented. Basically, the structure comprises two coaxial graphite sleeves integrally joined together by radial fins. Due to the high structural strength of graphite at high temperatures and the rigidity of this structure, nuclear fuel encased within the inner sleeve in contiguous relation therewith is supported and prevented from expanding radially at high temperatures. Thus, the necessity of relying on the usual cladding materials with relatively low temperature limitations for structural strength is removed. (AEC)

Design of a new reactor-like high temperature near ambient pressure scanning tunneling microscope for catalysis studies.

PubMed

Tao, Franklin Feng; Nguyen, Luan; Zhang, Shiran

2013-03-01

Here, we present the design of a new reactor-like high-temperature near ambient pressure scanning tunneling microscope (HT-NAP-STM) for catalysis studies. This HT-NAP-STM was designed for exploration of structures of catalyst surfaces at atomic scale during catalysis or under reaction conditions. In this HT-NAP-STM, the minimized reactor with a volume of reactant gases of ∼10 ml is thermally isolated from the STM room through a shielding dome installed between the reactor and STM room. An aperture on the dome was made to allow tip to approach to or retract from a catalyst surface in the reactor. This dome minimizes thermal diffusion from hot gas of the reactor to the STM room and thus remains STM head at a constant temperature near to room temperature, allowing observation of surface structures at atomic scale under reaction conditions or during catalysis with minimized thermal drift. The integrated quadrupole mass spectrometer can simultaneously measure products during visualization of surface structure of a catalyst. This synergy allows building an intrinsic correlation between surface structure and its catalytic performance. This correlation offers important insights for understanding of catalysis. Tests were done on graphite in ambient environment, Pt(111) in CO, graphene on Ru(0001) in UHV at high temperature and gaseous environment at high temperature. Atom-resolved surface structure of graphene on Ru(0001) at 500 K in a gaseous environment of 25 Torr was identified.

Small Reactor Designs Suitable for Direct Nuclear Thermal Propulsion: Interim Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bruce G. Schnitzler

Advancement of U.S. scientific, security, and economic interests requires high performance propulsion systems to support missions beyond low Earth orbit. A robust space exploration program will include robotic outer planet and crewed missions to a variety of destinations including the moon, near Earth objects, and eventually Mars. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. In NASA's recent Mars Design Reference Architecture (DRA) 5.0 study, nuclear thermal propulsion (NTP) was again selectedmore » over chemical propulsion as the preferred in-space transportation system option for the human exploration of Mars because of its high thrust and high specific impulse ({approx}900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. The recently announced national space policy2 supports the development and use of space nuclear power systems where such systems safely enable or significantly enhance space exploration or operational capabilities. An extensive nuclear thermal rocket technology development effort was conducted under the Rover/NERVA, GE-710 and ANL nuclear rocket programs (1955-1973). Both graphite and refractory metal alloy fuel types were pursued. The primary and significantly larger Rover/NERVA program focused on graphite type fuels. Research, development, and testing of high temperature graphite fuels was conducted. Reactors and engines employing these fuels were designed, built, and ground tested. The GE-710 and ANL programs focused on an alternative ceramic-metallic 'cermet' fuel type consisting of UO2 (or UN) fuel embedded in a refractory metal matrix such as tungsten. The General Electric program examined closed loop concepts for space or terrestrial applications as well as open loop systems for direct nuclear thermal propulsion. Although a number of fast spectrum reactor and engine designs suitable for direct nuclear thermal propulsion were proposed and designed, none were built. This report summarizes status results of evaluations of small nuclear reactor designs suitable for direct nuclear thermal propulsion.« less

Hybrid sulfur cycle operation for high-temperature gas-cooled reactors

DOEpatents

Gorensek, Maximilian B

2015-02-17

A hybrid sulfur (HyS) cycle process for the production of hydrogen is provided. The process uses a proton exchange membrane (PEM) SO.sub.2-depolarized electrolyzer (SDE) for the low-temperature, electrochemical reaction step and a bayonet reactor for the high-temperature decomposition step The process can be operated at lower temperature and pressure ranges while still providing an overall energy efficient cycle process.

STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Davenport, Michael; Petti, D. A.; Palmer, Joe

2016-11-01

The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experimentsmore » are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control and monitoring systems are very similar. The final experiment, AGR-5/6/7, is scheduled to begin irradiation in early summer 2017.« less

Multi-Purpose Thermal Hydraulic Loop: Advanced Reactor Technology Integral System Test (ARTIST) Facility for Support of Advanced Reactor Technologies

DOE Office of Scientific and Technical Information (OSTI.GOV)

James E. O'Brien; Piyush Sabharwall; SuJong Yoon

2001-11-01

Effective and robust high temperature heat transfer systems are fundamental to the successful deployment of advanced reactors for both power generation and non-electric applications. Plant designs often include an intermediate heat transfer loop (IHTL) with heat exchangers at either end to deliver thermal energy to the application while providing isolation of the primary reactor system. In order to address technical feasibility concerns and challenges a new high-temperature multi-fluid, multi-loop test facility “Advanced Reactor Technology Integral System Test facility” (ARTIST) is under development at the Idaho National Laboratory. The facility will include three flow loops: high-temperature helium, molten salt, and steam/water.more » Details of some of the design aspects and challenges of this facility, which is currently in the conceptual design phase, are discussed« less

In situ monitored in-pile creep testing of zirconium alloys

NASA Astrophysics Data System (ADS)

Kozar, R. W.; Jaworski, A. W.; Webb, T. W.; Smith, R. W.

2014-01-01

The experiments described herein were designed to investigate the detailed irradiation creep behavior of zirconium based alloys in the HALDEN Reactor spectrum. The HALDEN Test Reactor has the unique capability to control both applied stress and temperature independently and externally for each specimen while the specimen is in-reactor and under fast neutron flux. The ability to monitor in situ the creep rates following a stress and temperature change made possible the characterization of creep behavior over a wide stress-strain-rate-temperature design space for two model experimental heats, Zircaloy-2 and Zircaloy-2 + 1 wt%Nb, with only 12 test specimens in a 100-day in-pile creep test program. Zircaloy-2 specimens with and without 1 wt% Nb additions were tested at irradiation temperatures of 561 K and 616 K and stresses ranging from 69 MPa to 455 MPa. Various steady state creep models were evaluated against the experimental results. The irradiation creep model proposed by Nichols that separates creep behavior into low, intermediate, and high stress regimes was the best model for predicting steady-state creep rates. Dislocation-based primary creep, rather than diffusion-based transient irradiation creep, was identified as the mechanism controlling deformation during the transitional period of evolving creep rate following a step change to different test conditions.

Fuel leak detection apparatus for gas cooled nuclear reactors

DOEpatents

Burnette, Richard D.

1977-01-01

Apparatus is disclosed for detecting nuclear fuel leaks within nuclear power system reactors, such as high temperature gas cooled reactors. The apparatus includes a probe assembly that is inserted into the high temperature reactor coolant gaseous stream. The probe has an aperture adapted to communicate gaseous fluid between its inside and outside surfaces and also contains an inner tube for sampling gaseous fluid present near the aperture. A high pressure supply of noncontaminated gas is provided to selectively balance the pressure of the stream being sampled to prevent gas from entering the probe through the aperture. The apparatus includes valves that are operable to cause various directional flows and pressures, which valves are located outside of the reactor walls to permit maintenance work and the like to be performed without shutting down the reactor.

Development of OTM Syngas Process and Testing of Syngas Derived Ultra-clean Fuels in Diesel Engines and Fuel Cells

DOE Office of Scientific and Technical Information (OSTI.GOV)

E.T.; James P. Meagher; Prasad Apte

2002-12-31

This topical report summarizes work accomplished for the Program from November 1, 2001 to December 31, 2002 in the following task areas: Task 1: Materials Development; Task 2: Composite Development; Task 4: Reactor Design and Process Optimization; Task 8: Fuels and Engine Testing; 8.1 International Diesel Engine Program; 8.2 Nuvera Fuel Cell Program; and Task 10: Program Management. Major progress has been made towards developing high temperature, high performance, robust, oxygen transport elements. In addition, a novel reactor design has been proposed that co-produces hydrogen, lowers cost and improves system operability. Fuel and engine testing is progressing well, but wasmore » delayed somewhat due to the hiatus in program funding in 2002. The Nuvera fuel cell portion of the program was completed on schedule and delivered promising results regarding low emission fuels for transportation fuel cells. The evaluation of ultra-clean diesel fuels continues in single cylinder (SCTE) and multiple cylinder (MCTE) test rigs at International Truck and Engine. FT diesel and a BP oxygenate showed significant emissions reductions in comparison to baseline petroleum diesel fuels. Overall through the end of 2002 the program remains under budget, but behind schedule in some areas.« less

Plasma flow reactor for steady state monitoring of physical and chemical processes at high temperatures.

PubMed

Koroglu, Batikan; Mehl, Marco; Armstrong, Michael R; Crowhurst, Jonathan C; Weisz, David G; Zaug, Joseph M; Dai, Zurong; Radousky, Harry B; Chernov, Alex; Ramon, Erick; Stavrou, Elissaios; Knight, Kim; Fabris, Andrea L; Cappelli, Mark A; Rose, Timothy P

2017-09-01

We present the development of a steady state plasma flow reactor to investigate gas phase physical and chemical processes that occur at high temperature (1000 < T < 5000 K) and atmospheric pressure. The reactor consists of a glass tube that is attached to an inductively coupled argon plasma generator via an adaptor (ring flow injector). We have modeled the system using computational fluid dynamics simulations that are bounded by measured temperatures. In situ line-of-sight optical emission and absorption spectroscopy have been used to determine the structures and concentrations of molecules formed during rapid cooling of reactants after they pass through the plasma. Emission spectroscopy also enables us to determine the temperatures at which these dynamic processes occur. A sample collection probe inserted from the open end of the reactor is used to collect condensed materials and analyze them ex situ using electron microscopy. The preliminary results of two separate investigations involving the condensation of metal oxides and chemical kinetics of high-temperature gas reactions are discussed.

Monolithic catalyst beds for hydrazine reactors

NASA Technical Reports Server (NTRS)

1973-01-01

A monolithic catalyst bed for monopropellant hydrazine decomposition was evaluated. The program involved the evaluation of a new hydrazine catalyst concept wherein open-celled foamed materials are used as supports for the active catalysts. A high-surface-area material is deposited upon the open-celled foamed material and is then coated with an active metal to provide a spontaneous catalyst. Only a fraction of the amount of expensive active metal in currently available catalysts is needed to promote monolithic catalyst. Numerous parameters were evaluated during the program, and the importance of additional parameters became obvious only while the program was in progress. A demonstration firing (using a 2.2-Newton (N)(0.5-lbf) reactor) successfully accumulated 7,700 seconds of firing time and 16 ambient temperature starts without degradation. Based on the excellent results obtained throughout the program and the demonstrated life capability of the monolithic foam, it is recommended that additional studies be conducted to further exploit the advantages of this concept.

Automatic reactor model synthesis with genetic programming.

PubMed

Dürrenmatt, David J; Gujer, Willi

2012-01-01

Successful modeling of wastewater treatment plant (WWTP) processes requires an accurate description of the plant hydraulics. Common methods such as tracer experiments are difficult and costly and thus have limited applicability in practice; engineers are often forced to rely on their experience only. An implementation of grammar-based genetic programming with an encoding to represent hydraulic reactor models as program trees should fill this gap: The encoding enables the algorithm to construct arbitrary reactor models compatible with common software used for WWTP modeling by linking building blocks, such as continuous stirred-tank reactors. Discharge measurements and influent and effluent concentrations are the only required inputs. As shown in a synthetic example, the technique can be used to identify a set of reactor models that perform equally well. Instead of being guided by experience, the most suitable model can now be chosen by the engineer from the set. In a second example, temperature measurements at the influent and effluent of a primary clarifier are used to generate a reactor model. A virtual tracer experiment performed on the reactor model has good agreement with a tracer experiment performed on-site.

Combustion Chemistry of Fuels: Quantitative Speciation Data Obtained from an Atmospheric High-temperature Flow Reactor with Coupled Molecular-beam Mass Spectrometer.

PubMed

Köhler, Markus; Oßwald, Patrick; Krueger, Dominik; Whitside, Ryan

2018-02-19

This manuscript describes a high-temperature flow reactor experiment coupled to the powerful molecular beam mass spectrometry (MBMS) technique. This flexible tool offers a detailed observation of chemical gas-phase kinetics in reacting flows under well-controlled conditions. The vast range of operating conditions available in a laminar flow reactor enables access to extraordinary combustion applications that are typically not achievable by flame experiments. These include rich conditions at high temperatures relevant for gasification processes, the peroxy chemistry governing the low temperature oxidation regime or investigations of complex technical fuels. The presented setup allows measurements of quantitative speciation data for reaction model validation of combustion, gasification and pyrolysis processes, while enabling a systematic general understanding of the reaction chemistry. Validation of kinetic reaction models is generally performed by investigating combustion processes of pure compounds. The flow reactor has been enhanced to be suitable for technical fuels (e.g. multi-component mixtures like Jet A-1) to allow for phenomenological analysis of occurring combustion intermediates like soot precursors or pollutants. The controlled and comparable boundary conditions provided by the experimental design allow for predictions of pollutant formation tendencies. Cold reactants are fed premixed into the reactor that are highly diluted (in around 99 vol% in Ar) in order to suppress self-sustaining combustion reactions. The laminar flowing reactant mixture passes through a known temperature field, while the gas composition is determined at the reactors exhaust as a function of the oven temperature. The flow reactor is operated at atmospheric pressures with temperatures up to 1,800 K. The measurements themselves are performed by decreasing the temperature monotonically at a rate of -200 K/h. With the sensitive MBMS technique, detailed speciation data is acquired and quantified for almost all chemical species in the reactive process, including radical species.

Supplemental Thermal-Hydraulic Transient Analyses of BR2 in Support of Conversion to LEU Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Licht, J.; Dionne, B.; Sikik, E.

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The RELAP5/Mod 3.3 code has been used to perform transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. A RELAP5 model of BR2 has been validated against select transient BR2 reactor experiments performed in 1963 by showingmore » agreement with measured cladding temperatures. Following the validation, the RELAP5 model was then updated to represent the current use of the reactor; taking into account core configuration, neutronic parameters, trip settings, component changes, etc. Simulations of the 1963 experiments were repeated with this updated model to re-evaluate the boiling risks associated with the currently allowed maximum heat flux limit of 470 W/cm 2 and temporary heat flux limit of 600 W/cm 2. This document provides analysis of additional transient simulations that are required as part of a modern BR2 safety analysis report (SAR). The additional simulations included in this report are effect of pool temperature, reduced steady-state flow rate, in-pool loss of coolant accidents, and loss of external cooling. The simulations described in this document have been performed for both an HEU- and LEU-fueled core.« less

Safety philosophy of gas turbine high temperature reactor (GTHTR300)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shoji Katanishi; Kazuhiko Kunitomi; Shusaku Shiozawa

2002-07-01

Japan Atomic Energy Research Institute (JAERI) has undertaken the study of an original design concept of gas turbine high temperature reactor, the GTHTR300. The general concept of this study is development of a greatly simplified design that leads to substantially reduced technical and cost requirements. Newly proposed design features enable the GTHTR300 to be an efficient and economically competitive reactor in 2010's. Also, the GTHTR300 fully takes advantage of its inherent safety characteristics. The safety philosophy of the GTHTR300 is developed based on the HTTR (High Temperature Engineering Test Reactor) of JAERI which is the first HTGR in Japan. Majormore » features of the newly proposed safety philosophy for the GTHTR300 are described in this article. (authors)« less

Fundamental Mechanisms, Predictive Modeling, and Novel Aerospace Applications of Plasma Assisted Combustion: Laminar Flow Reactor and Nanoparticle Studies at Low to Intermediate Temperatures. Program Overview

DTIC Science & Technology

2009-11-04

plasma enhanced combustion in flow  reactors and flames Motivation •Nano‐ particles  are known to be ionized more easily than  molecules  and atoms (due to...aluminum nano‐ particles  at high  temperature (~1100 K), providing a strong driving force for ion  transport •Nano‐ particles  are chemically and catalytically...active in plasma  •Functionalized nano‐ particles  may enhance the effectiveness of  plasma  Functionalized graphene sheet colloids enhance fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aaron, Adam M.; Cunningham, Richard Burns; Fugate, David L.

Effective high-temperature thermal energy exchange and delivery at temperatures over 600°C has the potential of significant impact by reducing both the capital and operating cost of energy conversion and transport systems. It is one of the key technologies necessary for efficient hydrogen production and could potentially enhance efficiencies of high-temperature solar systems. Today, there are no standard commercially available high-performance heat transfer fluids above 600°C. High pressures associated with water and gaseous coolants (such as helium) at elevated temperatures impose limiting design conditions for the materials in most energy systems. Liquid salts offer high-temperature capabilities at low vapor pressures, goodmore » heat transport properties, and reasonable costs and are therefore leading candidate fluids for next-generation energy production. Liquid-fluoride-salt-cooled, graphite-moderated reactors, referred to as Fluoride Salt Reactors (FHRs), are specifically designed to exploit the excellent heat transfer properties of liquid fluoride salts while maximizing their thermal efficiency and minimizing cost. The FHR s outstanding heat transfer properties, combined with its fully passive safety, make this reactor the most technologically desirable nuclear power reactor class for next-generation energy production. Multiple FHR designs are presently being considered. These range from the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) [1] design originally developed by UC-Berkeley to the Small Advanced High-Temperature Reactor (SmAHTR) and the large scale FHR both being developed at ORNL [2]. The value of high-temperature, molten-salt-cooled reactors is also recognized internationally, and Czechoslovakia, France, India, and China all have salt-cooled reactor development under way. The liquid salt experiment presently being developed uses the PB-AHTR as its focus. One core design of the PB-AHTR features multiple 20 cm diameter, 3.2 m long fuel channels with 3 cm diameter graphite-based fuel pebbles slowly circulating up through the core. Molten salt coolant (FLiBe) at 700°C flows concurrently (at significantly higher velocity) with the pebbles and is used to remove heat generated in the reactor core (approximately 1280 W/pebble), and supply it to a power conversion system. Refueling equipment continuously sorts spent fuel pebbles and replaces spent or damaged pebbles with fresh fuel. By combining greater or fewer numbers of pebble channel assemblies, multiple reactor designs with varying power levels can be offered. The PB-AHTR design is discussed in detail in Reference [1] and is shown schematically in Fig. 1. Fig. 1. PB-AHTR concept (drawing taken from Peterson et al., Design and Development of the Modular PB-AHTR Proceedings of ICApp 08). Pebble behavior within the core is a key issue in proving the viability of this concept. This includes understanding the behavior of the pebbles thermally, hydraulically, and mechanically (quantifying pebble wear characteristics, flow channel wear, etc). The experiment being developed is an initial step in characterizing the pebble behavior under realistic PB-AHTR operating conditions. It focuses on thermal and hydraulic behavior of a static pebble bed using a convective salt loop to provide prototypic fluid conditions to the bed, and a unique inductive heating technique to provide prototypic heating in the pebbles. The facility design is sufficiently versatile to allow a variety of other experimentation to be performed in the future. The facility can accommodate testing of scaled reactor components or sub-components such as flow diodes, salt-to-salt heat exchangers, and improved pump designs as well as testing of refueling equipment, high temperature instrumentation, and other reactor core designs.« less

PRELIMINARY RESULTS OF THE AGC-4 IRRADIATION IN THE ADVANCED TEST REACTOR AND DESIGN OF AGC-5 (HTR16-18469)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Davenport, Michael; Petti, D. A.

The United States Department of Energy’s Advanced Reactor Technologies (ART) Program will irradiate up to six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments are being irradiated over an approximate eight year period to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Very High Temperature Gasmore » Reactor (VHTR), as well as other future gas reactors. The experiments each consist of a single capsule that contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens are not be subjected to a compressive load during irradiation. The six stacks have differing compressive loads applied to the top half of diametrically opposite pairs of specimen stacks. A seventh specimen stack in the center of the capsule does not have a compressive load. The specimens are being irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There are also samples taken of the sweep gas effluent to measure any oxidation or off-gassing of the specimens that may occur during initial start-up of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. The third experiment, AGC-3, started its irradiation in late November 2012 and completed in the April of 2014. AGC-4 is currently being irradiated in the ATR. This paper will briefly discuss the preliminary irradiation results of the AGC-4 experiment, as well as the design of AGC-5.« less

Alternative nuclear technologies

NASA Astrophysics Data System (ADS)

Schubert, E.

1981-10-01

The lead times required to develop a select group of nuclear fission reactor types and fuel cycles to the point of readiness for full commercialization are compared. Along with lead times, fuel material requirements and comparative costs of producing electric power were estimated. A conservative approach and consistent criteria for all systems were used in estimates of the steps required and the times involved in developing each technology. The impact of the inevitable exhaustion of the low- or reasonable-cost uranium reserves in the United States on the desirability of completing the breeder reactor program, with its favorable long-term result on fission fuel supplies, is discussed. The long times projected to bring the most advanced alternative converter reactor technologies the heavy water reactor and the high-temperature gas-cooled reactor into commercial deployment when compared to the time projected to bring the breeder reactor into equivalent status suggest that the country's best choice is to develop the breeder. The perceived diversion-proliferation problems with the uranium plutonium fuel cycle have workable solutions that can be developed which will enable the use of those materials at substantially reduced levels of diversion risk.

A Comparative Study of Welded ODS Cladding materials for AFCI/GNEP Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Indrajit Charit; Megan Frary; Darryl Butt

2011-03-31

This research project involved working on the pressure resistance welding of oxide dispersion strengthened (ODS) alloys which will have a large role to play in advanced nuclear reactors. The project also demonstrated the research collaboration between four universities and one nation laboratory (Idaho National Laboratory) with participation from an industry for developing for ODS alloys. These alloys contain a high number density of very fine oxide particles that can impart high temperature strength and radiation damage resistance suitable for in-core applications in advanced reactors. The conventional fusion welding techniques tend to produce porosity-laden microstructure in the weld region and leadmore » to the agglomeration and non-uniform distribution of the neededoxide particles. That is why two solid state welding methods - pressure resistance welding (PRW) and friction stir welding (FSW) - were chosen to be evaluated in this project. The proposal is expected to support the development of Advanced Burner Reactors (ABR) under the GNEP program (now incorporated in Fuel Cycle R&D program). The outcomes of the concluded research include training of graduate and undergraduate students and get them interested in nuclear related research.« less

Testing of Sapphire Optical Fiber and Sensors in Intense Radiation Fields When Subjected to Very High Temperatures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blue, Thomas; Windl, Wolfgang

The primary objective of this project was to determine the optical attenuation and signal degradation of sapphire optical fibers & sensors (temperature & strain), in-situ, operating at temperatures up to 1500°C during reactor irradiation through experiments and modeling. The results will determine the feasibility of extending sapphire optical fiber-based instrumentation to extremely high temperature radiation environments. This research will pave the way for future testing of sapphire optical fibers and fiber-based sensors under conditions expected in advanced high temperature reactors.

Characterization of Sodium Thermal Hydraulics with Optical Fiber Temperature Sensors

NASA Astrophysics Data System (ADS)

Weathered, Matthew Thomas

The thermal hydraulic properties of liquid sodium make it an attractive coolant for use in Generation IV reactors. The liquid metal's high thermal conductivity and low Prandtl number increases efficiency in heat transfer at fuel rods and heat exchangers, but can also cause features such as high magnitude temperature oscillations and gradients in the coolant. Currently, there exists a knowledge gap in the mechanisms which may create these features and their effect on mechanical structures in a sodium fast reactor. Two of these mechanisms include thermal striping and thermal stratification. Thermal striping is the oscillating temperature field created by the turbulent mixing of non-isothermal flows. Usually this occurs at the reactor core outlet or in piping junctions and can cause thermal fatigue in mechanical structures. Meanwhile, thermal stratification results from large volumes of non-isothermal sodium in a pool type reactor, usually caused by a loss of coolant flow accident. This stratification creates buoyancy driven flow transients and high temperature gradients which can also lead to thermal fatigue in reactor structures. In order to study these phenomena in sodium, a novel method for the deployment of optical fiber temperature sensors was developed. This method promotes rapid thermal response time and high spatial temperature resolution in the fluid. The thermal striping and stratification behavior in sodium may be experimentally analyzed with these sensors with greater fidelity than ever before. Thermal striping behavior at a junction of non-isothermal sodium was fully characterized with optical fibers. An experimental vessel was hydrodynamically scaled to model thermal stratification in a prototypical sodium reactor pool. Novel auxiliary applications of the optical fiber temperature sensors were developed throughout the course of this work. One such application includes local convection coefficient determination in a vessel with the corollary application of level sensing. Other applications were cross correlation velocimetry to determine bulk sodium flow rate and the characterization of coherent vortical structures in sodium with temperature frequency data. The data harvested, instrumentation developed and techniques refined in this work will help in the design of more robust reactors as well as validate computational models for licensing sodium fast reactors.

Focused technology: Nuclear propulsion

NASA Technical Reports Server (NTRS)

Miller, Thomas J.

1991-01-01

The topics presented are covered in viewgraph form and include: nuclear thermal propulsion (NTP), which challenges (1) high temperature fuel and materials, (2) hot hydrogen environment, (3) test facilities, (4) safety, (5) environmental impact compliance, and (6) concept development, and nuclear electric propulsion (NEP), which challenges (1) long operational lifetime, (2) high temperature reactors, turbines, and radiators, (3) high fuel burn-up reactor fuels, and designs, (4) efficient, high temperature power conditioning, (5) high efficiency, and long life thrusters, (6) safety, (7) environmental impact compliance, and (8) concept development.

Grey water treatment in upflow anaerobic sludge blanket (UASB) reactor at different temperatures.

PubMed

Elmitwalli, Tarek; Otterpohl, Ralf

2011-01-01

The treatment of grey water in two upflow anaerobic sludge blanket (UASB) reactors, operated at different hydraulic retention times (HRTs) and temperatures, was investigated. The first reactor (UASB-A) was operated at ambient temperature (14-25 degrees C) and HRT of 20, 12 and 8 h, while the second reactor (UASB-30) was operated at controlled temperature of 30 degrees C and HRT of 16, 10 and 6 h. The two reactors were fed with grey water from 'Flintenbreite' settlement in Luebeck, Germany. When the grey water was treated in the UASB reactor at 30 degrees C, total chemical oxygen demand (CODt) removal of 52-64% was achieved at HRT between 6 and 16 h, while at lower temperature lower removal (31-41%) was obtained at HRT between 8 and 20 h. Total nitrogen and phosphorous removal in the UASB reactors were limited (22-36 and 10-24%, respectively) at all operational conditions. The results showed that at increasing temperature or decreasing HRT of the reactors, maximum specific methanogenic activity of the sludge in the reactors improved. As the UASB reactor showed a significantly higher COD removal (31-64%) than the septic tank (11-14%) even at low temperature, it is recommended to use UASB reactor instead of septic tank (the most common system) for grey water pre-treatment. Based on the achieved results and due to high peak flow factor, a HRT between 8 and 12 h can be considered the suitable HRT for the UASB reactor treating grey water at temperature 20-30 degrees C, while a HRT of 12-24 h can be applied at temperature lower than 20 degrees C.

An Overview of INEL Fusion Safety R&D Facilities

NASA Astrophysics Data System (ADS)

McCarthy, K. A.; Smolik, G. R.; Anderl, R. A.; Carmack, W. J.; Longhurst, G. R.

1997-06-01

The Fusion Safety Program at the Idaho National Engineering Laboratory has the lead for fusion safety work in the United States. Over the years, we have developed several experimental facilities to provide data for fusion reactor safety analyses. We now have four major experimental facilities that provide data for use in safety assessments. The Steam-Reactivity Measurement System measures hydrogen generation rates and tritium mobilization rates in high-temperature (up to 1200°C) fusion relevant materials exposed to steam. The Volatilization of Activation Product Oxides Reactor Facility provides information on mobilization and transport and chemical reactivity of fusion relevant materials at high temperature (up to 1200°C) in an oxidizing environment (air or steam). The Fusion Aerosol Source Test Facility is a scaled-up version of VAPOR. The ion-implanta-tion/thermal-desorption system is dedicated to research into processes and phenomena associated with the interaction of hydrogen isotopes with fusion materials. In this paper we describe the capabilities of these facilities.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maloy, Stuart Andrew; Pestovich, Kimberly Shay; Anderoglu, Osman

The Fuel Cycle Research and Development program is investigating methods of transmuting minor actinides in various fuel cycle options. To achieve this goal, new fuels and cladding materials must be developed and tested to high burnup levels (e.g. >20%) requiring cladding to withstand very high doses (greater than 200 dpa) while in contact with the coolant and the fuel. To develop and qualify materials to a total fluence greater than 200 dpa requires development of advanced alloys and irradiations in fast reactors to test these alloys. Recent results from testing numerous ferritic/martensitic steels at low temperatures suggest that improvements inmore » low temperature radiation tolerance can be achieved through carefully controlling the nitrogen content in these alloys. Thus, four new heats of HT-9 were produced with controlled nitrogen content: two by Metalwerks and two by Sophisticated Alloys. Initial results on these new alloys are presented including microstructural analysis and hardness testing. Future testing will include irradiation testing with ions and in reactor.« less

High velocity continuous-flow reactor for the production of solar grade silicon

NASA Technical Reports Server (NTRS)

Woerner, L.

1977-01-01

The feasibility of a high volume, high velocity continuous reduction reactor as an economical means of producing solar grade silicon was tested. Bromosilanes and hydrogen were used as the feedstocks for the reactor along with preheated silicon particles which function both as nucleation and deposition sites. A complete reactor system was designed and fabricated. Initial preheating studies have shown the stability of tetrabromosilane to being heated as well as the ability to preheat hydrogen to the desired temperature range. Several test runs were made and some silicon was obtained from runs carried out at temperatures in excess of 1180 K.

NUCLEAR REACTOR

DOEpatents

Grebe, J.J.

1959-07-14

High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Glenn McCreery; Hugh McIlroy

2007-09-01

Advanced computer modeling and simulation tools and protocols will be heavily relied on for a wide variety of system studies, engineering design activities, and other aspects of the Next Generation Nuclear Power (NGNP) Very High Temperature Reactor (VHTR), the DOE Global Nuclear Energy Partnership (GNEP), and light-water reactors. The goal is for all modeling and simulation tools to be demonstrated accurate and reliable through a formal Verification and Validation (V&V) process, especially where such tools are to be used to establish safety margins and support regulatory compliance, or to design a system in a manner that reduces the role ofmore » expensive mockups and prototypes. Recent literature identifies specific experimental principles that must be followed in order to insure that experimental data meet the standards required for a “benchmark” database. Even for well conducted experiments, missing experimental details, such as geometrical definition, data reduction procedures, and manufacturing tolerances have led to poor Benchmark calculations. The INL has a long and deep history of research in thermal hydraulics, especially in the 1960s through 1980s when many programs such as LOFT and Semiscle were devoted to light-water reactor safety research, the EBRII fast reactor was in operation, and a strong geothermal energy program was established. The past can serve as a partial guide for reinvigorating thermal hydraulic research at the laboratory. However, new research programs need to fully incorporate modern experimental methods such as measurement techniques using the latest instrumentation, computerized data reduction, and scaling methodology. The path forward for establishing experimental research for code model validation will require benchmark experiments conducted in suitable facilities located at the INL. This document describes thermal hydraulic facility requirements and candidate buildings and presents examples of suitable validation experiments related to VHTRs, sodium-cooled fast reactors, and light-water reactors. These experiments range from relatively low-cost benchtop experiments for investigating individual phenomena to large electrically-heated integral facilities for investigating reactor accidents and transients.« less

PLASTIC-SASS--A COMPUTER PROGRAM FOR STRESSES AND DEFLECTIONS IN A REACTOR SUBASSEMBLY UNDER THERMAL, HYDRAULIC, AND FUEL EXPANSION LOADS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Friedrich, C.M.

1963-05-01

PLASTlC-SASS, an ALTAC-3 computer program that determines stresses and deflections in a flat-plate, rectangular reactor subassembly is described. Elastic, plastic, and creep properties are used to calculate the results of temperature, pressure, and fuel expansion. Plate deflections increase or decrease local channel thicknesses and thus produce a hydraulic load which is a function of fuel plate deflection. (auth)

Preliminary Safeguards Assessment for the Pebble-Bed Fluoride High-Temperature Reactor (PB-FHR) Concept

DOE Office of Scientific and Technical Information (OSTI.GOV)

Disser, Jay; Arthur, Edward; Lambert, Janine

2016-09-01

This report examines a preliminary design for a pebble bed fluoride salt-cooled high temperature reactor (PB-FHR) concept, assessing it from an international safeguards perspective. Safeguards features are defined, in a preliminary fashion, and suggestions are made for addressing further nuclear materials accountancy needs.

Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

2017-12-01

The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an `early design' variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

Preparation of high temperature gas-cooled reactor fuel element

DOEpatents

Bradley, Ronnie A.; Sease, John D.

1976-01-01

This invention relates to a method for the preparation of high temperature gas-cooled reactor (HTGR) fuel elements wherein uncarbonized fuel rods are inserted in appropriate channels of an HTGR fuel element block and the entire block is inserted in an autoclave for in situ carbonization under high pressure. The method is particularly applicable to remote handling techniques.

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

DOE PAGES

Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.; ...

2016-12-21

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would usemore » tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate disposal. It is recognized that non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.« less

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

DOE Office of Scientific and Technical Information (OSTI.GOV)

Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would usemore » tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate disposal. It is recognized that non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.« less

An experimental test plan for the characterization of molten salt thermochemical properties in heat transport systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pattrick Calderoni

2010-09-01

Molten salts are considered within the Very High Temperature Reactor program as heat transfer media because of their intrinsically favorable thermo-physical properties at temperatures starting from 300 C and extending up to 1200 C. In this context two main applications of molten salt are considered, both involving fluoride-based materials: as primary coolants for a heterogeneous fuel reactor core and as secondary heat transport medium to a helium power cycle for electricity generation or other processing plants, such as hydrogen production. The reference design concept here considered is the Advanced High Temperature Reactor (AHTR), which is a large passively safe reactormore » that uses solid graphite-matrix coated-particle fuel (similar to that used in gas-cooled reactors) and a molten salt primary and secondary coolant with peak temperatures between 700 and 1000 C, depending upon the application. However, the considerations included in this report apply to any high temperature system employing fluoride salts as heat transfer fluid, including intermediate heat exchangers for gas-cooled reactor concepts and homogenous molten salt concepts, and extending also to fast reactors, accelerator-driven systems and fusion energy systems. The purpose of this report is to identify the technical issues related to the thermo-physical and thermo-chemical properties of the molten salts that would require experimental characterization in order to proceed with a credible design of heat transfer systems and their subsequent safety evaluation and licensing. In particular, the report outlines an experimental R&D test plan that would have to be incorporated as part of the design and operation of an engineering scaled facility aimed at validating molten salt heat transfer components, such as Intermediate Heat Exchangers. This report builds on a previous review of thermo-physical properties and thermo-chemical characteristics of candidate molten salt coolants that was generated as part of the same project [1]. However, this work focuses on two materials: the LiF-BeF2 eutectic (67 and 33 mol%, respectively, also known as flibe) as primary coolant and the LiF-NaF-KF eutectic (46.5, 11.5, and 52 mol%, respectively, also known as flinak) as secondary heat transport fluid. At first common issues are identified, involving the preparation and purification of the materials as well as the development of suitable diagnostics. Than issues specific to each material and its application are considered, with focus on the compatibility with structural materials and the extension of the existing properties database.« less

An Experimental Test Facility to Support Development of the Fluoride Salt Cooled High Temperature Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yoder Jr, Graydon L; Aaron, Adam M; Cunningham, Richard Burns

2014-01-01

The need for high-temperature (greater than 600 C) energy exchange and delivery systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The Fluoride Salt-Cooled High-Temperature Reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during themore » development process. The experimental facility described here has been constructed to support the development of the Fluoride Salt Cooled High Temperature Reactor concept. The facility is capable of operating at up to 700 C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system; trace heating system; and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride salt heat transfer inside a heated pebble bed similar to that used for the core of the pebble bed advanced high-temperature reactor. This document describes the details of the loop design, auxiliary systems used to support the facility, the inductive heating system, and facility capabilities.« less

Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-3: High Temperature Gas Cooled Reactor Thermal-Hydraulics.

ERIC Educational Resources Information Center

Reihman, Thomas C.

This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical high temperature gas-cooled reactor (HTGR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module…

Burn Control in Fusion Reactors via Isotopic Fuel Tailoring

NASA Astrophysics Data System (ADS)

Boyer, Mark D.; Schuster, Eugenio

2011-10-01

The control of plasma density and temperature are among the most fundamental problems in fusion reactors and will be critical to the success of burning plasma experiments like ITER. Economic and technological constraints may require future commercial reactors to operate with low temperature, high-density plasma, for which the burn condition may be unstable. An active control system will be essential for stabilizing such operating points. In this work, a volume-averaged transport model for the energy and the densities of deuterium and tritium fuel ions, as well as the alpha particles, is used to synthesize a nonlinear feedback controller for stabilizing the burn condition. The controller makes use of ITER's planned isotopic fueling capability and controls the densities of these ions separately. The ability to modulate the DT fuel mix is exploited in order to reduce the fusion power during thermal excursions without the need for impurity injection. By moving the isotopic mix in the plasma away from the optimal 50:50 mix, the reaction rate is slowed and the alpha-particle heating is reduced to desired levels. Supported by the NSF CAREER award program (ECCS-0645086).

Investigation on the Core Bypass Flow in a Very High Temperature Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hassan, Yassin

2013-10-22

Uncertainties associated with the core bypass flow are some of the key issues that directly influence the coolant mass flow distribution and magnitude, and thus the operational core temperature profiles, in the very high-temperature reactor (VHTR). Designers will attempt to configure the core geometry so the core cooling flow rate magnitude and distribution conform to the design values. The objective of this project is to study the bypass flow both experimentally and computationally. Researchers will develop experimental data using state-of-the-art particle image velocimetry in a small test facility. The team will attempt to obtain full field temperature distribution using racksmore » of thermocouples. The experimental data are intended to benchmark computational fluid dynamics (CFD) codes by providing detailed information. These experimental data are urgently needed for validation of the CFD codes. The following are the project tasks: • Construct a small-scale bench-top experiment to resemble the bypass flow between the graphite blocks, varying parameters to address their impact on bypass flow. Wall roughness of the graphite block walls, spacing between the blocks, and temperature of the blocks are some of the parameters to be tested. • Perform CFD to evaluate pre- and post-test calculations and turbulence models, including sensitivity studies to achieve high accuracy. • Develop the state-of-the art large eddy simulation (LES) using appropriate subgrid modeling. • Develop models to be used in systems thermal hydraulics codes to account and estimate the bypass flows. These computer programs include, among others, RELAP3D, MELCOR, GAMMA, and GAS-NET. Actual core bypass flow rate may vary considerably from the design value. Although the uncertainty of the bypass flow rate is not known, some sources have stated that the bypass flow rates in the Fort St. Vrain reactor were between 8 and 25 percent of the total reactor mass flow rate. If bypass flow rates are on the high side, the quantity of cooling flow through the core may be considerably less than the nominal design value, causing some regions of the core to operate at temperatures in excess of the design values. These effects are postulated to lead to localized hot regions in the core that must be considered when evaluating the VHTR operational and accident scenarios.« less

NETL - Chemical Looping Reactor

ScienceCinema

None

2018-02-14

NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

Report on the FY17 Development of Computer Program for ASME Section III, Division 5, Subsection HB, Subpart B Rules

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swindeman, M. J.; Jetter, R. I.; Sham, T. -L.

One of the objectives of the high temperature design methodology activities is to develop and validate both improvements and the basic features of ASME Boiler and Pressure Vessel Code, Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors, Subsection HB, Subpart B (HBB). The overall scope of this task is to develop a computer program to aid assessment procedures of components under specified loading conditions in accordance with the elevated temperature design requirements for Division 5 Class A components. There are many features and alternative paths of varying complexity in HBB. The initial focus ofmore » this computer program is a basic path through the various options for a single reference material, 316H stainless steel. However, the computer program is being structured for eventual incorporation all of the features and permitted materials of HBB. This report will first provide a description of the overall computer program, particular challenges in developing numerical procedures for the assessment, and an overall approach to computer program development. This is followed by a more comprehensive appendix, which is the draft computer program manual for the program development. The strain limits rules have been implemented in the computer program. The evaluation of creep-fatigue damage will be implemented in future work scope.« less

Qualification of data obtained during a severe accident. Illustrative examples from TMI-2 evaluations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rempe, Joy L.; Knudson, Darrell L.

2015-02-01

The accidents at the Three Mile Island Unit 2 (TMI-2) Pressurized Water Reactor (PWR) and the Daiichi Units 1, 2, and 3 Boiling Water Reactors (BWRs) provide unique opportunities to evaluate instrumentation exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during the TMI-2 accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. Post-TMI-2 instrumentation evaluation programs focused on data required by TMI-2 operators to assess the condition of the reactor and containment and the effect of mitigating actions taken bymore » these operators. Prior efforts also focused on sensors providing data required for subsequent forensic evaluations and accident simulations. This paper provides additional details related to the formal process used to develop a qualified TMI-2 data base and presents data qualification details for three parameters: reactor coolant system (RCS) pressure; containment building temperature; and containment pressure. These selected examples illustrate the types of activities completed in the TMI-2 data qualification process and the importance of such a qualification effort. These details are described to facilitate implementation of a similar process using data and examinations at the Daiichi Units 1, 2, and 3 reactors so that BWR-specific benefits can be obtained.« less

Plasma flow reactor for steady state monitoring of physical and chemical processes at high temperatures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koroglu, Batikan; Mehl, Marco; Armstrong, Michael R.

Here, we present the development of a steady state plasma flow reactor to investigate gas phase physical and chemical processes that occur at high temperature (1000 < T < 5000 K) and atmospheric pressure. The reactor consists of a glass tube that is attached to an inductively coupled argon plasma generator via an adaptor (ring flow injector). We have modeled the system using computational fluid dynamics simulations that are bounded by measured temperatures. In situ line-of-sight optical emission and absorption spectroscopy have been used to determine the structures and concentrations of molecules formed during rapid cooling of reactants after theymore » pass through the plasma. Emission spectroscopy also enables us to determine the temperatures at which these dynamic processes occur. A sample collection probe inserted from the open end of the reactor is used to collect condensed materials and analyze them ex situ using electron microscopy. The preliminary results of two separate investigations involving the condensation of metal oxides and chemical kinetics of high-temperature gas reactions are discussed.« less

Effects of Surface Roughness, Oxidation, and Temperature on the Emissivity of Reactor Pressure Vessel Alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

King, J. L.; Jo, H.; Tirawat, R.

Thermal radiation will be an important mode of heat transfer in future high-temperature reactors and in off-normal high-temperature scenarios in present reactors. In this work, spectral directional emissivities of two reactor pressure vessel (RPV) candidate materials were measured at room temperature after exposure to high-temperature air. In the case of SA508 steel, significant increases in emissivity were observed due to oxidation. In the case of Grade 91 steel, only very small increases were observed under the tested conditions. Effects of roughness were also investigated. To study the effects of roughening, unexposed samples of SA508 and Grade 91 steel were roughenedmore » via one of either grinding or shot-peening before being measured. Significant increases were observed only in samples having roughness exceeding the roughness expected of RPV surfaces. While the emissivity increases for SA508 from oxidation were indeed significant, the measured emissivity coefficients were below that of values commonly used in heat transfer models. Based on the observed experimental data, recommendations for emissivity inputs for heat transfer simulations are provided.« less

Plasma flow reactor for steady state monitoring of physical and chemical processes at high temperatures

DOE PAGES

Koroglu, Batikan; Mehl, Marco; Armstrong, Michael R.; ...

2017-09-11

Here, we present the development of a steady state plasma flow reactor to investigate gas phase physical and chemical processes that occur at high temperature (1000 < T < 5000 K) and atmospheric pressure. The reactor consists of a glass tube that is attached to an inductively coupled argon plasma generator via an adaptor (ring flow injector). We have modeled the system using computational fluid dynamics simulations that are bounded by measured temperatures. In situ line-of-sight optical emission and absorption spectroscopy have been used to determine the structures and concentrations of molecules formed during rapid cooling of reactants after theymore » pass through the plasma. Emission spectroscopy also enables us to determine the temperatures at which these dynamic processes occur. A sample collection probe inserted from the open end of the reactor is used to collect condensed materials and analyze them ex situ using electron microscopy. The preliminary results of two separate investigations involving the condensation of metal oxides and chemical kinetics of high-temperature gas reactions are discussed.« less

Exploratory development of a glass ceramic automobile thermal reactor. [anti-pollution devices

NASA Technical Reports Server (NTRS)

Gould, R. E.; Petticrew, R. W.

1973-01-01

This report summarizes the design, fabrication and test results obtained for glass-ceramic (CER-VIT) automotive thermal reactors. Several reactor designs were evaluated using both engine-dynamometer and vehicle road tests. A maximum reactor life of about 330 hours was achieved in engine-dynamometer tests with peak gas temperatures of about 1065 C (1950 F). Reactor failures were mechanically induced. No evidence of chemical degradation was observed. It was concluded that to be useful for longer times, the CER-VIT parts would require a mounting system that was an improvement over those tested in this program. A reactor employing such a system was designed and fabricated.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mitchell K Meyer

Blister–threshold testing of fuel plates is a standard method through which the safety margin for operation of plate-type in research and test reactors is assessed. The blister-threshold temperature is indicative of the ability of fuel to operate at high temperatures for short periods of time (transient conditions) without failure. This method of testing was applied to the newly developed U-Mo monolithic fuel system. Blister annealing studies on the U-Mo monolithic fuel plates began in 2007, with the Reduced Enrichment for Research and Test Reactors (RERTR)-6 experiment, and they have continued as the U-Mo fuel system has evolved through the researchmore » and development process. Blister anneal threshold temperatures from early irradiation experiments (RERTR-6 through RERTR-10) ranged from 400 to 500°C. These temperatures were projected to be acceptable for NRC-licensed research reactors and the high-power Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR) based on current safety-analysis reports (SARs). Initial blister testing results from the RERTR-12 experiment capsules X1 and X2 showed a decrease in the blister-threshold temperatures. Blister threshold temperatures from this experiment ranged from 300 to 400°C. Selected plates from the AFIP-4 experiment, which was fabricated using a process similar to that used to fabricate the RERTR-12 experiment, also underwent blister testing to determine whether results would be similar. The measured blister-threshold temperatures from the AFIP-4 plates fell within the same blister-threshold temperature range measured in the RERTR-12 plates. Investigation of the cause of this decrease in bister threshold temperature is being conducted under the guidance of Idaho National Laboratory PLN-4155, “Analysis of Low Blister Threshold Temperatures in the RERTR-12 and AFIP-4 Experiments,” and is driven by hypotheses. The main focus of the investigation is in the following areas: 1. Fabrication variables 2. Pre-irradiation characterization 3. Irradiation conditions 4. Post-irradiation examination 5. Additional blister testing 6. Mechanical modeling This report documents the preliminary results of this investigation. Several hypotheses can be dismissed as a result of this investigation. Two primary categories of causes remain. The most prominent theory, supported by the data, is that low blister-threshold temperature is the result of mechanical energy imparted on the samples during the fabrication process (hot and cold rolling) without adequate post processing (annealing). The mechanisms are not clearly understood and require further investigation, but can be divided into two categories: • Residual Stress • Undesirable interaction boundary and/or U-Mo microstructure change A secondary theory that cannot be dismissed with the information that is currently available is that a change in the test conditions has resulted in a statistically significant downward shift of measured blister temperature. This report outlines the results of the forensic investigations conducted to date. The data and conclusions presented in this report are preliminary. Definitive cause and effect relationships will be established by future experimental programs.« less

Advanced high temperature thermoelectrics for space power

NASA Technical Reports Server (NTRS)

Lockwood, A.; Ewell, R.; Wood, C.

1981-01-01

Preliminary results from a spacecraft system study show that an optimum hot junction temperature is in the range of 1500 K for advanced nuclear reactor technology combined with thermoelectric conversion. Advanced silicon germanium thermoelectric conversion is feasible if hot junction temperatures can be raised roughly 100 C or if gallium phosphide can be used to improve the figure of merit, but the performance is marginal. Two new classes of refractory materials, rare earth sulfides and boron-carbon alloys, are being investigated to improve the specific weight of the generator system. Preliminary data on the sulfides have shown very high figures of merit over short temperature ranges. Both n- and p-type doping have been obtained. Pure boron-carbide may extrapolate to high figure of merit at temperatures well above 1500 K but not lower temperature; n-type conduction has been reported by others, but not yet observed in the JPL program. Inadvertant impurity doping may explain the divergence of results reported.

Supercritical Brayton Cycle Nuclear Power System Concepts

NASA Astrophysics Data System (ADS)

Wright, Steven A.

2007-01-01

Both the NASA and DOE have programs that are investigating advanced power conversion cycles for planetary surface power on the moon or Mars, and for next generation nuclear power plants on earth. The gas Brayton cycle offers many practical solutions for space nuclear power systems and was selected as the nuclear power system of choice for the NASA Prometheus project. An alternative Brayton cycle that offers high efficiency at a lower reactor coolant outlet temperature is the supercritical Brayton cycle (SCBC). The supercritical cycle is a true Brayton cycle because it uses a single phase fluid with a compressor inlet temperature that is just above the critical point of the fluid. This paper describes the use of a supercritical Brayton cycle that achieves a cycle efficiency of 26.6% with a peak coolant temperature of 750 K and for a compressor inlet temperature of 390 K. The working fluid uses a clear odorless, nontoxic refrigerant C318 perflurocarbon (C4F8) that always operates in the gas phase. This coolant was selected because it has a critical temperature and pressure of 388.38 K and 2.777 MPa. The relatively high critical temperature allows for efficient thermal radiation that keeps the radiator mass small. The SCBC achieves high efficiency because the loop design takes advantage of the non-ideal nature of the coolant equation of state just above the critical point. The lower coolant temperature means that metal fuels, uranium oxide fuels, and uranium zirconium hydride fuels with stainless steel, ferretic steel, or superalloy cladding can be used with little mass penalty or reduction in cycle efficiency. The reactor can use liquid-metal coolants and no high temperature heat exchangers need to be developed. Indirect gas cooling or perhaps even direct gas cooling can be used if the C4F8 coolant is found to be sufficiently radiation tolerant. Other fluids can also be used in the supercritical Brayton cycle including Propane (C3H8, Tcritical = 369 K) and Hexane (C6H14, Tcritical = 506.1 K) provided they have adequate chemical compatibility and stability. Overall the use of supercritical Brayton cycles may offer ``break through'' operating capabilities for space nuclear power plants because high efficiencies can be achieved a very low reactor operating temperatures which in turn allows for the use of available fuels, cladding, and structural materials.

THE ARMOUR DUST FUELED REACTOR (ADFR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krucoff, D.

1958-01-01

The A-DFR is based on the use of a fissionable dust carried in a gas. This fuel ferm offers promise of a major economic advance through the use of 2,000 to 3,000 F operating temperatures and a low cost fuel cycle. The development program is described that was initiated to investigate experimentally the proposed fuel and study analytically other reactor characteristics. A brief review of the reactor concept is presented. (W.D.M.)

DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-[TEMPERATURE GAS-COOLED TEST REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James; Bayless, Paul; Strydom, Gerhard

A point design for a graphite-moderated, high-temperature, gas-cooled test reactor (HTG TR) has been developed by Idaho National Laboratory (INL) as part of a United States (U.S.) Department of Energy (DOE) initiative to explore and potentially expand the existing U.S. test reactor capabilities. This paper provides a summary of the design and its main attributes. The 200 MW HTG TR is a thermal-neutron spectrum reactor composed of hexagonal prismatic fuel and graphite reflector blocks. Twelve fuel columns (96 fuel blocks total and 6.34 m active core height) are arranged in two hexagonal rings to form a relatively compact, high-power density,more » annular core sandwiched between inner, outer, top, and bottom graphite reflectors. The HTG-TR is designed to operate at 7 MPa with a coolant inlet/outlet temperature of 325°C/650°C, and utilizes TRISO particle fuel from the DOE AGR Program with 425 ?m uranium oxycarbide (UCO) kernels and an enrichment of 15.5 wt% 235U. The primary mission of the HTG TR is material irradiation and therefore the core has been specifically designed and optimized to provide the highest possible thermal and fast neutron fluxes. The highest thermal neutron flux (3.90E+14 n/cm2s) occurs in the outer reflector, and the maximum fast flux levels (1.17E+14 n/cm2s) are produced in the central reflector column where most of the graphite has been removed. Due to high core temperatures under accident conditions, all the irradiation test facilities have been located in the inner and outer reflectors where fast flux levels decline. The core features a large number of irradiation positions with large test volumes and long test lengths, ideal for thermal neutron irradiation of large test articles. The total available test volume is more than 1100 liters. Up to four test loop facilities can be accommodated with pressure tube boundaries to isolate test articles and test fluids (e.g., liquid metal, liquid salt, light water) from the helium primary coolant system.« less

Design and Application of a High-Temperature Linear Ion Trap Reactor

NASA Astrophysics Data System (ADS)

Jiang, Li-Xue; Liu, Qing-Yu; Li, Xiao-Na; He, Sheng-Gui

2018-01-01

A high-temperature linear ion trap reactor with hexapole design was homemade to study ion-molecule reactions at variable temperatures. The highest temperature for the trapped ions is up to 773 K, which is much higher than those in available reports. The reaction between V2O6 - cluster anions and CO at different temperatures was investigated to evaluate the performance of this reactor. The apparent activation energy was determined to be 0.10 ± 0.02 eV, which is consistent with the barrier of 0.12 eV calculated by density functional theory. This indicates that the current experimental apparatus is prospective to study ion-molecule reactions at variable temperatures, and more kinetic details can be obtained to have a better understanding of chemical reactions that have overall barriers. [Figure not available: see fulltext.

Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peterson, Per; Greenspan, Ehud

2015-02-09

This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designsmore » are used, the power density of salt- cooled reactors is limited to 10 MW/m 3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m 3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses novel digital x-ray tomography methods to track both the translational and rotational motion of spherical pebbles, which provides unique experimental results that can be used to validate discrete element method (DEM) simulations of pebble motion. The validation effort supported by the X-PREX facility provides a means to build confidence in analysis of pebble bed configuration and residence time distributions that impact the neutronics, thermal hydraulics, and safety analysis of pebble bed reactor cores. Experimental and DEM simulation results are reported for silo drainage, a classical problem in the granular flow literature, at several hopper angles. These studies include conventional converging and novel diverging geometries that provide additional flexibility in the design of pebble bed reactor cores. Excellent agreement is found between the X-PREX experimental and DEM simulation results. This report also includes results for additional studies relevant to the design and analysis of pebble bed reactor cores including the study of forces on shut down blades inserted directly into a packed bed and pebble flow in a cylindrical hopper that is representative of a small test reactor.« less

Alcohol synthesis in a high-temperature slurry reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Roberts, G.W.; Marquez, M.A.; McCutchen, M.S.

1995-12-31

The overall objective of this contract is to develop improved process and catalyst technology for producing higher alcohols from synthesis gas or its derivatives. Recent research has been focused on developing a slurry reactor that can operate at temperatures up to about 400{degrees}C and on evaluating the so-called {open_quotes}high pressure{close_quotes} methanol synthesis catalyst using this reactor. A laboratory stirred autoclave reactor has been developed that is capable of operating at temperatures up to 400{degrees}C and pressures of at least 170 atm. The overhead system on the reactor is designed so that the temperature of the gas leaving the system canmore » be closely controlled. An external liquid-level detector is installed on the gas/liquid separator and a pump is used to return condensed slurry liquid from the separator to the reactor. In order to ensure that gas/liquid mass transfer does not influence the observed reaction rate, it was necessary to feed the synthesis gas below the level of the agitator. The performance of a commercial {open_quotes}high pressure {close_quotes} methanol synthesis catalyst, the so-called {open_quotes}zinc chromite{close_quotes} catalyst, has been characterized over a range of temperature from 275 to 400{degrees}C, a range of pressure from 70 to 170 atm., a range of H{sub 2}/CO ratios from 0.5 to 2.0 and a range of space velocities from 2500 to 10,000 sL/kg.(catalyst),hr. Towards the lower end of the temperature range, methanol was the only significant product.« less

SPERT I DESTRUCTIVE TEST PROGRAM SAFETY ANALYSIS REPORT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spano, A.H.; Miller, R.W.

1962-06-15

The water-moderated core used for destructive experiments is mounted in the Spent I open-type reactor vessel, which has no provision for pressurization or forced coolant flow. The core is an array of highly enriched aluminum clad, plate-type fuel assemblies, using four bladetype, gang-operated control rods. Reactor transients are initiated at ambient temperature by step-insentions of reactivity, using a control rod which can be quickly ejected from the core. Following an initial series of static measurements to determine the basic- reactor properties of the test core, a series of nondestructive, self-limiting power excursion tests was performed, which covered a reactor periodmore » range down to the point where minor fuel plate damage first occurred -approximately for a 10- msec period test. These tests provided power, temperature, and pressure data. Additional kinetic teste in the period region between 10 and 5 msec were completed to explore the region of limited core damage. Fuel plate damage results included plate distortion, cladding cracking, and fuel melting. These exploratory tests were valuable in revealing unexpected changes in the dependence of pressure, temperature, burst energy, and burst shape parameters on reactor period, although the dependence of peak power on reactor period was not significantly changed. An evaluation of hazards involved in conducting the 2- msec test, based on pessimistic assumptions regarding fission product release and weather conditions, indicates that with the procedural controls normally exercised in the conduct of any transient test at Spent and the special controls to be in effect during the destructive test series, no significant hazard to personnel or to the general public will be obtained. All nuclear operation is conducted remotely approximately 1/2 mile from the reactor building. Discussion is also given of the supervision and control of personnel during and after each destructive test, and of the plans for re-entry, cleanup, and restoration of the facility. (auth)« less

Influence of temperature on the single-stage ATAD process predicted by a thermal equilibrium model.

PubMed

Cheng, Jiehong; Zhu, Jun; Kong, Feng; Zhang, Chunyong

2015-06-01

Autothermal thermophilic aerobic digestion (ATAD) is a promising biological process that will produce an effluent satisfying the Class A requirements on pathogen control and land application. The thermophilic temperature in an ATAD reactor is one of the critical factors that can affect the satisfactory operation of the ATAD process. This paper established a thermal equilibrium model to predict the effect of variables on the auto-rising temperature in an ATAD system. The reactors with volumes smaller than 10 m(3) could not achieve temperatures higher than 45 °C under ambient temperature of -5 °C. The results showed that for small reactors, the reactor volume played a key role in promoting auto-rising temperature in the winter. Thermophilic temperature achieved in small ATAD reactors did not entirely depend on the heat release from biological activities during degrading organic matters in sludges, but was related to the ambient temperature. The ratios of surface area-to-effective volume less than 2.0 had less impact on the auto-rising temperature of an ATAD reactor. The influence of ambient temperature on the auto-rising reactor temperature decreased with increasing reactor volumes. High oxygen transfer efficiency had a significant influence on the internal temperature rise in an ATAD system, indicating that improving the oxygen transfer efficiency of aeration devices was a key factor to achieve a higher removal rate of volatile solids (VS) during the ATAD process operation. Compared with aeration using cold air, hot air demonstrated a significant effect on maintaining the internal temperature (usually 4-5 °C higher). Copyright © 2015 Elsevier Ltd. All rights reserved.

Assessment of Silicon Carbide Composites for Advanced Salt-Cooled Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Katoh, Yutai; Wilson, Dane F; Forsberg, Charles W

2007-09-01

The Advanced High-Temperature Reactor (AHTR) is a new reactor concept that uses a liquid fluoride salt coolant and a solid high-temperature fuel. Several alternative fuel types are being considered for this reactor. One set of fuel options is the use of pin-type fuel assemblies with silicon carbide (SiC) cladding. This report provides (1) an initial viability assessment of using SiC as fuel cladding and other in-core components of the AHTR, (2) the current status of SiC technology, and (3) recommendations on the path forward. Based on the analysis of requirements, continuous SiC fiber-reinforced, chemically vapor-infiltrated SiC matrix (CVI SiC/SiC) compositesmore » are recommended as the primary option for further study on AHTR fuel cladding among various industrially available forms of SiC. Critical feasibility issues for the SiC-based AHTR fuel cladding are identified to be (1) corrosion of SiC in the candidate liquid salts, (2) high dose neutron radiation effects, (3) static fatigue failure of SiC/SiC, (4) long-term radiation effects including irradiation creep and radiation-enhanced static fatigue, and (5) fabrication technology of hermetic wall and sealing end caps. Considering the results of the issues analysis and the prospects of ongoing SiC research and development in other nuclear programs, recommendations on the path forward is provided in the order or priority as: (1) thermodynamic analysis and experimental examination of SiC corrosion in the candidate liquid salts, (2) assessment of long-term mechanical integrity issues using prototypical component sections, and (3) assessment of high dose radiation effects relevant to the anticipated operating condition.« less

Gas-cooled reactor programs. High-temperature gas-cooled reactor technology development program. Annual progress report, December 31, 1983

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.

1984-06-01

ORNL continues to make significant contributions to the national program. In the HTR fuels area, we are providing detailed statistical information on the fission product retention performance of irradiated fuel. Our studies are also providing basic data on the mechanical, physical, and chemical behavior of HTR materials, including metals, ceramics, graphite, and concrete. The ORNL has an important role in the development of improved HTR graphites and in the specification of criteria that need to be met by commercial products. We are also developing improved reactor physics design methods. Our work in component development and testing centers in the Componentmore » Flow Test Loop (CFTL), which is being used to evaluate the performance of the HTR core support structure. Other work includes experimental evaluation of the shielding effectiveness of the lower portions of an HTR core. This evaluation is being performed at the ORNL Tower Shielding Facility. Researchers at ORNL are developing welding techniques for attaching steam generator tubing to the tubesheets and are testing ceramic pads on which the core posts rest. They are also performing extensive testing of aggregate materials obtained from potential HTR site areas for possible use in prestressed concrete reactor vessels. During the past year we continued to serve as a peer reviewer of small modular reactor designs being developed by GA and GE with balance-of-plant layouts being developed by Bechtel Group, Inc. We have also evaluated the national need for developing HTRs with emphasis on the longer term applications of the HTRs to fossil conversion processes.« less

Core Dynamics Analysis for Reactivity Insertion and Loss of Coolant Flow Tests Using the High Temperature Engineering Test Reactor

NASA Astrophysics Data System (ADS)

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-temperature Gas-cooled Reactors (HTGRs). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named-ACCORD-, was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We have modified this code to use a model with four parallel channels and twenty temperature coefficients. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results. Moreover, the effect of the model is formulated quantitatively with our proposed equation. Finally, the pre-analytical result of the loss of coolant flow test by tripping all gas circulators is also discussed.

Process for making silicon from halosilanes and halosilicons

NASA Technical Reports Server (NTRS)

Levin, Harry (Inventor)

1988-01-01

A reactor apparatus (10) adapted for continuously producing molten, solar grade purity elemental silicon by thermal reaction of a suitable precursor gas, such as silane (SiH.sub.4), is disclosed. The reactor apparatus (10) includes an elongated reactor body (32) having graphite or carbon walls which are heated to a temperature exceeding the melting temperature of silicon. The precursor gas enters the reactor body (32) through an efficiently cooled inlet tube assembly (22) and a relatively thin carbon or graphite septum (44). The septum (44), being in contact on one side with the cooled inlet (22) and the heated interior of the reactor (32) on the other side, provides a sharp temperature gradient for the precursor gas entering the reactor (32) and renders the operation of the inlet tube assembly (22) substantially free of clogging. The precursor gas flows in the reactor (32) in a substantially smooth, substantially axial manner. Liquid silicon formed in the initial stages of the thermal reaction reacts with the graphite or carbon walls to provide a silicon carbide coating on the walls. The silicon carbide coated reactor is highly adapted for prolonged use for production of highly pure solar grade silicon. Liquid silicon (20) produced in the reactor apparatus (10) may be used directly in a Czochralski or other crystal shaping equipment.

Process for making silicon

NASA Technical Reports Server (NTRS)

Levin, Harry (Inventor)

1987-01-01

A reactor apparatus (10) adapted for continuously producing molten, solar grade purity elemental silicon by thermal reaction of a suitable precursor gas, such as silane (SiH.sub.4), is disclosed. The reactor apparatus (10) includes an elongated reactor body (32) having graphite or carbon walls which are heated to a temperature exceeding the melting temperature of silicon. The precursor gas enters the reactor body (32) through an efficiently cooled inlet tube assembly (22) and a relatively thin carbon or graphite septum (44). The septum (44), being in contact on one side with the cooled inlet (22) and the heated interior of the reactor (32) on the other side, provides a sharp temperature gradient for the precursor gas entering the reactor (32) and renders the operation of the inlet tube assembly (22) substantially free of clogging. The precursor gas flows in the reactor (32) in a substantially smooth, substantially axial manner. Liquid silicon formed in the initial stages of the thermal reaction reacts with the graphite or carbon walls to provide a silicon carbide coating on the walls. The silicon carbide coated reactor is highly adapted for prolonged use for production of highly pure solar grade silicon. Liquid silicon (20) produced in the reactor apparatus (10) may be used directly in a Czochralski or other crystal shaping equipment.

Biomethanation under psychrophilic conditions.

PubMed

Dhaked, Ram Kumar; Singh, Padma; Singh, Lokendra

2010-12-01

The biomethanation of organic matter represents a long-standing, well-established technology. Although at mesophilic and thermophilic temperatures the process is well understood, current knowledge on psychrophilic biomethanation is somewhat scarce. Methanogenesis is particularly sensitive to temperature, which not only affects the activity and structure of the microbial community, but also results in a change in the degradation pathway of organic matter. There is evidence of psychrophilic methanogenesis in natural environments, and a number of methanogenic archaea have been isolated with optimum growth temperatures of 15-25 °C. At psychrophilic temperatures, large amounts of heat are needed to operate reactors, thus resulting in a marginal or negative overall energy yield. Biomethanation at ambient temperature can alleviate this requirement, but for stable biogas production, a microbial consortium adapted to low temperatures or a psychrophilic consortium is required. Single-step or two-step high rate anaerobic reactors [expanded granular sludge bed (EGSB) and up flow anaerobic sludge bed (UASB)] have been used for the treatment of low strength wastewater. Simplified versions of these reactors, such as anaerobic sequencing batch reactors (ASBR) and anaerobic migrating blanket reactor (AMBR) have also been developed with the aim of reducing volume and cost. This technology has been further simplified and extended for the disposal of night soil in high altitude, low temperature areas of the Himalayas, where the hilly terrain, non-availability of conventional energy, harsh climate and space constraints limit the application of complicated reactors. Biomethanation at psychrophilic temperatures and the contribution made to night-soil degradation in the Himalayas are reviewed in this article. Copyright © 2010 Elsevier Ltd. All rights reserved.

Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

2017-12-01

A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

Post impact behavior of mobile reactor core containment systems

NASA Technical Reports Server (NTRS)

Puthoff, R. L.; Parker, W. G.; Vanbibber, L. E.

1972-01-01

The reactor core containment vessel temperatures after impact, and the design variables that affect the post impact survival of the system are analyzed. The heat transfer analysis includes conduction, radiation, and convection in addition to the core material heats of fusion and vaporization under partially burial conditions. Also, included is the fact that fission products vaporize and transport radially outward and condense outward and condense on cooler surfaces, resulting in a moving heat source. A computer program entitled Executive Subroutines for Afterheat Temperature Analysis (ESATA) was written to consider this complex heat transfer analysis. Seven cases were calculated of a reactor power system capable of delivering up to 300 MW of thermal power to a nuclear airplane.

Digital computer program for nuclear reactor design water properties (LWBR Development Program)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lynn, L.L.

1967-07-01

An edit program MO899 for the tabulation of thermodynamic and transport properties of liquid and vapor water, frequently used in design calculations for pressurized water nuclear reactors, is described. The data tabulated are obtained from a FORTRAN IV subroutine named HOH. Values of enthalpy, specific volume, viscosity, and thermal conductivity are given for the following ranges: pressure from one bar (14.5 psia) to 175 bars (2538 psia) and temperature from as much as 320 deg C (608 deg F) below saturation up to 500 deg C (932 deg F) above saturation. (NSA 21: 38472)

Chemical compatibility issues associated with use of SiC/SiC in advanced reactor concepts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wilson, Dane F.

2015-09-01

Silicon carbide/silicon carbide (SiC/SiC) composites are of interest for components that will experience high radiation fields in the High Temperature Gas Cooled Reactor (HTGR), the Very High Temperature Reactor (VHTR), the Sodium Fast Reactor (SFR), or the Fluoride-cooled High-temperature Reactor (FHR). In all of the reactor systems considered, reactions of SiC/SiC composites with the constituents of the coolant determine suitability of materials of construction. The material of interest is nuclear grade SiC/SiC composites, which consist of a SiC matrix [high-purity, chemical vapor deposition (CVD) SiC or liquid phase-sintered SiC that is crystalline beta-phase SiC containing small amounts of alumina-yttria impurity],more » a pyrolytic carbon interphase, and somewhat impure yet crystalline beta-phase SiC fibers. The interphase and fiber components may or may not be exposed, at least initially, to the reactor coolant. The chemical compatibility of SiC/SiC composites in the three reactor environments is highly dependent on thermodynamic stability with the pure coolant, and on reactions with impurities present in the environment including any ingress of oxygen and moisture. In general, there is a dearth of information on the performance of SiC in these environments. While there is little to no excess Si present in the new SiC/SiC composites, the reaction of Si with O 2 cannot be ignored, especially for the FHR, in which environment the product, SiO 2, can be readily removed by the fluoride salt. In all systems, reaction of the carbon interphase layer with oxygen is possible especially under abnormal conditions such as loss of coolant (resulting in increased temperature), and air and/ or steam ingress. A global outline of an approach to resolving SiC/SiC chemical compatibility concerns with the environments of the three reactors is presented along with ideas to quickly determine the baseline compatibility performance of SiC/SiC.« less

Status of Post Irradiation Examination of FCAB and FCAT Irradiation Capsules

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G.; Yamamoto, Yukinori; Howard, Richard H.

A series of irradiation programs are ongoing to address the need for determining the radiation tolerance of FeCrAl alloys. These irradiation programs, deemed the FCAT and FCAB irradiation programs, use the High Flux Isotope Reactor (HFIR) to irradiate second generation wrought FeCrAl alloys and early-generation powder-metallurgy (PM) oxide dispersion-strengthened (ODS) FeCrAl alloys. Irradiations have been or are being performed at temperatures of 200°C, 330°C, and 550°C from doses of 1.8 dpa up to 16 dpa. Preliminary post-irradiation examination (PIE) on low dose (<2 dpa) irradiation capsules of tensile specimens has been performed. Analysis of co-irradiated SiC thermometry have shown reasonablemore » matching between the nominal irradiation temperatures and the target irradiation temperatures. Room temperature tensile tests have shown typical radiation-induced hardening and embrittlement at irradiations of 200°C and 330°C, but a propensity for softening when irradiated to 550°C for the wrought alloys. The PM-ODS FeCrAl specimens showed less hardening compared to the wrought alloys. Future PIE includes high temperature tensile tests on the low dose irradiation capsules as well as the determination of reference fracture toughness transition temperature, T o, in alloys irradiated to 7 dpa and higher.« less

Mechanical Performance of Ferritic Martensitic Steels for High Dose Applications in Advanced Nuclear Reactors

NASA Astrophysics Data System (ADS)

Anderoglu, Osman; Byun, Thak Sang; Toloczko, Mychailo; Maloy, Stuart A.

2013-01-01

Ferritic/martensitic (F/M) steels are considered for core applications and pressure vessels in Generation IV reactors as well as first walls and blankets for fusion reactors. There are significant scientific data on testing and industrial experience in making this class of alloys worldwide. This experience makes F/M steels an attractive candidate. In this article, tensile behavior, fracture toughness and impact property, and creep behavior of the F/M steels under neutron irradiations to high doses with a focus on high Cr content (8 to 12) are reviewed. Tensile properties are very sensitive to irradiation temperature. Increase in yield and tensile strength (hardening) is accompanied with a loss of ductility and starts at very low doses under irradiation. The degradation of mechanical properties is most pronounced at <0.3 T M ( T M is melting temperature) and up to 10 dpa (displacement per atom). Ferritic/martensitic steels exhibit a high fracture toughness after irradiation at all temperatures even below 673 K (400 °C), except when tested at room temperature after irradiations below 673 K (400 °C), which shows a significant reduction in fracture toughness. Creep studies showed that for the range of expected stresses in a reactor environment, the stress exponent is expected to be approximately one and the steady state creep rate in the absence of swelling is usually better than austenitic stainless steels both in terms of the creep rate and the temperature sensitivity of creep. In short, F/M steels show excellent promise for high dose applications in nuclear reactors.

Transient modeling of the thermohydraulic behavior of high temperature heat pipes for space reactor applications

NASA Technical Reports Server (NTRS)

Hall, Michael L.; Doster, Joseph M.

1986-01-01

Many proposed space reactor designs employ heat pipes as a means of conveying heat. Previous researchers have been concerned with steady state operation, but the transient operation is of interest in space reactor applications due to the necessity of remote startup and shutdown. A model is being developed to study the dynamic behavior of high temperature heat pipes during startup, shutdown and normal operation under space environments. Model development and preliminary results for a hypothetical design of the system are presented.

ITER structural design criteria and their extension to advanced reactor blankets*1

NASA Astrophysics Data System (ADS)

Majumdar, S.; Kalinin, G.

2000-12-01

Applications of the recent ITER structural design criteria (ISDC) are illustrated by two components. First, the low-temperature-design rules are applied to copper alloys that are particularly prone to irradiation embrittlement at relatively low fluences at certain temperatures. Allowable stresses are derived and the impact of the embrittlement on allowable surface heat flux of a simple first-wall/limiter design is demonstrated. Next, the high-temperature-design rules of ISDC are applied to evaporation of lithium and vapor extraction (EVOLVE), a blanket design concept currently being investigated under the US Advanced Power Extraction (APEX) program. A single tungsten first-wall tube is considered for thermal and stress analyses by finite-element method.

System for thermochemical hydrogen production

DOEpatents

Werner, R.W.; Galloway, T.R.; Krikorian, O.H.

1981-05-22

Method and apparatus are described for joule boosting a SO/sub 3/ decomposer using electrical instead of thermal energy to heat the reactants of the high temperature SO/sub 3/ decomposition step of a thermochemical hydrogen production process driven by a tandem mirror reactor. Joule boosting the decomposer to a sufficiently high temperature from a lower temperature heat source eliminates the need for expensive catalysts and reduces the temperature and consequent materials requirements for the reactor blanket. A particular decomposer design utilizes electrically heated silicon carbide rods, at a temperature of 1250/sup 0/K, to decompose a cross flow of SO/sub 3/ gas.

METHOD AND APPARATUS FOR PRODUCING POWER

DOEpatents

Wollan, E.O.

1961-06-27

A neutronic reactor comprising two discrete zones; namely, an inner zone containing fissionable material and an outer zone containing fertile material is described. The inner zone is operated at a low temperature and is cooled by pressurized water. The outer zone is operated at a substantially higher temperature and is cooled by steam flashed from the inner zone. The reactor is particularly advantageous in that it produces high temperature steam; yet the materials of construction in the core (inner zone) are not restricted to materials capable of withstanding high temperature operation.

Method and apparatus for producing synthesis gas

DOEpatents

Hemmings, John William; Bonnell, Leo; Robinson, Earl T.

2010-03-03

A method and apparatus for reacting a hydrocarbon containing feed stream by steam methane reforming reactions to form a synthesis gas. The hydrocarbon containing feed is reacted within a reactor having stages in which the final stage from which a synthesis gas is discharged incorporates expensive high temperature materials such as oxide dispersed strengthened metals while upstream stages operate at a lower temperature allowing the use of more conventional high temperature alloys. Each of the reactor stages incorporate reactor elements having one or more separation zones to separate oxygen from an oxygen containing feed to support combustion of a fuel within adjacent combustion zones, thereby to generate heat to support the endothermic steam methane reforming reactions.

Correlative Microscopy of Neutron-Irradiated Materials

DOE PAGES

Briggs, Samuel A.; Sridharan, Kumar; Field, Kevin G.

2016-12-31

A nuclear reactor core is a highly demanding environment that presents several unique challenges for materials performance. Materials in modern light water reactor (LWR) cores must survive several decades in high-temperature (300-350°C) aqueous corrosion conditions while being subject to large amounts of high-energy neutron irradiation. Next-generation reactor designs seek to use more corrosive coolants (e.g., molten salts) and even greater temperatures and neutron doses. The high amounts of disorder and unique crystallographic defects and microchemical segregation effects induced by radiation inevitably lead to property degradation of materials. Thus, maintaining structural integrity and safety margins over the course of the reactor'smore » service life thus necessitates the ability to understand and predict these degradation phenomena in order to develop new, radiation-tolerant materials that can maintain the required performance in these extreme conditions.« less

Assessment of quasi-linear effect of RF power spectrum for enabling lower hybrid current drive in reactor plasmas

NASA Astrophysics Data System (ADS)

Cesario, Roberto; Cardinali, Alessandro; Castaldo, Carmine; Amicucci, Luca; Ceccuzzi, Silvio; Galli, Alessandro; Napoli, Francesco; Panaccione, Luigi; Santini, Franco; Schettini, Giuseppe; Tuccillo, Angelo Antonio

2017-10-01

The main research on the energy from thermonuclear fusion uses deuterium plasmas magnetically trapped in toroidal devices. To suppress the turbulent eddies that impair thermal insulation and pressure tight of the plasma, current drive (CD) is necessary, but tools envisaged so far are unable accomplishing this task while efficiently and flexibly matching the natural current profiles self-generated at large radii of the plasma column [1-5]. The lower hybrid current drive (LHCD) [6] can satisfy this important need of a reactor [1], but the LHCD system has been unexpectedly mothballed on JET. The problematic extrapolation of the LHCD tool at reactor graded high values of, respectively, density and temperatures of plasma has been now solved. The high density problem is solved by the FTU (Frascati Tokamak Upgrade) method [7], and solution of the high temperature one is presented here. Model results based on quasi-linear (QL) theory evidence the capability, w.r.t linear theory, of suitable operating parameters of reducing the wave damping in hot reactor plasmas. Namely, using higher RF power densities [8], or a narrower antenna power spectrum in refractive index [9,10], the obstacle for LHCD represented by too high temperature of reactor plasmas should be overcome. The former method cannot be used for routinely, safe antenna operations, Thus, only the latter key is really exploitable in a reactor. The proposed solutions are ultimately necessary for viability of an economic reactor.

Reactor for tracking catalyst nanoparticles in liquid at high temperature under a high-pressure gas phase with X-ray absorption spectroscopy.

PubMed

Nguyen, Luan; Tao, Franklin Feng

2018-02-01

Structure of catalyst nanoparticles dispersed in liquid phase at high temperature under gas phase of reactant(s) at higher pressure (≥5 bars) is important for fundamental understanding of catalytic reactions performed on these catalyst nanoparticles. Most structural characterizations of a catalyst performing catalysis in liquid at high temperature under gas phase at high pressure were performed in an ex situ condition in terms of characterizations before or after catalysis since, from technical point of view, access to the catalyst nanoparticles during catalysis in liquid phase at high temperature under high pressure reactant gas is challenging. Here we designed a reactor which allows us to perform structural characterization using X-ray absorption spectroscopy including X-ray absorption near edge structure spectroscopy and extended X-ray absorption fine structure spectroscopy to study catalyst nanoparticles under harsh catalysis conditions in terms of liquid up to 350 °C under gas phase with a pressure up to 50 bars. This reactor remains nanoparticles of a catalyst homogeneously dispersed in liquid during catalysis and X-ray absorption spectroscopy characterization.

Vapor Phase Catalytic Ammonia Reduction

NASA Technical Reports Server (NTRS)

Flynn, Michael T.; Harper, Lynn D. (Technical Monitor)

1994-01-01

This paper discusses the development of a Vapor Phase Catalytic Ammonia Reduction (VPCAR) teststand and the results of an experimental program designed to evaluate the potential of the technology as a water purification process. In the experimental program the technology is evaluated based upon product water purity, water recovery rate, and power consumption. The experimental work demonstrates that the technology produces high purity product water and attains high water recovery rates at a relatively high specific power consumption. The experimental program was conducted in 3 phases. In phase I an Igepon(TM) soap and water mixture was used to evaluate the performance of an innovative Wiped-Film Rotating-Disk evaporator and associated demister. In phase II a phenol-water solution was used to evaluate the performance of the high temperature catalytic oxidation reactor. In phase III a urine analog was used to evaluate the performance of the combined distillation/oxidation functions of the processor.

Overview of the US Fusion Materials Sciences Program

NASA Astrophysics Data System (ADS)

Zinkle, Steven

2004-11-01

The challenging fusion reactor environment (radiation, heat flux, chemical compatibility, thermo-mechanical stresses) requires utilization of advanced materials to fulfill the promise of fusion to provide safe, economical, and environmentally acceptable energy. This presentation reviews recent experimental and modeling highlights on structural materials for fusion energy. The materials requirements for fusion will be compared with other demanding technologies, including high temperature turbine components, proposed Generation IV fission reactors, and the current NASA space fission reactor project to explore the icy moons of Jupiter. A series of high-performance structural materials have been developed by fusion scientists over the past ten years with significantly improved properties compared to earlier materials. Recent advances in the development of high-performance ferritic/martensitic and bainitic steels, nanocomposited oxide dispersion strengthened ferritic steels, high-strength V alloys, improved-ductility Mo alloys, and radiation-resistant SiC composites will be reviewed. Multiscale modeling is providing important insight on radiation damage and plastic deformation mechanisms and fracture mechanics behavior. Electron microscope in-situ straining experiments are uncovering fundamental physical processes controlling deformation in irradiated metals. Fundamental modeling and experimental studies are determining the behavior of transmutant helium in metals, enabling design of materials with improved resistance to void swelling and helium embrittlement. Recent chemical compatibility tests have identified promising new candidates for magnetohydrodynamic insulators in lithium-cooled systems, and have established the basic compatibility of SiC with Pb-Li up to high temperature. Research on advanced joining techniques such as friction stir welding will be described. ITER materials research will be briefly summarized.

Fracture toughness evaluation of select advanced replacement alloys for LWR core internals

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tan, Lizhen; Chen, Xiang

Life extension of the existing nuclear reactors imposes irradiation of high fluences to structural materials, resulting in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to develop and test degradation resistant alloys from current commercial alloy specifications bymore » 2021 to a new advanced alloy with superior degradation resistance in light water reactor (LWR)-relevant environments by 2024. Fracture toughness is one of the key engineering properties required for core internal materials. Together with other properties, which are being examined such as high-temperature steam oxidation resistance, radiation hardening, and irradiation-assisted stress corrosion cracking resistance, the alloys will be down-selected for neutron irradiation study and comprehensive post-irradiation examinations. According to the candidate alloys selected under the ARRM program, ductile fracture toughness of eight alloys was evaluated at room temperature and the LWR-relevant temperatures. The tested alloys include two ferritic alloys (Grade 92 and an oxide-dispersion-strengthened alloy 14YWT), two austenitic stainless steels (316L and 310), four Ni-base superalloys (718A, 725, 690, and X750). Alloy 316L and X750 are included as reference alloys for low- and high-strength alloys, respectively. Compact tension specimens in 0.25T and 0.2T were machined from the alloys in the T-L and R-L orientations according to the product forms of the alloys. This report summarizes the final results of the specimens tested and analyzed per ASTM Standard E1820. Unlike the ferritic alloys showing slight decreases (Grade 92) or significant decreases (14YWT) in fracture toughness at elevated temperatures, the fracture toughness of the austenitic stainless steels and Ni-base superalloys were not strongly dependent upon the test temperatures. The fracture toughness of the alloys at the LWR-relevant temperatures was estimated by averaging the toughness values within 250– 350°C, which suggested the fracture toughness of the alloys in a descending order as 316L (752±98 MPa√m), 310 (513±66 MPa√m), 718A (313±43 MPa√m), 690 (267±48 MPa√m), 725 (218±55 MPa√m), X750 (145±16 MPa√m), Grade 92 (112±12 MPa√m), and 14YWT (63±3 MPa√m). Tearing modulus of the alloys was analyzed in the meantime, which were not strongly dependent upon the test temperatures. The high-strength alloys 718A, 725, X750, and 14YWT had the lowest tearing modulus, ranging from ~45 to ~7. Alloy 690 exhibited the highest tearing modulus on the order of 450, followed by 316L and 310 on the order of 260. Grade 92 had a noticeably lower tearing modulus on the order of 70.« less

Reactor cell assembly for use in spectroscopy and microscopy applications

DOEpatents

Grindstaff, Quirinus; Stowe, Ashley Clinton; Smyrl, Norm; Powell, Louis; McLane, Sam

2015-08-04

The present disclosure provides a reactor cell assembly that utilizes a novel design and that is wholly or partially manufactured from Aluminum, such that reactions involving Hydrogen, for example, including solid-gas reactions and thermal decomposition reactions, are not affected by any degree of Hydrogen outgassing. This reactor cell assembly can be utilized in a wide range of optical and laser spectroscopy applications, as well as optical microscopy applications, including high-temperature and high-pressure applications. The result is that the elucidation of the role of Hydrogen in the reactions studied can be achieved. Various window assemblies can be utilized, such that high temperatures and high pressures can be accommodated and the signals obtained can be optimized.

Study on Characteristic of Temperature Coefficient of Reactivity for Plutonium Core of Pebbled Bed Reactor

NASA Astrophysics Data System (ADS)

Zuhair; Suwoto; Setiadipura, T.; Bakhri, S.; Sunaryo, G. R.

2018-02-01

As a part of the solution searching for possibility to control the plutonium, a current effort is focused on mechanisms to maximize consumption of plutonium. Plutonium core solution is a unique case in the high temperature reactor which is intended to reduce the accumulation of plutonium. However, the safety performance of the plutonium core which tends to produce a positive temperature coefficient of reactivity should be examined. The pebble bed inherent safety features which are characterized by a negative temperature coefficient of reactivity must be maintained under any circumstances. The purpose of this study is to investigate the characteristic of temperature coefficient of reactivity for plutonium core of pebble bed reactor. A series of calculations with plutonium loading varied from 0.5 g to 1.5 g per fuel pebble were performed by the MCNPX code and ENDF/B-VII library. The calculation results show that the k eff curve of 0.5 g Pu/pebble declines sharply with the increase in fuel burnup while the greater Pu loading per pebble yields k eff curve declines slighter. The fuel with high Pu content per pebble may reach long burnup cycle. From the temperature coefficient point of view, it is concluded that the reactor containing 0.5 g-1.25 g Pu/pebble at high burnup has less favorable safety features if it is operated at high temperature. The use of fuel with Pu content of 1.5 g/pebble at high burnup should be considered carefully from core safety aspect because it could affect transient behavior into a fatal accident situation.

H2/O2 three-body rates at high temperatures

NASA Technical Reports Server (NTRS)

Marinelli, William J.; Kessler, William J.; Piper, Lawrence G.; Rawlins, W. Terry

1990-01-01

The extraction of thrust from air breathing hypersonic propulsion systems is critically dependent on the degree to which chemical equilibrium is reached in the combustion process. In the combustion of H2/Air mixtures, slow three-body chemical reactions involving H-atoms, O-atoms, and the OH radical play an important role in energy extraction. A first-generation high temperature and pressure flash-photolysis/laser-induced fluorescence reactor was designed and constructed to measure these important three-body rates. The system employs a high power excimer laser to produce these radicals via the photolysis of stable precursors. A novel two-photon laser-induced fluorescence technique is employed to detect H-atoms without optical thickness or O2 absorption problems. To demonstrate the feasibility of the technique the apparatus in the program is designed to perform preliminary measurements on the H + O2 + M reaction at temperatures from 300 to 835 K.

Optimization and Comparison of Direct and Indirect Supercritical Carbon Dioxide Power Plant Cycles for Nuclear Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Edwin A. Harvego; Michael G. McKellar

2011-11-01

There have been a number of studies involving the use of gases operating in the supercritical mode for power production and process heat applications. Supercritical carbon dioxide (CO2) is particularly attractive because it is capable of achieving relatively high power conversion cycle efficiencies in the temperature range between 550 C and 750 C. Therefore, it has the potential for use with any type of high-temperature nuclear reactor concept, assuming reactor core outlet temperatures of at least 550 C. The particular power cycle investigated in this paper is a supercritical CO2 Recompression Brayton Cycle. The CO2 Recompression Brayton Cycle can bemore » used as either a direct or indirect power conversion cycle, depending on the reactor type and reactor outlet temperature. The advantage of this cycle when compared to the helium Brayton cycle is the lower required operating temperature; 550 C versus 850 C. However, the supercritical CO2 Recompression Brayton Cycle requires an operating pressure in the range of 20 MPa, which is considerably higher than the required helium Brayton cycle operating pressure of 8 MPa. This paper presents results of analyses performed using the UniSim process analyses software to evaluate the performance of both a direct and indirect supercritical CO2 Brayton Recompression cycle for different reactor outlet temperatures. The direct supercritical CO2 cycle transferred heat directly from a 600 MWt reactor to the supercritical CO2 working fluid supplied to the turbine generator at approximately 20 MPa. The indirect supercritical CO2 cycle assumed a helium-cooled Very High Temperature Reactor (VHTR), operating at a primary system pressure of approximately 7.0 MPa, delivered heat through an intermediate heat exchanger to the secondary indirect supercritical CO2 Brayton Recompression cycle, again operating at a pressure of about 20 MPa. For both the direct and indirect cycles, sensitivity calculations were performed for reactor outlet temperature between 550 C and 850 C. The UniSim models used realistic component parameters and operating conditions to model the complete reactor and power conversion systems. CO2 properties were evaluated, and the operating ranges of the cycles were adjusted to take advantage of the rapidly changing properties of CO2 near the critical point. The results of the analyses showed that, for the direct supercritical CO2 power cycle, thermal efficiencies in the range of 40 to 50% can be achieved. For the indirect supercritical CO2 power cycle, thermal efficiencies were approximately 10% lower than those obtained for the direct cycle over the same reactor outlet temperature range.« less

Characteristic of molten fluoride salt system LiF-BeF2 (Flibe) and LiF-NaF-KF (Flinak) as coolant and fuel carrier in molten salt reactor (MSR)

NASA Astrophysics Data System (ADS)

Bahri, Che Nor Aniza Che Zainul; Al-Areqi, Wadee'ah Mohd; Ruf, Mohd'Izzat Fahmi Mohd; Majid, Amran Ab.

2017-01-01

Interest of fluoride salts have recently revived due to the high temperature application in nuclear reactors. Molten Salt Reactor (MSR) was designed to operate at high temperature in range 700 - 800°C and its fuel is dissolved in a circulating molten fluoride salt mixture. Molten fluoride salts are stable at high temperature, have good heat transfer properties and can dissolve high concentration of actinides and fission product. The aim of this paper was to discuss the physical properties (melting temperature, density and heat capacity) of two systems fluoride salt mixtures i.e; LiF-BeF2 (Flibe) and LiF-NaF-KF (Flinak) in terms of their application as coolant and fuel solvent in MSR. Both of these salts showed almost same physical properties but different applications in MSR. The advantages and the disadvantages of these fluoride salt systems will be discussed in this paper.

Method for fabricating wrought components for high-temperature gas-cooled reactors and product

DOEpatents

Thompson, Larry D.; Johnson, Jr., William R.

1985-01-01

A method and alloys for fabricating wrought components of a high-temperature gas-cooled reactor are disclosed. These wrought, nickel-based alloys, which exhibit strength and excellent resistance to carburization at elevated temperatures, include aluminum and titanium in amounts and ratios to promote the growth of carburization resistant films while preserving the wrought character of the alloys. These alloys also include substantial amounts of molybdenum and/or tungsten as solid-solution strengtheners. Chromium may be included in concentrations less than 10% to assist in fabrication. Minor amounts of carbon and one or more carbide-forming metals also contribute to high-temperature strength.

Microscale Heat Conduction Models and Doppler Feedback

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hawari, Ayman I.; Ougouag, Abderrafi

2015-01-22

The objective of this project is to establish an approach for providing the fundamental input that is needed to estimate the magnitude and time-dependence of the Doppler feedback mechanism in Very High Temperature reactors. This mechanism is the foremost contributor to the passive safety of gas-cooled, graphite-moderated high temperature reactors that use fuel based on Tristructural-Isotropic (TRISO) coated particles. Therefore, its correct prediction is essential to the conduct of safety analyses for these reactors. Since the effect is directly dependent on the actual temperature reached by the fuel during transients, the underlying phenomena of heat deposition, heat transfer and temperaturemore » rise must be correctly predicted. To achieve the above objective, this project will explore an approach that accounts for lattice effects as well as local temperature variations and the correct definition of temperature and related local effects.« less

Status and improvement of CLAM for nuclear application

NASA Astrophysics Data System (ADS)

Huang, Qunying

2017-08-01

A program for China low activation martensitic steel (CLAM) development has been underway since 2001 to satisfy the material requirements of the test blanket module (TBM) for ITER, China fusion engineering test reactor and China fusion demonstration reactor. It has been undertaken by the Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences under wide domestic and international collaborations. Extensive work and efforts are being devoted to the R&D of CLAM, such as mechanical property evaluation before and after neutron irradiation, fabrication of scaled TBM by welding and additive manufacturing, improvement of its irradiation resistance as well as high temperature properties by precipitate strengthening to achieve its final successful application in fusion systems. The status and improvement of CLAM are introduced in this paper.

DIFFUSE: a FORTRAN program for design computation of tritium transport through thermonuclear reactor components by combined ordinary and thermal diffusion when the principal resistance to diffusion is the bulk metal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pendergrass, J.H.

1977-10-01

Based on the theory developed in an earlier report, a FORTRAN computer program, DIFFUSE, was written. It computes, for design purposes, rates of transport of hydrogen isotopes by temperature-dependent quasi-unidirectional, and quasi-static combined ordinary and thermal diffusion through thin, hot thermonuclear reactor components that can be represented by composites of plane, cylindrical-shell, and spherical-shell elements when the dominant resistance to transfer is that of the bulk metal. The program is described, directions for its use are given, and a listing of the program, together with sample problem results, is presented.

Johnson Noise Thermometry for Advanced Small Modular Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Britton Jr, Charles L; Roberts, Michael; Bull, Nora D

Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor s physical condition. In and near core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurementmore » due to their fundamental natures. Small, Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of this project is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.« less

Characterization of the Performance of Sapphire Optical Fiber in Intense Radiation Fields, when Subjected to Very High Temperatures

NASA Astrophysics Data System (ADS)

Petrie, Christian M.

The U.S. Department of Energy is interested in extending optically-based instrumentation from non-extreme environments to extremely high temperature radiation environments for the purposes of developing in-pile instrumentation. The development of in-pile instrumentation would help support the ultimate goal of understanding the behavior and predicting the performance of nuclear fuel systems at a microstructural level. Single crystal sapphire optical fibers are a promising candidate for in-pile instrumentation due to the high melting temperature and radiation hardness of sapphire. In order to extend sapphire fiber-based optical instrumentation to high temperature radiation environments, the ability of sapphire fibers to adequately transmit light in such an environment must first be demonstrated. Broadband optical transmission measurements of sapphire optical fibers were made in-situ as the sapphire fibers were heated and/or irradiated. The damage processes in sapphire fibers were also modeled from the primary knock-on event from energetic neutrons to the resulting damage cascade in order to predict the formation of stable defects that ultimately determine the resulting change in optical properties. Sapphire optical fibers were shown to withstand temperatures as high as 1300 °C with minimal increases in optical attenuation. A broad absorption band was observed to grow over time without reaching a dynamic equilibrium when the sapphire fiber was heated at temperatures of 1400 °C and above. The growth of this absorption band limits the use of sapphire optical fibers, at least in air, to temperatures of 1300 °C and below. Irradiation of sapphire fibers with gamma rays caused saturation of a defect center located below 500 nm, and extending as far as ~1000 nm, with little effect on the transmission at 1300 and 1550 nm. Increasing temperature during gamma irradiation generally reduced the added attenuation. Reactor irradiation of sapphire fibers caused an initial rapid increase in attenuation, followed by a linear increase with continued irradiation time at constant reactor power. The linear increases were a result of displacement damage, and the rate of increase was proportional to the neutron flux. The transmission of sapphire fibers at 1300 and 1550 nm in a reactor radiation environment would ultimately be limited by the growth of low wavelength defect centers, whose tails extend into the near infrared. A model was proposed for the reactor radiation-induced attenuation that involves three previously reported color centers. The model accounts for gamma radiation-induced ionization of pre-existing defects, generation of new defects via displacement damage, and conversion between defect centers via ionization and charge recombination. Heated reactor irradiation experiments showed that the rate of increase of the added attenuation during constant power reactor irradiation monotonically decreases with increasing temperature up to 1000 °C, with the most significant decrease occurring between 300 and 600 °C. Testing of sapphire fiber-based sensors under irradiation at high temperatures is recommended as future work, along with advanced life irradiation testing, for example in the Advanced Test Reactor or the High Flux Isotope Reactor.

High-Temperature Fluid-Wall Reactor Technology Research, Test and Evaluation Performed at Naval Construction Battalion Center, Gulfport, MS, for the USAF Installation/Restoration Program

DTIC Science & Technology

1988-01-01

under field conditions. Sampling and analytical laboratory activities were performed by Ecology and Environment, Inc., and California Analytical...the proposed AER3 test conditions. All test samples would be obtained onsite by Ecology and Environment, Inc., of Buffalo, New York, and sent to...ensuring its safe operation. Ecology and Environment performed onsite verification sampling. This activity was coordinated with the Huber project team

Analyzing Flows In Rocket Nuclear Reactors

NASA Technical Reports Server (NTRS)

Clark, J. S.; Walton, J. T.; Mcguire, M.

1994-01-01

CAC is analytical prediction program to study heat-transfer and fluid-flow characteristics of circular coolant passage. Predicts, as function of time, axial and radial fluid conditions, temperatures of passage walls, rates of flow in each coolant passage, and approximate maximum material temperatures. Written in ANSI standard FORTRAN 77.

Solar Power Satellites - A Review of the Space Transportation Options.

DTIC Science & Technology

1980-03-01

already exists with such systems, gained mainly through liquid-metal breeder reactor programmes. 0 For example, inlet temperatures of 970 C can be handled...alternatives exist. In addition, there would be extreme reluctance on the part of most governments to allow large C- reactors , producing gigawatts of power, to...antenna. The reactors employed are high-temperature gas- cooled breeders , which convert U238 into fissile plutonium. Each of the modules includes a

New reactor cavity cooling system having passive safety features using novel shape for HTGRs and VHTRs

DOE PAGES

Takamatsu, Kuniyoshi; Hu, Rui

2014-11-27

A new, highly efficient reactor cavity cooling system (RCCS) with passive safety features without a requirement for electricity and mechanical drive is proposed for high temperature gas cooled reactors (HTGRs) and very high temperature reactors (VHTRs). The RCCS design consists of continuous closed regions; one is an ex-reactor pressure vessel (RPV) region and another is a cooling region having heat transfer area to ambient air assumed at 40 (°C). The RCCS uses a novel shape to efficiently remove the heat released from the RPV with radiation and natural convection. Employing the air as the working fluid and the ambient airmore » as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay heat removal. Therefore, HTGRs and VHTRs adopting the new RCCS design can avoid core melting due to overheating the fuels. The simulation results from a commercial CFD code, STAR-CCM+, show that the temperature distribution of the RCCS is within the temperature limits of the structures, such as the maximum operating temperature of the RPV, 713.15 (K) = 440 (°C), and the heat released from the RPV could be removed safely, even during a loss of coolant accident (LOCA). Finally, when the RCCS can remove 600 (kW) of the rated nominal state even during LOCA, the safety review for building the HTTR could confirm that the temperature distribution of the HTTR is within the temperature limits of the structures to secure structures and fuels after the shutdown because the large heat capacity of the graphite core can absorb heat from the fuel in a short period. Therefore, the capacity of the new RCCS design would be sufficient for decay heat removal.« less

A model to predict thermal conductivity of irradiated U–Mo dispersion fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burkes, Douglas E.; Huber, Tanja K.; Casella, Andrew M.

The Office of Materials Management and Minimization Reactor Conversion Program continues to develop existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. The program is focused on assisting with the development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layermore » formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of the modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.« less

New fixed-point mini-cell to investigate thermocouple drift in a high-temperature environment under neutron irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Laurie, M.; Vlahovic, L.; Rondinella, V.V.

Temperature measurements in the nuclear field require a high degree of reliability and accuracy. Despite their sheathed form, thermocouples subjected to nuclear radiations undergo changes due to radiation damage and transmutation that lead to significant EMF drift during long-term fuel irradiation experiment. For the purpose of a High Temperature Reactor fuel irradiation to take place in the High Flux Reactor Petten, a dedicated fixed-point cell was jointly developed by LNE-Cnam and JRC-IET. The developed cell to be housed in the irradiation rig was tailor made to quantify the thermocouple drift during the irradiation (about two year duration) and withstand highmore » temperature (in the range 950 deg. C - 1100 deg. C) in the presence of contaminated helium in a graphite environment. Considering the different levels of temperature achieved in the irradiation facility and the large palette of thermocouple types aimed at surveying the HTR fuel pebble during the qualification test both copper (1084.62 deg. C) and gold (1064.18 deg. C) fixed-point materials were considered. The aim of this paper is to first describe the fixed-point mini-cell designed to be embedded in the reactor rig and to discuss the preliminary results achieved during some out of pile tests as much as some robustness tests representative of the reactor scram scenarios. (authors)« less

Fluidized Bed Membrane Reactors for Ultra Pure H₂ Production--A Step forward towards Commercialization.

PubMed

Helmi, Arash; Fernandez, Ekain; Melendez, Jon; Pacheco Tanaka, David Alfredo; Gallucci, Fausto; van Sint Annaland, Martin

2016-03-19

In this research the performance of a fluidized bed membrane reactor for high temperature water gas shift and its long term stability was investigated to provide a proof-of-concept of the new system at lab scale. A demonstration unit with a capacity of 1 Nm³/h of ultra-pure H₂ was designed, built and operated over 900 h of continuous work. Firstly, the performance of the membranes were investigated at different inlet gas compositions and at different temperatures and H₂ partial pressure differences. The membranes showed very high H₂ fluxes (3.89 × 10(-6) mol·m(-2)·Pa(-1)·s(-1) at 400 °C and 1 atm pressure difference) with a H₂/N₂ ideal perm-selectivity (up to 21,000 when integrating five membranes in the module) beyond the DOE 2015 targets. Monitoring the performance of the membranes and the reactor confirmed a very stable performance of the unit for continuous high temperature water gas shift under bubbling fluidization conditions. Several experiments were carried out at different temperatures, pressures and various inlet compositions to determine the optimum operating window for the reactor. The obtained results showed high hydrogen recovery factors, and very low CO concentrations at the permeate side (in average <10 ppm), so that the produced hydrogen can be directly fed to a low temperature PEM fuel cell.

RELAP5-3D Results for Phase I (Exercise 2) of the OECD/NEA MHTGR-350 MW Benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerhard Strydom

2012-06-01

The coupling of the PHISICS code suite to the thermal hydraulics system code RELAP5-3D has recently been initiated at the Idaho National Laboratory (INL) to provide a fully coupled prismatic Very High Temperature Reactor (VHTR) system modeling capability as part of the NGNP methods development program. The PHISICS code consists of three modules: INSTANT (performing 3D nodal transport core calculations), MRTAU (depletion and decay heat generation) and a perturbation/mixer module. As part of the verification and validation activities, steady state results have been obtained for Exercise 2 of Phase I of the newly-defined OECD/NEA MHTGR-350 MW Benchmark. This exercise requiresmore » participants to calculate a steady-state solution for an End of Equilibrium Cycle 350 MW Modular High Temperature Reactor (MHTGR), using the provided geometry, material, and coolant bypass flow description. The paper provides an overview of the MHTGR Benchmark and presents typical steady state results (e.g. solid and gas temperatures, thermal conductivities) for Phase I Exercise 2. Preliminary results are also provided for the early test phase of Exercise 3 using a two-group cross-section library and the Relap5-3D model developed for Exercise 2.« less

RELAP5-3D results for phase I (Exercise 2) of the OECD/NEA MHTGR-350 MW benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strydom, G.; Epiney, A. S.

2012-07-01

The coupling of the PHISICS code suite to the thermal hydraulics system code RELAP5-3D has recently been initiated at the Idaho National Laboratory (INL) to provide a fully coupled prismatic Very High Temperature Reactor (VHTR) system modeling capability as part of the NGNP methods development program. The PHISICS code consists of three modules: INSTANT (performing 3D nodal transport core calculations), MRTAU (depletion and decay heat generation) and a perturbation/mixer module. As part of the verification and validation activities, steady state results have been obtained for Exercise 2 of Phase I of the newly-defined OECD/NEA MHTGR-350 MW Benchmark. This exercise requiresmore » participants to calculate a steady-state solution for an End of Equilibrium Cycle 350 MW Modular High Temperature Reactor (MHTGR), using the provided geometry, material, and coolant bypass flow description. The paper provides an overview of the MHTGR Benchmark and presents typical steady state results (e.g. solid and gas temperatures, thermal conductivities) for Phase I Exercise 2. Preliminary results are also provided for the early test phase of Exercise 3 using a two-group cross-section library and the Relap5-3D model developed for Exercise 2. (authors)« less

High-irradiance reactor design with practical unfolded optics

NASA Astrophysics Data System (ADS)

Feuermann, Daniel; Gordon, Jeffrey M.

2008-08-01

In the design of high-temperature chemical reactors and furnaces, as well as high-radiance light projection applications, reconstituting the ultra-high radiance of short-arc discharge lamps at maximum radiative efficiency constitutes a significant challenge. The difficulty is exacerbated by the high numerical aperture necessary at both the source and the target. Separating the optic from both the light source and the target allows practical operation, control, monitoring, diagnostics and maintenance. We present near-field unfolded aplanatic optics as a feasible solution. The concept is illustrated with a design customized to a high-temperature chemical reactor for nano-material synthesis, driven by an ultra-bright xenon short-arc discharge lamp, with near-unity numerical aperture for both light input and light output. We report preliminary optical measurements for the first prototype, which constitutes a double-ellipsoid solution. We also propose compound unfolded aplanats that collect the full angular extent of lamp emission (in lieu of light recycling optics) and additionally permit nearly full-circumference irradiation of the reactor.

Optimizing Glassy Polymer Network Morphology for Nano-particle Dispersion, Stabilization and Performance

DTIC Science & Technology

2016-10-03

dissolution, toughener dissolution and controlled chain-extension reactions in the continuous reactor high temperature “hot-zone” to advance conversion...rheology and tack. 2. Simultaneous MWCNT dispersion and stabilization in the continuous reactor low temperature “cold-zone” leading to an increased...Weight and Low Dispersity Polyacrylonitrile by Low Temperature RAFT Polymerization, Moskowitz, Jeremy, Abel, Brooks, McCormick, Charles, Wiggins

High rate mesophilic, thermophilic, and temperature phased anaerobic digestion of waste activated sludge: A pilot scale study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bolzonella, David, E-mail: david.bolzonella@univr.it; Cavinato, Cristina, E-mail: cavinato@unive.it; Fatone, Francesco, E-mail: francesco.fatone@univr.it

2012-06-15

Highlights: Black-Right-Pointing-Pointer High temperatures were tested in single and two-stage anaerobic digestion of waste activated sludge. Black-Right-Pointing-Pointer The increased temperature demonstrated the possibility of improving typical yields of the conventional mesophilic process. Black-Right-Pointing-Pointer The temperature phased anaerobic digestion process (65 + 55 Degree-Sign C) showed the best performances with yields of 0.49 m{sup 3}/kgVS{sub fed}. Black-Right-Pointing-Pointer Ammonia and phosphate released from solids destruction determined the precipitation of struvite in the reactor. - Abstract: The paper reports the findings of a two-year pilot scale experimental trial for the mesophilic (35 Degree-Sign C), thermophilic (55 Degree-Sign C) and temperature phased (65 +more » 55 Degree-Sign C) anaerobic digestion of waste activated sludge. During the mesophilic and thermophilic runs, the reactor operated at an organic loading rate of 2.2 kgVS/m{sup 3}d and a hydraulic retention time of 20 days. In the temperature phased run, the first reactor operated at an organic loading rate of 15 kgVS/m{sup 3}d and a hydraulic retention time of 2 days while the second reactor operated at an organic loading rate of 2.2 kgVS/m{sup 3}d and a hydraulic retention time of 18 days (20 days for the whole temperature phased system). The performance of the reactor improved with increases in temperature. The COD removal increased from 35% in mesophilic conditions, to 45% in thermophilic conditions, and 55% in the two stage temperature phased system. As a consequence, the specific biogas production increased from 0.33 to 0.45 and to 0.49 m{sup 3}/kgVS{sub fed} at 35, 55, and 65 + 55 Degree-Sign C, respectively. The extreme thermophilic reactor working at 65 Degree-Sign C showed a high hydrolytic capability and a specific yield of 0.33 gCOD (soluble) per gVS{sub fed}. The effluent of the extreme thermophilic reactor showed an average concentration of soluble COD and volatile fatty acids of 20 and 9 g/l, respectively. Acetic and propionic acids were the main compounds found in the acids mixture. Because of the improved digestion efficiency, organic nitrogen and phosphorus were solubilised in the bulk. Their concentration, however, did not increase as expected because of the formation of salts of hydroxyapatite and struvite inside the reactor.« less

PARTIAL ECONOMIC STUDY OF STEAM COOLED HEAVY WATER MODERATED REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1960-04-01

Steam-cooled reactors are compared with CAHDU for costs of Calandria tubes, pressure tubes. heavy water moderator, heavy water reflector, fuel supply, heat exchanger, and turbine generator. A direct-cycle lightsteam-cooled heavy- water-moderated pressure-tube reactor formed the basic reactor design for the study. Two methods of steam circulation through the reactor were examined. In both cases the steam was generated outside the reactor and superheated in the reactor core. One method consisted of a series of reactor and steam generator passes. The second method consisted of the Loeffler cycle and its modifications. The fuel was assumed to be natural cylindrical UO/sub 2/more » pellets sheathed in a hypothetical material with the nuclear properties of Zircaloy, but able to function at temperatures to 900 deg F. For the conditions assumed, the longer the rod, the higher the outlet temperature and therefore the higher the efficiency. The turbine cycle efficiency was calculated on the assumption that suitable steam generators are available. As the neutron losses to the pressure tubes were significant, an economic analysis of insulated pressure tubes is included. A description of the physics program for steam-cooled reactors is included. Results indicated that power from the steam-cooled reactor would cost 1.4 mills/ kwh compared with 1.25 mills/kwh for CANDU. (M.C.G.)« less

Fabrication of Monolithic RERTR Fuels by Hot Isostatic Pressing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jan-Fong Jue; Blair H. Park; Curtis R. Clark

2010-11-01

The RERTR (Reduced Enrichment for Research and Test Reactors) Program is developing advanced nuclear fuels for high-power test reactors. Monolithic fuel design provides higher uranium loading than that of the traditional dispersion fuel design. Hot isostatic pressing is a promising process for low-cost batch fabrication of monolithic RERTR fuel plates for these high-power reactors. Bonding U Mo fuel foil and 6061 Al cladding by hot isostatic press bonding was successfully developed at Idaho National Laboratory. Due to the relatively high processing temperature, the interaction between fuel meat and aluminum cladding is a concern. Two different methods were employed to mitigatemore » this effect: (1) a diffusion barrier and (2) a doping addition to the interface. Both types of fuel plates have been fabricated by hot isostatic press bonding. Preliminary results show that the direct fuel/cladding interaction during the bonding process was eliminated by introducing a thin zirconium diffusion barrier layer between the fuel and the cladding. Fuel plates were also produced and characterized with a silicon-rich interlayer between fuel and cladding. This paper reports the recent progress of this developmental effort and identifies the areas that need further attention.« less

Investigations on neutron irradiated 3D carbon fibre reinforced carbon composite material

NASA Astrophysics Data System (ADS)

Venugopalan, Ramani; Alur, V. D.; Patra, A. K.; Acharya, R.; Srivastava, D.

2018-04-01

As against conventional graphite materials carbon-carbon (C/C) composite materials are now being contemplated as the promising candidate materials for the high temperature and fusion reactor owing to their high thermal conductivity and high thermal resistance, better mechanical/thermal properties and irradiation stability. The current need is for focused research on novel carbon materials for future new generation nuclear reactors. The advantage of carbon-carbon composite is that the microstructure and the properties can be tailor made. The present study encompasses the irradiation of 3D carbon composite prepared by reinforcement using PAN carbon fibers for nuclear application. The carbon fiber reinforced composite was subjected to neutron irradiation in the research reactor DHRUVA. The irradiated samples were characterized by Differential Scanning Calorimetry (DSC), small angle neutron scattering (SANS), XRD and Raman spectroscopy. The DSC scans were taken in argon atmosphere under a linear heating program. The scanning was carried out at temperature range from 30 °C to 700 °C at different heating rates in argon atmosphere along with reference as unirradiated carbon composite. The Wigner energy spectrum of irradiated composite showed two peaks corresponding to 200 °C and 600 °C. The stored energy data for the samples were in the range 110-170 J/g for temperature ranging from 30 °C to 700 °C. The Wigner energy spectrum of irradiated carbon composite did not indicate spontaneous temperature rise during thermal annealing. Small angle neutron scattering (SANS) experiments have been carried out to investigate neutron irradiation induced changes in porosity of the composite samples. SANS data were recorded in the scattering wave vector range of 0.17 nm-1 to 3.5 nm-1. Comparison of SANS profiles of irradiated and unirradiated samples indicates significant change in pore morphology. Pore size distributions of the samples follow power law size distribution with different exponent. Narrowing of SANS profile of the irradiated sample indicates creation of significant number of larger pores due to neutron irradiation.

Structure and creep of Russian reactor steels with a BCC structure

NASA Astrophysics Data System (ADS)

Sagaradze, V. V.; Kochetkova, T. N.; Kataeva, N. V.; Kozlov, K. A.; Zavalishin, V. A.; Vil'danova, N. F.; Ageev, V. S.; Leont'eva-Smirnova, M. V.; Nikitina, A. A.

2017-05-01

The structural phase transformations have been revealed and the characteristics of the creep and long-term strength at 650, 670, and 700°C and 60-140 MPa have been determined in six Russian reactor steels with a bcc structure after quenching and high-temperature tempering. Creep tests were carried out using specially designed longitudinal and transverse microsamples, which were fabricated from the shells of the fuel elements used in the BN-600 fast neutron reactor. It has been found that the creep rate of the reactor bcc steels is determined by the stability of the lath martensitic and ferritic structures in relation to the diffusion processes of recovery and recrystallization. The highest-temperature oxide-free steel contains the maximum amount of the refractory elements and carbides. The steel strengthened by the thermally stable Y-Ti nanooxides has a record high-temperature strength. The creep rate at 700°C and 100 MPa in the samples of this steel is lower by an order of magnitude and the time to fracture is 100 times greater than that in the oxide-free reactor steels.

Versatile in situ gas analysis apparatus for nanomaterials reactors.

PubMed

Meysami, Seyyed Shayan; Snoek, Lavina C; Grobert, Nicole

2014-09-02

We report a newly developed technique for the in situ real-time gas analysis of reactors commonly used for the production of nanomaterials, by showing case-study results obtained using a dedicated apparatus for measuring the gas composition in reactors operating at high temperature (<1000 °C). The in situ gas-cooled sampling probe mapped the chemistry inside the high-temperature reactor, while suppressing the thermal decomposition of the analytes. It thus allows a more accurate study of the mechanism of progressive thermocatalytic cracking of precursors compared to previously reported conventional residual gas analyses of the reactor exhaust gas and hence paves the way for the controlled production of novel nanomaterials with tailored properties. Our studies demonstrate that the composition of the precursors dynamically changes as they travel inside of the reactor, causing a nonuniform growth of nanomaterials. Moreover, mapping of the nanomaterials reactor using quantitative gas analysis revealed the actual contribution of thermocatalytic cracking and a quantification of individual precursor fragments. This information is particularly important for quality control of the produced nanomaterials and for the recycling of exhaust residues, ultimately leading toward a more cost-effective continuous production of nanomaterials in large quantities. Our case study of multiwall carbon nanotube synthesis was conducted using the probe in conjunction with chemical vapor deposition (CVD) techniques. Given the similarities of this particular CVD setup to other CVD reactors and high-temperature setups generally used for nanomaterials synthesis, the concept and methodology of in situ gas analysis presented here does also apply to other systems, making it a versatile and widely applicable method across a wide range of materials/manufacturing methods, catalysis, as well as reactor design and engineering.

Corrosion of Structural Materials for Advanced Supercritical Carbon- Dioxide Brayton Cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sridharan, Kumar

The supercritical carbon-dioxide (referred to as SC-CO 2 hereon) Brayton cycle is being considered for power conversion systems for a number of nuclear reactor concepts, including the sodium fast reactor (SFR), fluoride saltcooled high temperature reactor (FHR), and high temperature gas reactor (HTGR), and several types of small modular reactors (SMR). The SC-CO 2 direct cycle gas fast reactor has also been recently proposed. The SC-CO 2 Brayton cycle (discussed in Chapter 1) provides higher efficiencies compared to the Rankine steam cycle due to less compression work stemming from higher SC-CO 2 densities, and allows for smaller components size, fewermore » components, and simpler cycle layout. For example, in the case of a SFR using a SC-CO 2 Brayton cycle instead of a steam cycle would also eliminate the possibility of sodium-water interactions. The SC-CO 2 cycle has a higher efficiency than the helium Brayton cycle, with the additional advantage of being able to operate at lower temperatures and higher pressures. In general, the SC-CO 2 Brayton cycle is well-suited for any type of nuclear reactor (including SMR) with core outlet temperature above ~ 500°C in either direct or indirect versions. In all the above applications, materials corrosion in high temperature SC-CO 2 is an important consideration, given their expected lifetimes of 20 years or longer. Our discussions with National Laboratories and private industry early on in this project indicated materials corrosion to be one of the significant gaps in the implementation of SC-CO 2 Brayton cycle. Corrosion can lead to a loss of effective load-bearing wall thickness of a component and can potentially lead to the generation of oxide particulate debris which can lead to three-body wear in turbomachinery components. Another environmental degradation effect that is rather unique to CO 2 environment is the possibility for simultaneous occurrence of carburization during oxidation of the material. Carburization can potentially lead to embrittlement of structural alloys in SC-CO 2 Brayton cycle. An important consideration in regards to corrosion is that the temperatures can vary widely across the various sections of the SC-CO 2 Brayton cycle, from room temperature to 750°C, with even higher temperatures being desirable for higher efficiencies. Thus the extent of corrosion and corrosion mechanisms in various components and SC-CO 2 Brayton cycle will be different, requiring a judicious selection of materials for different sections of the cycle. The goal of this project was to address materials corrosion-related challenges, identify appropriate materials, and advance the body of scientific knowledge in the area of high temperature SC-CO 2 corrosion. The focus was on corrosion of materials in SC-CO 2 environment in the temperature range of 450°C to 750°C at a pressure of 2900 psi for exposure duration for up to 1000 hours. The Table below lists the materials tested in the project. The materials were selected based on their high temperature strength, their code certification status, commercial availabilities, and their prior or current usage in the nuclear reactor industry. Additionally, pure Fe, Fe-12%Cr, and Ni-22%Cr were investigated as simple model materials to more clearly understand corrosion mechanisms. This first phase of the project involved testing in research grade SC-CO 2 (99.999% purity). Specially designed autoclaves with high fidelity temperature, pressure, and flow control capabilities were built or modified for this project.« less

Characterization of a novel micro-pressure swirl reactor for removal of chemical oxygen demand and total nitrogen from domestic wastewater at low temperature.

PubMed

Ren, Qingkai; Yu, Yang; Zhu, Suiyi; Bian, Dejun; Huo, Mingxin; Zhou, Dandan; Huo, Hongliang

2017-06-01

A novel micro-pressure swirl reactor (MPSR) was designed and applied to treat domestic wastewater at low temperature by acclimating microbial biomass with steadily decreasing temperature from 15 to 3 °C. Chemical oxygen demand (COD) was constantly removed by 85% and maintained below 50 mg L -1 in the effluent during the process. When the air flow was controlled at 0.2 m 3  h -1 , a swirl circulation was formed in the reactor, which created a dissolved oxygen (DO) gradient with a low DO zone in the center and a high DO zone in the periphery for denitrification and nitrification. 81% of total nitrogen was removed by this reactor, in which ammonium was reduced by over 90%. However, denitrification was less effective because of the presence of low levels of oxygen. The progressively decreasing temperature favored acclimation of psychrophilic bacteria in the reactor, which replaced mesophilic bacteria in the process of treatment.

Verification of combined thermal-hydraulic and heat conduction analysis code FLOWNET/TRUMP

NASA Astrophysics Data System (ADS)

Maruyama, Soh; Fujimoto, Nozomu; Kiso, Yoshihiro; Murakami, Tomoyuki; Sudo, Yukio

1988-09-01

This report presents the verification results of the combined thermal-hydraulic and heat conduction analysis code, FLOWNET/TRUMP which has been utilized for the core thermal hydraulic design, especially for the analysis of flow distribution among fuel block coolant channels, the determination of thermal boundary conditions for fuel block stress analysis and the estimation of fuel temperature in the case of fuel block coolant channel blockage accident in the design of the High Temperature Engineering Test Reactor(HTTR), which the Japan Atomic Energy Research Institute has been planning to construct in order to establish basic technologies for future advanced very high temperature gas-cooled reactors and to be served as an irradiation test reactor for promotion of innovative high temperature new frontier technologies. The verification of the code was done through the comparison between the analytical results and experimental results of the Helium Engineering Demonstration Loop Multi-channel Test Section(HENDEL T(sub 1-M)) with simulated fuel rods and fuel blocks.

Conceptual Design of Low-Temperature Hydrogen Production and High-Efficiency Nuclear Reactor Technology

NASA Astrophysics Data System (ADS)

Fukushima, Kimichika; Ogawa, Takashi

Hydrogen, a potential alternative energy source, is produced commercially by methane (or LPG) steam reforming, a process that requires high temperatures, which are produced by burning fossil fuels. However, as this process generates large amounts of CO2, replacement of the combustion heat source with a nuclear heat source for 773-1173K processes has been proposed in order to eliminate these CO2 emissions. In this paper, a novel method of nuclear hydrogen production by reforming dimethyl ether (DME) with steam at about 573K is proposed. From a thermodynamic equilibrium analysis of DME steam reforming, the authors identified conditions that provide high hydrogen production fraction at low pressure and temperatures of about 523-573K. By setting this low-temperature hydrogen production process upstream from a turbine and nuclear reactor at about 573K, the total energy utilization efficiency according to equilibrium mass and heat balance analysis is about 50%, and it is 75%for a fast breeder reactor (FBR), where turbine is upstream of the reformer.

Status and progress of the RERTR program in the year 2003.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Travelli, A.; Nuclear Engineering Division

2003-01-01

One of the most important events affecting the RERTR program during the past year was the decision by the U.S. Department of Energy to request the U.S. Congress to significantly increase RERTR program funding. This decision was prompted, at least in part, by the terrible events of September 11, 2001, and by a high-level U.S./Russian Joint Expert Group recommendation to immediately accelerate RERTR program activities in both countries, with the goal of converting all the world's research reactors to low-enriched fuel at the earliest possible time, and including both Soviet-designed and United States-designed research reactors. The U.S. Congress is expectedmore » to approve this request very soon, and the RERTR program has prepared itself well for the intense activities that the 'Accelerated RERTR Program' will require. Promising results have been obtained in the development of a fabrication process for monolithic LEU U-Mo fuel. Most existing and future research reactors could be converted to LEU with this fuel, which has a uranium density between 15.4 and 16.4 g/cm{sup 3} and yielded promising irradiation results in 2002. The most promising method hinges on producing the monolithic meat by cold-rolling a thin ingot produced by casting. The aluminum clad and the meat are bonded by friction stir welding and the cladding surface is finished by a light cold roll. This method can be applied to the production of miniplates and appears to be extendable to the production of full-size plates, possibly with intermediate anneals. Other methods planned for investigation include high temperature bonding and hot isostatic pressing. The progress achieved within the Russian RERTR program, both for the traditional tube-type elements and for the new 'universal' LEU U-Mo pin-type elements, promises to enable soon the conversion of many Russian-designed research and test reactors. Irradiation testing of both fuel types with LEU U-Mo dispersion fuels has begun. Detailed studies are in progress to define the feasibility of converting each Russian-designed research and test reactor to either fuel type. The plan for the Accelerated RERTR Program is structured to achieve LEU conversion of all HEU research reactors supplied by the United States and Russia during the next nine years. This effort will address, in addition to the fuel development and qualification, the analyses and performance/economic/safety evaluations needed to implement the conversions. In combination with this over-arching goal, the RERTR program plans to achieve at the earliest possible date qualification of LEU U-Mo dispersion fuels with uranium densities of 6 g/cm{sup 3} and 7 g/cm{sup 3}. Reactors currently using or planning to use LEU silicide fuel will rely on this fuel after termination of the FRRSNFA program, because it is acceptable to COGEMA for reprocessing. Qualification of LEU U-Mo dispersion fuels has suffered some unavoidable delays but, to accelerate it as much as possible, the RERTR program, the French CEA, and the Australian ANSTO have agreed to jointly pursue a two-element qualification test of LEU U-Mo dispersion fuel with uranium density of 7.0 g/cm{sup 3} to be performed in the Osiris reactor during 2004. The RERTR program also intends to eliminate all obstacles to the utilization of LEU in targets for isotope production, so that this important function can be performed without the need for weapons-grade materials. All of us, working together as we have for many years, can ensure that all these goals will be achieved. By promoting the efficiency and safety of research reactors while eliminating the traffic in weapons-grade uranium, we can prevent the possibility that some of this material might fall in the wrong hands. Few causes can be more deserving of our joint efforts.« less

Code qualification of structural materials for AFCI advanced recycling reactors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Natesan, K.; Li, M.; Majumdar, S.

2012-05-31

This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Codemore » Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the Power Reactor Innovative Small Module (PRISM), the NRC/Advisory Committee on Reactor Safeguards (ACRS) raised numerous safety-related issues regarding elevated-temperature structural integrity criteria. Most of these issues remained unresolved today. These critical licensing reviews provide a basis for the evaluation of underlying technical issues for future advanced sodium-cooled reactors. Major materials performance issues and high temperature design methodology issues pertinent to the ARR are addressed in the report. The report is organized as follows: the ARR reference design concepts proposed by the Argonne National Laboratory and four industrial consortia were reviewed first, followed by a summary of the major code qualification and licensing issues for the ARR structural materials. The available database is presented for the ASME Code-qualified structural alloys (e.g. 304, 316 stainless steels, 2.25Cr-1Mo, and mod.9Cr-1Mo), including physical properties, tensile properties, impact properties and fracture toughness, creep, fatigue, creep-fatigue interaction, microstructural stability during long-term thermal aging, material degradation in sodium environments and effects of neutron irradiation for both base metals and weld metals. An assessment of modified versions of Type 316 SS, i.e. Type 316LN and its Japanese version, 316FR, was conducted to provide a perspective for codification of 316LN or 316FR in Subsection NH. Current status and data availability of four new advanced alloys, i.e. NF616, NF616+TMT, NF709, and HT-UPS, are also addressed to identify the R&D needs for their code qualification for ARR applications. For both conventional and new alloys, issues related to high temperature design methodology are described to address the needs for improvements for the ARR design and licensing. Assessments have shown that there are significant data gaps for the full qualification and licensing of the ARR structural materials. Development and evaluation of structural materials require a variety of experimental facilities that have been seriously degraded in the past. The availability and additional needs for the key experimental facilities are summarized at the end of the report. Detailed information covered in each Chapter is given.« less

Final Environmental Impact Statement (EIS) for the Space Nuclear Thermal Propulsion (SNTP) program

NASA Astrophysics Data System (ADS)

1991-09-01

A program has been proposed to develop the technology and demonstrate the feasibility of a high-temperature particle bed reactor (PBR) propulsion system to be used to power an advanced second stage nuclear rocket engine. The purpose of this Final Environmental Impact Statement (FEIS) is to assess the potential environmental impacts of component development and testing, construction of ground test facilities, and ground testing. Major issues and goals of the program include the achievement and control of predicted nuclear power levels; the development of materials that can withstand the extremely high operating temperatures and hydrogen flow environments; and the reliable control of cryogenic hydrogen and hot gaseous hydrogen propellant. The testing process is designed to minimize radiation exposure to the environment. Environmental impact and mitigation planning are included for the following areas of concern: (1) Population and economy; (2) Land use and infrastructure; (3) Noise; (4) Cultural resources; (5) Safety (non-nuclear); (6) Waste; (7) Topography; (8) Geology; (9) Seismic activity; (10) Water resources; (11) Meteorology/Air quality; (12) Biological resources; (13) Radiological normal operations; (14) Radiological accidents; (15) Soils; and (16) Wildlife habitats.

Direct thermal water splitting by concentrated solar radiation for hydrogen production. Phase O: Proof of concept experiment

NASA Technical Reports Server (NTRS)

Genequand, P.

1980-01-01

The direct production of hydrogen from water and solar energy concentrated into a high temperature aperture is described. A solar powered reactor able to dissociate water vapor and to separate the reaction product at high temperature was developed, and direct water splitting has been achieved in a laboratory reactor. Water vapor and radiative heating from a carbon dioxide laser are fed into the reactor, and water vapor enriched in hydrogen and water vapor enriched in oxygen are produced. The enriched water vapors are separated through a separation membrane, a small disc of zirconium dioxide heated to a range of 1800 k to 2800 k. To avoid water vapor condensation within the reactor, the total pressure within the reactor was limited to 0.15 torr. A few modifications would enable the reactor to be operated at an increased pressure of a few torrs. More substantial modifications would allow for a reaction pressure of 0.1 atmosphere.

The Shock and Vibration Digest. Volume 17, Number 2

DTIC Science & Technology

1985-02-01

phenomena relative to A computer program has been developed to -.- buildings, bridges, dams, and other struc- simulate the motions of bodies subjected to...1982). (57) Ikushima, T., Honma, T., and Ishiz- uka, H., "Seismic Research on Block-Type (47) Kadle, D.S. and Chwang, A.T., "Hy- HTGR Core ," Nucl...T., "A Seismic Study of High Temperature Gas-Cooled Reactor Core - (48) Yang, C.Y., Chiarito, V., and Dressel, with Block-Type Fuel ; 2nd Rept: An Ana

Design data needs modular high-temperature gas-cooled reactor. Revision 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1987-03-01

The Design Data Needs (DDNs) provide summary statements for program management, of the designer`s need for experimental data to confirm or validate assumptions made in the design. These assumptions were developed using the Integrated Approach and are tabulated in the Functional Analysis Report. These assumptions were also necessary in the analyses or trade studies (A/TS) to develop selections of hardware design or design requirements. Each DDN includes statements providing traceability to the function and the associated assumption that requires the need.

High temperature UF6 RF plasma experiments applicable to uranium plasma core reactors

NASA Technical Reports Server (NTRS)

Roman, W. C.

1979-01-01

An investigation was conducted using a 1.2 MW RF induction heater facility to aid in developing the technology necessary for designing a self critical fissioning uranium plasma core reactor. Pure, high temperature uranium hexafluoride (UF6) was injected into an argon fluid mechanically confined, steady state, RF heated plasma while employing different exhaust systems and diagnostic techniques to simulate and investigate some potential characteristics of uranium plasma core nuclear reactors. The development of techniques and equipment for fluid mechanical confinement of RF heated uranium plasmas with a high density of uranium vapor within the plasma, while simultaneously minimizing deposition of uranium and uranium compounds on the test chamber peripheral wall, endwall surfaces, and primary exhaust ducts, is discussed. The material tests and handling techniques suitable for use with high temperature, high pressure, gaseous UF6 are described and the development of complementary diagnostic instrumentation and measurement techniques to characterize the uranium plasma, effluent exhaust gases, and residue deposited on the test chamber and exhaust system components is reported.

Effect of Temperature on the Desorption of Lithium from Molybdenum(110) Surfaces: Implications for Fusion Reactor First Wall Materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, Mohan; Roszell, John; Scoullos, Emanuel V.

2016-03-30

Determining the strength of Li binding to Mo is critical to assessing the survivability of Li as a potential first wall material in fusion reactors. Here, we present the results of a joint experimental and theoretical investigation into how Li desorbs from Mo(110) surfaces, based on what can be deduced from temperature-programmed desorption measurements and density functional theory (DFT). Li desorption peaks measured at temperatures ranging from 711 K (1 monolayer, ML) to 1030 K (0.04 ML), with corresponding desorption onsets from 489 to 878 K, follow a trend similar to predicted Gibbs free energies for Li adsorption. Bader chargemore » analysis of DFT densities reveals that repulsive forces between neighboring positively charged Li atoms increase with coverage and thus reduce the bond strength between Mo and Li, thereby lowering the desorption temperature as the coverage increases. In addition, DFT predicts that Li desorbs at higher temperatures from a surface with vacancies than from a perfect surface, offering an explanation for the anomalously high desorption temperatures for the last Li to desorb from Mo(110). Analysis of simulated local densities of states indicates that the stronger binding to the defective surface is correlated with enhanced interaction between Li and Mo, involving the Li 2s electrons and not only the Mo 4d electrons as in the case of the pristine surface, but also the Mo 5s electrons in the case with surface vacancies. We suggest that steps and kinks present on the Mo(110) surface behave similarly and contribute to the high desorption temperatures. These findings imply that roughened Mo surfaces may strengthen Li film adhesion at higher temperatures.« less

Magnetic nuclear core restraint and control

DOEpatents

Cooper, Martin H.

1979-01-01

A lateral restraint and control system for a nuclear reactor core adaptable to provide an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit are composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased a given amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction.

Magnetic nuclear core restraint and control

DOEpatents

Cooper, Martin H.

1978-01-01

A lateral restraint and control system for a nuclear reactor core adaptable to provide an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit are composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased a given amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction.

Materials, Turbomachinery and Heat Exchangers for Supercritical CO2 Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Anderson, Mark; Nellis, Greg; Corradini, Michael

2012-10-19

The objective of this project is to produce the necessary data to evaluate the performance of the supercritical carbon dioxide cycle. The activities include a study of materials compatibility of various alloys at high temperatures, the heat transfer and pressure drop in compact heat exchanger units, and turbomachinery issues, primarily leakage rates through dynamic seals. This experimental work will serve as a test bed for model development and design calculations, and will help define further tests necessary to develop high-efficiency power conversion cycles for use on a variety of reactor designs, including the sodium fast reactor (SFR) and very high-temperaturemore » gas reactor (VHTR). The research will be broken into three separate tasks. The first task deals with the analysis of materials related to the high-temperature S-CO{sub 2} Brayton cycle. The most taxing materials issues with regard to the cycle are associated with the high temperatures in the reactor side heat exchanger and in the high-temperature turbine. The system could experience pressures as high as 20MPa and temperatures as high as 650°C. The second task deals with optimization of the heat exchangers required by the S-CO{sub 2} cycle; the S-CO{sub 2} flow passages in these heat exchangers are required whether the cycle is coupled with a VHTR or an SFR. At least three heat exchangers will be required: the pre-cooler before compression, the recuperator, and the heat exchanger that interfaces with the reactor coolant. Each of these heat exchangers is unique and must be optimized separately. The most challenging heat exchanger is likely the pre-cooler, as there is only about a 40°C temperature change but it operates close to the CO{sub 2} critical point, therefore inducing substantial changes in properties. The proposed research will focus on this most challenging component. The third task examines seal leakage through various dynamic seal designs under the conditions expected in the S-CO{sub 2} cycle, including supercritical, choked, and two-phase flow conditions.« less

Proceedings of the 1994 international meeting on reduced enrichment for research and test reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1997-08-01

This meeting brought together participants in the international effort to minimize and eventually eliminate the use of highly enriched uranium in civilian nuclear programs. Papers cover the following topics: National programs; fuel cycle; nuclear fuels; analyses; advanced reactors; and reactor conversions. Selected papers have been indexed separately for inclusion to the Energy Science and Technology Database.

Neutron radiation characteristics of the IVth generation reactor spent fuel

NASA Astrophysics Data System (ADS)

Bedenko, Sergey; Shamanin, Igor; Grachev, Victor; Knyshev, Vladimir; Ukrainets, Olesya; Zorkin, Andrey

2018-03-01

Exploitation of nuclear power plants as well as construction of new generation reactors lead to great accumulation of spent fuel in interim storage facilities at nuclear power plants, and in spent fuel «wet» and «dry» long-term storages. Consequently, handling the fuel needs more attention. The paper is focused on the creation of an efficient computational model used for developing the procedures and regulations of spent nuclear fuel handling in nuclear fuel cycle of the new generation reactor. A Thorium High-temperature Gas-Cooled Reactor Unit (HGTRU, Russia) was used as an object for numerical research. Fuel isotopic composition of HGTRU was calculated using the verified code of the MCU-5 program. The analysis of alpha emitters and neutron radiation sources was made. The neutron yield resulting from (α,n)-reactions and at spontaneous fission was calculated. In this work it has been shown that contribution of (α,n)-neutrons is insignificant in case of such (Th,Pu)-fuel composition and HGTRU operation mode, and integral neutron yield can be approximated by the Watt spectral function. Spectral and standardized neutron distributions were achieved by approximation of the list of high-precision nuclear data. The distribution functions were prepared in group and continuous form for further use in calculations according to MNCP, MCU, and SCALE.

Synthesis of MgB2 at Low Temperature and Autogenous Pressure

PubMed Central

Mackinnon, Ian D. R.; Winnett, Abigail; Alarco, Jose A.; Talbot, Peter C.

2014-01-01

High quality, micron-sized interpenetrating grains of MgB2, with high density, are produced at low temperatures (~420 °C < T < ~500 °C) under autogenous pressure by pre-mixing Mg powder and NaBH4 and heating in an Inconel 601 alloy reactor for 5–15 h. Optimum production of MgB2, with yields greater than 75%, occurs for autogenous pressure in the range 1.0 MPa to 2.0 MPa, with the reactor at ~500 °C. Autogenous pressure is induced by the decomposition of NaBH4 in the presence of Mg and/or other Mg-based compounds. The morphology, transition temperature and magnetic properties of MgB2 are dependent on the heating regime. Significant improvement in physical properties accrues when the reactor temperature is held at 250 °C for >20 min prior to a hold at 500 °C. PMID:28788656

Solar-thermal reaction processing

DOEpatents

Weimer, Alan W; Dahl, Jaimee K; Lewandowski, Allan A; Bingham, Carl; Raska Buechler, Karen J; Grothe, Willy

2014-03-18

In an embodiment, a method of conducting a high temperature chemical reaction that produces hydrogen or synthesis gas is described. The high temperature chemical reaction is conducted in a reactor having at least two reactor shells, including an inner shell and an outer shell. Heat absorbing particles are included in a gas stream flowing in the inner shell. The reactor is heated at least in part by a source of concentrated sunlight. The inner shell is heated by the concentrated sunlight. The inner shell re-radiates from the inner wall and heats the heat absorbing particles in the gas stream flowing through the inner shell, and heat transfers from the heat absorbing particles to the first gas stream, thereby heating the reactants in the gas stream to a sufficiently high temperature so that the first gas stream undergoes the desired reaction(s), thereby producing hydrogen or synthesis gas in the gas stream.

Decommissioning of the Dragon High Temperature Reactor (HTR) Located at the Former United Kingdom Atomic Energy Authority (UKAEA) Research Site at Winfrith - 13180

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Anthony A.

2013-07-01

The Dragon Reactor was constructed at the United Kingdom Atomic Energy Research Establishment at Winfrith in Dorset through the late 1950's and into the early 1960's. It was a High Temperature Gas Cooled Reactor (HTR) with helium gas coolant and graphite moderation. It operated as a fuel testing and demonstration reactor at up to 20 MW (Thermal) from 1964 until 1975, when international funding for this project was terminated. The fuel was removed from the core in 1976 and the reactor was put into Safestore. To meet the UK's Nuclear Decommissioning Authority (NDA) objective to 'drive hazard reduction' [1] itmore » is necessary to decommission and remediate all the Research Sites Restoration Ltd (RSRL) facilities. This includes the Dragon Reactor where the activated core, pressure vessel and control rods and the contaminated primary circuit (including a {sup 90}Sr source) still remain. It is essential to remove these hazards at the appropriate time and return the area occupied by the reactor to a safe condition. (author)« less

Closed Brayton Cycle power system with a high temperature pellet bed reactor heat source for NEP applications

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.; El-Genk, Mohamed S.; Harper, William B., Jr.

1992-01-01

Capitalizing on past and future development of high temperature gas reactor (HTGR) technology, a low mass 15 MWe closed gas turbine cycle power system using a pellet bed reactor heating helium working fluid is proposed for Nuclear Electric Propulsion (NEP) applications. Although the design of this directly coupled system architecture, comprising the reactor/power system/space radiator subsystems, is presented in conceptual form, sufficient detail is included to permit an assessment of overall system performance and mass. Furthermore, an attempt is made to show how tailoring of the main subsystem design characteristics can be utilized to achieve synergistic system level advantages that can lead to improved reliability and enhanced system life while reducing the number of parasitic load driven peripheral subsystems.

A thermodynamic approach for advanced fuels of gas-cooled reactors

NASA Astrophysics Data System (ADS)

Guéneau, C.; Chatain, S.; Gossé, S.; Rado, C.; Rapaud, O.; Lechelle, J.; Dumas, J. C.; Chatillon, C.

2005-09-01

For both high temperature reactor (HTR) and gas cooled fast reactor (GFR) systems, the high operating temperature in normal and accidental conditions necessitates the assessment of the thermodynamic data and associated phase diagrams for the complex system constituted of the fuel kernel, the inert materials and the fission products. A classical CALPHAD approach, coupling experiments and thermodynamic calculations, is proposed. Some examples of studies are presented leading with the CO and CO 2 gas formation during the chemical interaction of [UO 2± x/C] in the HTR particle, and the chemical compatibility of the couples [UN/SiC], [(U, Pu)N/SiC], [(U, Pu)N/TiN] for the GFR system. A project of constitution of a thermodynamic database for advanced fuels of gas-cooled reactors is proposed.

Advanced sample environments for in situ neutron diffraction studies of nuclear materials

NASA Astrophysics Data System (ADS)

Reiche, Helmut Matthias

Generation IV nuclear reactor concepts, such as the supercritical-water-cooled nuclear reactor (SCWR), are actively researched internationally. Operating conditions above the critical point of water (374°C, 22.1 MPa) and fuel core temperature that potentially exceed 1850°C put a high demand on the surrounding materials. For their safe application, it is essential to characterize and understand the material properties on an atomic scale such as crystal structure and grain orientation (texture) changes as a function of temperature and stress. This permits the refinement of models predicting the macroscopic behavior of the material. Neutron diffraction is a powerful tool in characterizing such crystallographic properties due to their deep penetration depth into condensed matter. This leads to the ability to study bulk material properties, as opposed to surface effects, and allows for complex sample environments to study e.g. the individual contributions of thermo-mechanical processing steps during manufacturing, operating or accident scenarios. I present three sample environments for in situ neutron diffraction studies that provide such crystallographic information and have been successfully commissioned and integrated into the user program of the High Pressure -- Preferred Orientation (HIPPO) diffractometer at the Los Alamos Neutron Science Center (LANSCE) user facility. I adapted a sample changer for reliable and fast automated texture measurements of multiple specimens. I built a creep furnace combining a 2700 N load frame with a resistive vanadium furnace, capable of temperatures up to 1000°C, and manipulated by a pair of synchronized rotation stages. This combination allows following deformation and temperature dependent texture and strain evolutions in situ. Utilizing the presented sample changer and creep furnace we studied pressure tubes made of Zr-2.5wt%Nb currently employed in CANDURTM nuclear reactors and proposed for future SCWRs, acting as the primary containment vessel of high temperature heavy water (D2O) inside the reactor core. The measured sample texture shows that upon traversing the phase transition, which proceeded according to the Burger orientation relationship, variant selection occurred during heating and cooling of the zirconium alloy. Experimental results of lattice strains depending on the crystallographic orientation can be used to calculate strain pole figures which grant insight into the three-dimensional mechanical response of a polycrystalline aggregate and represent an extremely powerful material model validation tool. Lastly, I developed a resistive graphite high-temperature furnace with sample motion for in situ crystal structure and texture measurements of nuclear materials at steady-state temperatures up to at least 2200°C. This permits in situ observation of e.g. phase transitions and coefficients of thermal expansion, as well as phase formation and texture development during solidification. Utilizing this apparatus, I investigated the carbothermic reduction of UO2 nanopowder forming uranium carbide, a promising Generation IV reactor fuel. The onset of the UO2 + 2C → UC + CO2 reaction was observed at 1440°C with the bulk portion being complete at 1500°C. I describe the novel synthesis for this nanoparticle UO2 powder, which closely imitates observed nano grains in partially burnt reactor fuels. Of the three opposing structure models reported for the non-quenchable cubic UC2 phase, stable between 1769°C and 2560°C, the NaCl-type structure according to Bowman is found to be correct. This is deemed major progress as the CaF2-type structure was used for recent thermal modeling of safety critical factors in nuclear reactors. A temperature dependent increase in density due to carbon diffusion has been observed and quantified. I provide first experimental data of an unspecified, reversible order-disorder transition in this delta-phase with its onset at ˜1800°C which is likely due to rotating C2 molecules in the sublattice.

Recycling of hazardous solid waste material using high-temperature solar process heat. 2. Reactor design and experimentation.

PubMed

Schaffner, Beatrice; Meier, Anton; Wuillemin, Daniel; Hoffelner, Wolfgang; Steinfeld, Aldo

2003-01-01

A novel high-temperature solar chemical reactor is proposed for the thermal recycling of hazardous solid waste material using concentrated solar power. It features two cavities in series, with the inner one functioning as the solar absorber and the outer one functioning as the reaction chamber. The solar reactor can handle thermochemical processes at temperatures above 1,300 K involving multiphases and controlled atmospheres. It further allows for batch or continuous mode of operation and for easy adjustment of the residence time of the reactants to match the kinetics of the reaction. A 10-kW solar reactor prototype was designed and tested for the carbothermic reduction of electric arc furnace dusts (EAFD). The reactor was subjected to mean solar flux intensities of 2,000 kW m(-2) and operated in both batch and continuous mode within the temperature range of 1,120-1,400 K. Extraction of over 90% of the toxic compounds originally contained in the EAFD was achieved while the condensable products of the off-gas contained mainly Zn, Pb, and Cl. The use of concentrated solar energy as the source of process heat offers the possibility of converting hazardous solid waste material into valuable commodities for processes in closed and sustainable material cycles.

Synfuels from fusion: using the tandem mirror reactor and a thermochemical cycle to produce hydrogen

DOE Office of Scientific and Technical Information (OSTI.GOV)

Werner, R.W.

1982-11-01

This study is concerned with the following area: (1) the tandem mirror reactor and its physics; (2) energy balance; (3) the lithium oxide canister blanket system; (4) high-temperature blanket; (5) energy transport system-reactor to process; (6) thermochemical hydrogen processes; (7) interfacing the GA cycle; (8) matching power and temperature demands; (9) preliminary cost estimates; (10) synfuels beyond hydrogen; and (11) thermodynamics of the H/sub 2/SO/sub 4/-H/sub 2/O system. (MOW)

Advanced Electron Microscopy and Micro analytical technique development and application for Irradiated TRISO Coated Particles from the AGR-1 Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Rooyen, Isabella Johanna; Lillo, Thomas Martin; Wen, Haiming

2017-01-01

A series of up to seven irradiation experiments are planned for the Advanced Gas Reactor (AGR) Fuel Development and Quantification Program, with irradiation completed at the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for the first experiment (i.e., AGR-1) in November 2009 for an effective 620 full power days. The objective of the AGR-1 experiment was primarily to provide lessons learned on the multi-capsule test train design and to provide early data on fuel performance for use in fuel fabrication process development and post-irradiation safety testing data at high temperatures. This report describes the advanced microscopy and micro-analysismore » results on selected AGR-1 coated particles.« less

Nodal Diffusion Burnable Poison Treatment for Prismatic Reactor Cores

DOE Office of Scientific and Technical Information (OSTI.GOV)

A. M. Ougouag; R. M. Ferrer

2010-10-01

The prismatic block version of the High Temperature Reactor (HTR) considered as a candidate Very High Temperature Reactor (VHTR)design may use burnable poison pins in locations at some corners of the fuel blocks (i.e., assembly equivalent structures). The presence of any highly absorbing materials, such as these burnable poisons, within fuel blocks for hexagonal geometry, graphite-moderated High Temperature Reactors (HTRs) causes a local inter-block flux depression that most nodal diffusion-based method have failed to properly model or otherwise represent. The location of these burnable poisons near vertices results in an asymmetry in the morphology of the assemblies (or blocks). Hencemore » the resulting inadequacy of traditional homogenization methods, as these “spread” the actually local effect of the burnable poisons throughout the assembly. Furthermore, the actual effect of the burnable poison is primarily local with influence in its immediate vicinity, which happens to include a small region within the same assembly as well as similar regions in the adjacent assemblies. Traditional homogenization methods miss this artifact entirely. This paper presents a novel method for treating the local effect of the burnable poison explicitly in the context of a modern nodal method.« less

Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pocoima, CA; Benander, Robert E [Pacoima, CA

2010-02-23

Methods for manufacturing porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's). Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, a thin coating of nuclear fuel may be deposited inside of a highly porous skeletal structure made, for example, of reticulated vitreous carbon foam.

Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

2011-03-01

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

2013-09-03

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Analytical modeling of helium turbomachinery using FORTRAN 77

NASA Astrophysics Data System (ADS)

Balaji, Purushotham

Advanced Generation IV modular reactors, including Very High Temperature Reactors (VHTRs), utilize helium as the working fluid, with a potential for high efficiency power production utilizing helium turbomachinery. Helium is chemically inert and nonradioactive which makes the gas ideal for a nuclear power-plant environment where radioactive leaks are a high concern. These properties of helium gas helps to increase the safety features as well as to decrease the aging process of plant components. The lack of sufficient helium turbomachinery data has made it difficult to study the vital role played by the gas turbine components of these VHTR powered cycles. Therefore, this research work focuses on predicting the performance of helium compressors. A FORTRAN77 program is developed to simulate helium compressor operation, including surge line prediction. The resulting design point and off design performance data can be used to develop compressor map files readable by Numerical Propulsion Simulation Software (NPSS). This multi-physics simulation software that was developed for propulsion system analysis has found applications in simulating power-plant cycles.

A Plasma Reactor for the Synthesis of High-Temperature Materials: Electro Thermal, Processing and Service Life Characteristics

NASA Astrophysics Data System (ADS)

Galevskiy, G. V.; Rudneva, V. V.; Galevskiy, S. G.; Tomas, K. I.; Zubkov, M. S.

2016-08-01

The three-jet direct-flow plasma reactor with a channel diameter of 0.054 m was studied in terms of service life, thermal, technical, and functional capabilities. It was established that the near-optimal combination of thermal efficiency, required specific enthalpy of the plasma-forming gas and its mass flow rate is achieved at a reactor power of 150 kW. The bulk temperature of plasma flow over the rector of 12 gauges long varies within 5500÷3200 K and the wall temperature within 1900÷850 K, when a cylinder from zirconium dioxide of 0.005 m thick is used to thermally insulate the reactor. The specific electric power reaches a high of 1214 MW/m3. The rated service life of electrodes is 4700 hours for a copper anode and 111 hours for a tungsten cathode. The projected contamination of carbides and borides with elec-trode-erosion products doesn't exceed 0.0001% of copper and 0.00002% of tungsten.

An atmospheric pressure high-temperature laminar flow reactor for investigation of combustion and related gas phase reaction systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oßwald, Patrick; Köhler, Markus

A new high-temperature flow reactor experiment utilizing the powerful molecular beam mass spectrometry (MBMS) technique for detailed observation of gas phase kinetics in reacting flows is presented. The reactor design provides a consequent extension of the experimental portfolio of validation experiments for combustion reaction kinetics. Temperatures up to 1800 K are applicable by three individually controlled temperature zones with this atmospheric pressure flow reactor. Detailed speciation data are obtained using the sensitive MBMS technique, providing in situ access to almost all chemical species involved in the combustion process, including highly reactive species such as radicals. Strategies for quantifying the experimentalmore » data are presented alongside a careful analysis of the characterization of the experimental boundary conditions to enable precise numeric reproduction of the experimental results. The general capabilities of this new analytical tool for the investigation of reacting flows are demonstrated for a selected range of conditions, fuels, and applications. A detailed dataset for the well-known gaseous fuels, methane and ethylene, is provided and used to verify the experimental approach. Furthermore, application for liquid fuels and fuel components important for technical combustors like gas turbines and engines is demonstrated. Besides the detailed investigation of novel fuels and fuel components, the wide range of operation conditions gives access to extended combustion topics, such as super rich conditions at high temperature important for gasification processes, or the peroxy chemistry governing the low temperature oxidation regime. These demonstrations are accompanied by a first kinetic modeling approach, examining the opportunities for model validation purposes.« less

Biological oxidation of hydrogen sulfide in mineral media using a biofilm airlift suspension reactor.

PubMed

Moghanloo, G M Mojarrad; Fatehifar, E; Saedy, S; Aghaeifar, Z; Abbasnezhad, H

2010-11-01

Hydrogen sulfide (H(2)S) removal in mineral media using Thiobacillus thioparus TK-1 in a biofilm airlift suspension reactor (BAS) was investigated to evaluate the relationship between biofilm formation and changes in inlet loading rates. Aqueous sodium sulfide was fed as the substrate into the continuous BAS-reactor. The reactor was operated at a constant temperature of 30 degrees C and a pH of 7, the optimal temperature and pH for biomass growth. The startup of the reactor was performed with basalt carrier material. Optimal treatment performance was obtained at a loading rate of 4.8 mol S(2-) m(-3) h(-1) at a conversion efficiency as high as 100%. The main product of H(2)S oxidation in the BAS-reactor was sulfate because of high oxygen concentrations in the airlift reactor. The maximum sulfide oxidation rate was 6.7 mol S(2-) m(-3) h(-1) at a hydraulic residence time of 3.3 h in the mineral medium. The data showed that the BAS-reactor with this microorganism can be used for sulfide removal from industrial effluent. Copyright 2010 Elsevier Ltd. All rights reserved.

Thermodynamic Analysis of the Use a Chemical Heat Pump to Link a Supercritical Water-Cooled Nuclear Reactor and a Thermochemical Water-Splitting Cycle for Hydrogen Production

NASA Astrophysics Data System (ADS)

Granovskii, Mikhail; Dincer, Ibrahim; Rosen, Marc A.; Pioro, Igor

Increases in the power generation efficiency of nuclear power plants (NPPs) are mainly limited by the permissible temperatures in nuclear reactors and the corresponding temperatures and pressures of the coolants in reactors. Coolant parameters are limited by the corrosion rates of materials and nuclear-reactor safety constraints. The advanced construction materials for the next generation of CANDU reactors, which employ supercritical water (SCW) as a coolant and heat carrier, permit improved “steam” parameters (outlet temperatures up to 625°C and pressures of about 25 MPa). An increase in the temperature of steam allows it to be utilized in thermochemical water splitting cycles to produce hydrogen. These methods are considered by many to be among the most efficient ways to produce hydrogen from water and to have advantages over traditional low-temperature water electrolysis. However, even lower temperature water splitting cycles (Cu-Cl, UT-3, etc.) require an intensive heat supply at temperatures higher than 550-600°C. A sufficient increase in the heat transfer from the nuclear reactor to a thermochemical water splitting cycle, without jeopardizing nuclear reactor safety, might be effectively achieved by application of a heat pump, which increases the temperature of the heat supplied by virtue of a cyclic process driven by mechanical or electrical work. Here, a high-temperature chemical heat pump, which employs the reversible catalytic methane conversion reaction, is proposed. The reaction shift from exothermic to endothermic and back is achieved by a change of the steam concentration in the reaction mixture. This heat pump, coupled with the second steam cycle of a SCW nuclear power generation plant on one side and a thermochemical water splitting cycle on the other, increases the temperature of the “nuclear” heat and, consequently, the intensity of heat transfer into the water splitting cycle. A comparative preliminary thermodynamic analysis is conducted of the combined system comprising a SCW nuclear power generation plant and a chemical heat pump, which provides high-temperature heat to a thermochemical water splitting cycle for hydrogen production. It is concluded that the proposed chemical heat pump permits the utilization efficiency of nuclear energy to be improved by at least 2% without jeopardizing nuclear reactor safety. Based on this analysis, further research appears to be merited on the proposed advanced design of a nuclear power generation plant combined with a chemical heat pump, and implementation in appropriate applications seems worthwhile.

Spectral emissivity of candidate alloys for very high temperature reactors in high temperature air environment

NASA Astrophysics Data System (ADS)

Cao, G.; Weber, S. J.; Martin, S. O.; Sridharan, K.; Anderson, M. H.; Allen, T. R.

2013-10-01

Emissivity measurements for candidate alloys for very high temperature reactors were carried out in a custom-built experimental facility, capable of both efficient and reliable measurements of spectral emissivities of multiple samples at high temperatures. The alloys studied include 304 and 316 austenitic stainless steels, Alloy 617, and SA508 ferritic steel. The oxidation of alloys plays an important role in dictating emissivity values. The higher chromium content of 304 and 316 austenitic stainless steels, and Alloy 617 results in an oxide layer only of sub-micron thickness even at 700 °C and consequently the emissivity of these alloys remains low. In contrast, the low alloy SA508 ferritic steel which contains no chromium develops a thicker oxide layer, and consequently exhibits higher emissivity values.

Performance assessment of low pressure nuclear thermal propulsion

NASA Technical Reports Server (NTRS)

Gerrish, Harrold P., Jr.; Doughty, Glen E.

1993-01-01

An increase in Isp for nuclear thermal propulsion systems is desirable for reducing the propellant requirements and cost of future applications, such as the Mars Transfer Vehicle. Several previous design studies have suggested that the Isp could be increased substantially with hydrogen dissociation/recombination. Hydrogen molecules (H2), at high temperatures and low pressures, will dissociate to monatomic hydrogen (H). The reverse process (i.e., formation of H2 from H) is exothermic. The exothermic energy in a nozzle increases the kinetic energy and therefore, increases the Isp. The low pressure nuclear thermal propulsion system (LPNTP) system is expected to maximize the hydrogen dissociation/recombination and Isp by operating at high chamber temperatures and low chamber pressures. The process involves hydrogen flow through a high temperature, low pressure fission reactor, and out a nozzle. The high temperature (approximately 3000 K) of the hydrogen in the reactor is limited by the temperature limits of the reactor material. The minimum chamber pressure is about 1 atm because lower pressures decrease the engines thrust to weight ratio below acceptable limits. This study assumes that hydrogen leaves the reactor and enters the nozzle at the 3000 K equilibrium dissociation level. Hydrogen dissociation in the reactor does not affect LPNTP performance like dissociation in traditional chemical propulsion systems, because energy from the reactor resupplies energy lost due to hydrogen dissociation. Recombination takes place in the nozzle due primarily to a drop in temperature as the Mach number increases. However, as the Mach number increases beyond the nozzle throat, the static pressure and density of the flow decreases and minimizes the recombination. The ideal LPNTP Isp at 3000 K and 10 psia is 1160 seconds due to the added energy from fast recombination rates. The actual Isp depends on the finite kinetic reaction rates which affect the amount of monatomic hydrogen recombination before the flow exits the nozzle. A LPNTP system has other technical issues (e.g. flow instability and two-phase flow) besides hydrogen dissociation/recombination which affect the systems practicality. In this study, only the effects of hydrogen dissociation/recombination are examined.

Multi-cycle operation of enhanced biological phosphorus removal (EBPR) with different carbon sources under high temperature.

PubMed

Shen, Nan; Chen, Yun; Zhou, Yan

2017-05-01

Many studies reported that it is challenging to apply enhanced biological phosphorus removal (EBPR) process at high temperature. Glycogen accumulating organisms (GAOs) could easily gain their dominance over poly-phosphate accumulating organisms (PAOs) when the operating temperature was in the range of 25 °C-30 °C. However, a few successful EBPR processes operated at high temperature have been reported recently. This study aimed to have an in-depth understanding on the impact of feeding strategy and carbon source types on EBPR performance in tropical climate. P-removal performance of two EBPR systems was monitored through tracking effluent quality and cyclic studies. The results confirmed that EBPR was successfully obtained and maintained at high temperature with a multi-cycle strategy. More stable performance was observed with acetate as the sole carbon source compared to propionate. Stoichiometric ratios of phosphorus and carbon transformation during both anaerobic and aerobic phases were higher at high temperature than low temperature (20±1 °C) except anaerobic PHA/C ratios within most of the sub-cycles. Furthermore, the fractions of PHA and glycogen in biomass were lower compared with one-cycle pulse feed operation. The microbial community structure was more stable in acetate-fed sequencing batch reactor (C2-SBR) than that in propionate-fed reactor (C3-SBR). Accumulibacter Clade IIC was found to be highly abundant in both reactors. Copyright © 2017 Elsevier Ltd. All rights reserved.

Development of a Model and Computer Code to Describe Solar Grade Silicon Production Processes

NASA Technical Reports Server (NTRS)

Srivastava, R.; Gould, R. K.

1979-01-01

The program aims at developing mathematical models and computer codes based on these models, which allow prediction of the product distribution in chemical reactors for converting gaseous silicon compounds to condensed-phase silicon. The major interest is in collecting silicon as a liquid on the reactor walls and other collection surfaces. Two reactor systems are of major interest, a SiCl4/Na reactor in which Si(l) is collected on the flow tube reactor walls and a reactor in which Si(l) droplets formed by the SiCl4/Na reaction are collected by a jet impingement method. During this quarter the following tasks were accomplished: (1) particle deposition routines were added to the boundary layer code; and (2) Si droplet sizes in SiCl4/Na reactors at temperatures below the dew point of Si are being calculated.

A high temperature drop-tube and packed-bed solar reactor for continuous biomass gasification

NASA Astrophysics Data System (ADS)

Bellouard, Quentin; Abanades, Stéphane; Rodat, Sylvain; Dupassieux, Nathalie

2017-06-01

Biomass gasification is an attractive process to produce high-value syngas. Utilization of concentrated solar energy as the heat source for driving reactions increases the energy conversion efficiency, saves biomass resource, and eliminates the needs for gas cleaning and separation. A high-temperature tubular solar reactor combining drop tube and packed bed concepts was used for continuous solar-driven gasification of biomass. This 1 kW reactor was experimentally tested with biomass feeding under real solar irradiation conditions at the focus of a 2 m-diameter parabolic solar concentrator. Experiments were conducted at temperatures ranging from 1000°C to 1400°C using wood composed of a mix of pine and spruce (bark included) as biomass feedstock. The aim of this study was to demonstrate the feasibility of syngas production in this reactor concept and to prove the reliability of continuous biomass gasification processing using solar energy. The study first consisted of a parametric study of the gasification conditions to obtain an optimal gas yield. The influence of temperature and oxidizing agent (H2O or CO2) on the product gas composition was investigated. The study then focused on solar gasification during continuous biomass particle injection for demonstrating the feasibility of a continuous process. Regarding the energy conversion efficiency of the lab scale reactor, energy upgrade factor of 1.21 and solar-to-fuel thermochemical efficiency up to 28% were achieved using wood heated up to 1400°C.

FALCON reactor-pumped laser description and program overview

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1989-12-01

The FALCON (Fission Activated Laser CONcept) reactor-pumped laser program at Sandia National Laboratories is examining the feasibility of high-power systems pumped directly by the energy from a nuclear reactor. In this concept we use the highly energetic fission fragments from neutron induced fission to excite a large volume laser medium. This technology has the potential to scale to extremely large optical power outputs in a primarily self-powered device. A laser system of this type could also be relatively compact and capable of long run times without refueling.

Process optimization of an auger pyrolyzer with heat carrier using response surface methodology.

PubMed

Brown, J N; Brown, R C

2012-01-01

A 1 kg/h auger reactor utilizing mechanical mixing of steel shot heat carrier was used to pyrolyze red oak wood biomass. Response surface methodology was employed using a circumscribed central composite design of experiments to optimize the system. Factors investigated were: heat carrier inlet temperature and mass flow rate, rotational speed of screws in the reactor, and volumetric flow rate of sweep gas. Conditions for maximum bio-oil and minimum char yields were high flow rate of sweep gas (3.5 standard L/min), high heat carrier temperature (∼600 °C), high auger speeds (63 RPM) and high heat carrier mass flow rates (18 kg/h). Regression models for bio-oil and char yields are described including identification of a novel interaction effect between heat carrier mass flow rate and auger speed. Results suggest that auger reactors, which are rarely described in literature, are well suited for bio-oil production. The reactor achieved liquid yields greater than 73 wt.%. Copyright © 2011 Elsevier Ltd. All rights reserved.

Using SA508/533 for the HTGR Vessel Material

DOE Office of Scientific and Technical Information (OSTI.GOV)

Larry Demick

2012-06-01

This paper examines the influence of High Temperature Gas-cooled Reactor (HTGR) module power rating and normal operating temperatures on the use of SA508/533 material for the HTGR vessel system with emphasis on the calculated times at elevated temperatures approaching or exceeding ASME Code Service Limits (Levels B&C) to which the reactor pressure vessel could be exposed during postulated pressurized and depressurized conduction cooldown events over its design lifetime.

Numerical study of air ingress transition to natural circulation in a high temperature helium loop

DOE Office of Scientific and Technical Information (OSTI.GOV)

Franken, Daniel; Gould, Daniel; Jain, Prashant K.

Here, the generation-IV high temperature gas cooled reactors (HTGRs) are designed with many passive safety features, one of which is the ability to passively remove heat under a loss of coolant accident (LOCA). However, several common reactor designs do not prevent against a large break in the coolant system and may therefore experience a depressurized LOCA. This would lead to air entering into the reactor system via several potential modes of ingress: diffusion, gravity currents, and natural circulation. At the onset of a LOCA, the initial rate of air ingress is expected to be very slow because it is governedmore » by molecular diffusion. However, after several hours, natural circulation would commence, thus, bringing the air into the reactor system at a much higher rate. As a consequence, air ingress would cause the high temperature graphite matrix to oxidize, leading to its thermal degradation and decreased passive heat (decay) removal capability. Therefore, it is essential to understand the transition of air ingress from molecular diffusion to natural circulation in an HTGR system. This paper presents results from a computational fluid dynamics (CFD) model to study the air ingress transition behavior. These results are validated against an h-shaped high temperature helium loop experiment. Details are provided to quantitatively predict the transition time from molecular diffusion to natural circulation.« less

Numerical study of air ingress transition to natural circulation in a high temperature helium loop

DOE PAGES

Franken, Daniel; Gould, Daniel; Jain, Prashant K.; ...

2017-09-21

Here, the generation-IV high temperature gas cooled reactors (HTGRs) are designed with many passive safety features, one of which is the ability to passively remove heat under a loss of coolant accident (LOCA). However, several common reactor designs do not prevent against a large break in the coolant system and may therefore experience a depressurized LOCA. This would lead to air entering into the reactor system via several potential modes of ingress: diffusion, gravity currents, and natural circulation. At the onset of a LOCA, the initial rate of air ingress is expected to be very slow because it is governedmore » by molecular diffusion. However, after several hours, natural circulation would commence, thus, bringing the air into the reactor system at a much higher rate. As a consequence, air ingress would cause the high temperature graphite matrix to oxidize, leading to its thermal degradation and decreased passive heat (decay) removal capability. Therefore, it is essential to understand the transition of air ingress from molecular diffusion to natural circulation in an HTGR system. This paper presents results from a computational fluid dynamics (CFD) model to study the air ingress transition behavior. These results are validated against an h-shaped high temperature helium loop experiment. Details are provided to quantitatively predict the transition time from molecular diffusion to natural circulation.« less

Core Design Characteristics of the Fluoride Salt-Cooled High Temperature Demonstration Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas R; Qualls, A L; Betzler, Benjamin R

2016-01-01

Fluoride salt-cooled high temperature reactors (FHRs) are a promising reactor technology option with significant knowledge gaps to implementation. One potential approach to address those technology gaps is via a small-scale demonstration reactor with the goal of increasing the technology readiness level (TRL) of the overall system for the longer term. The objective of this paper is to outline a notional concept for such a system, and to address how the proposed concept would advance the TRL of FHR concepts. Development of the proposed FHR Demonstration Reactor (DR) will enable commercial FHR deployment through disruptive and rapid technology development and demonstration.more » The FHR DR will close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. Important capabilities that will be demonstrated by building and operating the FHR DR include core design methodologies; fabrication and operation of high temperature reactors; salt procurement, handling, maintenance, and ultimate disposal; salt chemistry control to maximize vessel life; tritium management; heat exchanger performance; pump performance; and reactivity control. The FHR DR is considered part of a broader set of FHR technology development and demonstration efforts, some of which are already underway. Nonreactor test efforts (e.g., heated salt loops or loops using simulant fluids) can demonstrate many technologies necessary for commercial deployment of FHRs. The FHR DR, however, fulfills a crucial role in FHR technology development by advancing the technical maturity and readiness level of the system as a whole.« less

Chemical Technology Division annual technical report, 1990

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1991-05-01

Highlights of the Chemical Technology (CMT) Division's activities during 1990 are presented. In this period, CMT conducted research and development in the following areas: (1) electrochemical technology, including advanced batteries and fuel cells; (2) technology for coal- fired magnetohydrodynamics and fluidized-bed combustion; (3) methods for recovery of energy from municipal waste and techniques for treatment of hazardous organic waste; (4) the reaction of nuclear waste glass and spent fuel under conditions expected for a high-level waste repository; (5) processes for separating and recovering transuranic elements from nuclear waste streams, concentrating plutonium solids in pyrochemical residues by aqueous biphase extraction, andmore » treating natural and process waters contaminated by volatile organic compounds; (6) recovery processes for discharged fuel and the uranium blanket in the Integral Fast Reactor (IFR); (7) processes for removal of actinides in spent fuel from commercial water-cooled nuclear reactors and burnup in IFRs; and (8) physical chemistry of selected materials in environments simulating those of fission and fusion energy systems. The Division also has a program in basic chemistry research in the areas of fluid catalysis for converting small molecules to desired products; materials chemistry for superconducting oxides and associated and ordered solutions at high temperatures; interfacial processes of importance to corrosion science, high-temperature superconductivity, and catalysis; and the geochemical processes responsible for trace-element migration within the earth's crust. The Analytical Chemistry Laboratory in CMT provides a broad range of analytical chemistry support services to the scientific and engineering programs at Argonne National Laboratory (ANL). 66 refs., 69 figs., 6 tabs.« less

A preliminary assessment of the effects of heat flux distribution and penetration on the creep rupture of a reactor vessel lower head

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chu, T.Y.; Bentz, J.; Simpson, R.

1997-02-01

The objective of the Lower Head Failure (LHF) Experiment Program is to experimentally investigate and characterize the failure of the reactor vessel lower head due to thermal and pressure loads under severe accident conditions. The experiment is performed using 1/5-scale models of a typical PWR pressure vessel. Experiments are performed for various internal pressure and imposed heat flux distributions with and without instrumentation guide tube penetrations. The experimental program is complemented by a modest modeling program based on the application of vessel creep rupture codes developed in the TMI Vessel Investigation Project. The first three experiments under the LHF programmore » investigated the creep rupture of simulated reactor pressure vessels without penetrations. The heat flux distributions for the three experiments are uniform (LHF-1), center-peaked (LHF-2), and side-peaked (LHF-3), respectively. For all the experiments, appreciable vessel deformation was observed to initiate at vessel wall temperatures above 900K and the vessel typically failed at approximately 1000K. The size of failure was always observed to be smaller than the heated region. For experiments with non-uniform heat flux distributions, failure typically occurs in the region of peak temperature. A brief discussion of the effect of penetration is also presented.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cappiello, M.; Hobbins, R.; Penny, K.

As part of the Department of Energy Advanced Fuel Cycle program, a series of fuels development irradiation tests have been performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. These tests are providing excellent data for advanced fuels development. The program is focused on the transmutation of higher actinides which best can be accomplished in a sodium-cooled fast reactor. Because a fast test reactor is no longer available in the US, a special test vehicle is used to achieve near-prototypic fast reactor conditions (neutron spectra and temperature) for use in ATR (a water-cooled thermal reactor). As partmore » of the testing program, there were many successful tests of advanced fuels including metals and ceramics. Recently however, there have been three experimental campaigns using metal fuels that experienced failure during irradiation. At the request of the program, an independent review committee was convened to review the post-test analyses performed by the fuels development team, to assess the conclusions of the team for the cause of the failures, to assess the adequacy and completeness of the analyses, to identify issues that were missed, and to make recommendations for improvements in the design and operation of future tests. Although there is some difference of opinion, the review committee largely agreed with the conclusions of the fuel development team regarding the cause of the failures. For the most part, the analyses that support the conclusions are sufficient.« less

Summary of Thermocouple Performance During Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor and Out-of-Pile Thermocouple Testing in Support of Such Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

A. J. Palmer; DC Haggard; J. W. Herter

High temperature gas reactor experiments create unique challenges for thermocouple based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B); and tungsten-rhenium thermocouples (Types C and W). For lower temperature applications, previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type Nmore » thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of these Nickel based thermocouples is limited when the temperature exceeds 1000°C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past ten years, three long-term Advanced Gas Reactor (AGR) experiments have been conducted with measured temperatures ranging from 700oC – 1200oC. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out of pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150oC and 1200oC for 2000 hours at each temperature, followed by 200 hours at 1250oC, and 200 hours at 1300oC. The standard Type N design utilizes high purity crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including Haynes 214 alloy sheath, spinel (MgAl2O4) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard fired alumina insulation and molybdenum sheath. The most current version of the High Temperature Irradiation Resistant Thermocouple (HTIR-TC) based on molybdenum/niobium alloys, and developed at Idaho National Laboratory, was also tested.« less

Summary of thermocouple performance during advanced gas reactor fuel irradiation experiments in the advanced test reactor and out-of-pile thermocouple testing in support of such experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Palmer, A. J.; Haggard, DC; Herter, J. W.

High temperature gas reactor experiments create unique challenges for thermocouple-based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time-dependent change in composition and, as a consequence, a time-dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C). For lower temperature applications, previous experiences with Type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly, Type N thermocouples are expected to bemore » only slightly affected by neutron fluence. Currently, the use of these nickel-based thermocouples is limited when the temperature exceeds 1000 deg. C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past 10 years, three long-term Advanced Gas Reactor experiments have been conducted with measured temperatures ranging from 700 deg. C - 1200 deg. C. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out-of-pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150 deg. C and 1200 deg. C for 2,000 hours at each temperature, followed by 200 hours at 1250 deg. C and 200 hours at 1300 deg. C. The standard Type N design utilizes high purity, crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including a Haynes 214 alloy sheath, spinel (MgAl{sub 2}O{sub 4}) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard-fired alumina insulation and a molybdenum sheath. The most current version of the High Temperature Irradiation Resistant Thermocouple, based on molybdenum/niobium alloys and developed at Idaho National Laboratory, was also tested. (authors)« less

A combined gas cooled nuclear reactor and fuel cell cycle

NASA Astrophysics Data System (ADS)

Palmer, David J.

Rising oil costs, global warming, national security concerns, economic concerns and escalating energy demands are forcing the engineering communities to explore methods to address these concerns. It is the intention of this thesis to offer a proposal for a novel design of a combined cycle, an advanced nuclear helium reactor/solid oxide fuel cell (SOFC) plant that will help to mitigate some of the above concerns. Moreover, the adoption of this proposal may help to reinvigorate the Nuclear Power industry while providing a practical method to foster the development of a hydrogen economy. Specifically, this thesis concentrates on the importance of the U.S. Nuclear Navy adopting this novel design for its nuclear electric vessels of the future with discussion on efficiency and thermodynamic performance characteristics related to the combined cycle. Thus, the goals and objectives are to develop an innovative combined cycle that provides a solution to the stated concerns and show that it provides superior performance. In order to show performance, it is necessary to develop a rigorous thermodynamic model and computer program to analyze the SOFC in relation with the overall cycle. A large increase in efficiency over the conventional pressurized water reactor cycle is realized. Both sides of the cycle achieve higher efficiencies at partial loads which is extremely important as most naval vessels operate at partial loads as well as the fact that traditional gas turbines operating alone have poor performance at reduced speeds. Furthermore, each side of the cycle provides important benefits to the other side. The high temperature exhaust from the overall exothermic reaction of the fuel cell provides heat for the reheater allowing for an overall increase in power on the nuclear side of the cycle. Likewise, the high temperature helium exiting the nuclear reactor provides a controllable method to stabilize the fuel cell at an optimal temperature band even during transients helping to increase performance and reduce degradation of the fuel cell. It also provides the high temperature needed to efficiently produce hydrogen for the fuel cell. Moreover, the inclusion of a highly reliable and electrically independent fuel cell is particularly important as the ship will have the ability to divert large amounts of power from the propulsion system to energize high energy weapon pulse loads without disturbing vital parts of the C4ISR systems or control panels. Ultimately, the thesis shows that the combined cycle is mutually beneficial to each side of the cycle and overall critically needed for our future.

Investigation of applications for high-power, self-critical fissioning uranium plasma reactors

NASA Technical Reports Server (NTRS)

Rodgers, R. J.; Latham, T. S.; Krascella, N. L.

1976-01-01

Analytical studies were conducted to investigate potentially attractive applications for gaseous nuclear cavity reactors fueled by uranium hexafluoride and its decomposition products at temperatures of 2000 to 6000 K and total pressures of a few hundred atmospheres. Approximate operating conditions and performance levels for a class of nuclear reactors in which fission energy removal is accomplished principally by radiant heat transfer from the high temperature gaseous nuclear fuel to surrounding absorbing media were determined. The results show the radiant energy deposited in the absorbing media may be efficiently utilized in energy conversion system applications which include (1) a primary energy source for high thrust, high specific impulse space propulsion, (2) an energy source for highly efficient generation of electricity, and (3) a source of high intensity photon flux for heating working fluid gases for hydrogen production or MHD power extraction.

Heat transfer analysis of cylindrical anaerobic reactors with different sizes: a heat transfer model.

PubMed

Liu, Jiawei; Zhou, Xingqiu; Wu, Jiangdong; Gao, Wen; Qian, Xu

2017-10-01

The temperature is the essential factor that influences the efficiency of anaerobic reactors. During the operation of the anaerobic reactor, the fluctuations of ambient temperature can cause a change in the internal temperature of the reactor. Therefore, insulation and heating measures are often used to maintain anaerobic reactor's internal temperature. In this paper, a simplified heat transfer model was developed to study heat transfer between cylindrical anaerobic reactors and their surroundings. Three cylindrical reactors of different sizes were studied, and the internal relations between ambient temperature, thickness of insulation, and temperature fluctuations of the reactors were obtained at different reactor sizes. The model was calibrated by a sensitivity analysis, and the calibrated model was well able to predict reactor temperature. The Nash-Sutcliffe model efficiency coefficient was used to assess the predictive power of heat transfer models. The Nash coefficients of the three reactors were 0.76, 0.60, and 0.45, respectively. The model can provide reference for the thermal insulation design of cylindrical anaerobic reactors.

A PC-based high temperature gas reactor simulator for Indonesian conceptual HTR reactor basic training

NASA Astrophysics Data System (ADS)

Syarip; Po, L. C. C.

2018-05-01

In planning for nuclear power plant construction in Indonesia, helium cooled high temperature reactor (HTR) is favorable for not relying upon water supply that might be interrupted by earthquake. In order to train its personnel, BATAN has cooperated with Micro-Simulation Technology of USA to develop a 200 MWt PC-based simulation model PCTRAN/HTR. It operates in Win10 environment with graphic user interface (GUI). Normal operation of startup, power maneuvering, shutdown and accidents including pipe breaks and complete loss of AC power have been conducted. A sample case of safety analysis simulation to demonstrate the inherent safety features of HTR was done for helium pipe break malfunction scenario. The analysis was done for the variation of primary coolant pipe break i.e. from 0,1% - 0,5 % and 1% - 10 % helium gas leakages, while the reactor was operated at the maximum constant power of 10 MWt. The result shows that the highest temperature of HTR fuel centerline and coolant were 1150 °C and 1296 °C respectively. With 10 kg/s of helium flow in the reactor core, the thermal power will back to the startup position after 1287 s of helium pipe break malfunction.

ASME code considerations for the compact heat exchanger

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nestell, James; Sham, Sam

2015-08-31

The mission of the U.S. Department of Energy (DOE), Office of Nuclear Energy is to advance nuclear power in order to meet the nation's energy, environmental, and energy security needs. Advanced high temperature reactor systems such as sodium fast reactors and high and very high temperature gas-cooled reactors are being considered for the next generation of nuclear reactor plant designs. The coolants for these high temperature reactor systems include liquid sodium and helium gas. Supercritical carbon dioxide (sCO₂), a fluid at a temperature and pressure above the supercritical point of CO₂, is currently being investigated by DOE as a workingmore » fluid for a nuclear or fossil-heated recompression closed Brayton cycle energy conversion system that operates at 550°C (1022°F) at 200 bar (2900 psi). Higher operating temperatures are envisioned in future developments. All of these design concepts require a highly effective heat exchanger that transfers heat from the nuclear or chemical reactor to the chemical process fluid or the to the power cycle. In the nuclear designs described above, heat is transferred from the primary to the secondary loop via an intermediate heat exchanger (IHX) and then from the intermediate loop to either a working process or a power cycle via a secondary heat exchanger (SHX). The IHX is a component in the primary coolant loop which will be classified as "safety related." The intermediate loop will likely be classified as "not safety related but important to safety." These safety classifications have a direct bearing on heat exchanger design approaches for the IHX and SHX. The very high temperatures being considered for the VHTR will require the use of very high temperature alloys for the IHX and SHX. Material cost considerations alone will dictate that the IHX and SHX be highly effective; that is, provide high heat transfer area in a small volume. This feature must be accompanied by low pressure drop and mechanical reliability and robustness. Classic shell and tube designs will be large and costly, and may only be appropriate in steam generator service in the SHX where boiling inside the tubes occurs. For other energy conversion systems, all of these features can be met in a compact heat exchanger design. This report will examine some of the ASME Code issues that will need to be addressed to allow use of a Code-qualified compact heat exchanger in IHX or SHX nuclear service. Most effort will focus on the IHX, since the safety-related (Class A) design rules are more extensive than those for important-to-safety (Class B) or commercial rules that are relevant to the SHX.« less

Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCulloch, R.W.; MacPherson, R.E.

1983-03-01

The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through cladmore » melting at 1370/sup 0/C.« less

Molten Slag Would Boost Coal Conversion

NASA Technical Reports Server (NTRS)

Ferrall, J. F.

1984-01-01

Reactor increases residence time of uncovered char. Near-100percent carbon conversion achievable in reactor incorporating moltenslag bath. Slag maintains unconverted carbon impinging on surface at high temperatures for longer period of time, enhancing conversion.

Fast quench reactor and method

DOEpatents

Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.

1998-05-12

A fast quench reactor includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This ``freezes`` the desired end product(s) in the heated equilibrium reaction stage. 7 figs.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stillman, J. A.; Feldman, E. E.; Wilson, E. H.

This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. This report contains themore » results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. In the framework of non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MURR. This report presents the results of a study of core behavior under a set of accident conditions for MURR cores fueled with HEU U-Alx dispersion fuel or LEU monolithic U-Mo alloy fuel with 10 wt% Mo (U-10Mo).« less

Summary of space nuclear reactor power systems, 1983--1992

DOE Office of Scientific and Technical Information (OSTI.GOV)

Buden, D.

1993-08-11

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressedmore » from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.« less

Summary of space nuclear reactor power systems, 1983 - 1992

NASA Astrophysics Data System (ADS)

Buden, D.

1993-08-01

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987-88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

EXPERIMENTAL STUDIES OF TRANSIENT EFFECTS IN FAST REACTOR FUELS. SERIES I. UO$sub 2$ IRRADIATIONS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, J.H.

1962-11-15

An experimental program to evaluate the performance of FCR and EFCR fuel during transient operation is outlined, and the initial series of tests are described in some detail. Test results from five experiments in the TREAT reactor, using 1-in. OD SS-clad UO/sub 2/ fuel specimens, are compared with regard to fuel temperatures, mechanical integrity, and post-irradiation appearance. Incipient fuel pin failure limits for transients are identified with maximum fuel temperatures in the range of 7000 deg F. Multiple transient damage to the cladding is likely for transients above the melting point of the fuel. (auth)

A Computational Fluid Dynamic and Heat Transfer Model for Gaseous Core and Gas Cooled Space Power and Propulsion Reactors

NASA Technical Reports Server (NTRS)

Anghaie, S.; Chen, G.

1996-01-01

A computational model based on the axisymmetric, thin-layer Navier-Stokes equations is developed to predict the convective, radiation and conductive heat transfer in high temperature space nuclear reactors. An implicit-explicit, finite volume, MacCormack method in conjunction with the Gauss-Seidel line iteration procedure is utilized to solve the thermal and fluid governing equations. Simulation of coolant and propellant flows in these reactors involves the subsonic and supersonic flows of hydrogen, helium and uranium tetrafluoride under variable boundary conditions. An enthalpy-rebalancing scheme is developed and implemented to enhance and accelerate the rate of convergence when a wall heat flux boundary condition is used. The model also incorporated the Baldwin and Lomax two-layer algebraic turbulence scheme for the calculation of the turbulent kinetic energy and eddy diffusivity of energy. The Rosseland diffusion approximation is used to simulate the radiative energy transfer in the optically thick environment of gas core reactors. The computational model is benchmarked with experimental data on flow separation angle and drag force acting on a suspended sphere in a cylindrical tube. The heat transfer is validated by comparing the computed results with the standard heat transfer correlations predictions. The model is used to simulate flow and heat transfer under a variety of design conditions. The effect of internal heat generation on the heat transfer in the gas core reactors is examined for a variety of power densities, 100 W/cc, 500 W/cc and 1000 W/cc. The maximum temperature, corresponding with the heat generation rates, are 2150 K, 2750 K and 3550 K, respectively. This analysis shows that the maximum temperature is strongly dependent on the value of heat generation rate. It also indicates that a heat generation rate higher than 1000 W/cc is necessary to maintain the gas temperature at about 3500 K, which is typical design temperature required to achieve high efficiency in the gas core reactors. The model is also used to predict the convective and radiation heat fluxes for the gas core reactors. The maximum value of heat flux occurs at the exit of the reactor core. Radiation heat flux increases with higher wall temperature. This behavior is due to the fact that the radiative heat flux is strongly dependent on wall temperature. This study also found that at temperature close to 3500 K the radiative heat flux is comparable with the convective heat flux in a uranium fluoride failed gas core reactor.

Optimization of 200 MWth and 250 MWt Ship Based Small Long Life NPP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fitriyani, Dian; Su'ud, Zaki

2010-06-22

Design optimization of ship-based 200 MWth and 250 MWt nuclear power reactors have been performed. The neutronic and thermo-hydraulic programs of the three-dimensional X-Y-Z geometry have been developed for the analysis of ship-based nuclear power plant. Quasi-static approach is adopted to treat seawater effect. The reactor are loop type lead bismuth cooled fast reactor with nitride fuel and with relatively large coolant pipe above reactor core, the heat from primary coolant system is directly transferred to watersteam loop through steam generators. Square core type are selected and optimized. As the optimization result, the core outlet temperature distribution is changing withmore » the elevation angle of the reactor system and the characteristics are discussed.« less

HYFIRE II: fusion/high-temperature electrolysis conceptual-design study. Annual report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fillo, J.A.

1983-08-01

As in the previous HYFIRE design study, the current study focuses on coupling a Tokamak fusion reactor with a high-temperature blanket to a High-Temperature Electrolyzer (HTE) process to produce hydrogen and oxygen. Scaling of the STARFIRE reactor to allow a blanket power to 6000 MW(th) is also assumed. The primary difference between the two studies is the maximum inlet steam temperature to the electrolyzer. This temperature is decreased from approx. 1300/sup 0/ to approx. 1150/sup 0/C, which is closer to the maximum projected temperature of the Westinghouse fuel cell design. The process flow conditions change but the basic design philosophymore » and approaches to process design remain the same as before. Westinghouse assisted in the study in the areas of systems design integration, plasma engineering, balance-of-plant design, and electrolyzer technology.« less

Analytical analyses of startup measurements associated with the first use of LEU fuel in Romania`s 14-MW TRIGA reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bretscher, M.M.; Snelgrove, J.L.; Ciocanescu, M.

1992-12-01

The 14-MW TRIGA steady state reactor (SSR) is located in Pitesti, Romania. Beginning with an HEU core (10 wt% U), the reactor first went critical in November 1979 but was shut down ten years later because of insufficient excess reactivity. Last November the Institute for Nuclear Research (INR), which operates the SSR, received from the ANL RERTR program a shipment of 125 LEU pins fabricated by General Atomics and of the same geometry as the original fuel but with an enrichment of 19.7% 235U and a loading of 45 wt% U. Using 100 of these pins, four LEU clusters, eachmore » containing a 5 x 5 square array of fuel rods, were assembled. These four LEU clusters replaced the four most highly burned HEU elements in the SSR. The reactor resumed operations last February with a 35-element mixed HEU/LEU core configuration. In preparation for full power operation of the SSR with this mixed HEU/LEU core, a number of measurements were made. These included control rod calibrations, excess reactivity determinations, worths of experiment facilities, reaction rate distributions, and themocouple measurements of fuel temperatures as a function of reactor power. This paper deals with a comparison of some of these measured reactor parameters with corresponding analytical calculations.« less

Fischer-Tropsch Wastewater Utilization

DOEpatents

Shah, Lalit S.

2003-03-18

The present invention is generally directed to handling the wastewater, or condensate, from a hydrocarbon synthesis reactor. More particularly, the present invention provides a process wherein the wastewater of a hydrocarbon synthesis reactor, such as a Fischer-Tropsch reactor, is sent to a gasifier and subsequently reacted with steam and oxygen at high temperatures and pressures so as to produce synthesis gas. The wastewater may also be recycled back to a slurry preparation stage, where solid combustible organic materials are pulverized and mixed with process water and the wastewater to form a slurry, after which the slurry fed to a gasifier where it is reacted with steam and oxygen at high temperatures and pressures so as to produce synthesis gas.

Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Nuclear Energy Research Initiative Project 2001-001, Westinghouse Electric Co. Grant Number: DE-FG07-02SF22533, Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Philip E. MacDonald

2005-01-01

The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% versus about 33% efficiency for current Light Water Reactors [LWRs]) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus, the need for a pressurizer, steam generators, steam separators, and dryers is eliminated. The main mission ofmore » the SCWR is generation of low-cost electricity. It is built upon two proven technologies: LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which are also in use around the world. The reference SCWR design for the U.S. program is a direct cycle system operating at 25.0 MPa, with core inlet and outlet temperatures of 280 and 500 C, respectively. The coolant density decreases from about 760 kg/m3 at the core inlet to about 90 kg/m3 at the core outlet. The inlet flow splits with about 10% of the inlet flow going down the space between the core barrel and the reactor pressure vessel (the downcomer) and about 90% of the inlet flow going to the plenum at the top of the rector pressure vessel, to then flow down through the core in special water rods to the inlet plenum. Here it mixes with the feedwater from the downcomer and flows upward to remove the heat in the fuel channels. This strategy is employed to provide good moderation at the top of the core. The coolant is heated to about 500 C and delivered to the turbine. The purpose of this NERI project was to assess the reference U.S. Generation IV SCWR design and explore alternatives to determine feasibility. The project was organized into three tasks: Task 1. Fuel-cycle Neutronic Analysis and Reactor Core Design Task 2. Fuel Cladding and Structural Material Corrosion and Stress Corrosion Cracking Task 3. Plant Engineering and Reactor Safety Analysis. moderator rods. materials.« less

Recovery of tritium from tritiated molecules

DOEpatents

Swansiger, William A.

1987-01-01

A method of recovering tritium from tritiated compounds comprises the steps of heating tritiated water and other co-injected tritiated compounds in a preheater to temperatures of about 600.degree. C. The mixture is injected into a reactor charged with a mixture of uranium and uranium dioxide. The injected mixture undergoes highly exothermic reactions with the uranium causing reaction temperatures to occur in excess of the melting point of uranium, and complete decomposition of the tritiated compounds to remove tritium therefrom. The uranium dioxide functions as an insulating material and heat sink preventing the reactor side walls from attaining reaction temperatures to thereby minimize tritium permeation rates. The uranium dioxide also functions as a diluent to allow for volumetric expansion of the uranium as it is converted to uranium dioxide. The reactor vessel is preferably stainless steel of sufficient mass so as to function as a heat sink preventing the reactor side walls from approaching high temperatures. A disposable copper liner extends between the reaction chamber and stainless steel outer vessel to prevent alloying of the uranium with the outer vessel. Apparatus used to carry out the method of the invention is also disclosed.

Normal operation and maintenance safety lessons from the ITER US PbLi test blanket module program for a US FNSF and DEMO

DOE Office of Scientific and Technical Information (OSTI.GOV)

L. C. Cadwallader; C. P. C. Wong; M. Abdou

2014-10-01

A leading power reactor breeding blanket candidate for a fusion demonstration power plant (DEMO) being pursued by the US Fusion Community is the Dual Coolant Lead Lithium (DCLL) concept. The safety hazards associated with the DCLL concept as a reactor blanket have been examined in several US design studies. These studies identify the largest radiological hazards as those associated with the dust generation by plasma erosion of plasma blanket module first walls, oxidation of blanket structures at high temperature in air or steam, inventories of tritium bred in or permeating through the ferritic steel structures of the blanket module andmore » blanket support systems, and the 210Po and 203Hg produced in the PbLi breeder/coolant. What these studies lack is the scrutiny associated with a licensing review of the DCLL concept. An insight into this process was gained during the US participation in the International Thermonuclear Experimental Reactor (ITER) Test Blanket Module (TBM) Program. In this paper we discuss the lessons learned during this activity and make safety proposals for the design of a Fusion Nuclear Science Facility (FNSF) or a DEMO that employs a lead lithium breeding blanket.« less

Development of Advanced 9Cr Ferritic-Martensitic Steels and Austenitic Stainless Steels for Sodium-Cooled Fast Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sham, Sam; Tan, Lizhen; Yamamoto, Yukinori

2013-01-01

Ferritic-martensitic (FM) steel Grade 92, with or without thermomechanical treatment (TMT), and austenitic stainless steels HT-UPS (high-temperature ultrafine precipitate strengthening) and NF709 were selected as potential candidate structural materials in the U.S. Sodium-cooled Fast Reactor (SFR) program. The objective is to develop advanced steels with improved properties as compared with reference materials such as Grade 91 and Type 316H steels that are currently in nuclear design codes. Composition modification and/or processing optimization (e.g., TMT and cold-work) were performed to improve properties such as resistance to thermal aging, creep, creep-fatigue, fracture, and sodium corrosion. Testings to characterize these properties for themore » advanced steels were conducted by the Idaho National Laboratory, the Argonne National Laboratory and the Oak Ridge National Laboratory under the U.S. SFR program. This paper focuses on the resistance to thermal aging and creep of the advanced steels. The advanced steels exhibited up to two orders of magnitude increase in creep life compared to the reference materials. Preliminary results on the weldment performance of the advanced steels are also presented. The superior performance of the advanced steels would improve reactor design flexibility, safety margins and economics.« less

Returning HEU Fuel from the Czech Republic to Russia

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael Tyacke; Dr. Igor Bolshinsky

In December 1999, representatives from the United States, Russian Federation, and International Atomic Energy Agency began working on a program to return Russian supplied, highly enriched, uranium fuel stored at foreign research reactors to Russia. Now, under the Global Threat Reduction Initiative’s Russian Research Reactor Fuel Return Program, this effort has repatriated over 800 kg of highly enriched uranium to Russia from over 10 countries. In May 2004, the “Agreement Between the Government of the United States of America and the Government of the Russian Federation Concerning Cooperation for the Transfer of Russian Produced Research Reactor Nuclear Fuel to themore » Russian Federation” was signed. This agreement provides legal authority for the Russian Research Reactor Fuel Return Program and establishes parameters whereby eligible countries may return highly enriched uranium spent and fresh fuel assemblies and other fissile materials to Russia. On December 8, 2007, one of the largest shipments of highly enriched uranium spent nuclear fuel was successfully made from a Russian-designed nuclear research reactor in the Czech Republic to the Russian Federation. This accomplishment is the culmination of years of planning, negotiations, and hard work. The United States, Russian Federation, and the International Atomic Energy Agency have been working together. In February 2003, Russian Research Reactor Fuel Return Program representatives met with the Nuclear Research Institute in Rež, Czech Republic, and discussed the return of their highly enriched uranium spent nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This article discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.« less

Thermal storage/discharge performances of Cu-Si alloy for solar thermochemical process

NASA Astrophysics Data System (ADS)

Gokon, Nobuyuki; Yamaguchi, Tomoya; Cho, Hyun-seok; Bellan, Selvan; Hatamachi, Tsuyoshi; Kodama, Tatsuya

2017-06-01

The present authors (Niigata University, Japan) have developed a tubular reactor system using novel "double-walled" reactor/receiver tubes with carbonate molten-salt thermal storage as a phase change material (PCM) for solar reforming of natural gas and with Al-Si alloy thermal storage as a PCM for solar air receiver to produce high-temperature air. For both of the cases, the high heat capacity and large latent heat (heat of solidification) of the PCM phase circumvents the rapid temperature change of the reactor/receiver tubes at high temperatures under variable and uncontinuous characteristics of solar radiation. In this study, we examined cyclic properties of thermal storage/discharge for Cu-Si alloy in air stream in order to evaluate a potentiality of Cu-Si alloy as a PCM thermal storage material. Temperature-increasing performances of Cu-Si alloy are measured during thermal storage (or heat-charge) mode and during cooling (or heat-discharge) mode. A oxidation state of the Cu-Si alloy after the cyclic reaction was evaluated by using electron probe micro analyzer (EPMA).

Nuclear fuel elements and method of making same

DOEpatents

Schweitzer, Donald G.

1992-01-01

A nuclear fuel element for a high temperature gas nuclear reactor that has an average operating temperature in excess of 2000.degree. C., and a method of making such a fuel element. The fuel element is characterized by having fissionable fuel material localized and stabilized within pores of a carbon or graphite member by melting the fissionable material to cause it to chemically react with the carbon walls of the pores. The fissionable fuel material is further stabilized and localized within the pores of the graphite member by providing one or more coatings of pyrolytic carbon or diamond surrounding the porous graphite member so that each layer defines a successive barrier against migration of the fissionable fuel from the pores, and so that the outermost layer of pyrolytic carbon or diamond forms a barrier between the fissionable material and the moderating gases used in an associated high temperature gas reactor. The method of the invention provides for making such new elements either as generally spherically elements, or as flexible filaments, or as other relatively small-sized fuel elements that are particularly suited for use in high temperature gas reactors.

KWU's high conversion reactor concept - An economical evolution of modern pressurized water reactor technology toward improved uranium ore utilization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Markl, H.; Goetzmann, C.A.; Moldaschl, H.

The Kraftwerk Union AG high conversion reactor represents a quasi-standard PWR with fuel assemblies of more or less uniformly enriched fuel rods, arranged in a tight hexagonal array with a pitch-to-diameter ratio p/d approx. = 1.12. High fuel enrichment as well as a high conversion ratio of --0.9 will provide the potential for high burnup values up to 70 000 MWd/tonne and a low fissile material consumption. The overall objective of the actual RandD program is to have the technical feasibility, including that for licensibility, established by the early 1990s as a prerequisite for deciding whether to enter a demonstrationmore » plant program.« less

Improved hydrocracker temperature control: Mobil quench zone technology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sarli, M.S.; McGovern, S.J.; Lewis, D.W.

1993-01-01

Hydrocracking is a well established process in the oil refining industry. There are over 2.7 million barrels of installed capacity world-wide. The hydrocracking process comprises several families of highly exothermic reactions and the total adiabatic temperature rise can easily exceed 200 F. Reactor temperature control is therefore very important. Hydrocracking reactors are typically constructed with multiple catalyst beds in series. Cold recycle gas is usually injected between the catalyst beds to quench the reactions, thereby controlling overall temperature rise. The design of this quench zone is the key to good reactor temperature control, particularly when processing poorer quality, i.e., highermore » heat release, feeds. Mobil Research and Development Corporation (MRDC) has developed a robust and very effective quench zone technology (QZT) package, which is now being licensed to the industry for hydrocracking applications.« less

Investigation of a para-ortho hydrogen reactor for application to spacecraft sensor cooling

NASA Technical Reports Server (NTRS)

Nast, T. C.

1983-01-01

The utilization of solid hydrogen in space for sensor and instrument cooling is a very efficient technique for long term cooling or for cooling at high heat rates. The solid hydrogen can provide temperatures as low as 7 to 8 K to instruments. Vapor cooling is utilized to reduce parasitic heat inputs to the 7 to 8 K stage and is effective in providing intermediate cooling for instrument components operating at higher temperatures. The use of solid hydrogen in place of helium may lead to weight reductions as large as a factor of ten and an attendent reduction in system volume. The results of an investigation of a catalytic reactor for use with a solid hydrogen cooling system is presented. Trade studies were performed on several configurations of reactor to meet the requirements of high reactor efficiency with low pressure drop. Results for the selected reactor design are presented for both liquid hydrogen systems operating at near atmospheric pressure and the solid hydrogen cooler operating as low as 1 torr.

The importance of Soret transport in the production of high purity silicon for solar cells

NASA Technical Reports Server (NTRS)

Srivastava, R.

1985-01-01

Temperature-gradient-driven diffusion, or Soret transport, of silicon vapor and liquid droplets is analyzed under conditions typical of current production reactors for obtaining high purity silicon for solar cells. Contrary to the common belief that Soret transport is negligible, it is concluded that some 15-20 percent of the silicon vapor mass flux to the reactor walls is caused by the high temperature gradients that prevail inside such reactors. Moreover, since collection of silicon is also achieved via deposition of silicon droplets onto the walls, the Soret transport mechanism becomes even more crucial due to size differences between diffusing species. It is shown that for droplets in the 0.01 to 1 micron diameter range, collection by Soret transport dominates both Brownian and turbulent mechanisms.

Dual reactor for in situ/operando fluorescent mode XAS studies of sample containing low-concentration 3d or 5d metal elements

NASA Astrophysics Data System (ADS)

Nguyen, Luan; Tang, Yu; Li, Yuting; Zhang, Xiaoyan; Wang, Ding; Tao, Franklin Feng

2018-05-01

Transition metal elements are the most important elements of heterogeneous catalysts used for chemical and energy transformations. Many of these catalysts are active at a temperature higher than 400 °C. For a catalyst containing a 3d or 5d metal element with a low concentration, typically their released fluorescence upon the K-edge or L-edge adsorption of X-rays is collected for the analysis of chemical and coordination environments of these elements. However, it is challenging to perform in situ/operando X-ray absorption spectroscopy (XAS) studies of elements of low-energy absorption edges at a low concentration in a catalyst during catalysis at a temperature higher than about 450 °C. Here a unique reaction system consisting two reactors, called a dual reactor system, was designed for performing in situ or operando XAS studies of these elements of low-energy absorption edges in a catalyst at a low concentration during catalysis at a temperature higher than 450 °C in a fluorescent mode. This dual-reactor system contains a quartz reactor for preforming high-temperature catalysis up to 950 °C and a Kapton reactor remaining at a temperature up to 450 °C for collecting data in the same gas of catalysis. With this dual reactor, chemical and coordination environments of low-concentration metal elements with low-energy absorption edges such as the K-edge of 3d metals including Ti, V, Cr, Mn, Fe, Co, Ni, and Cu and L edge of 5d metals including W, Re, Os, Ir, Pt, and Au can be examined through first performing catalysis at a temperature higher than 450 °C in the quartz reactor and then immediately flipping the catalyst in the same gas flow to the Kapton reactor remained up to 450 °C to collect data. The capability of this dual reactor was demonstrated by tracking the Mn K-edge of the MnOx/Na2WO4 catalyst during activation in the temperature range of 300-900 °C and catalysis at 850 °C.

Control of Advanced Reactor-Coupled Heat Exchanger System: Incorporation of Reactor Dynamics in System Response to Load Disturbances

DOE PAGES

Skavdahl, Isaac; Utgikar, Vivek; Christensen, Richard; ...

2016-05-24

We present an alternative control schemes for an Advanced High Temperature Reactor system consisting of a reactor, an intermediate heat exchanger, and a secondary heat exchanger (SHX) in this paper. One scheme is designed to control the cold outlet temperature of the SHX (T co) and the hot outlet temperature of the intermediate heat exchanger (T ho2) by manipulating the hot-side flow rates of the heat exchangers (F h/F h2) responding to the flow rate and temperature disturbances. The flow rate disturbances typically require a larger manipulation of the flow rates than temperature disturbances. An alternate strategy examines the controlmore » of the cold outlet temperature of the SHX (T co) only, since this temperature provides the driving force for energy production in the power conversion unit or the process application. The control can be achieved by three options: (1) flow rate manipulation; (2) reactor power manipulation; or (3) a combination of the two. The first option has a quicker response but requires a large flow rate change. The second option is the slowest but does not involve any change in the flow rates of streams. The final option appears preferable as it has an intermediate response time and requires only a minimal flow rate change.« less

ASME Material Challenges for Advanced Reactor Concepts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Piyush Sabharwall; Ali Siahpush

2013-07-01

This study presents the material Challenges associated with Advanced Reactor Concept (ARC) such as the Advanced High Temperature Reactor (AHTR). ACR are the next generation concepts focusing on power production and providing thermal energy for industrial applications. The efficient transfer of energy for industrial applications depends on the ability to incorporate cost-effective heat exchangers between the nuclear heat transport system and industrial process heat transport system. The heat exchanger required for AHTR is subjected to a unique set of conditions that bring with them several design challenges not encountered in standard heat exchangers. The corrosive molten salts, especially at highermore » temperatures, require materials throughout the system to avoid corrosion, and adverse high-temperature effects such as creep. Given the very high steam generator pressure of the supercritical steam cycle, it is anticipated that water tube and molten salt shell steam generators heat exchanger will be used. In this paper, the ASME Section III and the American Society of Mechanical Engineers (ASME) Section VIII requirements (acceptance criteria) are discussed. Also, the ASME material acceptance criteria (ASME Section II, Part D) for high temperature environment are presented. Finally, lack of ASME acceptance criteria for thermal design and analysis are discussed.« less

Studies Related to the Oregon State University High Temperature Test Facility: Scaling, the Validation Matrix, and Similarities to the Modular High Temperature Gas-Cooled Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Richard R. Schultz; Paul D. Bayless; Richard W. Johnson

2010-09-01

The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5 year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) beganmore » their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant project. Because the NRC interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC). Since DOE has incorporated the HTTF as an ingredient in the NGNP thermal-fluids validation program, several important outcomes should be noted: 1. The reference prismatic reactor design, that serves as the basis for scaling the HTTF, became the modular high temperature gas-cooled reactor (MHTGR). The MHTGR has also been chosen as the reference design for all of the other NGNP thermal-fluid experiments. 2. The NGNP validation matrix is being planned using the same scaling strategy that has been implemented to design the HTTF, i.e., the hierarchical two-tiered scaling methodology developed by Zuber in 1991. Using this approach a preliminary validation matrix has been designed that integrates the HTTF experiments with the other experiments planned for the NGNP thermal-fluids verification and validation project. 3. Initial analyses showed that the inherent power capability of the OSU infrastructure, which only allowed a total operational facility power capability of 0.6 MW, is inadequate to permit steady-state operation at reasonable conditions. 4. To enable the HTTF to operate at a more representative steady-state conditions, DOE recently allocated funding via a DOE subcontract to HTTF to permit an OSU infrastructure upgrade such that 2.2 MW will become available for HTTF experiments. 5. Analyses have been performed to study the relationship between HTTF and MHTGR via the hierarchical two-tiered scaling methodology which has been used successfully in the past, e.g., APEX facility scaling to the Westinghouse AP600 plant. These analyses have focused on the relationship between key variables that will be measured in the HTTF to the counterpart variables in the MHTGR with a focus on natural circulation, using nitrogen as a working fluid, and core heat transfer. 6. Both RELAP5-3D and computational fluid dynamics (CD-Adapco’s STAR-CCM+) numerical models of the MHTGR and the HTTF have been constructed and analyses are underway to study the relationship between the reference reactor and the HTTF. The HTTF is presently being designed. It has ¼-scaling relationship to the MHTGR in both the height and the diameter. Decisions have been made to design the reactor cavity cooling system (RCCS) simulation as a boundary condition for the HTTF to ensure that (a) the boundary condition is well defined and (b) the boundary condition can be modified easily to achieve the desired heat transfer sink for HTTF experimental operations.« less

Beyond the classical kinetic model for chronic graphite oxidation by moisture in high temperature gas-cooled reactors

DOE PAGES

Contescu, Cristian I.; Mee, Robert W.; Lee, Yoonjo; ...

2017-11-03

Four grades of nuclear graphite with various microstructures were subjected to accelerated oxidation tests in helium with traces of moisture and hydrogen in order to evaluate the effects of chronic oxidation on graphite components in high temperature gas cooled reactors. Kinetic analysis showed that the Langmuir-Hinshelwood (LH) model cannot consistently reproduce all results. In particular, at high temperatures and water partial pressures oxidation was always faster than the LH model predicts, with stronger deviations for superfine grain graphite than for medium grain grades. It was also found empirically that the apparent reaction order for water has a sigmoid-type variation withmore » temperature which follows the integral Boltzmann distribution function. This suggests that the apparent activation with temperature of graphite reactive sites that causes deviations from the LH model is rooted in specific structural and electronic properties of surface sites on graphite. A semi-global kinetic model was proposed, whereby the classical LH model was modified with a temperature-dependent reaction order for water. The new Boltzmann-enhanced model (BLH) was shown to consistently predict experimental oxidation rates over large ranges of temperature (800-1100 oC) and partial pressures of water (3-1200 Pa) and hydrogen (0-300 Pa), not only for the four grades of graphite but also for the historic grade H-451. The BLH model offers as more reliable input for modeling the chemical environment effects during the life-time operation of new grades of graphite in advanced nuclear reactors operating at high and very high temperatures.« less

Beyond the classical kinetic model for chronic graphite oxidation by moisture in high temperature gas-cooled reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Contescu, Cristian I.; Mee, Robert W.; Lee, Yoonjo

Four grades of nuclear graphite with various microstructures were subjected to accelerated oxidation tests in helium with traces of moisture and hydrogen in order to evaluate the effects of chronic oxidation on graphite components in high temperature gas cooled reactors. Kinetic analysis showed that the Langmuir-Hinshelwood (LH) model cannot consistently reproduce all results. In particular, at high temperatures and water partial pressures oxidation was always faster than the LH model predicts, with stronger deviations for superfine grain graphite than for medium grain grades. It was also found empirically that the apparent reaction order for water has a sigmoid-type variation withmore » temperature which follows the integral Boltzmann distribution function. This suggests that the apparent activation with temperature of graphite reactive sites that causes deviations from the LH model is rooted in specific structural and electronic properties of surface sites on graphite. A semi-global kinetic model was proposed, whereby the classical LH model was modified with a temperature-dependent reaction order for water. The new Boltzmann-enhanced model (BLH) was shown to consistently predict experimental oxidation rates over large ranges of temperature (800-1100 oC) and partial pressures of water (3-1200 Pa) and hydrogen (0-300 Pa), not only for the four grades of graphite but also for the historic grade H-451. The BLH model offers as more reliable input for modeling the chemical environment effects during the life-time operation of new grades of graphite in advanced nuclear reactors operating at high and very high temperatures.« less

Gap Size Uncertainty Quantification in Advanced Gas Reactor TRISO Fuel Irradiation Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pham, Binh T.; Einerson, Jeffrey J.; Hawkes, Grant L.

The Advanced Gas Reactor (AGR)-3/4 experiment is the combination of the third and fourth tests conducted within the tristructural isotropic fuel development and qualification research program. The AGR-3/4 test consists of twelve independent capsules containing a fuel stack in the center surrounded by three graphite cylinders and shrouded by a stainless steel shell. This capsule design enables temperature control of both the fuel and the graphite rings by varying the neon/helium gas mixture flowing through the four resulting gaps. Knowledge of fuel and graphite temperatures is crucial for establishing the functional relationship between fission product release and irradiation thermal conditions.more » These temperatures are predicted for each capsule using the commercial finite-element heat transfer code ABAQUS. Uncertainty quantification reveals that the gap size uncertainties are among the dominant factors contributing to predicted temperature uncertainty due to high input sensitivity and uncertainty. Gap size uncertainty originates from the fact that all gap sizes vary with time due to dimensional changes of the fuel compacts and three graphite rings caused by extended exposure to high temperatures and fast neutron irradiation. Gap sizes are estimated using as-fabricated dimensional measurements at the start of irradiation and post irradiation examination dimensional measurements at the end of irradiation. Uncertainties in these measurements provide a basis for quantifying gap size uncertainty. However, lack of gap size measurements during irradiation and lack of knowledge about the dimension change rates lead to gap size modeling assumptions, which could increase gap size uncertainty. In addition, the dimensional measurements are performed at room temperature, and must be corrected to account for thermal expansion of the materials at high irradiation temperatures. Uncertainty in the thermal expansion coefficients for the graphite materials used in the AGR-3/4 capsules also increases gap size uncertainty. This study focuses on analysis of modeling assumptions and uncertainty sources to evaluate their impacts on the gap size uncertainty.« less

Research Program of a Super Fast Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie

2006-07-01

Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is notmore » breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)« less

Investigation of Liquid Metal Embrittlement of Materials for use in Fusion Reactors

NASA Astrophysics Data System (ADS)

Kennedy, Daniel; Jaworski, Michael

2014-10-01

Liquid metals can provide a continually replenished material for the first wall and extraction blankets of fusion reactors. However, research has shown that solid metal surfaces will experience embrittlement when exposed to liquid metals under stress. Therefore, it is important to understand the changes in structural strength of the solid metal materials and test different surface treatments that can limit embrittlement. Research was conducted to design and build an apparatus for exposing solid metal samples to liquid metal under high stress and temperature. The apparatus design, results of tensile testing, and surface imaging of fractured samples will be presented. This work was supported in part by the U.S. Department of Energy, Office of Science, Office of Workforce Development for Teachers and Scientists (WDTS) under the Science Undergraduate Laboratory Internships Program (SULI).

Reactor Materials Program - Baseline Material Property Handbook - Mechanical Properties of 1950's Vintage Stainless Steel Weldment Components, Task Number 89-23-A-1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stoner, K.J.

1999-11-05

The Process Water System (primary coolant) piping of the nuclear production reactors constructed in the 1950''s at Savannah River Site is comprised primarily of Type 304 stainless steel with Type 308 stainless steel weld filler. A program to measure the mechanical properties of archival PWS piping and weld materials (having approximately six years of service at temperatures between 25 and 100 degrees C) has been completed. The results from the mechanical testing has been synthesized to provide a mechanical properties database for structural analyses of the SRS piping.

Solid Polymer Electrolyte Fuel Cell Technology Program

NASA Technical Reports Server (NTRS)

1980-01-01

Work is reported on phase 5 of the Solid Polymer Electrolyte (SPE) Fuel Cell Technology Development program. The SPE fuel cell life and performance was established at temperatures, pressures, and current densities significantly higher than those previously demonstrated in sub-scale hardware. Operation of single-cell Buildup No. 1 to establish life capabilities of the full-scale hardware was continued. A multi-cell full-scale unit (Buildup No. 2) was designed, fabricated, and test evaluated laying the groundwork for the construction of a reactor stack. A reactor stack was then designed, fabricated, and successfully test-evaluated to demonstrate the readiness of SPE fuel cell technology for future space applications.

Highly Selective Nuclide Removal from the R-Reactor Disassembly Basin at the SRS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pickett, J. B.; Austin, W. E.; Dukes, H. H.

This paper describes the results of a deployment of highly selective ion-exchange resin technologies for the in-situ removal of Cs-137 and Sr-90 from the Savannah River Site (SRS) R-Reactor Disassembly Basin. The deployment was supported by the DOE Office of Science and Technology's (OST, EM-50) National Engineering Technology Laboratory (NETL), as a part of an Accelerated Site Technology Deployment (ASTD) project. The Facilities Decontamination and Decommissioning (FDD) Program at the SRS conducted this deployment as a part of an overall program to deactivate three of the site's five reactor disassembly basins.

Highly Selective Nuclide Removal from the R-Reactor Disassembly Basin at SRS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pickett, J.B.

This paper describes the results of a deployment of highly selective ion-exchange resin technologies for the in-situ removal of Cs-137 and Sr-90 from the Savannah River Site (SRS) R-Reactor Disassembly Basin. The deployment was supported by the DOE Office of Science and Technology's (OST, EM-50) National Engineering Technology Laboratory (NETL), as a part of an Accelerated Site Technology Deployment (ASTD) project. The Facilities Decontamination and Decommissioning (FDD) Program at the SRS conducted this deployment as a part of an overall program to deactivate three of the site's five reactor disassembly basins

NGNP Data Management and Analysis System Analysis and Web Delivery Capabilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cynthia D. Gentillon

2011-09-01

Projects for the Very High Temperature Reactor (VHTR) Technology Development Office provide data in support of Nuclear Regulatory Commission licensing of the very high temperature reactor. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high-temperature and high-fluence environments. The NGNP Data Management and Analysis System (NDMAS) at the Idaho National Laboratory has been established to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the third NDMAS objective. It describes capabilities formore » displaying the data in meaningful ways and for data analysis to identify useful relationships among the measured quantities. The capabilities are described from the perspective of NDMAS users, starting with those who just view experimental data and analytical results on the INL NDMAS web portal. Web display and delivery capabilities are described in detail. Also the current web pages that show Advanced Gas Reactor, Advanced Graphite Capsule, and High Temperature Materials test results are itemized. Capabilities available to NDMAS developers are more extensive, and are described using a second series of examples. Much of the data analysis efforts focus on understanding how thermocouple measurements relate to simulated temperatures and other experimental parameters. Statistical control charts and correlation monitoring provide an ongoing assessment of instrument accuracy. Data analysis capabilities are virtually unlimited for those who use the NDMAS web data download capabilities and the analysis software of their choice. Overall, the NDMAS provides convenient data analysis and web delivery capabilities for studying a very large and rapidly increasing database of well-documented, pedigreed data.« less

Thermal energy storage material thermophysical property measurement and heat transfer impact

NASA Technical Reports Server (NTRS)

Tye, R. P.; Bourne, J. G.; Destarlais, A. O.

1976-01-01

The thermophysical properties of salts having potential for thermal energy storage to provide peaking energy in conventional electric utility power plants were investigated. The power plants studied were the pressurized water reactor, boiling water reactor, supercritical steam reactor, and high temperature gas reactor. The salts considered were LiNO3, 63LiOH/37 LiCl eutectic, LiOH, and Na2B4O7. The thermal conductivity, specific heat (including latent heat of fusion), and density of each salt were measured for a temperature range of at least + or - 100 K of the measured melting point. Measurements were made with both reagent and commercial grades of each salt.

Gaseous fuel nuclear reactor research

NASA Technical Reports Server (NTRS)

Schwenk, F. C.; Thom, K.

1975-01-01

Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. Kokkinos

2005-04-28

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophymore » on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.« less

Nuclear Radiation Tolerance of Single Crystal Aluminum Nitride Ultrasonic Transducer

NASA Astrophysics Data System (ADS)

Reinhard, Brian; Tittmann, Bernhard R.; Suprock, Andrew

Ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of ultrasonic technologies to provide enhanced sensors for in-pile instrumentation during irradiation testing. For example, the ability of small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models, (Rempe et al., 2011; Kazys et al., 2005). These efforts are limited by the lack of identified ultrasonic transducer materials capable of long term performance under irradiation test conditions. To address this need, the Pennsylvania State University (PSU) was awarded an Advanced Test Reactor National Scientific User Facility (ATR NSUF) project to evaluate the performance of promising magnetostrictive and piezoelectric transducers in the Massachusetts Institute of Technology Research Reactor (MITR) up to a fast fluence of at least 1021 n/cm2. The irradiation is also supported by a multi-National Laboratory collaboration funded by the Nuclear Energy Enabling Technologies Advanced Sensors and Instrumentation (NEET ASI) program. The results from this irradiation, which started in February 2014, offer the potential to enable the development of novel radiation tolerant ultrasonic sensors for use in Material Testing Reactors (MTRs). As such, this test is an instrumented lead test and real-time transducer performance data is collected along with temperature and neutron and gamma flux data. Hence, results from this irradiation offer the potential to bridge the gap between proven out-of-pile ultrasonic techniques and in-pile deployment of ultrasonic sensors by acquiring the data necessary to demonstrate the performance of ultrasonic transducers. To date, very encouraging results have been attained as several transducers have continued to operate under irradiation. The irradiation is ongoing and will continue to approximately mid-2015.

Master Curve and Conventional Fracture Toughness of Modified 9Cr-1Mo Steel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ji-Hyun, Yoon; Sung-Ho, Kim; Bong-Sang, Lee

2006-07-01

Modified 9Cr-1Mo steel is a primary candidate material for reactor pressure vessel of Very High Temperature Gas-Cooled Reactor (VHTR) in Korean Nuclear Hydrogen Development and Demonstration (NHDD) program. In this study, T0 reference temperature, J-R fracture resistance and Charpy impact properties were evaluated for commercial Grade 91 steel as preliminary tests for the selection of the RPV material for VHTR. The fracture toughness of the modified 9Cr-1Mo steel was compared with those of SA508-Gr.3. The objective of this study was to obtain pre-irradiation fracture toughness properties of modified 9Cr-1Mo steel as reference data for the radiation effects investigation. The resultsmore » are as follows. Charpy impact properties of the modified 9Cr-1Mo steel were similar to those of SA508-Gr.3. T0 reference temperatures were measured as -67.7 deg C and -72.4 deg C from the tests with standard PCVN (pre-cracked Charpy V-notch) and half sized PCVN specimens respectively, which were similar to results for SA508-Gr.3. The K{sub Jc} values of modified 9Cr-1Mo with test temperatures are successfully expressed with the Master Curve. The J-R fracture resistance of modified 9Cr-1Mo steel at room temperature was almost the same as that of SA508-Gr.3. On the other hand it was a little bit higher at an elevated temperature. (authors)« less

Fracture toughness and the master curve for modified 9Cr-1Mo steel

NASA Astrophysics Data System (ADS)

Yoon, Ji-Hyun; Yoon, Eui-Pak

2006-12-01

Modified 9Cr-1Mo steel is a primary candidate material for the reactor pressure vessel of a Very High Temperature Gas-Cooled Reactor (VHTR) in the Korean Nuclear Hydrogen Development and Demonstration (NHDD) program. In this study, the T0 reference temperature, J-R fracture resistance and Charpy impact properties were evaluated for commercial Grade 91 steel as part of the preliminary testing for a selection of the RPV material for the VHTR. The fracture toughness of the modified 9Cr-1Mo steel was compared with that of SA508-Gr.3. The objective of this study was to obtain the pre-irradiation fracture toughness properties of the modified 9Cr-1Mo steel as reference data for an investigation of radiation effects. Charpy impact properties of the modified 9Cr-1Mo steel were similar to those of SA508-Gr.3. T0 reference temperatures were measured as -67.7 and -72.4°C from the tests with standard PCVN (pre-cracked Charpy V-notch) and half-sized PCVN specimens respectively, which were similar to the results for SA508-Gr.3. The KJc values of the modified 9Cr-1Mo steel with the test temperatures are successfully expressed by the Master Curve. The J-R fracture resistance of the modified 9Cr-1Mo steel at room temperature was nearly identical to that of SA508-Gr.3; in contrast, it was slightly higher at an elevated temperature.

Machine‐Assisted Organic Synthesis

PubMed Central

Fitzpatrick, Daniel E.; Myers, Rebecca M.; Battilocchio, Claudio; Ingham, Richard. J.

2015-01-01

Abstract In this Review we describe how the advent of machines is impacting on organic synthesis programs, with particular emphasis on the practical issues associated with the design of chemical reactors. In the rapidly changing, multivariant environment of the research laboratory, equipment needs to be modular to accommodate high and low temperatures and pressures, enzymes, multiphase systems, slurries, gases, and organometallic compounds. Additional technologies have been developed to facilitate more specialized reaction techniques such as electrochemical and photochemical methods. All of these areas create both opportunities and challenges during adoption as enabling technologies. PMID:26193360

Calculation evaluation of multiplying properties of LWR with thorium fuel

NASA Astrophysics Data System (ADS)

Shamanin, I. V.; Grachev, V. M.; Knyshev, V. V.; Bedenko, S. V.; Novikova, N. G.

2017-01-01

The results of multiplying properties design research of the unit cell and LWR fuel assembly with the high temperature gas-cooled thorium reactor fuel pellet are presented in the work. The calculation evaluation showed the possibility of using thorium in LWR effectively. In this case the amount of fissile isotope is 2.45 times smaller in comparison with the standard loading of LWR. The research and numerical experiments were carried out using the verified accounting code of the program MCU5, modern libraries of evaluated nuclear data and multigroup approximations.

Power and Thermal Technologies for Air and Space-Scientific Research Program. Delivery Order 0012: High-Temperature Superconductor Performance Enhancement

DTIC Science & Technology

2010-06-01

house to grow CNTs. Initially the CNTs were grown at atmospheric pressure using C2H2/Ar mixtures. Prior to deposition, the quartz tube of the reactor...imaged clearly. It appears that there could be some amorphous carbon present on the surface of the tubes with the present set of conditions used and...chip cooling with CNT microfin architectures have been recently proposed by Kordas et al. [5]. CNT films as thermal interface materials were also

HIGH TEMPERATURE SULFATION STUDIES IN AN ISOTHERMAL REACTOR: A COMPARISON OF THEORY AND EXPERIMENT

EPA Science Inventory

The paper gives high-temperature isothermal data on sulfur dioxide (SO2) capture, obtained as a function of temperature, SO2 partial pressure, and Ca/S molar ratio for a pulverized dolomite (34 micrometer mean size) and a high-purity calcite (11 micrometer mean size). The experim...

West Europe Report.

DTIC Science & Technology

1986-05-23

Kraftwerk Union Power Plant... DER SPIEGEL: ...a 100-percent Siemens daughter enterprise... Kaske: ...to companies which are participating in the...major competitor, Kraftwerk Union AG (KWU) at Muelheim on the Ruhr, with its mass-produced light-water reactors. The High Temperature Reactor

ASME Code Efforts Supporting HTGRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

D.K. Morton

2010-09-01

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

ASME Code Efforts Supporting HTGRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

D.K. Morton

2011-09-01

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

ASME Code Efforts Supporting HTGRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

D.K. Morton

2012-09-01

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

Nuclear fuels - Present and future

NASA Astrophysics Data System (ADS)

Olander, D.

2009-06-01

The important developments in nuclear fuels and their problems are reviewed and compared with the status of present light-water reactor fuels. The limitations of LWR fuels are reviewed with respect to important recent concerns, namely provision of outlet coolant temperatures high enough for use in H 2 production, destruction of plutonium to eliminate proliferation concerns, and burning of the minor actinides to reduce the waste repository heat load and long-term radiation hazard. In addition to current oxide-based fuel rod designs, the hydride fuel with liquid-metal thermal bonding of the fuel-cladding gap is covered. Finally, two of the most promising Generation IV reactor concepts, the very high temperature reactor and the sodium fast reactor, and the accompanying reprocessing technologies, aqueous-based UREX+1a and pyrometallurgical, are summarized. In all of the topics covered, the thermodynamics involved in the fuel's behavior under irradiation and in the reprocessing schemes are emphasized.

Development of a high-temperature durable catalyst for use in catalytic combustors for advanced automotive gas turbine engines

NASA Astrophysics Data System (ADS)

Tong, H.; Snow, G. C.; Chu, E. K.; Chang, R. L. S.; Angwin, M. J.; Pessagno, S. L.

1981-09-01

Durable catalytic reactors for advanced gas turbine engines were developed. Objectives were: to evaluate furnace aging as a cost effective catalytic reactor screening test, measure reactor degradation as a function of furnace aging, demonstrate 1,000 hours of combustion durability, and define a catalytic reactor system with a high probability of successful integration into an automotive gas turbine engine. Fourteen different catalytic reactor concepts were evaluated, leading to the selection of one for a durability combustion test with diesel fuel for combustion conditions. Eight additional catalytic reactors were evaluated and one of these was successfully combustion tested on propane fuel. This durability reactor used graded cell honeycombs and a combination of noble metal and metal oxide catalysts. The reactor was catalytically active and structurally sound at the end of the durability test.

Development of a high-temperature durable catalyst for use in catalytic combustors for advanced automotive gas turbine engines

NASA Technical Reports Server (NTRS)

Tong, H.; Snow, G. C.; Chu, E. K.; Chang, R. L. S.; Angwin, M. J.; Pessagno, S. L.

1981-01-01

Durable catalytic reactors for advanced gas turbine engines were developed. Objectives were: to evaluate furnace aging as a cost effective catalytic reactor screening test, measure reactor degradation as a function of furnace aging, demonstrate 1,000 hours of combustion durability, and define a catalytic reactor system with a high probability of successful integration into an automotive gas turbine engine. Fourteen different catalytic reactor concepts were evaluated, leading to the selection of one for a durability combustion test with diesel fuel for combustion conditions. Eight additional catalytic reactors were evaluated and one of these was successfully combustion tested on propane fuel. This durability reactor used graded cell honeycombs and a combination of noble metal and metal oxide catalysts. The reactor was catalytically active and structurally sound at the end of the durability test.

Tritium Control and Capture in Salt-Cooled Fission and Fusion Reactors: Status, Challenges, and Path Forward

DOE Office of Scientific and Technical Information (OSTI.GOV)

Forsberg, Charles W.; Lam, Stephen; Carpenter, David M.

Three advanced nuclear power systems use liquid salt coolants that generate tritium and thus face the common challenges of containing and capturing tritium to prevent its release to the environment. The fluoride salt–cooled high-temperature reactor (FHR) uses clean fluoride salt coolants and the same graphite-matrix coated-particle fuel as high-temperature gas-cooled reactors. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt with release of fission product tritium into the salt. In most FHR and MSR systems, the baseline salts contain lithium where isotopically separated 7Li is proposed to minimize tritium production from neutron interactions with the salt.more » The Chinese Academy of Sciences plans to start operation of a 2-MW(thermal) molten salt test reactor by 2020. For high-magnetic-field fusion machines, the use of lithium enriched in 6Li is proposed to maximize tritium generation—the fuel for a fusion machine. Advances in superconductors that enable higher power densities may require the use of molten lithium salts for fusion blankets and as coolants. Recent technical advances in these three reactor classes have resulted in increased government and private interest and the beginning of a coordinated effort to address the tritium control challenges in 700°C liquid salt systems. In this paper, we describe characteristics of salt-cooled fission and fusion machines, the basis for growing interest in these technologies, tritium generation in molten salts, the environment for tritium capture, models for high-temperature tritium transport in salt systems, alternative strategies for tritium control, and ongoing experimental work. Several methods to control tritium appear viable. Finally, limited experimental data are the primary constraint for designing efficient cost-effective methods of tritium control.« less

Tritium Control and Capture in Salt-Cooled Fission and Fusion Reactors: Status, Challenges, and Path Forward

DOE PAGES

Forsberg, Charles W.; Lam, Stephen; Carpenter, David M.; ...

2017-02-26

Three advanced nuclear power systems use liquid salt coolants that generate tritium and thus face the common challenges of containing and capturing tritium to prevent its release to the environment. The fluoride salt–cooled high-temperature reactor (FHR) uses clean fluoride salt coolants and the same graphite-matrix coated-particle fuel as high-temperature gas-cooled reactors. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt with release of fission product tritium into the salt. In most FHR and MSR systems, the baseline salts contain lithium where isotopically separated 7Li is proposed to minimize tritium production from neutron interactions with the salt.more » The Chinese Academy of Sciences plans to start operation of a 2-MW(thermal) molten salt test reactor by 2020. For high-magnetic-field fusion machines, the use of lithium enriched in 6Li is proposed to maximize tritium generation—the fuel for a fusion machine. Advances in superconductors that enable higher power densities may require the use of molten lithium salts for fusion blankets and as coolants. Recent technical advances in these three reactor classes have resulted in increased government and private interest and the beginning of a coordinated effort to address the tritium control challenges in 700°C liquid salt systems. In this paper, we describe characteristics of salt-cooled fission and fusion machines, the basis for growing interest in these technologies, tritium generation in molten salts, the environment for tritium capture, models for high-temperature tritium transport in salt systems, alternative strategies for tritium control, and ongoing experimental work. Several methods to control tritium appear viable. Finally, limited experimental data are the primary constraint for designing efficient cost-effective methods of tritium control.« less

Supercell Depletion Studies for Prismatic High Temperature Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Ortensi

2012-10-01

The traditional two-step method of analysis is not accurate enough to represent the neutronic effects present in the prismatic high temperature reactor concept. The long range coupling of the various regions in high temperature reactors poses a set of challenges that are not seen in either LWRs or fast reactors. Unlike LWRs, which exhibit large, localized effects, the dominant effects in PMRs are, for the most part, distributed over larger regions, but with lower magnitude. The 1-D in-line treatment currently used in pebble bed reactor analysis is not sufficient because of the 2-D nature of the prismatic blocks. Considerable challengesmore » exist in the modeling of blocks in the vicinity of reflectors, which, for current small modular reactor designs with thin annular cores, include the majority of the blocks. Additional challenges involve the treatment of burnable poisons, operational and shutdown control rods. The use of a large domain for cross section preparation provides a better representation of the neutron spectrum, enables the proper modeling of BPs and CRs, allows the calculation of generalized equivalence theory parameters, and generates a relative power distribution that can be used in compact power reconstruction. The purpose of this paper is to quantify the effects of the reflector, burnable poison, and operational control rods on an LEU design and to delineate an analysis approach for the Idaho National Laboratory. This work concludes that the use of supercells should capture these long-range effects in the preparation of cross sections and along with a set of triangular meshes to treat BPs, and CRs a high fidelity neutronics computation is attainable.« less

Optimization of tritium breeding and shielding analysis to plasma in ITER fusion reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Indah Rosidah, M., E-mail: indah.maymunah@gmail.com; Suud, Zaki, E-mail: szaki@fi.itb.ac.id; Yazid, Putranto Ilham

The development of fusion energy is one of the important International energy strategies with the important milestone is ITER (International Thermonuclear Experimental Reactor) project, initiated by many countries, such as: America, Europe, and Japan who agreed to set up TOKAMAK type fusion reactor in France. In ideal fusion reactor the fuel is purely deuterium, but it need higher temperature of reactor. In ITER project the fuels are deuterium and tritium which need lower temperature of the reactor. In this study tritium for fusion reactor can be produced by using reaction of lithium with neutron in the blanket region. With themore » tritium breeding blanket which react between Li-6 in the blanket with neutron resulted from the plasma region. In this research the material used in each layer surrounding the plasma in the reactor is optimized. Moreover, achieving self-sufficiency condition in the reactor in order tritium has enough availability to be consumed for a long time. In order to optimize Tritium Breeding Ratio (TBR) value in the fusion reactor, there are several strategies considered here. The first requirement is making variation in Li-6 enrichment to be 60%, 70%, and 90%. But, the result of that condition can not reach TBR value better than with no enrichment. Because there is reduction of Li-7 percent when increasing Li-6 percent. The other way is converting neutron multiplier material with Pb. From this, we get TBR value better with the Be as neutron multiplier. Beside of TBR value, fusion reactor can analyze the distribution of neutron flux and dose rate of neutron to know the change of neutron concentration for each layer in reactor. From the simulation in this study, 97% neutron concentration can be absorbed by material in reactor, so it is good enough. In addition, it is required to analyze spectrum neutron energy in many layers in the fusion reactor such as in blanket, coolant, and divertor. Actually material in that layer can resist in high temperature and high pressure condition for more than ten years.« less

HYBRID SULFUR PROCESS REFERENCE DESIGN AND COST ANALYSIS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gorensek, M.; Summers, W.; Boltrunis, C.

2009-05-12

This report documents a detailed study to determine the expected efficiency and product costs for producing hydrogen via water-splitting using energy from an advanced nuclear reactor. It was determined that the overall efficiency from nuclear heat to hydrogen is high, and the cost of hydrogen is competitive under a high energy cost scenario. It would require over 40% more nuclear energy to generate an equivalent amount of hydrogen using conventional water-cooled nuclear reactors combined with water electrolysis compared to the proposed plant design described herein. There is a great deal of interest worldwide in reducing dependence on fossil fuels, whilemore » also minimizing the impact of the energy sector on global climate change. One potential opportunity to contribute to this effort is to replace the use of fossil fuels for hydrogen production by the use of water-splitting powered by nuclear energy. Hydrogen production is required for fertilizer (e.g. ammonia) production, oil refining, synfuels production, and other important industrial applications. It is typically produced by reacting natural gas, naphtha or coal with steam, which consumes significant amounts of energy and produces carbon dioxide as a byproduct. In the future, hydrogen could also be used as a transportation fuel, replacing petroleum. New processes are being developed that would permit hydrogen to be produced from water using only heat or a combination of heat and electricity produced by advanced, high temperature nuclear reactors. The U.S. Department of Energy (DOE) is developing these processes under a program known as the Nuclear Hydrogen Initiative (NHI). The Republic of South Africa (RSA) also is interested in developing advanced high temperature nuclear reactors and related chemical processes that could produce hydrogen fuel via water-splitting. This report focuses on the analysis of a nuclear hydrogen production system that combines the Pebble Bed Modular Reactor (PBMR), under development by PBMR (Pty.) Ltd. in the RSA, with the Hybrid Sulfur (HyS) Process, under development by the Savannah River National Laboratory (SRNL) in the US as part of the NHI. This work was performed by SRNL, Westinghouse Electric Company, Shaw, PBMR (Pty) Ltd., and Technology Insights under a Technical Consulting Agreement (TCA). Westinghouse Electric, serving as the lead for the PBMR process heat application team, established a cost-shared TCA with SRNL to prepare an updated HyS thermochemical water-splitting process flowsheet, a nuclear hydrogen plant preconceptual design and a cost estimate, including the cost of hydrogen production. SRNL was funded by DOE under the NHI program, and the Westinghouse team was self-funded. The results of this work are presented in this Final Report. Appendices have been attached to provide a detailed source of information in order to document the work under the TCA contract.« less

An overview of optical diagnostics developed for the Lockheed Martin compact fusion reactor

NASA Astrophysics Data System (ADS)

Sommers, Bradley; Raymond, Anthony; Gucker, Sarah; Lockheed Martin Compact Fusion Reactor Team

2017-10-01

The T4B experiment is a linear, encapsulated ring cusp confinement device, designed to develop a physics and technology basis for a follow-on high beta machine as part of the compact fusion reactor program. Toward this end, a collection of non-invasive optical diagnostics have been developed to investigate confinement, neutral beam heating, and source behavior on the T4B device. These diagnostics include: (1) a multipoint Thomson scattering system employing a 532 nm Nd:YAG laser and high throughput spectrometer to measure 1D profiles of electron density and temperature, (2) a dispersion interferometer utilizing a continuous-wave CO2 laser (10.6 μm) to measure time resolved, line-integrated electron density, and (3) a bolometer suite utilizing four AXUV photodiodes with 64 lines of sight to generate 2D reconstructions of total radiative power and soft x-ray emission (via beryllium filters). An overview of design methods, including laser systems, detection schemes, and data analysis techniques is presented as well as results to date.

(High temperature flaw assessment procedure)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ruggles, M.B.

1990-06-01

The Electric Power Research Institute (EPRI), the Japanese Central Research Institute of Electric Power Industry (CRIEPI), and the British Nuclear Electric (NE) are conducting joint studies in the field of liquid metal reactor development. The traveler is currently responsible for the EPRI/CRIEPI/NE High-Temperature Flaw Assessment Procedure activities at the Oak Ridge National Laboratory (ORNL). The traveler participated, on behalf of EPRI, in the EPRI/CRIEPI/NE specialist working session, the purpose of which was to produce the interim High-Temperature Flaw Assessment guide. The traveler also led discussions on the High-Temperature Flaw Assessment Procedure Phase 2 program plan, and on the plan formore » a new joint EPRI/CRIEPI/NE study in Inelastic Behavior and Failure Criteria for Modified 9Cr--1Mo Steel. The traveler visited Profs. K. Ikegami, Y. Asada, N. Ohno, T. Inoue, and K. Kaneko at the Tokyo Institute of Technology, the University of Tokyo, Nagoya University, Kyoto University, and Science University of Tokyo, respectively to hold discussions on research advances in the areas of high-temperature fracture mechanics, inelastic material behavior, and constitutive modeling. In addition, the traveler visited Kajima Corp. and Ohbayashi Corp. Technical Research Institute to collect information on research in the area of fiber reinforced concrete.« less

Injector nozzle for molten salt destruction of energetic waste materials

DOEpatents

Brummond, William A.; Upadhye, Ravindra S.

1996-01-01

An injector nozzle has been designed for safely injecting energetic waste materials, such as high explosives, propellants, and rocket fuels, into a molten salt reactor in a molten salt destruction process without premature detonation or back burn in the injection system. The energetic waste material is typically diluted to form a fluid fuel mixture that is injected rapidly into the reactor. A carrier gas used in the nozzle serves as a carrier for the fuel mixture, and further dilutes the energetic material and increases its injection velocity into the reactor. The injector nozzle is cooled to keep the fuel mixture below the decomposition temperature to prevent spontaneous detonation of the explosive materials before contact with the high-temperature molten salt bath.

Injector nozzle for molten salt destruction of energetic waste materials

DOEpatents

Brummond, W.A.; Upadhye, R.S.

1996-02-13

An injector nozzle has been designed for safely injecting energetic waste materials, such as high explosives, propellants, and rocket fuels, into a molten salt reactor in a molten salt destruction process without premature detonation or back burn in the injection system. The energetic waste material is typically diluted to form a fluid fuel mixture that is injected rapidly into the reactor. A carrier gas used in the nozzle serves as a carrier for the fuel mixture, and further dilutes the energetic material and increases its injection velocity into the reactor. The injector nozzle is cooled to keep the fuel mixture below the decomposition temperature to prevent spontaneous detonation of the explosive materials before contact with the high-temperature molten salt bath. 2 figs.

Evaluation and Optimization of a Supercritical Carbon Dioxide Power Conversion Cycle for Nuclear Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Edwin A. Harvego; Michael G. McKellar

2011-05-01

There have been a number of studies involving the use of gases operating in the supercritical mode for power production and process heat applications. Supercritical carbon dioxide (CO2) is particularly attractive because it is capable of achieving relatively high power conversion cycle efficiencies in the temperature range between 550°C and 750°C. Therefore, it has the potential for use with any type of high-temperature nuclear reactor concept, assuming reactor core outlet temperatures of at least 550°C. The particular power cycle investigated in this paper is a supercritical CO2 Recompression Brayton Cycle. The CO2 Recompression Brayton Cycle can be used as eithermore » a direct or indirect power conversion cycle, depending on the reactor type and reactor outlet temperature. The advantage of this cycle when compared to the helium Brayton Cycle is the lower required operating temperature; 550°C versus 850°C. However, the supercritical CO2 Recompression Brayton Cycle requires an operating pressure in the range of 20 MPa, which is considerably higher than the required helium Brayton cycle operating pressure of 8 MPa. This paper presents results of analyses performed using the UniSim process analyses software to evaluate the performance of the supercritical CO2 Brayton Recompression Cycle for different reactor outlet temperatures. The UniSim model assumed a 600 MWt reactor power source, which provides heat to the power cycle at a maximum temperature of between 550°C and 750°C. The UniSim model used realistic component parameters and operating conditions to model the complete power conversion system. CO2 properties were evaluated, and the operating range for the cycle was adjusted to take advantage of the rapidly changing conditions near the critical point. The UniSim model was then optimized to maximize the power cycle thermal efficiency at the different maximum power cycle operating temperatures. The results of the analyses showed that power cycle thermal efficiencies in the range of 40 to 50% can be achieved.« less

Fission fragment assisted reactor concept for space propulsion: Foil reactor

NASA Technical Reports Server (NTRS)

Wright, Steven A.

1991-01-01

The concept is to fabricate a reactor using thin films or foils of uranium, uranium oxide and then to coat them on substrates. These coatings would be made so thin as to allow the escaping fission fragments to directly heat a hydrogen propellant. The idea was studied of direct gas heating and direct gas pumping in a nuclear pumped laser program. Fission fragments were used to pump lasers. In this concept two substrates are placed opposite each other. The internal faces are coated with thin foil of uranium oxide. A few of the advantages of this technology are listed. In general, however, it is felt that if one look at all solid core nuclear thermal rockets or nuclear thermal propulsion methods, one is going to find that they all pretty much look the same. It is felt that this reactor has higher potential reliability. It has low structural operating temperatures, very short burn times, with graceful failure modes, and it has reduced potential for energetic accidents. Going to a design like this would take the NTP community part way to some of the very advanced engine designs, such as the gas core reactor, but with reduced risk because of the much lower temperatures.

Analysis of ORNL site temperature and humidity data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Willis, B.E.

1989-08-01

The Advanced Neutron Source (ANS) is planned as a new state-of-the-art facility for neutron research and is currently undergoing conceptual design at the Oak Ridge National Laboratory (ORNL). The current concept calls for a nuclear research reactor with an operating power near 350 MW and extensive experiment and user support facilities. Analyses have been undertaken to determine an acceptable design basis wet-bulb temperature range for the facility. Comparisons are drawn with the design wet-bulb temperature previously used for the High Flux Isotope Reactor (HFIR), which is located on an adjacent site a Oak Ridge. This report explains the importance ofmore » wet-bulb temperature to the reactor cooling system performance, and describes the analysis of available meteorological data, and presents the results and the recommendations for a wet-bulb temperature range for use as a part of the plant design basis conditions. 1 ref., 6 figs.« less

Method of and apparatus for removing silicon from a high temperature sodium coolant

DOEpatents

Yunker, W.H.; Christiansen, D.W.

1983-11-25

This patent discloses a method of and system for removing silicon from a high temperature liquid sodium coolant system for a nuclear reactor. The sodium is cooled to a temperature below the silicon saturation temperature and retained at such reduced temperature while inducing high turbulence into the sodium flow for promoting precipitation of silicon compounds and ultimate separation of silicon compound particles from the liquid sodium.

Method of and apparatus for removing silicon from a high temperature sodium coolant

DOEpatents

Yunker, Wayne H.; Christiansen, David W.

1987-05-05

A method of and system for removing silicon from a high temperature liquid sodium coolant system for a nuclear reactor. The sodium is cooled to a temperature below the silicon saturation temperature and retained at such reduced temperature while inducing high turbulence into the sodium flow for promoting precipitation of silicon compounds and ultimate separation of silicon compound particles from the liquid sodium.

Method of and apparatus for removing silicon from a high temperature sodium coolant

DOEpatents

Yunker, Wayne H.; Christiansen, David W.

1987-01-01

A method of and system for removing silicon from a high temperature liquid sodium coolant system for a nuclear reactor. The sodium is cooled to a temperature below the silicon saturation temperature and retained at such reduced temperature while inducing high turbulence into the sodium flow for promoting precipitation of silicon compounds and ultimate separation of silicon compound particles from the liquid sodium.

Apparatus to recover tritium from tritiated molecules

DOEpatents

Swansiger, William A.

1988-01-01

An apparatus for recovering tritium from tritiated compounds is provided, including a preheater for heating tritiated water and other co-injected tritiated compounds to temperatures of about 600.degree. C. and a reactor charged with a mixture of uranium and uranium dioxide for receiving the preheated mixture. The reactor vessel is preferably stainless steel of sufficient mass so as to function as a heat sink preventing the reactor side walls from approaching high temperatures. A disposable copper liner extends between the reaction chamber and stainless steel outer vessel to prevent alloying of the uranium with the outer vessel. The uranium dioxide functions as an insulating material and heat sink preventing the reactor side walls from attaining reaction temperatures to thereby minimize tritium permeation rates. The uranium dioxide also functions as a diluent to allow for volumetric expansion of the uranium as it is converted to uranium dioxide.

Effect of reactor temperature on direct growth of carbon nanomaterials on stainless steel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Edzatty, A. N., E-mail: nuredzatty@gmail.com; Syazwan, S. M., E-mail: mdsyazwan.sanusi@gmail.com; Norzilah, A. H., E-mail: norzilah@unimap.edu.my

Currently, carbon nanomaterials (CNMs) are widely used for various applications due to their extraordinary electrical, thermal and mechanical properties. In this work, CNMs were directly grown on the stainless steel (SS316) via chemical vapor deposition (CVD). Acetone was used as a carbon source and argon was used as carrier gas, to transport the acetone vapor into the reactor when the reaction occurred. Different reactor temperature such as 700, 750, 800, 850 and 900 °C were used to study their effect on CNMs growth. The growth time and argon flow rate were fixed at 30 minutes and 200 ml/min, respectively. Characterizationmore » of the morphology of the SS316 surface after CNMs growth using Scanning Electron Microscopy (SEM) showed that the diameter of grown-CNMs increased with the reactor temperature. Energy Dispersive X-ray (EDX) was used to analyze the chemical composition of the SS316 before and after CNMs growth, where the results showed that reduction of catalyst elements such as iron (Fe) and nickel (Ni) at high temperature (700 – 900 °C). Atomic Force Microscopy (AFM) analysis showed that the nano-sized hills were in the range from 21 to 80 nm. The best reactor temperature to produce CNMs was at 800 °C.« less

Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Kurt R.; Howard, Richard H.; Daily, Charles R.

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designsmore » allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.« less

Characteristics of DO, organic matter, and ammonium profile for practical-scale DHS reactor under various organic load and temperature conditions.

PubMed

Nomoto, Naoki; Ali, Muntjeer; Jayaswal, Komal; Iguchi, Akinori; Hatamoto, Masashi; Okubo, Tsutomu; Takahashi, Masanobu; Kubota, Kengo; Tagawa, Tadashi; Uemura, Shigeki; Yamaguchi, Takashi; Harada, Hideki

2018-04-01

Profile analysis of the down-flow hanging sponge (DHS) reactor was conducted under various temperature and organic load conditions to understand the organic removal and nitrification process for sewage treatment. Under high organic load conditions (3.21-7.89 kg-COD m -3  day -1 ), dissolved oxygen (DO) on the upper layer of the reactor was affected by organic matter concentration and water temperature, and sometimes reaches around zero. Almost half of the COD Cr was removed by the first layer, which could be attributed to the adsorption of organic matter on sponge media. After the first layer, organic removal proceeded along the first-order reaction equation from the second to the fourth layers. The ammoniacal nitrogen removal ratio decreased under high organic matter concentration (above 100 mg L -1 ) and low DO (less than 1 mg L -1 ) condition. Ammoniacal nitrogen removal proceeded via a zero-order reaction equation along the reactor height. In addition, the profile results of DO, COD Cr , and NH 3 -N were different in the horizontal direction. Thus, it is thought the concentration of these items and microbial activities were not in a uniform state even in the same sponge layer of the DHS reactor.

Preliminary Design of Critical Function Monitoring System of PGSFR

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

2015-07-01

A PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor) is under development at Korea Atomic Energy Research Institute. A critical function monitoring system of the PGSFR is preliminarily studied. The functions of CFMS are to display critical plant variables related to the safety of the plant during normal and accident conditions and guide the operators corrective actions to keep the plant in a safe condition and mitigate the consequences of accidents. The minimal critical functions of the PGSFR are composed of reactivity control, reactor core cooling, reactor coolant system integrity, primary heat transfer system(PHTS) heat removal, sodium water reaction mitigation, radiation controlmore » and containment conditions. The variables and alarm legs of each critical function of the PGSFR are as follows; - Reactivity control: The variables of reactivity control function are power range neutron flux instrumentation, intermediate range neutron flux instrumentation, source range neutron flux instrumentation, and control rod bottom contacts. The alarm leg to display the reactivity controls consists of status of control drop malfunction, high post trip power and thermal reactivity addition. - Reactor core cooling: The variables are PHTS sodium level, hot pool temperature of PHTS, subassembly exit temperature, cold pool temperature of the PHTS, PHTS pump current, and PHTS pump breaker status. The alarm leg consists of high core delta temperature, low sodium level of the PHTS, high subassembly exit temperature, and low PHTS pump load. - Reactor coolant system integrity: The variables are PHTS sodium level, cover gas pressure, and safeguard vessel sodium level. The alarm leg is composed of low sodium level of PHTS, high cover gas pressure and high sodium level of the safety guard vessel. - PHTS heat removal: The variables are PHTS sodium level, hot pool temperature of PHTS, core exit temperature, cold pool temperature of the PHTS, flow rate of passive residual heat removal system, flow rate of active residual heat removal system, and temperatures of air heat exchanger temperature of residual heat removal systems. The alarm legs are composed of two legs of a 'passive residual heat removal system not cooling' and 'active residual heat removal system not cooling'. - Sodium water reaction mitigation: The variables are intermediate heat transfer system(IHTS) pressure, pressure and temperature and level of sodium dump tank, the status of rupture disk, hydrogen concentration in IHTS and direct variable of sodium-water-reaction measure. The alarm leg consists of high IHTS pressure, the status of sodium water reaction mitigation system and the indication of direct measure. - Radiation control: The variables are radiation of PHTS, radiation of IHTS, and radiation of containment purge. The alarm leg is composed of high radiation of PHTS and IHTS, and containment purge system. - Containment condition: The variables are containment pressure, containment isolation status, and sodium fire. The alarm leg consists of high containment pressure, status of containment isolation and status of sodium fire. (authors)« less

Heat Pipe Solar Receiver for Oxygen Production of Lunar Regolith

NASA Astrophysics Data System (ADS)

Hartenstine, John R.; Anderson, William G.; Walker, Kara L.; Ellis, Michael C.

2009-03-01

A heat pipe solar receiver operating in the 1050° C range is proposed for use in the hydrogen reduction process for the extraction of oxygen from the lunar soil. The heat pipe solar receiver is designed to accept, isothermalize and transfer solar thermal energy to reactors for oxygen production. This increases the available area for heat transfer, and increases throughput and efficiency. The heat pipe uses sodium as the working fluid, and Haynes 230 as the heat pipe envelope material. Initial design requirements have been established for the heat pipe solar receiver design based on information from the NASA In-Situ Resource Utilization (ISRU) program. Multiple heat pipe solar receiver designs were evaluated based on thermal performance, temperature uniformity, and integration with the solar concentrator and the regolith reactor(s). Two designs were selected based on these criteria: an annular heat pipe contained within the regolith reactor and an annular heat pipe with a remote location for the reactor. Additional design concepts have been developed that would use a single concentrator with a single solar receiver to supply and regulate power to multiple reactors. These designs use variable conductance or pressure controlled heat pipes for passive power distribution management between reactors. Following the design study, a demonstration heat pipe solar receiver was fabricated and tested. Test results demonstrated near uniform temperature on the outer surface of the pipe, which will ultimately be in contact with the regolith reactor.

Development and characterization of lubricants for use near nuclear reactors in space vehicles

NASA Technical Reports Server (NTRS)

Robinson, G. L.; Akawie, R. I.; Gardos, M. N.; Krening, K. C.

1972-01-01

The synthesis and evaluation program was conducted to develop wide-temperature range lubricants suitable for use in space vehicles particularly in the vicinity of nuclear reactors. Synthetic approaches resulted in nonpolymeric, large molecular weight materials, all based on some combination of siloxane and aromatic groups. Evaluation of these materials indicated that certain tetramethyl and hexamethyl disiloxanes containing phenyl thiophenyl substituents are extremely promising with respect to radiation stability, wide temperature range, good lubricity, oxidation resistance and additive acceptance. The synthesis of fluids is discussed, and the equipment and methods used in evaluation are described, some of which were designed to evaluate micro-quantities of the synthesized lubricants.

LION4; LION; three-dimensional temperature distribution program. [CDC6600,7600; UNIVAC1108; IBM360,370; FORTRAN IV and ASCENT (CDC6600,7600), FORTRAN IV (UNIVAC1108A,B and IBM360,370)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Binney, E.J.

LION4 is a computer program for calculating one-, two-, or three-dimensional transient and steady-state temperature distributions in reactor and reactor plant components. It is used primarily for thermal-structural analyses. It utilizes finite difference techniques with first-order forward difference integration and is capable of handling a wide variety of bounding conditions. Heat transfer situations accommodated include forced and free convection in both reduced and fully-automated temperature dependent forms, coolant flow effects, a limited thermal radiation capability, a stationary or stagnant fluid gap, a dual dependency (temperature difference and temperature level) heat transfer, an alternative heat transfer mode comparison and selection facilitymore » combined with heat flux direction sensor, and any form of time-dependent boundary temperatures. The program, which handles time and space dependent internal heat generation, can also provide temperature dependent material properties with limited non-isotropic properties. User-oriented capabilities available include temperature means with various weightings and a complete heat flow rate surveillance system.CDC6600,7600;UNIVAC1108;IBM360,370; FORTRAN IV and ASCENT (CDC6600,7600), FORTRAN IV (UNIVAC1108A,B and IBM360,370); SCOPE (CDC6600,7600), EXEC8 (UNIVAC1108A,B), OS/360,370 (IBM360,370); The CDC6600 version plotter routine LAPL4 is used to produce the input required by the associated CalComp plotter for graphical output. The IBM360 version requires 350K for execution and one additional input/output unit besides the standard units.« less

THETRIS: A MICRO-SCALE TEMPERATURE AND GAS RELEASE MODEL FOR TRISO FUEL

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Ortensi; A.M. Ougouag

2011-12-01

The dominating mechanism in the passive safety of gas-cooled, graphite-moderated, high-temperature reactors (HTRs) is the Doppler feedback effect. These reactor designs are fueled with sub-millimeter sized kernels formed into TRISO particles that are imbedded in a graphite matrix. The best spatial and temporal representation of the feedback effect is obtained from an accurate approximation of the fuel temperature. Most accident scenarios in HTRs are characterized by large time constants and slow changes in the fuel and moderator temperature fields. In these situations a meso-scale, pebble and compact scale, solution provides a good approximation of the fuel temperature. Micro-scale models aremore » necessary in order to obtain accurate predictions in faster transients or when parameters internal to the TRISO are needed. Since these coated particles constitute one of the fundamental design barriers for the release of fission products, it becomes important to understand the transient behavior inside this containment system. An explicit TRISO fuel temperature model named THETRIS has been developed and incorporated into the CYNOD-THERMIX-KONVEK suite of coupled codes. The code includes gas release models that provide a simple predictive capability of the internal pressure during transients. The new model yields similar results to those obtained with other micro-scale fuel models, but with the added capability to analyze gas release, internal pressure buildup, and effects of a gap in the TRISO. The analyses show the instances when the micro-scale models improve the predictions of the fuel temperature and Doppler feedback. In addition, a sensitivity study of the potential effects on the transient behavior of high-temperature reactors due to the presence of a gap is included. Although the formation of a gap occurs under special conditions, its consequences on the dynamic behavior of the reactor can cause unexpected responses during fast transients. Nevertheless, the strong Doppler feedback forces the reactor to quickly stabilize.« less

Microstructural analysis of W-SiCf/SiC composite

NASA Astrophysics Data System (ADS)

Yoon, Hanki; Oh, Jeongseok; Kim, Gonho; Kim, Hyunsu; Takahashi, Heishichiro; Kohyama, Akira

2015-03-01

Continuous silicon carbide fiber-reinforced silicon carbide (SiCf/SiC) composites are promising structure candidates for future fusion power systems such as gas coolant fast channels, extreme high temperature reactor and fusion reactors, because of their intrinsic properties such as excellent mechanical properties, high thermal conductivity, good thermal-shock resistance as well as excellent physical and chemical stability in various environments under elevated temperature conditions. In this study, bonding of tungsten and SiCf/SiC was produced by hot-press method. Microstructure analyses were performed using SEM and TEM.

Solar-thermal fluid-wall reaction processing

DOEpatents

Weimer, Alan W.; Dahl, Jaimee K.; Lewandowski, Allan A.; Bingham, Carl; Buechler, Karen J.; Grothe, Willy

2006-04-25

The present invention provides a method for carrying out high temperature thermal dissociation reactions requiring rapid-heating and short residence times using solar energy. In particular, the present invention provides a method for carrying out high temperature thermal reactions such as dissociation of hydrocarbon containing gases and hydrogen sulfide to produce hydrogen and dry reforming of hydrocarbon containing gases with carbon dioxide. In the methods of the invention where hydrocarbon containing gases are dissociated, fine carbon black particles are also produced. The present invention also provides solar-thermal reactors and solar-thermal reactor systems.

Solar-Thermal Fluid-Wall Reaction Processing

DOEpatents

Weimer, A. W.; Dahl, J. K.; Lewandowski, A. A.; Bingham, C.; Raska Buechler, K. J.; Grothe, W.

2006-04-25

The present invention provides a method for carrying out high temperature thermal dissociation reactions requiring rapid-heating and short residence times using solar energy. In particular, the present invention provides a method for carrying out high temperature thermal reactions such as dissociation of hydrocarbon containing gases and hydrogen sulfide to produce hydrogen and dry reforming of hydrocarbon containing gases with carbon dioxide. In the methods of the invention where hydrocarbon containing gases are dissociated, fine carbon black particles are also produced. The present invention also provides solar-thermal reactors and solar-thermal reactor systems.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yung, Matthew M.; Stanton, Alexander R.; Iisa, Kristiina

Metal-impregnated (Ni or Ga) ZSM-5 catalysts were studied for biomass pyrolysis vapor upgrading to produce hydrocarbons using three reactors constituting a 100 000x change in the amount of catalyst used in experiments. Catalysts were screened for pyrolysis vapor phase upgrading activity in two small-scale reactors: (i) a Pyroprobe with a 10 mg catalyst in a fixed bed and (ii) a fixed-bed reactor with 500 mg of catalyst. The best performing catalysts were then validated with a larger scale fluidized-bed reactor (using ~1 kg of catalyst) that produced measurable quantities of bio-oil for analysis and evaluation of mass balances. Despite somemore » inherent differences across the reactor systems (such as residence time, reactor type, analytical techniques, mode of catalyst and biomass feed) there was good agreement of reaction results for production of aromatic hydrocarbons, light gases, and coke deposition. Relative to ZSM-5, Ni or Ga addition to ZSM-5 increased production of fully deoxygenated aromatic hydrocarbons and light gases. In the fluidized bed reactor, Ga/ZSM-5 slightly enhanced carbon efficiency to condensed oil, which includes oxygenates in addition to aromatic hydrocarbons, and reduced oil oxygen content compared to ZSM-5. Ni/ZSM-5, while giving the highest yield of fully deoxygenated aromatic hydrocarbons, gave lower overall carbon efficiency to oil but with the lowest oxygen content. Reaction product analysis coupled with fresh and spent catalyst characterization indicated that the improved performance of Ni/ZSM-5 is related to decreasing deactivation by coking, which keeps the active acid sites accessible for the deoxygenation and aromatization reactions that produce fully deoxygenated aromatic hydrocarbons. The addition of Ga enhances the dehydrogenation activity of the catalyst, which leads to enhanced olefin formation and higher fully deoxygenated aromatic hydrocarbon yields compared to unmodified ZSM-5. Catalyst characterization by ammonia temperature programmed desorption, surface area measurements, and postreaction temperature-programmed oxidation (TPO) also showed that the metal-modified zeolites retained a greater percentage of their initial acidity and surface area, which was consistent between the reactor scales. These results demonstrate that the trends observed with smaller (milligram to gram) catalyst reactors are applicable to larger, more industrially relevant (kg) scales to help guide catalyst research toward application.« less

Joint tests at INL and CEA of a transient hot wire needle probe for in-pile thermal conductivity measurement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daw, J.E.; Knudson, D.L.; Villard, J.F.

2015-07-01

Thermal conductivity is a key property that must be known for proper design, testing, and deployment of new fuels and structural materials in nuclear reactors. Thermal conductivity is highly dependent on the physical structure, chemical composition, and the state of the material. Typically, thermal conductivity changes that occur during irradiation are currently measured out-of-pile using a 'cook and look' approach. But repeatedly removing samples from a test reactor to make measurements is expensive, has the potential to disturb phenomena of interest, and only provides understanding of the sample's end state when each measurement is made. There are also limited thermo-physicalmore » property data available for advanced fuels; and such data are needed for simulation codes, the development of next generation reactors, and advanced fuels for existing nuclear plants. Being able to quickly characterize fuel thermal conductivity during irradiation can improve the fidelity of data, reduce costs of post-irradiation examinations, increase understanding of how fuels behave under irradiation, and confirm or improve existing thermal conductivity measurement techniques. This paper discusses efforts to develop and evaluate an innovative in-pile thermal conductivity sensor based on the transient hot wire thermal conductivity method (THWM), using a single needle probe (NP) containing a line heat source and thermocouple embedded in the fuel. The sensor that has been designed and manufactured by the Idaho National Laboratory (INL) includes a unique combination of materials, geometry, and fabrication techniques that make the hot wire method suitable for in-pile applications. In particular, efforts were made to minimize the influence of the sensor and maximize fuel hot-wire heating. The probe has a thermocouple-like construction with high temperature resistant materials that remain ductile while resisting transmutation and materials interactions. THWM-NP prototypes were fabricated for both room temperature proof-of-concept evaluations and high temperature testing. Evaluations have been performed jointly by the INL and the French Alternative Energies and Atomic Energy Commission (CEA), both in Idaho Falls (USA) and in Cadarache (France), in the framework of a collaborative program for instrumentation of Material Testing Reactors. Initial tests were conducted on samples with a large range of thermal conductivities and temperatures ranging from 20 deg. C to 600 deg. C. Particularly, tests were recently performed on a sample having thermal conductivity and dimensions similar to UO{sub 2} and MOX nuclear fuels, in order to validate the ability of this sensor to operate for in-pile characterization of Light Water Reactors fuels. The results of the tests already completed at INL and CEA indicate that the Transient Hot Wire Needle Probe offers an enhanced method for in-pile detection of thermal conductivity. (authors)« less

Fast quench reactor and method

DOEpatents

Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

2002-01-01

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

Fast quench reactor and method

DOEpatents

Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

1998-01-01

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

Fast quench reactor and method

DOEpatents

Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

2002-09-24

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

Design Report for the ½ Scale Air-Cooled RCCS Tests in the Natural convection Shutdown heat removal Test Facility (NSTF)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lisowski, D. D.; Farmer, M. T.; Lomperski, S.

The Natural convection Shutdown heat removal Test Facility (NSTF) is a large scale thermal hydraulics test facility that has been built at Argonne National Laboratory (ANL). The facility was constructed in order to carry out highly instrumented experiments that can be used to validate the performance of passive safety systems for advanced reactor designs. The facility has principally been designed for testing of Reactor Cavity Cooling System (RCCS) concepts that rely on natural convection cooling for either air or water-based systems. Standing 25-m in height, the facility is able to supply up to 220 kW at 21 kW/m 2 tomore » accurately simulate the heat fluxes at the walls of a reactor pressure vessel. A suite of nearly 400 data acquisition channels, including a sophisticated fiber optic system for high density temperature measurements, guides test operations and provides data to support scaling analysis and modeling efforts. Measurements of system mass flow rate, air and surface temperatures, heat flux, humidity, and pressure differentials, among others; are part of this total generated data set. The following report provides an introduction to the top level-objectives of the program related to passively safe decay heat removal, a detailed description of the engineering specifications, design features, and dimensions of the test facility at Argonne. Specifications of the sensors and their placement on the test facility will be provided, along with a complete channel listing of the data acquisition system.« less

DEMONSTRATION BULLETIN: FLAME REACTOR - HORSEHEAD RESOURCE DEVELOPMENT COMPANY, INC.

EPA Science Inventory

The Horsehead Resource Development Company, Inc. (HRD) Flame Reactor is a patented and proven high temperature thermal process designed to safely treat industrial residues and wastes containing metals. During processing, the waste material is introduced into the hottest portio...

Progress in the Production of JP-8 Based Hydrogen and Advanced Tactical Fuels for Military Applications

DTIC Science & Technology

2011-02-01

of a multi- year program to develop, optimize, and demonstrate the military viability of a technology for on-demand production of high...continuous reactor system used for kinetic rate data experiment 86 52 Schematic of a differential reactor. The catalyst bed is kept small , and...program to develop, optimize, and demonstrate the military viability of a technology for on-demand production of high-pressure hydrogen for fuel

Can high fields save the tokamak? The challenge of steady-state operation for low cost compact reactors

NASA Astrophysics Data System (ADS)

Freidberg, Jeffrey; Dogra, Akshunna; Redman, William; Cerfon, Antoine

2016-10-01

The development of high field, high temperature superconductors is thought to be a game changer for the development of fusion power based on the tokamak concept. We test the validity of this assertion for pilot plant scale reactors (Q 10) for two different but related missions: pulsed operation and steady-state operation. Specifically, we derive a set of analytic criteria that determines the basic design parameters of a given fusion reactor mission. As expected there are far more constraints than degrees of freedom in any given design application. However, by defining the mission of the reactor under consideration, we have been able to determine the subset of constraints that drive the design, and calculate the values for the key parameters characterizing the tokamak. Our conclusions are as follows: 1) for pulsed reactors, high field leads to more compact designs and thus cheaper reactors - high B is the way to go; 2) steady-state reactors with H-mode like transport are large, even with high fields. The steady-state constraint is hard to satisfy in compact designs - high B helps but is not enough; 3) I-mode like transport, when combined with high fields, yields relatively compact steady-state reactors - why is there not more research on this favorable transport regime?

Silver (Ag) Transport Mechanisms in TRISO Coated Particles: A Critical Review

DOE Office of Scientific and Technical Information (OSTI.GOV)

IJ van Rooyen; ML Dunzik-Gougar; PM van Rooyen

2014-05-01

Transport of 110mAg in the intact SiC layer of TRISO coated particles has been studied for approximately 30 years without arriving at a satisfactory explanation of the transport mechanism. In this paper the possible mechanisms postulated in previous experimental studies, both in-reactor and out-of reactor research environment studies are critically reviewed and of particular interest are relevance to very high temperature gas reactor operating and accident conditions. Among the factors thought to influence Ag transport are grain boundary stoichiometry, SiC grain size and shape, the presence of free silicon, nano-cracks, thermal decomposition, palladium attack, transmutation products, layer thinning and coatedmore » particle shape. Additionally new insight to nature and location of fission products has been gained via recent post irradiation electron microscopy examination of TRISO coated particles from the DOE’s fuel development program. The combined effect of critical review and new analyses indicates a direction for investigating possible the Ag transport mechanism including the confidence level with which these mechanisms may be experimentally verified.« less

Silver (Ag) Transport Mechanisms in TRISO coated particles: A Critical Review

DOE Office of Scientific and Technical Information (OSTI.GOV)

I J van Rooyen; J H Neethling; J A A Engelbrecht

2012-10-01

Transport of 110mAg in the intact SiC layer of TRISO coated particles has been studied for approximately 30 years without arriving at a satisfactory explanation of the transport mechanism. In this paper the possible mechanisms postulated in previous experimental studies, both in-reactor and out-of reactor research environment studies are critically reviewed and of particular interest are relevance to very high temperature gas reactor operating and accident conditions. Among the factors thought to influence Ag transport are grain boundary stoichiometry, SiC grain size and shape, the presence of free silicon, nano-cracks, thermal decomposition, palladium attack, transmutation products, layer thinning and coatedmore » particle shape. Additionally new insight to nature and location of fission products has been gained via recent post irradiation electron microscopy examination of TRISO coated particles from the DOE’s fuel development program. The combined effect of critical review and new analyses indicates a direction for investigating possible the Ag transport mechanism including the confidence level with which these mechanisms may be experimentally verified.« less

Solution of heat removal from nuclear reactors by natural convection

NASA Astrophysics Data System (ADS)

Zitek, Pavel; Valenta, Vaclav

2014-03-01

This paper summarizes the basis for the solution of heat removal by natural convection from both conventional nuclear reactors and reactors with fuel flowing coolant (such as reactors with molten fluoride salts MSR).The possibility of intensification of heat removal through gas lift is focused on. It might be used in an MSR (Molten Salt Reactor) for cleaning the salt mixture of degassed fission products and therefore eliminating problems with iodine pitting. Heat removal by natural convection and its intensification increases significantly the safety of nuclear reactors. Simultaneously the heat removal also solves problems with lifetime of pumps in the primary circuit of high-temperature reactors.

Experimental and numerical investigations of high temperature gas heat transfer and flow in a VHTR reactor core

NASA Astrophysics Data System (ADS)

Valentin Rodriguez, Francisco Ivan

High pressure/high temperature forced and natural convection experiments have been conducted in support of the development of a Very High Temperature Reactor (VHTR) with a prismatic core. VHTRs are designed with the capability to withstand accidents by preventing nuclear fuel meltdown, using passive safety mechanisms; a product of advanced reactor designs including the implementation of inert gases like helium as coolants. The present experiments utilize a high temperature/high pressure gas flow test facility constructed for forced and natural circulation experiments. This work examines fundamental aspects of high temperature gas heat transfer applied to VHTR operational and accident scenarios. Two different types of experiments, forced convection and natural circulation, were conducted under high pressure and high temperature conditions using three different gases: air, nitrogen and helium. The experimental data were analyzed to obtain heat transfer coefficient data in the form of Nusselt numbers as a function of Reynolds, Grashof and Prandtl numbers. This work also examines the flow laminarization phenomenon (turbulent flows displaying much lower heat transfer parameters than expected due to intense heating conditions) in detail for a full range of Reynolds numbers including: laminar, transition and turbulent flows under forced convection and its impact on heat transfer. This phenomenon could give rise to deterioration in convection heat transfer and occurrence of hot spots in the reactor core. Forced and mixed convection data analyzed indicated the occurrence of flow laminarization phenomenon due to the buoyancy and acceleration effects induced by strong heating. Turbulence parameters were also measured using a hot wire anemometer in forced convection experiments to confirm the existence of the flow laminarization phenomenon. In particular, these results demonstrated the influence of pressure on delayed transition between laminar and turbulent flow. The heat dissipating capabilities of helium flow, due to natural circulation in the system at both high and low pressure, were also examined. These experimental results are useful for the development and validation of VHTR design and safety analysis codes. Numerical simulations were performed using a Multiphysics computer code, COMSOL, displaying less than 5% error between the measured graphite temperatures in both the heated and cooled channels. Finally, new correlations have been proposed describing the thermal-hydraulic phenomena in buoyancy driven flows in both heated and cooled channels.

Further Development of Crack Growth Detection Techniques for US Test and Research Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kohse, Gordon; Carpenter, David M.; Ostrovsky, Yakov

One of the key issues facing Light Water Reactors (LWRs) in extending lifetimes beyond 60 years is characterizing the combined effect of irradiation and water chemistry on material degradation and failure. Irradiation Assisted Stress Corrosion Cracking (IASCC), in which a crack propagates in a susceptible material under stress in an aggressive environment, is a mechanism of particular concern. Full understanding of IASCC depends on real time crack growth data acquired under relevant irradiation conditions. Techniques to measure crack growth in actively loaded samples under irradiation have been developed outside the US - at the Halden Boiling Water Reactor, for example.more » Several types of IASCC tests have also been deployed at the MITR, including passively loaded crack growth measurements and actively loaded slow strain rate tests. However, there is not currently a facility available in the US to measure crack growth on actively loaded, pre-cracked specimens in LWR irradiation environments. A joint program between the Idaho National Laboratory (INL) and the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory (NRL) is currently underway to develop and demonstrate such a capability for US test and research reactors. Based on the Halden design, the samples will be loaded using miniature high pressure bellows and a compact loading mechanism, with crack length measured in real time using the switched Direct Current Potential Drop (DCPD) method. The basic design and initial mechanical testing of the load system and implementation of the DCPD method have been previously reported. This paper presents the results of initial autoclave testing at INL and the adaptation of the design for use in the high pressure, high temperature water loop at the MITR 6 MW research reactor, where an initial demonstration is planned in mid-2015. Materials considerations for the high pressure bellows are addressed. Design modifications to the loading mechanism required by the size constraints of the MITR water loop are described. The safety case for operation of the high pressure gas-driven bellows mechanism is also presented. Key issues are the design and response of systems to limit gas flow in the event of a high pressure gas leak in the in-core autoclave. Integrity of the autoclave must be maintained and reactivity effects due to voiding of the loop coolant must be shown to be within the reactor technical specifications. The technical development of the crack growth monitor for application in the INL Advanced Test Reactor or the MITR can act as a template for adaptation of this technology in other reactors. (authors)« less

Large-break LOCA, in-reactor fuel bundle Materials Test MT-6A

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wilson, C.L.; Hesson, G.M.; Pilger, J.P.

1993-09-01

This is a report on one of a series of experiments to simulates a loss-of-coolant accident (LOCA) using full-length fuel rods for pressurized water reactors (PWR). The experiments were conducted by Pacific Northwest Laboratory (PNL) under the LOCA simulation Program sponsored by the US Nuclear Regulatory Commission (NRC). The major objective of this program was causing the maximum possible expansion of the cladding on the fuel rods from a short-term adiabatic temperature transient to 1200 K (1700 F) leading to the rupture of the cladding; and second, by reflooding the fuel rods to determine the rate at which the fuelmore » bundle is cooled.« less

The RERTR Program status and progress

DOE Office of Scientific and Technical Information (OSTI.GOV)

Travelli, A.

1995-12-01

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1995 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 1994. The revelation that Iraq was on the verge of developing a nuclear weapon at the time of the Gulf War, and that it was planning to do so by extracting HEU from the fuel of its research reactors, has given new impetus and urgency to the RERTR commitment of eliminating HEU use in research and test reactors worldwide.more » Development of advanced LEU research reactor fuels is scheduled to begin in October 1995. The Russian RERTR program, which aims to develop and demonstrate within the next five years the technical means needed to convert Russian-supplied research reactors to LEU fuels, is now in operation. A Statement of Intent was signed by high US and Chinese officials, endorsing cooperative activities between the RERTR program and Chinese laboratories involved in similar activities. Joint studies of LEU technical feasibility were completed for the SAFARI-I reactor in South Africa and for the ANS reactor in the US. A new study has been initiated for the FRM-II reactor in Germany. Significant progress was made on several aspects of producing {sup 99}Mo from fission targets utilizing LEU instead of HEU. A cooperation agreements is in place with the Indonesian BATAN. The first prototypical irradiation of an LEU metal-foil target for {sup 99}Mo production was accomplished in Indonesia. The TR-2 reactor, in Turkey, began conversion. SAPHIR, in Switzerland, was shut down. LEU fuel fabrication has begun for the conversion of two more US reactors. Twelve foreign reactors and nine domestic reactors have been fully converted. Approximately 60 % of the work required to eliminate the use of HEU in US-supplied research reactors has been accomplished.« less

METALLURGY DIVISION QUARTERLY REPORT FOR JULY, AUGUST, AND SEPTEMBER 1957

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1958-10-01

Advanced Water Reactor Program. Three firings were made of initial closed-porosity fuel pellet bodies. Each firing coatained pellets of the composition 90 wt.% ThO/sub 2/-10 wt.%fl U0/sub 2/ with various additives and firing variables. Fast Power Breeder Reactor Program. To determine the potential usefulness of a Zr-5 wt. % Pu alloy, the fabricability of the alloy was tested. The manufacture of rod stock from which fuel and blanket elements for the Mark III loading of the EBR-1 were prcduced has been completed. The effect of irradiation on extruded and heat-treated U-2 wt.% Zr alloy for the EBR- 1 is reported.more » Fabrication procedures for making graphite-U/sub 3/O/sub 8/ test specimens for the TREAT Reactor were investigated. Advanced Engineering and Development. Ultrasonic bond tests were conducted on 590 EBR-1 Mark III blanket fuel elemeats. The blanket rods and part of the fuel rcds for the EBR-1 Mark III loading are being checked for cladding thickness by an eddy current system. Investigations of corrosionresistant Zr-Nb alloy were coatinued. Corrosion of MR alloys is being studied Ln support of the Mighty Mouse reactor program. Dynamic corrosion tests were performed on aluminum alloys, and results are included. Prcduction, Treatment, and Properties of Materials. The progress of the program of preparing highpurity Pu by fused salt electrolysis is summarized. Velocities of ultrasonic waves propagated in directions suitable for determining the room- temperature elastic moduli C/sub 12/, C/sub 13/, and C/sub 23/ of alpha U were determined. investigation of recrystallization in heavily coldrolled alpha- uranium sheet without a texture change was essentially concluded during this quarter. Selfdiffasion runs in polycrystalline uranium in the gamma phase, using the sputtering technique, have yielded a tentative value for the diffusion coefficient between 10/sup -8/ and 10/sup -7/ cm/sup 2/second. The preparation of high-purity U-Pan alloys is reponted. The data for the alpha-tobeta transformation temperatures in high-purity U and U-C alloys were confirmed by experiments on new specimens. Microstructure, density, and thermal arrest data were obtained for several injection cast, nominal U-5 wt.%fl fissium and U-8 wt.%fl fissium alloys. Phase diagrams are preseated for U-Mo and U-Ru alloys. Alloy Theory and The Nature of Solids. Four new isomorphs of Ti/sub 2/Ni have been discovered. Effects of Irradiation on Materials. The experimental and analytical work on the radial distribution of thermal neutrons within cylindrically shaped fuel specimens during irradiation was completed. (For preceding period see ANL-5790.) (W.L.H.)« less

Comprehensive investigation of HgCdTe metalorganic chemical vapor deposition

NASA Technical Reports Server (NTRS)

Raupp, Gregory B.

1993-01-01

The principal objective of this experimental and theoretical research program was to explore the possibility of depositing high quality epitaxial CdTe and HgCdTe at very low pressures through metalorganic chemical vapor deposition (MOCVD). We explored two important aspects of this potential process: (1) the interaction of molecular flow transport and deposition in an MOCVD reactor with a commercial configuration, and (2) the kinetics of metal alkyl source gas adsorption, decomposition and desorption from the growing film surface using ultra high vacuum surface science reaction techniques. To explore the transport-reaction issue, we have developed a reaction engineering analysis of a multiple wafer-in-tube ultrahigh vacuum chemical vapor deposition (UHV/CVD) reactor which allows an estimate of wafer or substrate throughput for a reactor of fixed geometry and a given deposition chemistry with specified film thickness uniformity constraints. The model employs a description of ballistic transport and reaction based on the pseudo-steady approximation to the Boltzmann equation in the limit of pure molecular flow. The model representation takes the form of an integral equation for the flux of each reactant or intermediate species to the wafer surfaces. Expressions for the reactive sticking coefficients (RSC) for each species must be incorporated in the term which represents reemission from a wafer surface. The interactions of MOCVD precursors with Si and CdTe were investigated using temperature programmed desorption (TPD) in ultra high vacuum combined with Auger electron spectroscopy (AES). These studies revealed that diethyltellurium (DETe) and dimethylcadmium (DMCd) adsorb weakly on clean Si(100) and desorb upon heating without decomposing. These precursors adsorb both weakly and strongly on CdTe(111)A, with DMCd exhibiting the stronger interaction with the surface than DETe.

A model to predict thermal conductivity of irradiated U-Mo dispersion fuel

NASA Astrophysics Data System (ADS)

Burkes, Douglas E.; Huber, Tanja K.; Casella, Andrew M.

2016-05-01

Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world's remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on assisting with the development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layer formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of the modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.

Development of Yield and Tensile Strength Design Curves for Alloy 617

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nancy Lybeck; T. -L. Sham

2013-10-01

The U.S. Department of Energy Very High Temperature Reactor Program is acquiring data in preparation for developing an Alloy 617 Code Case for inclusion in the nuclear section of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code. A draft code case was previously developed, but effort was suspended before acceptance by ASME. As part of the draft code case effort, a database was compiled of yield and tensile strength data from tests performed in air. Yield strength and tensile strength at temperature are used to set time independent allowable stress for construction materials in B&PVmore » Code, Section III, Subsection NH. The yield and tensile strength data used for the draft code case has been augmented with additional data generated by Idaho National Laboratory and Oak Ridge National Laboratory in the U.S. and CEA in France. The standard ASME Section II procedure for generating yield and tensile strength at temperature is presented, along with alternate methods that accommodate the change in temperature trends seen at high temperatures, resulting in a more consistent design margin over the temperature range of interest.« less

Light water reactor lower head failure analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rempe, J.L.; Chavez, S.A.; Thinnes, G.L.

1993-10-01

This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broadermore » range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.« less

Nuclear Fuels & Materials Spotlight Volume 5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Petti, David Andrew

2016-10-01

As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • Evaluation and modeling of light water reactor accident tolerant fuel concepts • Status and results of recent TRISO-coated particle fuel irradiations, post-irradiation examinations, high-temperature safety testing to demonstrate the accident performance of this fuel system, and advanced microscopy to improve the understanding of fission product transport in this fuel system.more » • Improvements in and applications of meso and engineering scale modeling of light water reactor fuel behavior under a range of operating conditions and postulated accidents (e.g., power ramping, loss of coolant accident, and reactivity initiated accidents) using the MARMOT and BISON codes. • Novel measurements of the properties of nuclear (actinide) materials under extreme conditions, (e.g. high pressure, low/high temperatures, high magnetic field) to improve the scientific understanding of these materials. • Modeling reactor pressure vessel behavior using the GRIZZLY code. • New methods using sound to sense temperature inside a reactor core. • Improved experimental capabilities to study the response of fusion reactor materials to a tritium plasma. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at Idaho National Laboratory, and hope that you find this issue informative.« less

Modeling a Packed Bed Reactor Utilizing the Sabatier Process

NASA Technical Reports Server (NTRS)

Shah, Malay G.; Meier, Anne J.; Hintze, Paul E.

2017-01-01

A numerical model is being developed using Python which characterizes the conversion and temperature profiles of a packed bed reactor (PBR) that utilizes the Sabatier process; the reaction produces methane and water from carbon dioxide and hydrogen. While the specific kinetics of the Sabatier reaction on the RuAl2O3 catalyst pellets are unknown, an empirical reaction rate equation1 is used for the overall reaction. As this reaction is highly exothermic, proper thermal control is of the utmost importance to ensure maximum conversion and to avoid reactor runaway. It is therefore necessary to determine what wall temperature profile will ensure safe and efficient operation of the reactor. This wall temperature will be maintained by active thermal controls on the outer surface of the reactor. Two cylindrical PBRs are currently being tested experimentally and will be used for validation of the Python model. They are similar in design except one of them is larger and incorporates a preheat loop by feeding the reactant gas through a pipe along the center of the catalyst bed. The further complexity of adding a preheat pipe to the model to mimic the larger reactor is yet to be implemented and validated; preliminary validation is done using the smaller PBR with no reactant preheating. When mapping experimental values of the wall temperature from the smaller PBR into the Python model, a good approximation of the total conversion and temperature profile has been achieved. A separate CFD model incorporates more complex three-dimensional effects by including the solid catalyst pellets within the domain. The goal is to improve the Python model to the point where the results of other reactor geometry can be reasonably predicted relatively quickly when compared to the much more computationally expensive CFD approach. Once a reactor size is narrowed down using the Python approach, CFD will be used to generate a more thorough prediction of the reactors performance.

Generating unstructured nuclear reactor core meshes in parallel

DOE PAGES

Jain, Rajeev; Tautges, Timothy J.

2014-10-24

Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor coremore » examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.« less

A reactor for high-throughput high-pressure nuclear magnetic resonance spectroscopy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beach, N. J.; Knapp, S. M. M.; Landis, C. R., E-mail: landis@chem.wisc.edu

The design of a reactor for operando nuclear magnetic resonance (NMR) monitoring of high-pressure gas-liquid reactions is described. The Wisconsin High Pressure NMR Reactor (WiHP-NMRR) design comprises four modules: a sapphire NMR tube with titanium tube holder rated for pressures as high as 1000 psig (68 atm) and temperatures ranging from −90 to 90 °C, a gas circulation system that maintains equilibrium concentrations of dissolved gases during gas-consuming or gas-releasing reactions, a liquid injection apparatus that is capable of adding measured amounts of solutions to the reactor under high pressure conditions, and a rapid wash system that enables the reactor tomore » be cleaned without removal from the NMR instrument. The WiHP-NMRR is compatible with commercial 10 mm NMR probes. Reactions performed in the WiHP-NMRR yield high quality, information-rich, and multinuclear NMR data over the entire reaction time course with rapid experimental turnaround.« less

Generic Stellarator-like Magnetic Fusion Reactor

NASA Astrophysics Data System (ADS)

Sheffield, John; Spong, Donald

2015-11-01

The Generic Magnetic Fusion Reactor paper, published in 1985, has been updated, reflecting the improved science and technology base in the magnetic fusion program. Key changes beyond inflation are driven by important benchmark numbers for technologies and costs from ITER construction, and the use of a more conservative neutron wall flux and fluence in modern fusion reactor designs. In this paper the generic approach is applied to a catalyzed D-D stellarator-like reactor. It is shown that an interesting power plant might be possible if the following parameters could be achieved for a reference reactor: R/ < a > ~ 4 , confinement factor, fren = 0.9-1.15, < β > ~ 8 . 0 -11.5 %, Zeff ~ 1.45 plus a relativistic temperature correction, fraction of fast ions lost ~ 0.07, Bm ~ 14-16 T, and R ~ 18-24 m. J. Sheffield was supported under ORNL subcontract 4000088999 with the University of Tennessee.

Development of Advanced ISS-WPA Catalysts for Organic Oxidation at Reduced Pressure/Temperature

NASA Technical Reports Server (NTRS)

Yu, Ping; Nalette, Tim; Kayatin, Matthew

2016-01-01

The Water Processor Assembly (WPA) at International Space Station (ISS) processes a waste stream via multi-filtration beds, where inorganic and non-volatile organic contaminants are removed, and a catalytic reactor, where low molecular weight organics not removed by the adsorption process are oxidized at elevated pressure in the presence of oxygen and elevated temperature above the normal water boiling point. Operation at an elevated pressure requires a more complex system design compared to a reactor that could operate at ambient pressure. However, catalysts currently available have insufficient activity to achieve complete oxidation of the organic load at a temperature less than the water boiling point and ambient pressure. Therefore, it is highly desirable to develop a more active and efficient catalyst at ambient pressure and a moderate temperature that is less than water boiling temperature. This paper describes our efforts in developing high efficiency water processing catalysts. Different catalyst support structures and coating metals were investigated in subscale reactors and results were compared against the flight WPA catalyst. Detailed improvements achieved on alternate metal catalysts at ambient pressure and 200 F will also be presented in the paper.

Considerations of Alloy 617 Application in the Gen IV Nuclear Reactor Systems - Part I: Mechanical Property Challenges

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ren, Weiju

2010-01-01

Alloy 617 is currently considered as a leading candidate material for high temperature components in the Gen IV Nuclear Reactor Systems. Because of the unprecedented severe working conditions beyond its commercial service experience required by the Gen IV systems, the alloy faces various challenges in both mechanical and metallurgical properties. This paper, as Part I of the discussion, is focused on the challenges and issues in the mechanical properties of Alloy 617 for the intended nuclear application. Considerations are given in details in its mechanical property data scatter, low creep strength in the desired high temperature range, lack of longtermmore » creep curves, high loading rate dependency, and preponderant tertiary creep. Some research and development activities are suggested with discussions on their viability to satisfy the Gen IV Nuclear Reactor System needs in near future and in the long run.« less

Jet pump-drive system for heat removal

NASA Technical Reports Server (NTRS)

French, James R. (Inventor)

1987-01-01

The invention does away with the necessity of moving parts such as a check valve in a nuclear reactor cooling system. Instead, a jet pump, in combination with a TEMP, is employed to assure safe cooling of a nuclear reactor after shutdown. A main flow exists for a reactor coolant. A point of withdrawal is provided for a secondary flow. A TEMP, responsive to the heat from said coolant in the secondary flow path, automatically pumps said withdrawn coolant to a higher pressure and thus higher velocity compared to the main flow. The high velocity coolant is applied as a driver flow for the jet pump which has a main flow chamber located in the main flow circulation pump. Upon nuclear shutdown and loss of power for the main reactor pumping system, the TEMP/jet pump combination continues to boost the coolant flow in the direction it is already circulating. During the decay time for the nuclear reactor, the jet pump keeps running until the coolant temperature drops to a lower and safe temperature where the heat is no longer a problem. At this lower temperature, the TEMP/jet pump combination ceases its circulation boosting operation. When the nuclear reactor is restarted and the coolant again exceeds the lower temperature setting, the TEMP/jet pump automatically resumes operation. The TEMP/jet pump combination is thus automatic, self-regulating and provides an emergency pumping system free of moving parts.

Multi-Megawatt Power System Trade Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Longhurst, Glen Reed; Schnitzler, Bruce Gordon; Parks, Benjamin Travis

2001-11-01

As part of a larger task, the Idaho National Engineering and Environmental Laboratory (INEEL) was tasked to perform a trade study comparing liquid-metal cooled reactors having Rankine power conversion systems with gas-cooled reactors having Brayton power conversion systems. This report summarizes the approach, the methodology, and the results of that trade study. Findings suggest that either approach has the possibility to approach the target specific mass of 3-5 kg/kWe for the power system, though it appears either will require improvements to achieve that. Higher reactor temperatures have the most potential for reducing the specific mass of gas-cooled reactors but domore » not necessarily have a similar effect for liquid-cooled Rankine systems. Fuels development will be the key to higher reactor operating temperatures. Higher temperature turbines will be important for Brayton systems. Both replacing lithium coolant in the primary circuit with gallium and replacing potassium with sodium in the power loop for liquid systems increase system specific mass. Changing the feed pump turbine to an electric motor in Rankine systems has little effect. Key technologies in reducing specific mass are high reactor and radiator operating temperatures, low radiator areal density, and low turbine/generator system masses. Turbine/generator mass tends to dominate overall power system mass for Rankine systems. Radiator mass was dominant for Brayton systems.« less

Local Limit Phenomena, Flow Compression, and Fuel Cracking Effects in High-Speed Turbulent Flames

DTIC Science & Technology

2015-06-01

e.g. local extinction and re- ignition , interactions between flow compression and fast-reaction induced dilatation (reaction compression ), and to...time as a function of initial temperature in constant-pressure auto - ignition , and (b) the S-curves of perfectly stirred reactors (PSRs), for n...mechanism. The reduction covered auto - ignition and perfectly stirred reactors for equivalence ratio range of 0.5~1.5, initial temperature higher than

Down-selection of candidate alloys for further testing of advanced replacement materials for LWR core internals

DOE Office of Scientific and Technical Information (OSTI.GOV)

Was, Gary; Leonard, Keith J.; Tan, Lizhen

Life extension of the existing nuclear reactors imposes irradiation of high fluences to structural materials, resulting in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) Light Water Reactor Sustainability Program to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to identify and develop advanced alloys with superiormore » degradation resistance in light water reactor (LWR)-relevant environments by 2024.« less

Novel inorganic nanomaterials generated with highly concentrated sunlight

NASA Astrophysics Data System (ADS)

Gordon, Jeffrey M.; Katz, Eugene A.; Feuermann, Daniel; Albu-Yaron, Ana; Levy, Moshe; Tenne, Reshef

2008-08-01

Reactors driven by highly concentrated sunlight can create conditions well suited to the synthesis of inorganic nanomaterials. We report the experimental realization of a broad range of closed-cage (fullerene-like) nanostructures, nanotubes and/or nanowires for MoS2, SiO2 and Si, achieved via solar ablation. The solar technique generates the strong temperature and radiative gradients - in addition to the extensive high-temperature annealing environment - conducive to producing such nanostructures. The identity of the nanostructures was established with TEM, HRTEM and EDS. The fullerene-like and nanotube MoS2 configurations achieved fundamentally minimum sizes predicted by molecular structural theory. Furthermore, our experiments represent the first time SiO2 nanofibers and nanospheres have been produced purely from quartz. The solar route is far less energy intensive than laser ablation and other high-temperature chemical reactors, simpler and less costly.

Conceptual design of quadriso particles with europium burnable absorber in HTRS.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Talamo, A.; Nuclear Engineering Division

2010-05-18

In High Temperature Reactors, burnable absorbers are utilized to manage the excess reactivity at the early stage of the fuel cycle. In this study QUADRISO particles are proposed to manage the initial xcess reactivity of High Temperature Reactors. The QUADRISO concept synergistically couples the decrease of the burnable poison with the decrease of the fissile materials at the fuel particle level. This echanism is set up by introducing a burnable poison layer around the fuel kernel in ordinary TRISO particles or by mixing the burnable poison with any of the TRISO coated layers. At the beginning of life, the nitialmore » excess reactivity is small because some neutrons are absorbed in the burnable poison and they are prevented from entering the fuel kernel. At the end of life, when the absorber is almost depleted, ore eutrons stream into the fuel kernel of QUADRISO particles causing fission reactions. The mechanism has been applied to a prismatic High Temperature Reactor with europium or erbium burnable absorbers, showing a significant reduction in the initial excess reactivity of the core.« less

A novel concept of QUADRISO particles. Part II: Utilization for excess reactivity control.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Talamo, A.

2010-07-01

In high temperature reactors, burnable absorbers are utilized to manage the excess reactivity at the early stage of the fuel cycle. In this paper QUADRISO particles are proposed to manage the initial excess reactivity of high temperature reactors. The QUADRISO concept synergistically couples the decrease of the burnable poison with the decrease of the fissile materials at the fuel particle level. This mechanism is set up by introducing a burnable poison layer around the fuel kernel in ordinary TRISO particles or by mixing the burnable poison with any of the TRISO coated layers. At the beginning of life, the initialmore » excess reactivity is small because some neutrons are absorbed in the burnable poison and they are prevented from entering the fuel kernel. At the end of life, when the absorber is almost depleted, more neutrons stream into the fuel kernel of QUADRISO particles causing fission reactions. The mechanism has been applied to a prismatic high temperature reactor with europium or erbium burnable absorbers, showing a significant reduction in the initial excess reactivity of the core.« less

Sodium effects on mechanical performance and consideration in high temperature structural design for advanced reactors

NASA Astrophysics Data System (ADS)

Natesan, K.; Li, Meimei; Chopra, O. K.; Majumdar, S.

2009-07-01

Sodium environmental effects are key limiting factors in the high temperature structural design of advanced sodium-cooled reactors. A guideline is needed to incorporate environmental effects in the ASME design rules to improve the performance reliability over long operating times. This paper summarizes the influence of sodium exposure on mechanical performance of selected austenitic stainless and ferritic/martensitic steels. Focus is on Type 316SS and mod.9Cr-1Mo. The sodium effects were evaluated by comparing the mechanical properties data in air and sodium. Carburization and decarburization were found to be the key factors that determine the tensile and creep properties of the steels. A beneficial effect of sodium exposure on fatigue life was observed under fully reversed cyclic loading in both austenitic stainless steels and ferritic/martensitic steels. However, when hold time was applied during cyclic loading, the fatigue life was significantly reduced. Based on the mechanical performance of the steels in sodium, consideration of sodium effects in high temperature structural design of advanced fast reactors is discussed.

A novel concept of QUADRISO particles : Part II Utilization for excess reactivity control.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Talamo, A.

2011-01-01

In high temperature reactors, burnable absorbers are utilized to manage the excess reactivity at the early stage of the fuel cycle. In this paper QUADRISO particles are proposed to manage the initial excess reactivity of high temperature reactors. The QUADRISO concept synergistically couples the decrease of the burnable poison with the decrease of the fissile materials at the fuel particle level. This mechanism is set up by introducing a burnable poison layer around the fuel kernel in ordinary TRISO particles or by mixing the burnable poison with any of the TRISO coated layers. At the beginning of life, the initialmore » excess reactivity is small because some neutrons are absorbed in the burnable poison and they are prevented from entering the fuel kernel. At the end of life, when the absorber is almost depleted, more neutrons stream into the fuel kernel of QUADRISO particles causing fission reactions. The mechanism has been applied to a prismatic high temperature reactor with europium or erbium burnable absorbers, showing a significant reduction in the initial excess reactivity of the core.« less

Development and applications of methodologies for the neutronic design of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR)

NASA Astrophysics Data System (ADS)

Fratoni, Massimiliano

This study investigated the neutronic characteristics of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a novel nuclear reactor concept that combines liquid salt (7LiF-BeF2---flibe) cooling and TRISO coated-particle fuel technology. The use of flibe enables operation at high power density and atmospheric pressure and improves passive decay-heat removal capabilities, but flibe, unlike conventional helium coolant, is not transparent to neutrons. The flibe occupies 40% of the PB-AHTR core volume and absorbs ˜8% of the neutrons, but also acts as an effective neutron moderator. Two novel methodologies were developed for calculating the time dependent and equilibrium core composition: (1) a simplified single pebble model that is relatively fast; (2) a full 3D core model that is accurate and flexible but computationally intensive. A parametric analysis was performed spanning a wide range of fuel kernel diameters and graphite-to-heavy metal atom ratios to determine the attainable burnup and reactivity coefficients. Using 10% enriched uranium ˜130 GWd/tHM burnup was found to be attainable, when the graphite-to-heavy metal atom ratio (C/HM) is in the range of 300 to 400. At this or smaller C/HM ratio all reactivity coefficients examined---coolant temperature, coolant small and full void, fuel temperature, and moderator temperature, were found to be negative. The PB-AHTR performance was compared to that of alternative options for HTRs, including the helium-cooled pebble-bed reactor and prismatic fuel reactors, both gas-cooled and flibe-cooled. The attainable burnup of all designs was found to be similar. The PB-AHTR generates at least 30% more energy per pebble than the He-cooled pebble-bed reactor. Compared to LWRs the PB-AHTR requires 30% less natural uranium and 20% less separative work per unit of electricity generated. For deep burn TRU fuel made from recycled LWR spent fuel, it was found that in a single pass through the core ˜66% of the TRU can be transmuted; this burnup is slightly superior to that attainable in helium-cooled reactors. A preliminary analysis of the modular variant for the PB-AHTR investigated the triple heterogeneity of this design and determined its performance characteristics.

Correlation of electrical reactor cable failure with materials degradation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stuetzer, O.M.

1986-03-01

Complete circuit failure (shortout) of electrical cables typically used in nuclear power plant containments is investigated. Failure modes are correlated with the mechanical deterioration of the elastomeric cable materials. It is found that for normal reactor operation, electrical cables are reliable and safe over very long periods. During high temperature excursions, however, cables pulled across corners under high stress may short out due to conductor creep. Severe cracking will occur in short times during high temperatures (>150/sup 0/C) and in times of the order of years at elevated temperatures (100/sup 0/C to 140/sup 0/C). A theoretical treatment of stress distributionmore » responsible for creep and for cracking by J.E. Reaugh of Science Applications, Inc. is contained in the Appendix. 29 refs., 32 figs.« less

Preliminary design and hazards report. Boiling Reactor Experiment V (BORAX V)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rice, R. E.

1960-02-01

The preliminary objectives of the proposed BORAX V program are to test nuclear superheating concepts and to advance the technology of boiling-water-reactor design by performing experiments which will improve the understanding of factors limiting the stability of boiling reactors at high power densities. The reactor vessel is a cylinder with ellipsoidal heads, made of carbon steel clad internally with stainless steel. Each of the three cores is 24 in. high and has an effective diameter of 39 in. This is a preliminary report. (W.D.M.)

Radiation intensification of the reactor pressure vessels recovery by low temperature heat treatment (wet annealing)

NASA Astrophysics Data System (ADS)

Krasikov, E.

2015-04-01

As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of NPP safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation. There are two approaches to annealing. The first one is so-called «dry» high temperature (∼475°C) annealing. It allows obtaining practically complete recovery, but requires the removal of the reactor core and internals. External heat source (furnace) is required to carry out RPV heat treatment. The alternative approach is to anneal RPV at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps while operating within the RPV design limits. This low temperature «wet» annealing, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible.

NUCLEAR REACTOR FUEL SYSTEMS

DOEpatents

Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

1959-09-15

Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

Estimating the Temperature Experienced by Biomass Particles during Fast Pyrolysis Using Microscopic Analysis of Biochars

DOE PAGES

Thompson, Logan C.; Ciesielski, Peter N.; Jarvis, Mark W.; ...

2017-07-12

Here, biomass particles can experience variable thermal conditions during fast pyrolysis due to differences in their size and morphology, and from local temperature variations within a reactor. These differences lead to increased heterogeneity of the chemical products obtained in the pyrolysis vapors and bio-oil. Here we present a simple, high-throughput method to investigate the thermal history experienced by large ensembles of particles during fast pyrolysis by imaging and quantitative image analysis. We present a correlation between the surface luminance (darkness) of the biochar particle and the highest temperature that it experienced during pyrolysis. Next, we apply this correlation to large,more » heterogeneous ensembles of char particles produced in a laminar entrained flow reactor (LEFR). The results are used to interpret the actual temperature distributions delivered by the reactor over a range of operating conditions.« less

Nuclear reactor shutdown system

DOEpatents

Bhate, Suresh K.; Cooper, Martin H.; Riffe, Delmar R.; Kinney, Calvin L.

1981-01-01

An inherent shutdown system for a nuclear reactor having neutron absorbing rods affixed to an armature which is held in an upper position by a magnetic flux flowing through a Curie temperature material. The Curie temperature material is fixedly positioned about the exterior of an inner duct in an annular region through which reactor coolant flows. Elongated fuel rods extending from within the core upwardly toward the Curie temperature material are preferably disposed within the annular region. Upon abnormal conditions which result in high neutron flux and coolant temperature, the Curie material loses its magnetic permeability, breaking the magnetic flux path and allowing the armature and absorber rods to drop into the core, thus shutting down the fissioning reaction. The armature and absorber rods are retrieved by lowering the housing for the electromagnet forming coils which create a magnetic flux path which includes the inner duct wall. The coil housing then is raised, resetting the armature.

Study of Compatibility of Stainless Steel Weld Joints with Liquid Sodium-Potassium Coolants for Fission Surface Power Reactors for Lunar and Space Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grossbeck, Martin; Qualls, Louis

To make a manned mission to the surface of the moon or to Mars with any significant residence time, the power requirements will make a nuclear reactor the most feasible source of energy. To prepare for such a mission, NASA has teamed with the DOE to develop Fission Surface Power technology with the goal of developing viable options. The Fission Surface Power System (FSPS) recommended as the initial baseline design includes a liquid metal reactor and primary coolant system that transfers heat to two intermediate liquid metal heat transfer loops. Each intermediate loop transfers heat to two Stirling heat exchangersmore » that each power two Stirling converters. Both the primary and the intermediate loops will use sodium-potassium (NaK) as the liquid metal coolant, and the primary loop will operate at temperatures exceeding 600°C. The alloy selected for the heat exchangers and piping is AISI Type 316L stainless steel. The extensive experience with NaK in breeder reactor programs and with earlier space reactors for unmanned missions lends considerable confidence in using NaK as a coolant in contact with stainless steel alloys. However, the microstructure, chemical segregation, and stress state of a weld leads to the potential for corrosion and cracking. Such failures have been experienced in NaK systems that have operated for times less than the eight year goal for the FSPS. For this reason, it was necessary to evaluate candidate weld techniques and expose welds to high-temperature, flowing NaK in a closed, closely controlled system. The goal of this project was to determine the optimum weld configuration for a NaK system that will withstand service for eight years under FSPS conditions. Since the most difficult weld to make and to evaluate is the tube to tube sheet weld in the intermediate heat exchangers, it was the focus of this research. A pumped loop of flowing NaK was fabricated for exposure of candidate weld specimens at temperatures of 600°C, the expected temperature within the intermediate heat exchangers. Since metal transfer from a high-temperature region to a cooler region is a predominant mode of corrosion in liquid metal systems, specimens were placed at zones in the loop at the above temperature to evaluate the effects of both alloy component leaching and metal deposition. Microstructural analysis was performed to evaluate weld performance on control weld specimens. The research was coordinated with Oak Ridge National Laboratory (ORNL) where most of the weld samples were prepared. In addition, ORNL participated in the loop operation to assist in keeping the testing relevant to the project and to take advantage of the extensive experience in liquid metal research at ORNL.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Thompson, Logan C.; Ciesielski, Peter N.; Jarvis, Mark W.

Here, biomass particles can experience variable thermal conditions during fast pyrolysis due to differences in their size and morphology, and from local temperature variations within a reactor. These differences lead to increased heterogeneity of the chemical products obtained in the pyrolysis vapors and bio-oil. Here we present a simple, high-throughput method to investigate the thermal history experienced by large ensembles of particles during fast pyrolysis by imaging and quantitative image analysis. We present a correlation between the surface luminance (darkness) of the biochar particle and the highest temperature that it experienced during pyrolysis. Next, we apply this correlation to large,more » heterogeneous ensembles of char particles produced in a laminar entrained flow reactor (LEFR). The results are used to interpret the actual temperature distributions delivered by the reactor over a range of operating conditions.« less

AGR-2 and AGR-3/4 Release-to-Birth Ratio Data Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pham, Binh T.; Einerson, Jeffrey J.; Scates, Dawn M.

A series of Advanced Gas Reactor (AGR) irradiation tests is being conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) low enriched fuel used in the High Temperature Gas-cooled Reactor (HTGR). Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. These capsules are instrumented with thermocouples embedded in the graphite enabling temperature control. AGR configuration and irradiation conditions are based on prismatic HTGR technology that is distinguished primarily through use of heliummore » coolant, a low-power-density ceramic core capable of withstanding very high temperatures, and TRISO coated particle fuel. Thus, these tests provide valuable irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, and support development and validation of fuel performance and fission product transport models and codes.« less

10 CFR 72.218 - Termination of licenses.

Code of Federal Regulations, 2011 CFR

2011-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.218 Termination of licenses. (a) The notification regarding the program for the management of spent fuel at the reactor required by § 50.54(bb) of...

10 CFR 72.218 - Termination of licenses.

Code of Federal Regulations, 2010 CFR

2010-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.218 Termination of licenses. (a) The notification regarding the program for the management of spent fuel at the reactor required by § 50.54(bb) of...

DESIGN CRITERIA FOR HIGH TEMPERATURE LATTICE TEST REACTOR PROJECT CAH-100

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ballard, D.L.; Brown, W.W.; Harrison, C.W.

Design and construction specifications to be followed in the development of the reactor, its associated systems and experimental facilities, and the housing and required services for the facility are presented. The testing procedures to be used are outlined. (D.C.W.)

Evaluation of the performance of high temperature conversion reactors for compound-specific oxygen stable isotope analysis.

PubMed

Hitzfeld, Kristina L; Gehre, Matthias; Richnow, Hans-Hermann

2017-05-01

In this study conversion conditions for oxygen gas chromatography high temperature conversion (HTC) isotope ratio mass spectrometry (IRMS) are characterised using qualitative mass spectrometry (IonTrap). It is shown that physical and chemical properties of a given reactor design impact HTC and thus the ability to accurately measure oxygen isotope ratios. Commercially available and custom-built tube-in-tube reactors were used to elucidate (i) by-product formation (carbon dioxide, water, small organic molecules), (ii) 2nd sources of oxygen (leakage, metal oxides, ceramic material), and (iii) required reactor conditions (conditioning, reduction, stability). The suitability of the available HTC approach for compound-specific isotope analysis of oxygen in volatile organic molecules like methyl tert-butyl ether is assessed. Main problems impeding accurate analysis are non-quantitative HTC and significant carbon dioxide by-product formation. An evaluation strategy combining mass spectrometric analysis of HTC products and IRMS 18 O/ 16 O monitoring for future method development is proposed.

Enhanced In-Pile Instrumentation at the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Rempe, Joy L.; Knudson, Darrell L.; Daw, Joshua E.; Unruh, Troy; Chase, Benjamin M.; Palmer, Joe; Condie, Keith G.; Davis, Kurt L.

2012-08-01

Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper provides an update on this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

GRAFEC: A New Spanish Program to Investigate Waste Management Options for Radioactive Graphite - 12399

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marquez, Eva; Pina, Gabriel; Rodriguez, Marina

Spain has to manage about 3700 tons of irradiated graphite from the reactor Vandellos I as radioactive waste. 2700 tons are the stack of the reactor and are still in the reactor core waiting for retrieval. The rest of the quantities, 1000 tons, are the graphite sleeves which have been already retrieved from the reactor. During operation the graphite sleeves were stored in a silo and during the dismantling stage a retrieval process was carried out separating the wires from the graphite, which were crushed and introduced into 220 cubic containers of 6 m{sup 3} each and placed in interimmore » storage. The graphite is an intermediate level radioactive waste but it contains long lived radionuclides like {sup 14}C which disqualifies disposal at the low level waste repository of El Cabril. Therefore, a new project has been started in order to investigate two new options for the management of this waste type. The first one is based on a selective decontamination of {sup 14}C by thermal methods. This method is based on results obtained at the Research Centre Juelich (FZJ) in the Frame of the EC programs 'Raphael' and 'Carbowaste'. The process developed at FZJ is based on a preferential oxidation of {sup 14}C in comparison to the bulk {sup 12}C. Explanations for this effect are the inhomogeneous distribution and a weaker bounding of {sup 14}C which is not incorporated in the graphite lattice. However these investigations have only been performed with graphite from the high temperature reactor Arbeitsgemeinschaft Versuchsreaktor Juelich AVR which has been operated in a non-oxidising condition or research reactor graphite operated at room temperature. The reactor Vandellos I has been operated with CO{sub 2} as coolant and significant amounts of graphite have been already oxidised. The aim of the project is to validate whether a {sup 14}C decontamination can also been achieved with graphite from Vandellos I. A second possibility under investigation is the encapsulation of the graphite in a long term stable glass matrix. The principal applicability has been already proved by FNAG. Crushed graphite mixed with a suitable glass powder has been pressed at elevated temperature under vacuum. The vacuum is required to avoid gas enclosures in the obtained product. The obtained products, named IGM for 'Impermeable Graphite Matrix', have densities above 99% of theoretical density. The amount of glass has been chosen with respect to the pore volume of the former graphite parts. The method allows the production of encapsulated graphite without increasing the disposal volume. This paper will give a short overview of characterisation results of different irradiated graphite materials obtained at CIEMAT and in the Carbowaste project as well as the proposed methods and the actual status of the program including first results about leaching of non-radioactive IGM samples and hopefully first tendencies concerning the C-14 separation from graphite of Vandellos I by thermal treatment. Both processes, the thermal treatment as well as the IGM, have the potential to solve problems related to the management of irradiated graphite in Spain. However the methods have only been tested with different types of i-graphite and virgin graphite, respectively. Only investigations with real i-graphite from Spain will reveal whether the described methods are applicable to graphite from Vandellos I. However all partners are convinced that one of these new methods or a combination of them will lead to a feasible option to manage i-graphite in Spain on an industrial scale. (authors)« less

Characterization Report on Fuels for NEAMS Model Validation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gofryk, Krzysztof

Nearly 20% of the world’s electricity today is generated by nuclear energy from uranium dioxide (UO 2) fuel. The thermal conductivity of UO 2 governs the conversion of heat produced from fission events into electricity and it is an important parameter in reactor design and safety. While nuclear fuel operates at high to very high temperatures, thermal conductivity and other materials properties lack sensitivity to temperature variations and to material variations at reactor temperatures. As a result, both the uncertainties in laboratory measurements at high temperatures and the small differences in properties of different materials inevitably lead to large uncertaintiesmore » in models and little predictive power. Conversely, properties measured at low to moderate temperatures have more sensitivity, less uncertainty, and have larger differences in properties for different materials. These variations need to be characterized as they will afford the highest predictive capability in modeling and offer best assurances for validation and verification at all temperatures. This is well emphasized in the temperature variation of the thermal conductivity of UO 2.« less

Assembly and Delivery of Rabbit Capsules for Irradiation of Silicon Carbide Cladding Tube Specimens in the High Flux Isotope Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koyanagi, Takaaki; Petrie, Christian M.

Neutron irradiation of silicon carbide (SiC)-based fuel cladding under a high radial heat flux presents a critical challenge for SiC cladding concepts in light water reactors (LWRs). Fission heating in the fuel provides a high heat flux through the cladding, which, combined with the degraded thermal conductivity of SiC under irradiation, results in a large temperature gradient through the thickness of the cladding. The strong temperature dependence of swelling in SiC creates a complex stress profile in SiCbased cladding tubes as a result of differential swelling. The Nuclear Science User Facilities (NSUF) Program within the US Department of Energy Officemore » of Nuclear Energy is supporting research efforts to improve the scientific understanding of the effects of irradiation on SiC cladding tubes. Ultimately, the results of this project will provide experimental validation of multi-physics models for SiC-based fuel cladding during LWR operation. The first objective of this project is to irradiate tube specimens using a previously developed design that allows for irradiation testing of miniature SiC tube specimens subjected to a high radial heat flux. The previous “rabbit” capsule design uses the gamma heating in the core of the High Flux Isotope Reactor (HFIR) to drive a high heat flux through the cladding tube specimens. A compressible aluminum foil allows for a constant thermal contact conductance between the cladding tubes and the rabbit housing despite swelling of the SiC tubes. To allow separation of the effects of irradiation from those due to differential swelling under a high heat flux, a new design was developed under the NSUF program. This design allows for irradiation of similar SiC cladding tube specimens without a high radial heat flux. This report briefly describes the irradiation experiment design concepts, summarizes the irradiation test matrix, and reports on the successful delivery of six rabbit capsules to the HFIR. Rabbits of both low and high heat flux configurations have been assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. These rabbits contain a wide variety of specimens including monolith tubes, SiC fiber SiC matrix (SiC/SiC) composites, duplex specimens (inner composite, outer monolith), and specimens with a variety of metallic or ceramic coatings on the outer surface. The rabbits are targeted for insertion during HFIR cycle 475, which is scheduled for September 2017.« less

Thermal synthesis apparatus

DOEpatents

Fincke, James R [Idaho Falls, ID; Detering, Brent A [Idaho Falls, ID

2009-08-18

An apparatus for thermal conversion of one or more reactants to desired end products includes an insulated reactor chamber having a high temperature heater such as a plasma torch at its inlet end and, optionally, a restrictive convergent-divergent nozzle at its outlet end. In a thermal conversion method, reactants are injected upstream from the reactor chamber and thoroughly mixed with the plasma stream before entering the reactor chamber. The reactor chamber has a reaction zone that is maintained at a substantially uniform temperature. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle, which "freezes" the desired end product(s) in the heated equilibrium reaction stage, or is discharged through an outlet pipe without the convergent-divergent nozzle. The desired end products are then separated from the gaseous stream.

Report on FY15 alloy 617 code rules development

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sham, Sam; Jetter, Robert I; Hollinger, Greg

2015-09-01

Due to its strength at very high temperatures, up to 950°C (1742°F), Alloy 617 is the reference construction material for structural components that operate at or near the outlet temperature of the very high temperature gas-cooled reactors. However, the current rules in the ASME Section III, Division 5 Subsection HB, Subpart B for the evaluation of strain limits and creep-fatigue damage using simplified methods based on elastic analysis have been deemed inappropriate for Alloy 617 at temperatures above 650°C (1200°F) (Corum and Brass, Proceedings of ASME 1991 Pressure Vessels and Piping Conference, PVP-Vol. 215, p.147, ASME, NY, 1991). The rationalemore » for this exclusion is that at higher temperatures it is not feasible to decouple plasticity and creep, which is the basis for the current simplified rules. This temperature, 650°C (1200°F), is well below the temperature range of interest for this material for the high temperature gas-cooled reactors and the very high temperature gas-cooled reactors. The only current alternative is, thus, a full inelastic analysis requiring sophisticated material models that have not yet been formulated and verified. To address these issues, proposed code rules have been developed which are based on the use of elastic-perfectly plastic (EPP) analysis methods applicable to very high temperatures. The proposed rules for strain limits and creep-fatigue evaluation were initially documented in the technical literature (Carter, Jetter and Sham, Proceedings of ASME 2012 Pressure Vessels and Piping Conference, papers PVP 2012 28082 and PVP 2012 28083, ASME, NY, 2012), and have been recently revised to incorporate comments and simplify their application. Background documents have been developed for these two code cases to support the ASME Code committee approval process. These background documents for the EPP strain limits and creep-fatigue code cases are documented in this report.« less

Wet air oxidation for the treatment of industrial wastes. Chemical aspects, reactor design and industrial applications in Europe

DOE Office of Scientific and Technical Information (OSTI.GOV)

Debellefontaine, H.; Foussard, J.N.

2000-07-01

Aqueous wastes containing organic pollutants can be efficiently treated by wet air oxidation (WAO), i.e., oxidation (or combustion) by molecular oxygen in the liquid phase, at high temperature (200--325 C) and pressure (up to 175 bar). This method is suited to the elimination of special aqueous wastes from the chemical industry as well as to the treatment of domestic sludge. It is an enclosed process, with a limited interaction with the environment, as opposed to incineration. Usually, the operating cost is lower than 95 Euro M{sup {minus}3} and the preferred COD load ranges from 10 to 80 kg m{sup {minus}3}.more » Only a handful of industrial reactors are in operation world-wide, mainly because of the high capital investment they require. This paper reviews the major results obtained with the WAO process and assess its field of possible application to industrial wastes. In addition, as only a very few studies have been devoted to the scientific design of such reactors (bubble columns), what needs to be known for this scientific design is discussed. At present, a computer program aimed at determining the performance of a wet air oxidation reactor depending on the various operating parameters has been implemented at the laboratory. Some typical results are presented, pointing out the most important parameters and the specific behavior of these units.« less

Modifications to the NRAD Reactor, 1977 to present

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weeks, A.A.; Pruett, D.P.; Heidel, C.C.

1986-01-01

Argonne National Laboratory-West, operated by the University of Chicago, is located near Idaho Falls, ID, on the Idaho National Engineering laboratory Site. ANL-West performs work in support of the Liquid Metal Fast Breeder Reactor Program (LMFBR) sponsored by the United States Department of Energy. The NRAD reactor is located at the Argonne Site within the Hot Fuel Examination Facility/North, a large hot cell facility where both non-destructive and destructive examinations are performed on highly irradiated reactor fuels and materials in support of the LMFBR program. The NRAD facility utilizes a 250-kW TRIGA reactor and is completely dedicated to neutron radiographymore » and the development of radiography techniques. Criticality was first achieved at the NRAD reactor in October of 1977. Since that time, a number of modifications have been implemented to improve operational efficiency and radiography production. This paper describes the modifications and changes that significantly improved operational efficiency and reliability of the reactor and the essential auxiliary reactor systems.« less

Influence of Alumina Binder Content on Catalytic Performance of Ni/HZSM-5 for Hydrodeoxygenation of Cyclohexanone

PubMed Central

Kong, Xiangjin; Liu, Junhai

2014-01-01

The influence of the amount of alumina binders on the catalytic performance of Ni/HZSM-5 for hydrodeoxygenation of cyclohexanone was investigated in a fixed-bed reactor. N2 sorption, X-ray diffraction, H2-chemisorption and temperature-programmed desorption of ammonia were used to characterize the catalysts. It can be observed that the Ni/HZSM-5 catalyst bound with 30 wt.% alumina binder exhibited the best catalytic performance. The high catalytic performance may be due to relatively good Ni metal dispersion, moderate mesoporosity, and proper acidity of the catalyst. PMID:25009974

Influence of alumina binder content on catalytic performance of Ni/HZSM-5 for hydrodeoxygenation of cyclohexanone.

PubMed

Kong, Xiangjin; Liu, Junhai

2014-01-01

The influence of the amount of alumina binders on the catalytic performance of Ni/HZSM-5 for hydrodeoxygenation of cyclohexanone was investigated in a fixed-bed reactor. N2 sorption, X-ray diffraction, H2-chemisorption and temperature-programmed desorption of ammonia were used to characterize the catalysts. It can be observed that the Ni/HZSM-5 catalyst bound with 30 wt.% alumina binder exhibited the best catalytic performance. The high catalytic performance may be due to relatively good Ni metal dispersion, moderate mesoporosity, and proper acidity of the catalyst.

Methods and codes for neutronic calculations of the MARIA research reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrzejewski, K.; Kulikowska, T.; Bretscher, M. M.

2002-02-18

The core of the MARIA high flux multipurpose research reactor is highly heterogeneous. It consists of beryllium blocks arranged in 6 x 8 matrix, tubular fuel assemblies, control rods and irradiation channels. The reflector is also heterogeneous and consists of graphite blocks clad with aluminum. Its structure is perturbed by the experimental beam tubes. This paper presents methods and codes used to calculate the MARIA reactor neutronics characteristics and experience gained thus far at IAE and ANL. At ANL the methods of MARIA calculations were developed in connection with the RERTR program. At IAE the package of programs was developedmore » to help its operator in optimization of fuel utilization.« less

Recommended high-tank temperatures for maintenance of high-tank backup support, Revision 3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Greager, O.H.

1964-05-20

Purpose of this note is to recommend revised curves for the high-tank temperature required to maintain adequate high-tank backup support at the six small reactors. Compliance with the conditions shown on these curves will ensure adequate high-tank flow rates following the simultaneous loss of electric and steam power.

MODELING THE AMBIENT CONDITION EFFECTS OF AN AIR-COOLED NATURAL CIRCULATION SYSTEM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Rui; Lisowski, Darius D.; Bucknor, Matthew

The Reactor Cavity Cooling System (RCCS) is a passive safety concept under consideration for the overall safety strategy of advanced reactors such as the High Temperature Gas-Cooled Reactor (HTGR). One such variant, air-cooled RCCS, uses natural convection to drive the flow of air from outside the reactor building to remove decay heat during normal operation and accident scenarios. The Natural convection Shutdown heat removal Test Facility (NSTF) at Argonne National Laboratory (“Argonne”) is a half-scale model of the primary features of one conceptual air-cooled RCCS design. The facility was constructed to carry out highly instrumented experiments to study the performancemore » of the RCCS concept for reactor decay heat removal that relies on natural convection cooling. Parallel modeling and simulation efforts were performed to support the design, operation, and analysis of the natural convection system. Throughout the testing program, strong influences of ambient conditions were observed in the experimental data when baseline tests were repeated under the same test procedures. Thus, significant analysis efforts were devoted to gaining a better understanding of these influences and the subsequent response of the NSTF to ambient conditions. It was determined that air humidity had negligible impacts on NSTF system performance and therefore did not warrant consideration in the models. However, temperature differences between the building exterior and interior air, along with the outside wind speed, were shown to be dominant factors. Combining the stack and wind effects together, an empirical model was developed based on theoretical considerations and using experimental data to correlate zero-power system flow rates with ambient meteorological conditions. Some coefficients in the model were obtained based on best fitting the experimental data. The predictive capability of the empirical model was demonstrated by applying it to the new set of experimental data. The empirical model was also implemented in the computational models of the NSTF using both RELAP5-3D and STARCCM+ codes. Accounting for the effects of ambient conditions, simulations from both codes predicted the natural circulation flow rates very well.« less

Thermal analysis of cylindrical natural-gas steam reformer for 5 kW PEMFC

NASA Astrophysics Data System (ADS)

Jo, Taehyun; Han, Junhee; Koo, Bonchan; Lee, Dohyung

2016-11-01

The thermal characteristics of a natural-gas based cylindrical steam reformer coupled with a combustor are investigated for the use with a 5 kW polymer electrolyte membrane fuel cell. A reactor unit equipped with nickel-based catalysts was designed to activate the steam reforming reaction without the inclusion of high-temperature shift and low-temperature shift processes. Reactor temperature distribution and its overall thermal efficiency depend on various inlet conditions such as the equivalence ratio, the steam to carbon ratio (SCR), and the fuel distribution ratio (FDR) into the reactor and the combustor components. These experiments attempted to analyze the reformer's thermal and chemical properties through quantitative evaluation of product composition and heat exchange between the combustor and the reactor. FDR is critical factor in determining the overall performance as unbalanced fuel injection into the reactor and the combustor deteriorates overall thermal efficiency. Local temperature distribution also influences greatly on the fuel conversion rate and thermal efficiency. For the experiments, the operation conditions were set as SCR was in range of 2.5-4.0 and FDR was in 0.4-0.7 along with equivalence ratio of 0.9-1.1; optimum results were observed for FDR of 0.63 and SCR of 3.0 in the cylindrical steam reformer.

Qualification of Daiichi Units 1, 2, and 3 Data for Severe Accident Evaluations - Process and Illustrative Examples from Prior TMI-2 Evaluations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rempe, Joy Lynn; Knudson, Darrell Lee

2014-09-01

The accidents at the Three Mile Island Unit 2 (TMI-2) Pressurized Water Reactor (PWR) and the Daiichi Units 1, 2, and 3 Boiling Water Reactors (BWRs) provide unique opportunities to evaluate instrumentation exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during the TMI-2 accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2more » sensor survivability and data qualification efforts. This initial review focused on the set of sensors deemed most important by post-TMI-2 instrumentation evaluation programs. Instrumentation evaluation programs focused on data required by TMI-2 operators to assess the condition of the reactor and containment and the effect of mitigating actions taken by these operators. In addition, prior efforts focused on sensors providing data required for subsequent forensic evaluations and accident simulations. To encourage the potential for similar activities to be completed for qualifying data from Daiichi Units 1, 2, and 3, this report provides additional details related to the formal process used to develop a qualified TMI-2 data base and presents data qualification details for three parameters: primary system pressure; containment building temperature; and containment pressure. As described within this report, sensor evaluations and data qualification required implementation of various processes, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design to instruments easily removed from the TMI-2 plant for evaluations. As documented in this report, results from qualifying data for these parameters led to key insights related to TMI-2 accident progression. Hence, these selected examples illustrate the types of activities completed in the TMI-2 data qualification process and the importance of such a qualification effort. These details are documented in this report to facilitate implementation of similar process using data and examinations at the Daiichi Units 1, 2, and 3 reactors so that BWR-specific benefits can be obtained.« less

Simulation of Oil Palm Shell Pyrolysis to Produce Bio-Oil with Self-Pyrolysis Reactor

NASA Astrophysics Data System (ADS)

Fika, R.; Nelwan, L. O.; Yulianto, M.

2018-05-01

A new self-pyrolysis reactor was designed to reduce the utilization of electric heater due to the energy saving for the production of bio-oil from oil palm shell. The yield of the bio- oil was then evaluated with the developed mathematical model by Sharma [1] with the characteristic of oil palm shell [2]. During the simulation, the temperature on the combustion chamber on the release of the bio-oil was utilized to determine the volatile composition from the combustion of the oil palm shell as fuel. The mass flow was assumed constant for three experiments. The model resulted in a significant difference between the simulated bio-oil and experiments. The bio-oil yields from the simulation were 22.01, 16.36, and 21.89 % (d.b.) meanwhile the experimental yields were 10.23, 9.82, and 8.41% (d.b.). The char yield varied from 30.7 % (d.b.) from the simulation to 40.9 % (d.b.) from the experiment. This phenomenon was due to the development of process temperature over time which was not considered as one of the influential factors in producing volatile matters on the simulation model. Meanwhile the real experiments highly relied on the process conditions (reactor type, temperature over time, gas flow). There was also possibilities of the occurrence of the gasification inside the reactor which caused the liquid yield was not as high as simulated. Further simulation model research on producing the bio-oil yield will be needed to predict the optimum condition and temperature development on the newly self-pyrolysis reactor.

Scaling and design analyses of a scaled-down, high-temperature test facility for experimental investigation of the initial stages of a VHTR air-ingress accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arcilesi, David J.; Ham, Tae Kyu; Kim, In Hun

2015-07-01

A critical event in the safety analysis of the very high-temperature gas-cooled reactor (VHTR) is an air-ingress accident. This accident is initiated, in its worst case scenario, by a double-ended guillotine break of the coaxial cross vessel, which leads to a rapid reactor vessel depressurization. In a VHTR, the reactor vessel is located within a reactor cavity that is filled with air during normal operating conditions. Following the vessel depressurization, the dominant mode of ingress of an air–helium mixture into the reactor vessel will either be molecular diffusion or density-driven stratified flow. The mode of ingress is hypothesized to dependmore » largely on the break conditions of the cross vessel. Since the time scales of these two ingress phenomena differ by orders of magnitude, it is imperative to understand under which conditions each of these mechanisms will dominate in the air ingress process. Computer models have been developed to analyze this type of accident scenario. There are, however, limited experimental data available to understand the phenomenology of the air-ingress accident and to validate these models. Therefore, there is a need to design and construct a scaled-down experimental test facility to simulate the air-ingress accident scenarios and to collect experimental data. The current paper focuses on the analyses performed for the design and operation of a 1/8th geometric scale (by height and diameter), high-temperature test facility. A geometric scaling analysis for the VHTR, a time scale analysis of the air-ingress phenomenon, a transient depressurization analysis of the reactor vessel, a hydraulic similarity analysis of the test facility, a heat transfer characterization of the hot plenum, a power scaling analysis for the reactor system, and a design analysis of the containment vessel are discussed.« less

Severe Accident Test Station Activity Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pint, Bruce A.; Terrani, Kurt A.

2015-06-01

Enhancing safety margins in light water reactor (LWR) severe accidents is currently the focus of a number of international R&D programs. The current UO2/Zr-based alloy fuel system is particularly susceptible since the Zr-based cladding experiences rapid oxidation kinetics in steam at elevated temperatures. Therefore, alternative cladding materials that offer slower oxidation kinetics and a smaller enthalpy of oxidation can significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. In the U.S. program, the high temperature steam oxidation performance of accident tolerant fuel (ATF) cladding solutions has been evaluated in the Severe Accidentmore » Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012. This report summarizes the capabilities of the SATS and provides an overview of the oxidation kinetics of several candidate cladding materials. A suggested baseline for evaluating ATF candidates is a two order of magnitude reduction in the steam oxidation resistance above 1000ºC compared to Zr-based alloys. The ATF candidates are categorized based on the protective external oxide or scale that forms during exposure to steam at high temperature: chromia, alumina, and silica. Comparisons are made to literature and SATS data for Zr-based alloys and other less-protective materials.« less

Method for improving performance of irradiated structural materials

DOEpatents

Megusar, Janez; Harling, Otto K.; Grant, Nicholas J.

1989-01-01

Method for extending service life of nuclear reactor components prepared from ductile, high strength crystalline alloys obtained by devitrification of metallic glasses. Two variations of the method are described: (1) cycling the temperature of the nuclear reactor between the operating temperature which leads to irradiation damage and a l The U.S. Government has rights in this invention by virtue of Department of Energy, Office of Fusion Energy, Grant No. DE-AC02-78ER-10107.

Overview of the multifaceted activities towards development and deployment of nuclear-grade FeCrAl Alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G; Yamamoto, Yukinori; Pint, Bruce A

2016-01-01

A large effort is underway under the leadership of US DOE Fuel Cycle R&D program to develop advanced FeCrAl alloys as accident tolerant fuel (ATF) cladding to replace Zr-based alloys in light water reactors. The primary motivation is the excellent oxidation resistance of these alloys in high-temperature steam environments right up to their melting point (roughly three orders of magnitude slower oxidation kinetics than zirconium). A multifaceted effort is ongoing to rapidly advance FeCrAl alloys as a mature ATF concept. The activities span the broad spectrum of alloy development, environmental testing (high-temperature high-pressure water and elevated temperature steam), detailed mechanicalmore » characterization, material property database development, neutron irradiation, thin tube production, and multiple integral fuel test campaigns. Instead of off-the-shelf commercial alloys that might not prove optimal for the LWR fuel cladding application, a large amount of effort has been placed on the alloy development to identify the most optimum composition and microstructure for this application. The development program is targeting a cladding that offers performance comparable to or better than modern Zr-based alloys under normal operating and off-normal conditions. This paper provides a comprehensive overview of the systematic effort to advance nuclear-grade FeCrAl alloys as an ATF cladding in commercial LWRs.« less

Development of a High Temperature Microbial Fermentation Processfor Butanol Production

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jeor, Jeffery D.; Reed, David W.; Daubaras, Dayna L.

2016-06-01

Transforming renewable biomass into cost competitive high-performance biofuels and bioproducts is key to US energy security. Butanol production by microbial fermentation and chemical conversion to polyolefins, elastomers, drop-in jet or diesel fuel, and other chemicals is a promising solution. A high temperature fermentation process can facilitate butanol recovery up to 40%, by using gas stripping. Other benefits of fermentation at high temperatures are optimal hydrolysis rates in the saccharification of biomass which leads to maximized butanol production, decrease in energy costs associated with reactor cooling and capital cost associated with reactor design, and a decrease in contamination and cost formore » maintaining a sterile environment. Butanol stripping at elevated temperatures gives higher butanol production through constant removal and continuous fermentation. We describe methods used in an attempt to genetically prepare Geobacillus caldoxylosiliticus for insertion of a butanol pathway. Methods used were electroporation of electrocompetent cells, ternary conjugation with E. coli, and protoplast fusion.« less

Hydrothermal treatment of hazardous energetic materials waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brill, T.B.; Schoppelrei, J.W.; Maiella, P.G.

1995-12-31

Destruction of energetic materials by hydrothermal methods presents a potential for strongly exothermic oxidation-reduction reactions, which, if localized at a site in the reactor, create {open_quotes}hot spots{close_quotes}. To investigate highly exothermic hydrothermal reactions, real-time spectroscopic measurements in the stream by infrared and Raman spectroscopy offer opportunities. Flow reactor-spectroscopy cells were developed for such studies, focusing on approximately oxygen-balanced nitrate salts for which highly exothermic reactions can occur. In addition, the kinetics of formation of later stage products were studied because these products are likely to be released to the environment and to be regulated. An experiment was designed to simulatemore » the occurence of a phase separation in a reactor followed by rapid exothermic reaction. By varying the pressure, water content, and hydrogen content in the reaction volume of the cell, the freeze out temperatures required to set the carbon monoxide/carbon dioxide ratio were determined to be 1300 to 1470 K. Such high temperatures suggest that localized hot spots can exist which greatly exceed the overall set temperature of the reactor. This scenario can occur if a phase separation occurs to isolate ethylenediammonium dinitrate in quantities as small as tenths of milligrams. Studies of the oxidation-reduction reactions of nitrate ion with the counter ion show that the oxidizing power of the nitrate ion is realized provided a readily oxidizable cation such as hydroxylammonium is present. When the cation has a low reactivity, such as quanidinium, a much higher reaction temperature is required before the nitrate ion reacts. At this temperature, the cation may have already begun to decompose by a hydrothermal route.« less

Moon base reactor system

NASA Technical Reports Server (NTRS)

Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

1989-01-01

The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

High aspect ratio catalytic reactor and catalyst inserts therefor

DOEpatents

Lin, Jiefeng; Kelly, Sean M.

2018-04-10

The present invention relates to high efficient tubular catalytic steam reforming reactor configured from about 0.2 inch to about 2 inch inside diameter high temperature metal alloy tube or pipe and loaded with a plurality of rolled catalyst inserts comprising metallic monoliths. The catalyst insert substrate is formed from a single metal foil without a central supporting structure in the form of a spiral monolith. The single metal foil is treated to have 3-dimensional surface features that provide mechanical support and establish open gas channels between each of the rolled layers. This unique geometry accelerates gas mixing and heat transfer and provides a high catalytic active surface area. The small diameter, high aspect ratio tubular catalytic steam reforming reactors loaded with rolled catalyst inserts can be arranged in a multi-pass non-vertical parallel configuration thermally coupled with a heat source to carry out steam reforming of hydrocarbon-containing feeds. The rolled catalyst inserts are self-supported on the reactor wall and enable efficient heat transfer from the reactor wall to the reactor interior, and lower pressure drop than known particulate catalysts. The heat source can be oxygen transport membrane reactors.

Stabilized three-stage oxidation of DME/air mixture in a micro flow reactor with a controlled temperature profile

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oshibe, Hiroshi; Nakamura, Hisashi; Tezuka, Takuya

Ignition and combustion characteristics of a stoichiometric dimethyl ether (DME)/air mixture in a micro flow reactor with a controlled temperature profile which was smoothly ramped from room temperature to ignition temperature were investigated. Special attention was paid to the multi-stage oxidation in low temperature condition. Normal stable flames in a mixture flow in the high velocity region, and non-stationary pulsating flames and/or repetitive extinction and ignition (FREI) in the medium velocity region were experimentally confirmed as expected from our previous study on a methane/air mixture. In addition, stable double weak flames were observed in the low velocity region for themore » present DME/air mixture case. It is the first observation of stable double flames by the present methodology. Gas sampling was conducted to obtain major species distributions in the flow reactor. The results indicated that existence of low-temperature oxidation was conjectured by the production of CH{sub 2}O occured in the upstream side of the experimental first luminous flame, while no chemiluminescence from it was seen. One-dimensional computation with detailed chemistry and transport was conducted. At low mixture velocities, three-stage oxidation was confirmed from profiles of the heat release rate and major chemical species, which was broadly in agreement with the experimental results. Since the present micro flow reactor with a controlled temperature profile successfully presented the multi-stage oxidations as spatially separated flames, it is shown that this flow reactor can be utilized as a methodology to separate sets of reactions, even for other practical fuels, at different temperature. (author)« less

PR-EDB: Power Reactor Embrittlement Database - Version 3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Subramani, Ranjit

2008-03-01

The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industrymore » standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for backend data storage, and Microsoft Excel for plotting graphs. This software package is compatible with Windows (98 or higher) and has been built with a highly versatile user interface. PR-EDB Version 3.0 also contains an "Evaluated Residual File" utility for generating the evaluated processed files used for radiation embrittlement study.« less

Pressurized reactor system and a method of operating the same

DOEpatents

Isaksson, J.M.

1996-06-18

A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Super-atmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gasification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor. 2 figs.

Pressurized reactor system and a method of operating the same

DOEpatents

Isaksson, Juhani M.

1996-01-01

A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Superatmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gassification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor.

Liquid fuel molten salt reactors for thorium utilization

DOE PAGES

Gehin, Jess C.; Powers, Jeffrey J.

2016-04-08

Molten salt reactors (MSRs) represent a class of reactors that use liquid salt, usually fluoride- or chloride-based, as either a coolant with a solid fuel (such as fluoride salt-cooled high temperature reactors) or as a combined coolant and fuel with fuel dissolved in a carrier salt. For liquid-fuelled MSRs, the salt can be processed online or in a batch mode to allow for removal of fission products as well as introduction of fissile fuel and fertile materials during reactor operation. The MSR is most commonly associated with the 233U/thorium fuel cycle, as the nuclear properties of 233U combined with themore » online removal of parasitic absorbers allow for the ability to design a thermal-spectrum breeder reactor; however, MSR concepts have been developed using all neutron energy spectra (thermal, intermediate, fast, and mixed-spectrum zoned concepts) and with a variety of fuels including uranium, thorium, plutonium, and minor actinides. Early MSR work was supported by a significant research and development (R&D) program that resulted in two experimental systems operating at ORNL in the 1960s, the Aircraft Reactor Experiment and the Molten Salt Reactor Experiment. Subsequent design studies in the 1970s focusing on thermal-spectrum thorium-fueled systems established reference concepts for two major design variants: (1) a molten salt breeder reactor (MSBR), with multiple configurations that could breed additional fissile material or maintain self-sustaining operation; and (2) a denatured molten salt reactor (DMSR) with enhanced proliferation-resistance. T MSRs has been selected as one of six most promising Generation IV systems and development activities have been seen in fast-spectrum MSRs, waste-burning MSRs, MSRs fueled with low-enriched uranium (LEU), as well as more traditional thorium fuel cycle-based MSRs. This study provides an historical background of MSR R&D efforts, surveys and summarizes many of the recent development, and provides analysis comparing thorium-based MSRs.« less

Off-design performance of a chemical looping combustion (CLC) combined cycle: effects of ambient temperature

NASA Astrophysics Data System (ADS)

Chi, Jinling; Wang, Bo; Zhang, Shijie; Xiao, Yunhan

2010-02-01

The present work investigates the influence of ambient temperature on the steady-state off-design thermodynamic performance of a chemical looping combustion (CLC) combined cycle. A sensitivity analysis of the CLC reactor system was conducted, which shows that the parameters that influence the temperatures of the CLC reactors most are the flow rate and temperature of air entering the air reactor. For the ambient temperature variation, three off-design control strategies have been assumed and compared: 1) without any Inlet Guide Vane (IGV) control, 2) IGV control to maintain air reactor temperature and 3) IGV control to maintain constant fuel reactor temperature, aside from fuel flow rate adjusting. Results indicate that, compared with the conventional combined cycle, due to the requirement of pressure balance at outlet of the two CLC reactors, CLC combined cycle shows completely different off-design thermodynamic characteristics regardless of the control strategy adopted. For the first control strategy, temperatures of the two CLC reactors both rise obviously as ambient temperature increases. IGV control adopted by the second and the third strategy has the effect to maintain one of the two reactors' temperatures at design condition when ambient temperature is above design point. Compare with the second strategy, the third would induce more severe decrease of efficiency and output power of the CLC combined cycle.

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael A. Pope

2011-10-01

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physicsmore » design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.« less

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Francesco Venneri; Chang-Keun Jo; Jae-Man Noh

2010-09-01

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physicsmore » design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.« less

Reactor and method for hydrocracking carbonaceous material

DOEpatents

Duncan, Dennis A.; Beeson, Justin L.; Oberle, R. Donald; Dirksen, Henry A.

1980-01-01

Solid, carbonaceous material is cracked in the presence of hydrogen or other reducing gas to provide aliphatic and aromatic hydrocarbons of lower molecular weight for gaseous and liquid fuels. The carbonaceous material, such as coal, is entrained as finely divided particles in a flow of reducing gas and preheated to near the decomposition temperature of the high molecular weight polymers. Within the reactor, small quantities of oxygen containing gas are injected at a plurality of discrete points to burn corresponding amounts of the hydrogen or other fuel and elevate the mixture to high temperatures sufficient to decompose the high molecular weight, carbonaceous solids. Turbulent mixing at each injection point rapidly quenches the material to a more moderate bulk temperature. Additional quenching after the final injection point can be performed by direct contact with quench gas or oil. The reactions are carried out in the presence of a hydrogen-containing reducing gas at moderate to high pressure which stabilizes the products.

Accelerated Irradiations for High Dose Microstructures in Fast Reactor Alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiao, Zhijie

The objective of this project is to determine the extent to which high dose rate, self-ion irradiation can be used as an accelerated irradiation tool to understand microstructure evolution at high doses and temperatures relevant to advanced fast reactors. We will accomplish the goal by evaluating phase stability and swelling of F-M alloys relevant to SFR systems at very high dose by combining experiment and modeling in an effort to obtain a quantitative description of the processes at high and low damage rates.

U.S. Department of Energy Accident Resistant SiC Clad Nuclear Fuel Development

DOE Office of Scientific and Technical Information (OSTI.GOV)

George W. Griffith

2011-10-01

A significant effort is being placed on silicon carbide ceramic matrix composite (SiC CMC) nuclear fuel cladding by Light Water Reactor Sustainability (LWRS) Advanced Light Water Reactor Nuclear Fuels Pathway. The intent of this work is to invest in a high-risk, high-reward technology that can be introduced in a relatively short time. The LWRS goal is to demonstrate successful advanced fuels technology that suitable for commercial development to support nuclear relicensing. Ceramic matrix composites are an established non-nuclear technology that utilizes ceramic fibers embedded in a ceramic matrix. A thin interfacial layer between the fibers and the matrix allows formore » ductile behavior. The SiC CMC has relatively high strength at high reactor accident temperatures when compared to metallic cladding. SiC also has a very low chemical reactivity and doesn't react exothermically with the reactor cooling water. The radiation behavior of SiC has also been studied extensively as structural fusion system components. The SiC CMC technology is in the early stages of development and will need to mature before confidence in the developed designs can created. The advanced SiC CMC materials do offer the potential for greatly improved safety because of their high temperature strength, chemical stability and reduced hydrogen generation.« less

System and method for air temperature control in an oxygen transport membrane based reactor

DOEpatents

Kelly, Sean M

2016-09-27

A system and method for air temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

System and method for temperature control in an oxygen transport membrane based reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelly, Sean M.

A system and method for temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

Secondary Heat Exchanger Design and Comparison for Advanced High Temperature Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Piyush Sabharwall; Ali Siahpush; Michael McKellar

2012-06-01

The goals of next generation nuclear reactors, such as the high temperature gas-cooled reactor and advance high temperature reactor (AHTR), are to increase energy efficiency in the production of electricity and provide high temperature heat for industrial processes. The efficient transfer of energy for industrial applications depends on the ability to incorporate effective heat exchangers between the nuclear heat transport system and the industrial process heat transport system. The need for efficiency, compactness, and safety challenge the boundaries of existing heat exchanger technology, giving rise to the following study. Various studies have been performed in attempts to update the secondarymore » heat exchanger that is downstream of the primary heat exchanger, mostly because its performance is strongly tied to the ability to employ more efficient conversion cycles, such as the Rankine super critical and subcritical cycles. This study considers two different types of heat exchangers—helical coiled heat exchanger and printed circuit heat exchanger—as possible options for the AHTR secondary heat exchangers with the following three different options: (1) A single heat exchanger transfers all the heat (3,400 MW(t)) from the intermediate heat transfer loop to the power conversion system or process plants; (2) Two heat exchangers share heat to transfer total heat of 3,400 MW(t) from the intermediate heat transfer loop to the power conversion system or process plants, each exchanger transfers 1,700 MW(t) with a parallel configuration; and (3) Three heat exchangers share heat to transfer total heat of 3,400 MW(t) from the intermediate heat transfer loop to the power conversion system or process plants. Each heat exchanger transfers 1,130 MW(t) with a parallel configuration. A preliminary cost comparison will be provided for all different cases along with challenges and recommendations.« less

Investigation of Abnormal Heat Transfer and Flow in a VHTR Reactor Core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kawaji, Masahiro; Valentin, Francisco I.; Artoun, Narbeh

2015-12-21

The main objective of this project was to identify and characterize the conditions under which abnormal heat transfer phenomena would occur in a Very High Temperature Reactor (VHTR) with a prismatic core. High pressure/high temperature experiments have been conducted to obtain data that could be used for validation of VHTR design and safety analysis codes. The focus of these experiments was on the generation of benchmark data for design and off-design heat transfer for forced, mixed and natural circulation in a VHTR core. In particular, a flow laminarization phenomenon was intensely investigated since it could give rise to hot spotsmore » in the VHTR core.« less

Multiple model approach to evaluation of accelerated carbonation for steelmaking slag in a slurry reactor.

PubMed

Pan, Shu-Yuan; Liu, Hsing-Lu; Chang, E-E; Kim, Hyunook; Chen, Yi-Hung; Chiang, Pen-Chi

2016-07-01

Basic oxygen furnace slag (BOFS) exhibits highly alkaline properties due to its high calcium content, which is beneficial to carbonation reaction. In this study, accelerated carbonation of BOFS was evaluated under different reaction times, temperatures, and liquid-to-solid (L/S) ratios in a slurry reactor. CO2 mass balance within the slurry reactor was carried out to validate the technical feasibility of fixing gaseous CO2 into solid precipitates. After that, a multiple model approach, i.e., theoretical kinetics and empirical surface model, for carbonation reaction was presented to determine the maximal carbonation conversion of BOFS in a slurry reactor. On one hand, the reaction kinetics of BOFS carbonation was evaluated by the shrinking core model (SCM). Calcite (CaCO3) was identified as a reaction product through the scanning electronic microscopy and X-ray diffraction analyses, which provided the rationale of applying the SCM in this study. The rate-limiting step of carbonation was found to be ash-diffusion controlled, and the effective diffusivity for carbonation of BOFS in a slurry reactor were determined accordingly. On the other hand, the carbonation conversion of BOFS was predicted by the response surface methodology (RSM) via a nonlinear mathematical programming. According to the experimental data, the highest carbonation conversion of BOFS achieved was 57% under an L/S ratio of 20 mL g(-1), a CO2 flow rate of 0.1 L min(-1), and a pressure of 101.3 kPa at 50 °C for 120 min. Furthermore, the applications and limitations of SCM and RSM were examined and exemplified by the carbonation of steelmaking slags. Copyright © 2016 Elsevier Ltd. All rights reserved.

Assessment of the Use of Nitrogen Trifluoride for Purifying Coolant and Heat Transfer Salts in the Fluoride Salt-Cooled High-Temperature Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scheele, Randall D.; Casella, Andrew M.

2010-09-28

This report provides an assessment of the use of nitrogen trifluoride for removing oxide and water-caused contaminants in the fluoride salts that will be used as coolants in a molten salt cooled reactor.

CHLORINE ABSORPTION IN S(IV) SOLUTIONS

EPA Science Inventory

The report gives results of measurements of the rate of Chlorine (Cl2) absorption into aqueous sulfite/bisulfite -- S(IV) -- solutions at ambient temperature using a highly characterized stirred-cell reactor. The reactor media were 0 to 10 mM S(IV) with pHs of 3.5-8.5. Experiment...

Irradiation Testing of Ultrasonic Transducers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daw, Joshua; Tittmann, Bernhard; Reinhardt, Brian

2014-07-30

Ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of ultrasonic technologies to provide enhanced sensors for in-pile instrumentation during irradiation testing. For example, the ability of single, small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. Other efforts include an ultrasonic technique to detect morphologymore » changes (such as crack initiation and growth) and acoustic techniques to evaluate fission gas composition and pressure. These efforts are limited by the lack of existing knowledge of ultrasonic transducer material survivability under irradiation conditions. For this reason, the Pennsylvania State University (PSU) was awarded an Advanced Test Reactor National Scientific User Facility (ATR NSUF) project to evaluate promising magnetostrictive and piezoelectric transducer performance in the Massachusetts Institute of Technology Research Reactor (MITR) up to a fast fluence of at least 1021 n/cm2 (E> 0.1 MeV). The goal of this research is to characterize magnetostrictive and piezoelectric transducer survivability during irradiation, enabling the development of novel radiation tolerant ultrasonic sensors for use in Material and Test Reactors (MTRs). As such, this test will be an instrumented lead test and real-time transducer performance data will be collected along with temperature and neutron and gamma flux data. The current work bridges the gap between proven out-of-pile ultrasonic techniques and in-pile deployment of ultrasonic sensors by acquiring the data necessary to demonstrate the performance of ultrasonic transducers.« less

Atomistic modeling of high temperature uranium-zirconium alloy structure and thermodynamics

NASA Astrophysics Data System (ADS)

Moore, A. P.; Beeler, B.; Deo, C.; Baskes, M. I.; Okuniewski, M. A.

2015-12-01

A semi-empirical Modified Embedded Atom Method (MEAM) potential is developed for application to the high temperature body-centered-cubic uranium-zirconium alloy (γ-U-Zr) phase and employed with molecular dynamics (MD) simulations to investigate the high temperature thermo-physical properties of U-Zr alloys. Uranium-rich U-Zr alloys (e.g. U-10Zr) have been tested and qualified for use as metallic nuclear fuel in U.S. fast reactors such as the Integral Fast Reactor and the Experimental Breeder Reactors, and are a common sub-system of ternary metallic alloys like U-Pu-Zr and U-Zr-Nb. The potential was constructed to ensure that basic properties (e.g., elastic constants, bulk modulus, and formation energies) were in agreement with first principles calculations and experimental results. After which, slight adjustments were made to the potential to fit the known thermal properties and thermodynamics of the system. The potentials successfully reproduce the experimental melting point, enthalpy of fusion, volume change upon melting, thermal expansion, and the heat capacity of pure U and Zr. Simulations of the U-Zr system are found to be in good agreement with experimental thermal expansion values, Vegard's law for the lattice constants, and the experimental enthalpy of mixing. This is the first simulation to reproduce the experimental thermodynamics of the high temperature γ-U-Zr metallic alloy system. The MEAM potential is then used to explore thermodynamics properties of the high temperature U-Zr system including the constant volume heat capacity, isothermal compressibility, adiabatic index, and the Grüneisen parameters.

An overview of research activities on materials for nuclear applications at the INL Safety, Tritium and Applied Research facility

NASA Astrophysics Data System (ADS)

Calderoni, P.; Sharpe, J.; Shimada, M.; Denny, B.; Pawelko, B.; Schuetz, S.; Longhurst, G.; Hatano, Y.; Hara, M.; Oya, Y.; Otsuka, T.; Katayama, K.; Konishi, S.; Noborio, K.; Yamamoto, Y.

2011-10-01

The Safety, Tritium and Applied Research facility at the Idaho National Laboratory is a US Department of Energy National User Facility engaged in various aspects of materials research for nuclear applications related to fusion and advanced fission systems. Research activities are mainly focused on the interaction of tritium with materials, in particular plasma facing components, liquid breeders, high temperature coolants, fuel cladding, cooling and blanket structures and heat exchangers. Other activities include validation and verification experiments in support of the Fusion Safety Program, such as beryllium dust reactivity and dust transport in vacuum vessels, and support of Advanced Test Reactor irradiation experiments. This paper presents an overview of the programs engaged in the activities, which include the US-Japan TITAN collaboration, the US ITER program, the Next Generation Power Plant program and the tritium production program, and a presentation of ongoing experiments as well as a summary of recent results with emphasis on fusion relevant materials.

Molten salt destruction of energetic waste materials

DOEpatents

Brummond, W.A.; Upadhye, R.S.; Pruneda, C.O.

1995-07-18

A molten salt destruction process is used to treat and destroy energetic waste materials such as high explosives, propellants, and rocket fuels. The energetic material is pre-blended with a solid or fluid diluent in safe proportions to form a fluid fuel mixture. The fuel mixture is rapidly introduced into a high temperature molten salt bath. A stream of molten salt is removed from the vessel and may be recycled as diluent. Additionally, the molten salt stream may be pumped from the reactor, circulated outside the reactor for further processing, and delivered back into the reactor or cooled and circulated to the feed delivery system to further dilute the fuel mixture entering the reactor. 4 figs.

Molten salt destruction of energetic waste materials

DOEpatents

Brummond, William A.; Upadhye, Ravindra S.; Pruneda, Cesar O.

1995-01-01

A molten salt destruction process is used to treat and destroy energetic waste materials such as high explosives, propellants, and rocket fuels. The energetic material is pre-blended with a solid or fluid diluent in safe proportions to form a fluid fuel mixture. The fuel mixture is rapidly introduced into a high temperature molten salt bath. A stream of molten salt is removed from the vessel and may be recycled as diluent. Additionally, the molten salt stream may be pumped from the reactor, circulated outside the reactor for further processing, and delivered back into the reactor or cooled and circulated to the feed delivery system to further dilute the fuel mixture entering the reactor.

Hydrogen Generation by Koh-Ethanol Plasma Electrolysis Using Double Compartement Reactor

NASA Astrophysics Data System (ADS)

Saksono, Nelson; Sasiang, Johannes; Dewi Rosalina, Chandra; Budikania, Trisutanti

2018-03-01

This study has successfully investigated the generation of hydrogen using double compartment reactor with plasma electrolysis process. Double compartment reactor is designed to achieve high discharged voltage, high concentration, and also reduce the energy consumption. The experimental results showed the use of double compartment reactor increased the productivity ratio 90 times higher compared to Faraday electrolysis process. The highest hydrogen production obtained is 26.50 mmol/min while the energy consumption can reach up 1.71 kJ/mmol H2 at 0.01 M KOH solution. It was shown that KOH concentration, addition of ethanol, cathode depth, and temperature have important effects on hydrogen production, energy consumption, and process efficiency.

Investigation of high temperature annealing effectiveness for recovery of radiation-induced structural changes and properties of 18Cr-10Ni-Ti austenitic stainless steels

NASA Astrophysics Data System (ADS)

Gurovich, B. A.; Kuleshova, E. A.; Frolov, A. S.; Maltsev, D. A.; Prikhodko, K. E.; Fedotova, S. V.; Margolin, B. Z.; Sorokin, A. A.

2015-10-01

A complex study of structural state and properties of 18Cr-10Ni-Ti austenitic stainless steel after irradiation in BOR-60 fast research reactor (in the temperature range 330-400 °С up to damaging doses of 145 dpa) and in VVER-1000 light water reactor (at temperature ∼320 °С and damaging doses ∼12-14 dpa) was performed. The possibility of recovery of structural-phase state and mechanical properties to the level almost corresponding to the initial state by the recovery annealing was studied. The principal possibility of the recovery annealing of pressurized water reactor internals that ensures almost complete recovery of its mechanical properties and microstructure was shown. The optimal mode of recovery annealing was established: 1000 °C during 120 h.

Lessons Learned about Liquid Metal Reactors from FFTF Experience

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wootan, David W.; Casella, Andrew M.; Omberg, Ronald P.

2016-09-20

The Fast Flux Test Facility (FFTF) is the most recent liquid-metal reactor (LMR) to operate in the United States, from 1982 to 1992. FFTF is located on the DOE Hanford Site near Richland, Washington. The 400-MWt sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission test reactor was designed specifically to irradiate Liquid Metal Fast Breeder Reactor (LMFBR) fuel and components in prototypical temperature and flux conditions. FFTF played a key role in LMFBR development and testing activities. The reactor provided extensive capability for in-core irradiation testing, including eight core positions that could be used with independent instrumentation for the test specimens.more » In addition to irradiation testing capabilities, FFTF provided long-term testing and evaluation of plant components and systems for LMFBRs. The FFTF was highly successful and demonstrated outstanding performance during its nearly 10 years of operation. The technology employed in designing and constructing this reactor, as well as information obtained from tests conducted during its operation, can significantly influence the development of new advanced reactor designs in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor operations. The FFTF complex included the reactor, as well as equipment and structures for heat removal, containment, core component handling and examination, instrumentation and control, and for supplying utilities and other essential services. The FFTF Plant was designed using a “system” concept. All drawings, specifications and other engineering documentation were organized by these systems. Efforts have been made to preserve important lessons learned during the nearly 10 years of reactor operation. A brief summary of Lessons Learned in the following areas will be discussed: Acceptance and Startup Testing of FFTF FFTF Cycle Reports« less

Chemisorption studies of Pt/SnO2 catalysts

NASA Technical Reports Server (NTRS)

Brown, Kenneth G.; Ohorodnik, Susan K.; Vannorman, John D.; Schryer, Jacqueline; Upchurch, Billy T.; Schryer, David R.

1990-01-01

The low temperature CO oxidation catalysts that are being developed and tested at NASA-Langley are fairly unique in their ability to efficiently oxidize CO at low temperatures (approx. 303 K). The bulk of the reaction data that has been collected in the laboratory has been determined using plug flow reactors with a low mass of Pt/SnO2/SiO2 catalyst (approx. 0.1 g) and a modest flow rate (5 to 10 sc sm). The researchers have previously characterized the surface solely in terms of N2 BET surface areas. These surface areas have not been that indicative of reaction rate. Indeed, some of the formulations with high BET surface area have yielded lower reaction rates than those with lower BET surface areas. As a result researchers began a program of determining the chemisorption of the various species involved in the reaction; CO, O2 and CO2. Such a determination of will lead to a better understanding of the mechanism and overall kinetics of the reaction. The pulsed-reactor technique, initially described by Freel, is used to determine the amount of a particular molecule that is adsorbed on the catalyst. Since there is some reaction of CO with the surface to produce CO2, the pulsed reactor had to be coupled with a gas chromatograph in order to distinguish between the loss of CO that is due to adsorption by the surface and the loss that is due to reaction with the surface.

Innovations for In-Pile Measurements in the Framework of the CEA-SCK•CEN Joint Instrumentation Laboratory

NASA Astrophysics Data System (ADS)

Villard, Jean-Francois; Schyns, Marc

2010-12-01

Optimizing the life cycle of nuclear systems under safety constraints requires high-performance experimental programs to reduce uncertainties on margins and limits. In addition to improvement in modeling and simulation, innovation in instrumentation is crucial for analytical and integral experiments conducted in research reactors. The quality of nuclear research programs relies obviously on an excellent knowledge of their experimental environment which constantly calls for better online determination of neutron and gamma flux. But the combination of continuously increasing scientific requirements and new experimental domains -brought for example by Generation IV programsnecessitates also major innovations for in-pile measurements of temperature, dimensions, pressure or chemical analysis in innovative mediums. At the same time, the recent arising of a European platform around the building of the Jules Horowitz Reactor offers new opportunities for research institutes and organizations to pool their resources in order to face these technical challenges. In this situation, CEA (French Nuclear Energy Commission) and SCK'CEN (Belgian Nuclear Research Centre) have combined their efforts and now share common developments through a Joint Instrumentation Laboratory. Significant progresses have thus been obtained recently in the field of in-pile measurements, on one hand by improvement of existing measurement methods, and on the other hand by introduction in research reactors of original measurement techniques. This paper highlights the state-of-the-art and the main requirements regarding in-pile measurements, particularly for the needs of current and future irradiation programs performed in material testing reactors. Some of the main on-going developments performed in the framework of the Joint Instrumentation Laboratory are also described, such as: - a unique fast neutron flux measurement system using fission chambers with 242Pu deposit and a specific online data processing, - an optical system designed to perform in-pile dimensional measurements of material samples under irradiation, - an acoustical instrumentation allowing the online characterization of fission gas release in Pressurized Water Reactor fuel rods. For each example, the obtained results, expected impacts and development status are detailed.

Regenerative Carbonate-Based Thermochemical Energy Storage System for Concentrating Solar Power

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gangwal, Santosh; Muto, Andrew

Southern Research has developed a thermochemical energy storage (TCES) technology that utilizes the endothermic-exothermic reversible carbonation of calcium oxide (lime) to store thermal energy at high-temperatures, such as those achieved by next generation concentrating solar power (CSP) facilities. The major challenges addressed in the development of this system include refining a high capacity, yet durable sorbent material and designing a low thermal resistance low-cost heat exchanger reactor system to move heat between the sorbent and a heat transfer fluid under conditions relevant for CSP operation (e.g., energy density, reaction kinetics, heat flow). The proprietary stabilized sorbent was developed by Precisionmore » Combustion, Inc. (PCI). A factorial matrix of sorbent compositions covering the design space was tested using accelerated high throughput screening in a thermo-gravimetric analyzer. Several promising formulations were selected for more thorough evaluation and one formulation with high capacity (0.38 g CO 2/g sorbent) and durability (>99.7% capacity retention over 100 cycles) was chosen as a basis for further development of the energy storage reactor system. In parallel with this effort, a full range of currently available commercial and developmental heat exchange reactor systems and sorbent loading methods were examined through literature research and contacts with commercial vendors. Process models were developed to examine if a heat exchange reactor system and balance of plant can meet required TCES performance and cost targets, optimizing tradeoffs between thermal performance, exergetic efficiency, and cost. Reactor types evaluated included many forms, from microchannel reactor, to diffusion bonded heat exchanger, to shell and tube heat exchangers. The most viable design for application to a supercritical CO 2 power cycle operating at 200-300 bar pressure and >700°C was determined to be a combination of a diffusion bonded heat exchanger with a shell and tube reactor. A bench scale reactor system was then designed and constructed to test sorbent performance under more commercially relevant conditions. This system utilizes a tube-in tube reactor design containing approximately 250 grams sorbent and is able to operate under a wide range of temperature, pressure and flow conditions as needed to explore system performance under a variety of operating conditions. A variety of sorbent loading methods may be tested using the reactor design. Initial bench test results over 25 cycles showed very high sorbent stability (>99%) and sufficient capacity (>0.28 g CO 2/g sorbent) for an economical commercial-scale system. Initial technoeconomic evaluation of the proposed storage system show that the sorbent cost should not have a significant impact on overall system cost, and that the largest cost impacts come from the heat exchanger reactor and balance of plant equipment, including compressors and gas storage, due to the high temperatures for sCO 2 cycles. Current estimated system costs are $47/kWhth based on current material and equipment cost estimates.« less

HyPEP FY06 Report: Models and Methods

DOE Office of Scientific and Technical Information (OSTI.GOV)

DOE report

2006-09-01

The Department of Energy envisions the next generation very high-temperature gas-cooled reactor (VHTR) as a single-purpose or dual-purpose facility that produces hydrogen and electricity. The Ministry of Science and Technology (MOST) of the Republic of Korea also selected VHTR for the Nuclear Hydrogen Development and Demonstration (NHDD) Project. This research project aims at developing a user-friendly program for evaluating and optimizing cycle efficiencies of producing hydrogen and electricity in a Very-High-Temperature Reactor (VHTR). Systems for producing electricity and hydrogen are complex and the calculations associated with optimizing these systems are intensive, involving a large number of operating parameter variations andmore » many different system configurations. This research project will produce the HyPEP computer model, which is specifically designed to be an easy-to-use and fast running tool for evaluating nuclear hydrogen and electricity production facilities. The model accommodates flexible system layouts and its cost models will enable HyPEP to be well-suited for system optimization. Specific activities of this research are designed to develop the HyPEP model into a working tool, including (a) identifying major systems and components for modeling, (b) establishing system operating parameters and calculation scope, (c) establishing the overall calculation scheme, (d) developing component models, (e) developing cost and optimization models, and (f) verifying and validating the program. Once the HyPEP model is fully developed and validated, it will be used to execute calculations on candidate system configurations. FY-06 report includes a description of reference designs, methods used in this study, models and computational strategies developed for the first year effort. Results from computer codes such as HYSYS and GASS/PASS-H used by Idaho National Laboratory and Argonne National Laboratory, respectively will be benchmarked with HyPEP results in the following years.« less

Comparison of irradiation behaviour of HTR graphite grades

NASA Astrophysics Data System (ADS)

Heijna, M. C. R.; de Groot, S.; Vreeling, J. A.

2017-08-01

The INNOGRAPH irradiations were executed in the High Flux Reactor (HFR) in Petten by NRG supported by the European Framework programs HTR-M, RAPHAEL, and ARCHER to generate data on the irradiation behaviour of graphite grades for High Temperature Reactor (HTR) application available at that time. Samples of the graphite grades NBG-10, NBG-17, NBG-18, NBG-20, NBG-25, PCEA, PPEA, PCIB, and IG-110 have been irradiated at 750 °C and 950 °C. The inherent scatter induced by the probabilistic material behaviour of graphite requires uncertainty and scatter induced by test conditions and post-irradiation examination to be minimized. The INNOGRAPH irradiations supplied an adequate number of irradiated samples to enable accurate determination of material properties and their evolution under irradiation. This allows comparison of different graphite grades and a qualitative assessment of their appropriateness for HTR applications, as a basis of selection, design and core component lifetime. The results indicate that coarse grained graphite grades exhibit more favourable behaviour for application in HTRs due to their low dimensional anisotropy and fracture propagation resilience.

Investigation of Iron Oxide Morphology in a Cyclic Redox Water Splitting Process for Hydrogen Generation

PubMed Central

Bobek, Michael M.; Stehle, Richard C.; Hahn, David W.

2012-01-01

A solar fuels generation research program is focused on hydrogen production by means of reactive metal water splitting in a cyclic iron-based redox process. Iron-based oxides are explored as an intermediary reactive material to dissociate water molecules at significantly reduced thermal energies. With a goal of studying the resulting oxide chemistry and morphology, chemical assistance via CO is used to complete the redox cycle. In order to exploit the unique characteristics of highly reactive materials at the solar reactor scale, a monolithic laboratory scale reactor has been designed to explore the redox cycle at temperatures ranging from 675 to 875 K. Using high resolution scanning electron microscope (SEM) and electron dispersive X-ray spectroscopy (EDS), the oxide morphology and the oxide state are quantified, including spatial distributions. These images show the change of the oxide layers directly after oxidation and after reduction. The findings show a significant non-stoichiometric O/Fe gradient in the atomic ratio following oxidation, which is consistent with a previous kinetics model, and a relatively constant, non-stoichiometric O/Fe atomic ratio following reduction.

Analysis on Operating Parameter Design to Steam Methane Reforming in Heat Application RDE

NASA Astrophysics Data System (ADS)

Dibyo, Sukmanto; Sunaryo, Geni Rina; Bakhri, Syaiful; Zuhair; Irianto, Ign. Djoko

2018-02-01

The high temperature reactor has been developed with various power capacities and can produce electricity and heat application. One of heat application is used for hydrogen production. Most hydrogen production occurs by steam reforming that operated at high temperature. This study aims to analyze the feasibility of heat application design of RDE reactor in the steam methane reforming for hydrogen production using the ChemCAD software. The outlet temperature of cogeneration heat exchanger is analyzed to be applied as a feed of steam reformer. Furthermore, the additional heater and calculating amount of fuel usage are described. Results show that at a low mass flow rate of feed, its can produce a temperature up to 480°C. To achieve the temperature of steam methane reforming of 850°C the additional fired heater was required. By the fired heater, an amount of fuel usage is required depending on the Reformer feed temperature produced from the heat exchanger of the cogeneration system.

Prospective scenarios of nuclear energy evolution over the 21. century

DOE Office of Scientific and Technical Information (OSTI.GOV)

Massara, S.; Tetart, P.; Garzenne, C.

2006-07-01

In this paper, different world scenarios of nuclear energy development over the 21. century are analyzed, by means of the EDF fuel cycle simulation code for nuclear scenario studies, TIRELIRE - STRATEGIE. Three nuclear demand scenarios are considered, and the performance of different nuclear strategies in satisfying these scenarios is analyzed and discussed, focusing on natural uranium consumption and industrial requirements related to the nuclear reactors and the associated fuel cycle facilities. Both thermal-spectrum systems (Pressurized Water Reactor and High Temperature Gas-cooled Reactor) and Fast Reactors are investigated. (authors)

Multiscale Evaluation of Catalytic Upgrading of Biomass Pyrolysis Vapors on Ni- and Ga-Modified ZSM-5

DOE PAGES

Yung, Matthew M.; Stanton, Alexander R.; Iisa, Kristiina; ...

2016-10-07

Metal-impregnated (Ni or Ga) ZSM-5 catalysts were studied for biomass pyrolysis vapor upgrading to produce hydrocarbons using three reactors constituting a 100 000x change in the amount of catalyst used in experiments. Catalysts were screened for pyrolysis vapor phase upgrading activity in two small-scale reactors: (i) a Pyroprobe with a 10 mg catalyst in a fixed bed and (ii) a fixed-bed reactor with 500 mg of catalyst. The best performing catalysts were then validated with a larger scale fluidized-bed reactor (using ~1 kg of catalyst) that produced measurable quantities of bio-oil for analysis and evaluation of mass balances. Despite somemore » inherent differences across the reactor systems (such as residence time, reactor type, analytical techniques, mode of catalyst and biomass feed) there was good agreement of reaction results for production of aromatic hydrocarbons, light gases, and coke deposition. Relative to ZSM-5, Ni or Ga addition to ZSM-5 increased production of fully deoxygenated aromatic hydrocarbons and light gases. In the fluidized bed reactor, Ga/ZSM-5 slightly enhanced carbon efficiency to condensed oil, which includes oxygenates in addition to aromatic hydrocarbons, and reduced oil oxygen content compared to ZSM-5. Ni/ZSM-5, while giving the highest yield of fully deoxygenated aromatic hydrocarbons, gave lower overall carbon efficiency to oil but with the lowest oxygen content. Reaction product analysis coupled with fresh and spent catalyst characterization indicated that the improved performance of Ni/ZSM-5 is related to decreasing deactivation by coking, which keeps the active acid sites accessible for the deoxygenation and aromatization reactions that produce fully deoxygenated aromatic hydrocarbons. The addition of Ga enhances the dehydrogenation activity of the catalyst, which leads to enhanced olefin formation and higher fully deoxygenated aromatic hydrocarbon yields compared to unmodified ZSM-5. Catalyst characterization by ammonia temperature programmed desorption, surface area measurements, and postreaction temperature-programmed oxidation (TPO) also showed that the metal-modified zeolites retained a greater percentage of their initial acidity and surface area, which was consistent between the reactor scales. These results demonstrate that the trends observed with smaller (milligram to gram) catalyst reactors are applicable to larger, more industrially relevant (kg) scales to help guide catalyst research toward application.« less

MOLTEN FLUORIDE NUCLEAR REACTOR FUEL

DOEpatents

Barton, C.J.; Grimes, W.R.

1960-01-01

Molten-salt reactor fuel compositions consisting of mixtures of fluoride salts are reported. In its broadest form, the composition contains an alkali fluoride such as sodium fluoride, zirconium tetrafluoride, and a uranium fluoride, the latter being the tetrafluoride or trifluoride or a mixture of the two. An outstanding property of these fuel compositions is a high coeffieient of thermal expansion which provides a negative temperature coefficient of reactivity in reactors in which they are used.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stillman, J. A.; Feldman, E. E.; Jaluvka, D.

This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members in the Research and Test Reactor Department at the Argonne National Laboratory (ANL) and the MURR Facility. MURR LEU conversion is part of an overall effort to develop and qualify high-density fuel within the U.S. High Performance Research Reactor Conversion (USHPRR) program conducted by the U.S. Department of Energy National Nuclearmore » Security Administration’s Office of Material Management and Minimization (M 3).« less

The Nuclear Renaissance — Implications on Quantitative Nondestructive Evaluations

NASA Astrophysics Data System (ADS)

Matzie, Regis A.

2007-03-01

The world demand for energy is growing rapidly, particularly in developing countries that are trying to raise the standard of living for billions of people, many of whom do not even have access to electricity. With this increased energy demand and the high and volatile price of fossil fuels, nuclear energy is experiencing resurgence. This so-called nuclear renaissance is broad based, reaching across Asia, the United States, Europe, as well as selected countries in Africa and South America. Some countries, such as Italy, that have actually turned away from nuclear energy are reconsidering the advisability of this design. This renaissance provides the opportunity to deploy more advanced reactor designs that are operating today, with improved safety, economy, and operations. In this keynote address, I will briefly present three such advanced reactor designs in whose development Westinghouse is participating. These designs include the advanced passive PWR, AP1000, which recently received design certification for the US Nuclear Regulatory Commission; the Pebble Bed Modular reactor (PBMR) which is being demonstrated in South Africa; and the International Reactor Innovative and Secure (IRIS), which was showcased in the US Department of Energy's recently announced Global Nuclear Energy Partnership (GNEP), program. The salient features of these designs that impact future requirements on quantitative nondestructive evaluations will be discussed. Such features as reactor vessel materials, operating temperature regimes, and new geometric configurations will be described, and mention will be made of the impact on quantitative nondestructive evaluation (NDE) approaches.

Proceedings of the Conference on High-temperature Electronics

NASA Technical Reports Server (NTRS)

1981-01-01

The development of electronic devices for use in high temperature environments is addressed. The instrumentational needs of planetary exploration, fossil and nuclear power reactors, turbine engine monitoring, and well logging are defined. Emphasis is place on the fabrication and performance of materials and semiconductor devices, circuits and systems and packaging.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burkes, Douglas E.; Senor, David J.; Casella, Andrew M.

Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. The current paper extends a failure model originally developed for UO2-stainless steel dispersion fuels and used currently available thermal-mechanical property information for the materials ofmore » interest in the current proposed design. A number of fabrication and irradiation parameters were investigated to understand the conditions at which failure of the matrix, classified as pore formation in the matrix, might occur. The results compared well with experimental observations published as part of the Reduced Enrichment for Research and Test Reactors (RERTR)-6 and -7 mini-plate experiments. Fission rate, a function of the 235U enrichment, appeared to be the most influential parameter in premature failure, mainly as a result of increased interaction layer formation and operational temperature, which coincidentally decreased the yield strength of the matrix and caused more rapid fission gas production and recoil into the surrounding matrix material. Addition of silicon to the matrix appeared effective at reducing the rate of interaction layer formation and can extend the performance of a fuel plate under a certain set of irradiation conditions, primarily moderate heat flux and burnup. Increasing the dispersed fuel particle diameter may also be effective, but only when combined with other parameters, e.g., lower enrichment and increased Si concentration. The model may serve as a valuable tool in initial experimental design.« less

Hydrogen Permeability of Incoloy 800H, Inconel 617, and Haynes 230 Alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pattrick Calderoni

A potential issue in the design of the NGNP reactor and high-temperature components is the permeation of fission generated tritium and hydrogen product from downstream hydrogen generation through high-temperature components. Such permeation can result in the loss of fission-generated tritium to the environment and the potential contamination of the helium coolant by permeation of product hydrogen into the coolant system. The issue will be addressed in the engineering design phase, and requires knowledge of permeation characteristics of the candidate alloys. Of three potential candidates for high-temperature components of the NGNP reactor design, the hydrogen permeability has been documented well onlymore » for Incoloy 800H, but at relatively high partial pressures of hydrogen. Hydrogen permeability data have been published for Inconel 617, but only in two literature reports and for partial pressures of hydrogen greater than one atmosphere, far higher than anticipated in the NGNP reactor. The hydrogen permeability of Haynes 230 has not been published. To support engineering design of the NGNP reactor components, the hydrogen permeability of Inconel 617 and Haynes 230 were determined using a measurement system designed and fabricated at the Idaho National Laboratory. The performance of the system was validated using Incoloy 800H as reference material, for which the permeability has been published in several journal articles. The permeability of Incoloy 800H, Inconel 617 and Haynes 230 was measured in the temperature range 650 to 950 °C and at hydrogen partial pressures of 10-3 and 10-2 atm, substantially lower pressures than used in the published reports. The measured hydrogen permeability of Incoloy 800H and Inconel 617 were in good agreement with published values obtained at higher partial pressures of hydrogen. The hydrogen permeability of Inconel 617 and Haynes 230 were similar, about 50% greater than for Incoloy 800H and with similar temperature dependence.« less

Surfactant studies for bench-scale operation

NASA Technical Reports Server (NTRS)

Hickey, Gregory S.; Sharma, Pramod K.

1992-01-01

A phase 2 study was initiated to investigate surfactant-assisted coal liquefaction, with the objective of quantifying the enhancement in liquid yields and product quality. This publication covers the first quarter of work. The major accomplishments were: the refurbishment of the high-pressure, high-temperature reactor autoclave, the completion of four coal liquefaction runs with Pittsburgh #8 coal, two each with and without sodium lignosulfonate surfactant, and the development of an analysis scheme for the product liquid filtrate and filter cake. Initial results at low reactor temperatures show that the addition of the surfactant produces an improvement in conversion yields and an increase in lighter boiling point fractions for the filtrate.

Thermionic switched self-actuating reactor shutdown system

DOEpatents

Barrus, Donald M.; Shires, Charles D.; Brummond, William A.

1989-01-01

A self-actuating reactor shutdown system incorporating a thermionic switched electromagnetic latch arrangement which is responsive to reactor neutron flux changes and to reactor coolant temperature changes. The system is self-actuating in that the sensing thermionic device acts directly to release (scram) the control rod (absorber) without reference or signal from the main reactor plant protective and control systems. To be responsive to both temperature and neutron flux effects, two detectors are used, one responsive to reactor coolant temperatures, and the other responsive to reactor neutron flux increase. The detectors are incorporated into a thermionic diode connected electrically with an electromagnetic mechanism which under normal reactor operating conditions holds the the control rod in its ready position (exterior of the reactor core). Upon reaching either a specified temperature or neutron flux, the thermionic diode functions to short-circuit the electromagnetic mechanism causing same to lose its holding power and release the control rod, which drops into the reactor core region under gravitational force.

Status of the US RERTR Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Travelli, A.

1995-02-01

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1994 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 1993 in collaboration with its many international partners. The RERTR Program has moved aggressively to support President Clinton`s nonproliferation policy and his goal {open_quotes}to minimize the use of highly-enriched uranium in civil nuclear programs{close_quotes}. An Environmental Assessment which addresses the urgent-relief acceptance of 409 spent fuel elements was completed, and the first shipment of spent fuel elements is scheduledmore » for this month. An Environmental Impact Statement addressing the acceptance of spent research reactor fuel containing enriched uranium of U.S. origin is scheduled for completion by the end of June 1995. The U.S. administration has decided to resume development of high-density LEU research reactor fuels. DOE funding and guidance are expected to begin soon. A preliminary plan for the resumption of fuel development has been prepared and is ready for implementation. The scope and main technical activities of a plan to develop and demonstrate within the next five years the technical means needed to convert Russian-supplied research reactors to LEU fuels was agreed upon by the RERTR Program and four Russian institutes lead by RDIPE. Both Secretary O`Leary and Minister Michailov have expressed strong support for this initiative. Joint studies have made significant progress, especially in assessing the technical and economic feasibility of using reduced enrichment fuels in the SAFARI-I reactor in South Africa and in the Advanced Neutron Source reactor under design at ORNL. Significant progress was achieved on several aspects of producing {sup 99}Mo from fission targets utilizing LEU instead of HEU to the achievement of the common goal.« less

Design and Laboratory Evaluation of Future Elongation and Diameter Measurements at the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

K. L. Davis; D. L. Knudson; J. L. Rempe

New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. Such materials can undergo significant dimensional and physical changes during high temperature irradiations. In order to accurately predict these changes, real-time data must be obtained under prototypic irradiation conditions for model development and validation. To provide such data, researchers at the Idaho National Laboratory (INL) High Temperature Test Laboratory (HTTL) are developing several instrumented test rigs to obtain data real-time from specimens irradiated in well-controlled pressurized water reactor (PWR) coolant conditions in the Advanced Test Reactor (ATR). This paper reports the status ofmore » INL efforts to develop and evaluate prototype test rigs that rely on Linear Variable Differential Transformers (LVDTs) in laboratory settings. Although similar LVDT-based test rigs have been deployed in lower flux Materials Testing Reactors (MTRs), this effort is unique because it relies on robust LVDTs that can withstand higher temperatures and higher fluxes than often found in other MTR irradiations. Specifically, the test rigs are designed for detecting changes in length and diameter of specimens irradiated in ATR PWR loops. Once implemented, these test rigs will provide ATR users with unique capabilities that are sorely needed to obtain measurements such as elongation caused by thermal expansion and/or creep loading and diameter changes associated with fuel and cladding swelling, pellet-clad interaction, and crud buildup.« less

Atmospheric reentry of the in-core thermionic SP-100 reactor system

NASA Technical Reports Server (NTRS)

Stamatelatos, M. G.; Barsell, A. W.; Harris, P. A.; Francisco, J.

1987-01-01

Presumed end-of-life atmospheric reentry of the GA SP-100 system was studied to assess dispersal feasibility and associated hazards. Reentry was studied by sequential use of an orbital trajectory and a heat analysis computer program. Two heating models were used. The first model assumed a thermal equilibrium condition between the stagnation point aerodynamic heating and the radiative cooling of the skin material surface. The second model allowed for infinite conductivity of the skin material. Four reentering configurations were studied representing stages of increased SP-100 breakup: (1) radiator, shield and reactor, (2) shield and reactor, (3) reactor with control drums, and (4) reactor without control drums. Each reentering configuration was started from a circular orbit at 116 km having an inertial velocity near Mach 25. The assumed failing criterion was the attainment of melting temperature of a critical system component. The reentry analysis revealed breakup of the vessel in the neighborhood of 61 km altitude and scattering of the fuel elements. Subsequent breakup of the fuel elements was not predicted. Oxidation of the niobium skin material was calculated to cause an increase in surface temperature of less than ten percent. The concept of thermite analogs for enhancing reactor reentry dispersal was assessed and found to be feasible in principle. A conservative worst-case hazards analysis was performed for radioactive and nonradioactive toxic SP-100 materials assumed to be dispersed during end-of-life reentry. The hazards associated with this phase of the SP-100 mission were calculated to be insignificant.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gehin, Jess C.; Powers, Jeffrey J.

Molten salt reactors (MSRs) represent a class of reactors that use liquid salt, usually fluoride- or chloride-based, as either a coolant with a solid fuel (such as fluoride salt-cooled high temperature reactors) or as a combined coolant and fuel with fuel dissolved in a carrier salt. For liquid-fuelled MSRs, the salt can be processed online or in a batch mode to allow for removal of fission products as well as introduction of fissile fuel and fertile materials during reactor operation. The MSR is most commonly associated with the 233U/thorium fuel cycle, as the nuclear properties of 233U combined with themore » online removal of parasitic absorbers allow for the ability to design a thermal-spectrum breeder reactor; however, MSR concepts have been developed using all neutron energy spectra (thermal, intermediate, fast, and mixed-spectrum zoned concepts) and with a variety of fuels including uranium, thorium, plutonium, and minor actinides. Early MSR work was supported by a significant research and development (R&D) program that resulted in two experimental systems operating at ORNL in the 1960s, the Aircraft Reactor Experiment and the Molten Salt Reactor Experiment. Subsequent design studies in the 1970s focusing on thermal-spectrum thorium-fueled systems established reference concepts for two major design variants: (1) a molten salt breeder reactor (MSBR), with multiple configurations that could breed additional fissile material or maintain self-sustaining operation; and (2) a denatured molten salt reactor (DMSR) with enhanced proliferation-resistance. T MSRs has been selected as one of six most promising Generation IV systems and development activities have been seen in fast-spectrum MSRs, waste-burning MSRs, MSRs fueled with low-enriched uranium (LEU), as well as more traditional thorium fuel cycle-based MSRs. This study provides an historical background of MSR R&D efforts, surveys and summarizes many of the recent development, and provides analysis comparing thorium-based MSRs.« less

Creep Strength of Nb-1Zr for SP-100 Applications

NASA Astrophysics Data System (ADS)

Horak, James A.; Egner, Larry K.

1994-07-01

Power systems that are used to provide electrical power in space are designed to optimize conversion of thermal energy to electrical energy and to minimize the mass and volume that must be launched. Only refractory metals and their alloys have sufficient long-term strength for several years of uninterrupted operation at the required temperatures of 1200 K and above. The high power densities and temperatures at which these reactors must operate require the use of liquid-metal coolants. The alloy Nb-1 wt % Zr (Nb-lZr), which exhibits excellent corrosion resistance to alkali liquid-metals at high temperatures, is being considered for the fuel cladding, reactor structural, and heat-transport systems for the SP-100 reactor system. Useful lifetime of this system is limited by creep deformation in the reactor core. Nb-lZr sheet procured to American Society for Testing and Materials (ASTM) specifications for reactor grade and commercial grade has been processed by several different cold work and annealing treatments to attempt to produce the grain structure (size, shape, and distribution of sizes) that provides the maximum creep strength of this alloy at temperatures from 1250 to 1450 K. The effects of grain size, differences in oxygen concentrations, tungsten concentrations, and electron beam and gas tungsten arc weldments on creep strength were studied. Grain size has a large effect on creep strength at 1450 K but only material with a very large grain size (150 μm) exhibits significantly higher creep strength at 1350 K. Differences in oxygen or tungsten concentrations did not affect creep strength, and the creep strengths of weldments were equal to, or greater than, those for base metal.

Development of ASTM Standard for SiC-SiC Joint Testing Final Scientific/Technical Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jacobsen, George; Back, Christina

2015-10-30

As the nuclear industry moves to advanced ceramic based materials for cladding and core structural materials for a variety of advanced reactors, new standards and test methods are required for material development and licensing purposes. For example, General Atomics (GA) is actively developing silicon carbide (SiC) based composite cladding (SiC-SiC) for its Energy Multiplier Module (EM2), a high efficiency gas cooled fast reactor. Through DOE funding via the advanced reactor concept program, GA developed a new test method for the nominal joint strength of an endplug sealed to advanced ceramic tubes, Fig. 1-1, at ambient and elevated temperatures called themore » endplug pushout (EPPO) test. This test utilizes widely available universal mechanical testers coupled with clam shell heaters, and specimen size is relatively small, making it a viable post irradiation test method. The culmination of this effort was a draft of an ASTM test standard that will be submitted for approval to the ASTM C28 ceramic committee. Once the standard has been vetted by the ceramics test community, an industry wide standard methodology to test joined tubular ceramic components will be available for the entire nuclear materials community.« less

Comparison between instrumented precracked Charpy and compact specimen tests of carbon steels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nanstad, R.K.

1980-01-01

The General Atomic Company High Temperature Gas-Cooled Reactor (HTGR) is housed within a prestressed concrete reactor vessel (PCRV). Various carbon steel structural members serve as closures at penetrations in the vessel. A program of testing and evaluation is underway to determine the need for reference fracture toughness (K/sub IR/) and indexing procedures for these materials as described in Appendix G to Section III, ASME Code for light water reactor steels. The materials of interest are carbon steel forgings (SA508, Class 1) and plates (SA537, Classes 1 and 2) as well as weldments of these steels. The fracture toughness behavior ismore » characterized with instrumented precracked Charpy V-votch specimens (PCVN) - slow-bend and dynamic - and compact specimens (10-mm and 25-mm thicknesses) using both linear elastic (ASTM E399) and elastic-plastic (equivalent Energy and J-Integral) analytical procedures. For the dynamic PCVN tests, force-time traces are analyzed according to the procedures of the Pressure Vessel Research Council (PVRC)/Metal Properties Council (MPC). Testing and analytical procedures are discussed and PCVN results are compared to those obtained with compact specimens.« less