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.

NEUTRONIC REACTOR CONSTRUCTION AND OPERATION

DOEpatents

West, J.M.; Weills, J.T.

1960-03-15

A method is given for operating a nuclear reactor having a negative coefficient of reactivity to compensate for the change in reactor reactivity due to the burn-up of the xenon peak following start-up of the reactor. When it is desired to start up the reactor within less than 72 hours after shutdown, the temperature of the reactor is lowered prior to start-up, and then gradually raised after start-up.

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

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.

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.

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. .

Evaluation of infrared thermography as a diagnostic tool in CVD applications

NASA Astrophysics Data System (ADS)

Johnson, E. J.; Hyer, P. V.; Culotta, P. W.; Clark, I. O.

1998-05-01

This research is focused on the feasibility of using infrared temperature measurements on the exterior of a chemical vapor deposition (CVD) reactor to ascertain both real-time information on the operating characteristics of a CVD system and provide data which could be post-processed to provide quantitative information for research and development on CVD processes. Infrared thermography techniques were used to measure temperatures on a horizontal CVD reactor of rectangular cross section which were correlated with the internal gas flow field, as measured with the laser velocimetry (LV) techniques. For the reactor tested, thermal profiles were well correlated with the gas flow field inside the reactor. Correlations are presented for nitrogen and hydrogen carrier gas flows. The infrared data were available to the operators in real time with sufficient sensitivity to the internal flow field so that small variations such as misalignment of the reactor inlet could be observed. The same data were post-processed to yield temperature measurements at known locations on the reactor surface. For the experiments described herein, temperatures associated with approximately 3.3 mm 2 areas on the reactor surface were obtained with a precision of ±2°C. These temperature measurements were well suited for monitoring a CVD production reactor, development of improved thermal boundary conditions for use in CFD models of reactors, and for verification of expected thermal conditions.

PRELIMINARY HAZARDS SUMMARY REPORT FOR THE VALLECITOS SUPERHEAT REACTOR

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

Murray, J.L.

1961-02-01

BS>The Vallecitos Superheat Reactor (VSR) is a light-watermoderated, thermal-spectrum reactor, cooled by a combination of moderator boiling and forced convection cooling with saturated steam. The reactor core consists of 32 fuel hurdles containing 5300 lb of UO/sub 2/ enriched in U/sub 235/ to 3.6%. The fuel elements are arranged in individual process tubes that direct the cooling steam flow and separate the steam from the water moderator. The reactor vessel is designed for 1250 psig and operates at 960 to 1000 psig. With the reactor operating at 12.5 Mw(t), the maximum fuel cladding temperature is 1250 deg F and themore » cooling steam is superheated to an average temperature of about 810 deg F at 905 psig. Nu clear operation of the reactor is controlled by 12 control rods, actuated by drives mounted on the bottom of the reactor vessel. The water moderator recirculates inside the reactor vessel and through the core region by natural convection. Inherent safety features of the reactor include the negative core reactivity effects upon heating the UO/sub 2/ fuel (Doppler effect), upon increasing the temperature or void content of the moderator in the operating condition, and upon unflooding the fuel process tubes in the hot condition. Snfety features designed into the reactor and plant systems include a system of sensors and devices to detect petentially unsafe operating conditions and to initiate automatically the appropriate countermeasures, a set of fast and reliable control rods for scramming the reactor if a potentially unsafe condition occurs, a manually-actuated liquid neutron poison system, and an emergency cooling system to provide continued steam flow through the reactor core in the event the reactor becomes isolated from either its normal source of steam supply or discharge. The release of radioactivity to unrestricted areas is maintained within permissible limits by monitoring the radioactivity of wastes and controlling their release. The reactor and many of its auxiliaries are housed within a high-integrity essentially leak-tight containment vessel. (auth)« less

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

Entropy Production in Chemical Reactors

NASA Astrophysics Data System (ADS)

Kingston, Diego; Razzitte, Adrián C.

2017-06-01

We have analyzed entropy production in chemically reacting systems and extended previous results to the two limiting cases of ideal reactors, namely continuous stirred tank reactor (CSTR) and plug flow reactor (PFR). We have found upper and lower bounds for the entropy production in isothermal systems and given expressions for non-isothermal operation and analyzed the influence of pressure and temperature in entropy generation minimization in reactors with a fixed volume and production. We also give a graphical picture of entropy production in chemical reactions subject to constant volume, which allows us to easily assess different options. We show that by dividing a reactor into two smaller ones, operating at different temperatures, the entropy production is lowered, going as near as 48 % less in the case of a CSTR and PFR in series, and reaching 58 % with two CSTR. Finally, we study the optimal pressure and temperature for a single isothermal PFR, taking into account the irreversibility introduced by a compressor and a heat exchanger, decreasing the entropy generation by as much as 30 %.

Kinetic study on the effect of temperature on biogas production using a lab scale batch reactor.

PubMed

Deepanraj, B; Sivasubramanian, V; Jayaraj, S

2015-11-01

In the present study, biogas production from food waste through anaerobic digestion was carried out in a 2l laboratory-scale batch reactor operating at different temperatures with a hydraulic retention time of 30 days. The reactors were operated with a solid concentration of 7.5% of total solids and pH 7. The food wastes used in this experiment were subjected to characterization studies before and after digestion. Modified Gompertz model and Logistic model were used for kinetic study of biogas production. The kinetic parameters, biogas yield potential of the substrate (B), the maximum biogas production rate (Rb) and the duration of lag phase (λ), coefficient of determination (R(2)) and root mean square error (RMSE) were estimated in each case. The effect of temperature on biogas production was evaluated experimentally and compared with the results of kinetic study. The results demonstrated that the reactor with operating temperature of 50°C achieved maximum cumulative biogas production of 7556ml with better biodegradation efficiency. Copyright © 2015 Elsevier Inc. All rights reserved.

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

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.

HOMOGENEOUS NUCLEAR POWER REACTOR

DOEpatents

King, L.D.P.

1959-09-01

A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

Pilot plant operation of a nonadiabatic methanation reactor. [15 refs. ; Raney nickel catalyst

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

Schehl, R.R.; Pennline, H.W.; Strakey, J.P.

The design and operation of a pilot plant scale hybrid methanation reactor is discussed. The hybrid methanator, utilizing a finned, Raney nickel coated insert, consolidates features of the tube-wall and hot-gas-recycle methanation reactors. Data are presented from four tests lasting from 3/sup 1///sub 2/ weeks to three months. Topics discussed include conversion, product yields, catalyst properties, and reactor temperature profiles. A one-dimensional mathematical model capable of explaining reactor performance trends is employed.

REACTOR CONTROL DEVICE

DOEpatents

Graham, R.H.

1962-09-01

A wholly mechanical compact control device is designed for automatically rendering the core of a fission reactor subcritical in response to core temperatures in excess of the design operating temperature limit. The control device comprises an expansible bellows interposed between the base of a channel in a reactor core and the inner end of a fuel cylinder therein which is normally resiliently urged inwardly. The bellows contains a working fluid which undergoes a liquid to vapor phase change at a temperature substantially equal to the design temperature limit. Hence, the bellows abruptiy expands at this limiting temperature to force the fuel cylinder outward and render the core subcritical. The control device is particularly applicable to aircraft propulsion reactor service. (AEC)

Dynamic System Simulation of the KRUSTY Experiment

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

Klein, Steven Karl; Kimpland, Robert Herbert

2016-05-09

The proposed KRUSTY experiment is a demonstration of a reactor operating at power. The planned experimental configuration includes a highly enriched uranium (HEU) reflected core, cooled by multiple heat pipes leading to Stirling engines for primary heat rejection. Operating power is expected to be approximately four (4) to five (5) kilowatts with a core temperature above 1,000 K. No data is available on any historical reactor employing HEU metal that operated over the temperature range required for the KRUSTY experiment. Further, no reactor has operated with heat pipes as the primary cooling mechanism. Historic power reactors have employed either naturalmore » or forced convection so data on their operation is not directly applicable to the KRUSTY experiment. The primary purpose of the system model once developed and refined by data from these component experiments, will be used to plan the KRUSTY experiment. This planning will include expected behavior of the reactor from start-up, through various transient conditions where cooling begins to become present and effective, and finally establishment of steady-state. In addition, the model can provide indicators of anticipated off-normal events and appropriate operator response to those conditions. This information can be used to develop specific experiment operating procedures and aids to guide the operators in conduct of the experiment.« less

Design of a self-tuning regulator for temperature control of a polymerization reactor.

PubMed

Vasanthi, D; Pranavamoorthy, B; Pappa, N

2012-01-01

The temperature control of a polymerization reactor described by Chylla and Haase, a control engineering benchmark problem, is used to illustrate the potential of adaptive control design by employing a self-tuning regulator concept. In the benchmark scenario, the operation of the reactor must be guaranteed under various disturbing influences, e.g., changing ambient temperatures or impurity of the monomer. The conventional cascade control provides a robust operation, but often lacks in control performance concerning the required strict temperature tolerances. The self-tuning control concept presented in this contribution solves the problem. This design calculates a trajectory for the cooling jacket temperature in order to follow a predefined trajectory of the reactor temperature. The reaction heat and the heat transfer coefficient in the energy balance are estimated online by using an unscented Kalman filter (UKF). Two simple physically motivated relations are employed, which allow the non-delayed estimation of both quantities. Simulation results under model uncertainties show the effectiveness of the self-tuning control concept. Copyright © 2011 ISA. Published by Elsevier Ltd. All rights reserved.

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

Reactor safety method

DOEpatents

Vachon, Lawrence J.

1980-03-11

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

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

Assessment of Sensor Technologies for Advanced Reactors

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

Korsah, Kofi; Kisner, R. A.; Britton Jr., C. L.

This paper provides an assessment of sensor technologies and a determination of measurement needs for advanced reactors (AdvRx). It is a summary of a study performed to provide the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program. The study covered two broad reactor technology categories: High Temperature Reactors and Fast Reactors. The scope of “High temperature reactors” included Gen IV reactors whose coolant exit temperatures exceed ≈650 °C and are moderated (as opposed to fast reactors). To bound the scope formore » fast reactors, this report reviewed relevant operating experience from US-operated Sodium Fast Reactor (SFR) and relevant test experience from the Fast Flux Test Facility (FFTF). For high temperature reactors the study showed that in many cases instrumentation have performed reasonably well in research and demonstration reactors. However, even in cases where the technology is “mature” (such as thermocouples), HTGRs can benefit from improved technologies. Current HTGR instrumentation is generally based on decades-old technology and adapting newer technologies could provide significant advantages. For sodium fast reactors, the study found that several key research needs arise around (1) radiation-tolerant sensor design for in-vessel or in-core applications, where possible non-invasive sensing approaches for key parameters that minimize the need to deploy sensors in-vessel, (2) approaches to exfiltrating data from in-vessel sensors while minimizing penetrations, (3) calibration of sensors in-situ, and (4) optimizing sensor placements to maximize the information content while minimizing the number of sensors needed.« less

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.

Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

NASA Technical Reports Server (NTRS)

Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

1991-01-01

The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

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.

CONVECTION REACTOR

DOEpatents

Hammond, R.P.; King, L.D.P.

1960-03-22

An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

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

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.

Performance and emissions of a catalytic reactor with propane, diesel, and Jet A fuels

NASA Technical Reports Server (NTRS)

Anderson, D. N.

1977-01-01

Tests were made to determine the performance and emissions of a catalytic reactor operated with propane, No. 2 diesel, and Jet A fuels. A 12-cm diameter and 16-cm long catalytic reactor using a proprietary noble metal catalyst was operated at an inlet temperature of 800 K, a pressure of 300,000 Pa and reference velocities of 10 to 15 m/s. No significant differences between the performance of the three fuels were observed when 98.5 percent purity propane was used. The combustion efficiency for 99.8-percent purity propane tested later was significantly lower, however. The diesel fuel contained 135 ppm of bound nitrogen and consequently produced the highest NOx emissions of the three fuels. As much as 85 percent of the bound nitrogen was converted to NOx. Steady-state emissions goals based on half the most stringent proposed automotive standards were met when the reactor was operated at an adiabatic combustion temperature higher than 1350 K with all fuels except the 99.8-percent purity propane. With that fuel, a minimum temperature of 1480 K was required.

Self-teaching neural network learns difficult reactor control problem

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

Jouse, W.C.

1989-01-01

A self-teaching neural network used as an adaptive controller quickly learns to control an unstable reactor configuration. The network models the behavior of a human operator. It is trained by allowing it to operate the reactivity control impulsively. It is punished whenever either the power or fuel temperature stray outside technical limits. Using a simple paradigm, the network constructs an internal representation of the punishment and of the reactor system. The reactor is constrained to small power orbits.

Process for operating equilibrium controlled reactions

DOEpatents

Nataraj, Shankar; Carvill, Brian Thomas; Hufton, Jeffrey Raymond; Mayorga, Steven Gerard; Gaffney, Thomas Richard; Brzozowski, Jeffrey Richard

2001-01-01

A cyclic process for operating an equilibrium controlled reaction in a plurality of reactors containing an admixture of an adsorbent and a reaction catalyst suitable for performing the desired reaction which is operated in a predetermined timed sequence wherein the heating and cooling requirements in a moving reaction mass transfer zone within each reactor are provided by indirect heat exchange with a fluid capable of phase change at temperatures maintained in each reactor during sorpreaction, depressurization, purging and pressurization steps during each process cycle.

Operating characteristic analysis of a 400 mH class HTS DC reactor in connection with a laboratory scale LCC type HVDC system

NASA Astrophysics Data System (ADS)

Kim, Sung-Kyu; Kim, Kwangmin; Park, Minwon; Yu, In-Keun; Lee, Sangjin

2015-11-01

High temperature superconducting (HTS) devices are being developed due to their advantages. Most line commutated converter based high voltage direct current (HVDC) transmission systems for long-distance transmission require large inductance of DC reactor; however, generally, copper-based reactors cause a lot of electrical losses during the system operation. This is driving researchers to develop a new type of DC reactor using HTS wire. The authors have developed a 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC system and applied the HTS DC reactor to the HVDC system to investigate their operating characteristics. The 400 mH class HTS DC reactor is designed using a toroid type magnet. The HVDC system is designed in the form of a mono-pole system with thyristor-based 12-pulse power converters. In this paper, the investigation results of the HTS DC reactor in connection with the HVDC system are described. The operating characteristics of the HTS DC reactor are analyzed under various operating conditions of the system. Through the results, applicability of an HTS DC reactor in an HVDC system is discussed in detail.

Low temperature pre-treatment of domestic sewage in an anaerobic hybrid or an anaerobic filter reactor.

PubMed

Elmitwalli, Tarek A; Sklyar, Vladimir; Zeeman, Grietje; Lettinga, Gatze

2002-05-01

The pre-treatment of domestic sewage for removal of suspended solids (SS) at a process temperature of 13 degrees C and an hydraulic retention time (HRT) of 4 h was investigated in an anaerobic filter (AF) and anaerobic hybrid (AH) reactor. The AF and the top of the AH reactor consisted of vertical sheets of reticulated polyurethane foam (RPF) with knobs. All biomass in the AF was only in attached form to avoid clogging and sludge washout. The AF reactor showed a significantly higher removal of total and suspended chemical oxygen demand (COD) than the AH reactor, respectively, 55% and 82% in the AF reactor and 34% and 53% in the AH reactor. Because the reactors were operated at a short HRT and low temperature, the hydrolysis, acidification and methanogenesis based on the influent COD were limited to, respectively, 12%, 21% and 23% for the AF reactor and 12%, 17% and 16% for the AH reactor. The excess sludge from the AH reactor was more stabilised and had a better settling capacity and dewaterability. However, the excess sludge from both the AH and AF reactors needed stabilisation. Therefore, the AF reactor is recommended for the pretreatment of domestic sewage at low temperatures.

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

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

Su'ud, Zaki; Anshari, Rio

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environmentmore » such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.« less

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

NASA Astrophysics Data System (ADS)

Su'ud, Zaki; Anshari, Rio

2012-06-01

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

Catalytic Tar Reduction for Assistance in Thermal Conversion of Space Waste for Energy Production

NASA Technical Reports Server (NTRS)

Caraccio, Anne Joan; Devor, Robert William; Hintze, Paul E.; Muscatello, Anthony C.; Nur, Mononita

2014-01-01

The Trash to Gas (TtG) project investigates technologies for converting waste generated during spaceflight into various resources. One of these technologies was gasification, which employed a downdraft reactor designed and manufactured at NASA's Kennedy Space Center (KSC) for the conversion of simulated space trash to carbon dioxide. The carbon dioxide would then be converted to methane for propulsion and water for life support systems. A minor byproduct of gasification includes large hydrocarbons, also known as tars. Tars are unwanted byproducts that add contamination to the product stream, clog the reactor and cause complications in analysis instrumentation. The objective of this research was to perform reduction studies of a mock tar using select catalysts and choose the most effective for primary treatment within the KSC downdraft gasification reactor. Because the KSC reactor is operated at temperatures below typical gasification reactors, this study evaluates catalyst performance below recommended catalytic operating temperatures. The tar reduction experimentation was observed by passing a model tar vapor stream over the catalysts at similar conditions to that of the KSC reactor. Reduction in tar was determined using gas chromatography. Tar reduction efficiency and catalyst performances were evaluated at different temperatures.

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.

Two-phase anaerobic digestion of source sorted OFMSW (organic fraction of municipal solid waste): performance and kinetic study.

PubMed

Pavan, P; Battistoni, P; Cecchi, F; Mata-Alvarez, J

2000-01-01

The results of a two-phase system operated in different conditions, treating the source-sorted organic fraction of municipal solid waste (SS-OFMSW), coming mainly from fruit and vegetable markets, are presented. Hydraulic retention time (HRT) in the hydrolytic reactor and in the methanogenic reactor and also the temperature in the hydrolytic reactor (mesophilic and thermophilic conditions) are varied in order to evaluate the effect of these factors. The methanogenic reactor is always operated within the thermophilic range. Optimum operating conditions are found to be around 12 days (total system) using the mesophilic range of temperature in the first reactor. Specific gas production (SGP) in these conditions is around 0.6 m3/kg TVS. A kinetic study is also carried out, using the first and the step diffusional models. The latter gives much better results, with fitted constants comparable to other studies. Finally, a comparison with a one-phase system is carried out, showing that a two-phase system is much more appropriate for the digestion of this kind of highly biodegradable substrate in thermophilic conditions.

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

Response of a continuous anaerobic digester to temperature transitions: A critical range for restructuring the microbial community structure and function.

PubMed

Kim, Jaai; Lee, Changsoo

2016-02-01

Temperature is a crucial factor that significantly influences the microbial activity and so the methanation performance of an anaerobic digestion (AD) process. Therefore, how to control the operating temperature for optimal activity of the microbes involved is a key to stable AD. This study examined the response of a continuous anaerobic reactor to a series of temperature shifts over a wide range of 35-65 °C using a dairy-processing byproduct as model wastewater. During the long-term experiment for approximately 16 months, the reactor was subjected to stepwise temperature increases by 5 °C at a fixed HRT of 15 days. The reactor showed stable performance within the temperature range of 35-45 °C, with the methane production rate and yield being maximum at 45 °C (18% and 26% greater, respectively, than at 35 °C). However, the subsequent increase to 50 °C induced a sudden performance deterioration with a complete cessation of methane recovery, indicating that the temperature range between 45 °C and 50 °C had a critical impact on the transition of the reactor's methanogenic activity from mesophilic to thermophilic. This serious process perturbation was associated with a severe restructuring of the reactor microbial community structure, particularly of methanogens, quantitatively as well as qualitatively. Once restored by interrupted feeding for about two months, the reactor maintained fairly stable performance under thermophilic conditions until it was upset again at 65 °C. Interestingly, in contrast to most previous reports, hydrogenotrophs largely dominated the methanogen community at mesophilic temperatures while acetotrophs emerged as a major group at thermophilic temperature. This implies that the primary methanogenesis route of the reactor shifted from hydrogen- to acetate-utilizing pathways with the temperature shifts from mesophilic to thermophilic temperatures. Our observations suggest that a mesophilic digester may not need to be cooled at up to 45 °C in case of undesired temperature rise, for example, by excessive self-heating, which offers a possibility to reduce operating costs. Copyright © 2015 Elsevier Ltd. All rights reserved.

Off-design temperature effects on nuclear fuel pins for an advanced space-power-reactor concept

NASA Technical Reports Server (NTRS)

Bowles, K. J.

1974-01-01

An exploratory out-of-reactor investigation was made of the effects of short-time temperature excursions above the nominal operating temperature of 990 C on the compatibility of advanced nuclear space-power reactor fuel pin materials. This information is required for formulating a reliable reactor safety analysis and designing an emergency core cooling system. Simulated uranium mononitride (UN) fuel pins, clad with tungsten-lined T-111 (Ta-8W-2Hf) showed no compatibility problems after heating for 8 hours at 2400 C. At 2520 C and above, reactions occurred in 1 hour or less. Under these conditions free uranium formed, redistributed, and attacked the cladding.

Cavity temperature and flow characteristics in a gas-core test reactor

NASA Technical Reports Server (NTRS)

Putre, H. A.

1973-01-01

A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

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.

Technologies for Upgrading Light Water Reactor Outlet Temperature

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

Daniel S. Wendt; Piyush Sabharwall; Vivek Utgikar

Nuclear energy could potentially be utilized in hybrid energy systems to produce synthetic fuels and feedstocks from indigenous carbon sources such as coal and biomass. First generation nuclear hybrid energy system (NHES) technology will most likely be based on conventional light water reactors (LWRs). However, these LWRs provide thermal energy at temperatures of approximately 300°C, while the desired temperatures for many chemical processes are much higher. In order to realize the benefits of nuclear hybrid energy systems with the current LWR reactor fleets, selection and development of a complimentary temperature upgrading technology is necessary. This paper provides an initial assessmentmore » of technologies that may be well suited toward LWR outlet temperature upgrading for powering elevated temperature industrial and chemical processes during periods of off-peak power demand. Chemical heat transformers (CHTs) are a technology with the potential to meet LWR temperature upgrading requirements for NHESs. CHTs utilize chemical heat of reaction to change the temperature at which selected heat sources supply or consume thermal energy. CHTs could directly utilize LWR heat output without intermediate mechanical or electrical power conversion operations and the associated thermodynamic losses. CHT thermal characteristics are determined by selection of the chemical working pair and operating conditions. This paper discusses the chemical working pairs applicable to LWR outlet temperature upgrading and the CHT operating conditions required for providing process heat in NHES applications.« 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

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

Passive heat-transfer means for nuclear reactors. [LMFBR

DOEpatents

Burelbach, J.P.

1982-06-10

An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

A ``NEW'' Solid-Core Reactor Fuel Form that Maximizes the Performance of Nuclear Thermal and Electric Rockets

NASA Astrophysics Data System (ADS)

Rom, Frank E.; Finnegan, Patrick M.

1994-07-01

The ``NEW'' solid-core fuel form is the old Vapor Transport (VT) fuel pin investigated at NASA about 30 years ago. It is simply a tube sealed at both ends partially filled with UO2. During operation the UO2 forms an annular layer on the inside of the tube by vaporization and condensation. This form is an ideal structure for overall strength and retention of fission products. All of the structural material lies between the fuel (including fission products) and the reactor coolant. The isothermal inside fuel surface temperature that results from the vaporization and condensation of fuel during operation eliminates hotspots, significantly increasing the design fuel pin surface temperature. For NTP, W-UO2 fuel pins yield higher operating temperatures than for other fuel forms, because W has about a ten-fold lower vaporization rate compared to any other known material. The use of perigee propulsion using W-UO2 fuel pins can result in a more than ten-fold reduction in reactor power. Lower reactor power, together with zero fission product release potential, and the simplicity of fabrication of VT fuel pins should greatly simplify and reduce the cost of development of NTP. For NEP, VT fuel pins can increase fast neutron spectrum reactor life with no fission product release. Thermal spectrum NEP reactors using W184 or Mo VT fuel pins, with only small amounts of high neutron absorbing additives, offer benefits because of much lower fissionable fuel requirements. The VT fuel pin has application to commercial power reactors with similar benefits.

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

Reactivity Coefficient Calculation for AP1000 Reactor Using the NODAL3 Code

NASA Astrophysics Data System (ADS)

Pinem, Surian; Malem Sembiring, Tagor; Tukiran; Deswandri; Sunaryo, Geni Rina

2018-02-01

The reactivity coefficient is a very important parameter for inherent safety and stability of nuclear reactors operation. To provide the safety analysis of the reactor, the calculation of changes in reactivity caused by temperature is necessary because it is related to the reactor operation. In this paper, the temperature reactivity coefficients of fuel and moderator of the AP1000 core are calculated, as well as the moderator density and boron concentration. All of these coefficients are calculated at the hot full power condition (HFP). All neutron diffusion constant as a function of temperature, water density and boron concentration were generated by the SRAC2006 code. The core calculations for determination of the reactivity coefficient parameter are done by using NODAL3 code. The calculation results show that the fuel temperature, moderator temperature and boron reactivity coefficients are in the range between -2.613 pcm/°C to -4.657pcm/°C, -1.00518 pcm/°C to 1.00649 pcm/°C and -9.11361 pcm/ppm to -8.0751 pcm/ppm, respectively. For the water density reactivity coefficients, the positive reactivity occurs at the water temperature less than 190 °C. The calculation results show that the reactivity coefficients are accurate because the results have a very good agreement with the design value.

Reactor for exothermic reactions

DOEpatents

Smith, Jr., Lawrence A.; Hearn, Dennis; Jones, Jr., Edward M.

1993-01-01

A liquid phase process for oligomerization of C.sub.4 and C.sub.5 isoolefins or the etherification thereof with C.sub.1 to C.sub.6 alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120.degree. to 300.degree. F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

Reactor for exothermic reactions

DOEpatents

Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

1993-03-02

A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

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

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.

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.

Drivers of microbial community composition in mesophilic and thermophilic temperature-phased anaerobic digestion pre-treatment reactors.

PubMed

Pervin, Hasina M; Dennis, Paul G; Lim, Hui J; Tyson, Gene W; Batstone, Damien J; Bond, Philip L

2013-12-01

Temperature-phased anaerobic digestion (TPAD) is an emerging technology that facilitates improved performance and pathogen destruction in anaerobic sewage sludge digestion by optimising conditions for 1) hydrolytic and acidogenic organisms in a first-stage/pre-treatment reactor and then 2) methogenic populations in a second stage reactor. Pre-treatment reactors are typically operated at 55-65 °C and as such select for thermophilic bacterial communities. However, details of key microbial populations in hydrolytic communities and links to functionality are very limited. In this study, experimental thermophilic pre-treatment (TP) and control mesophilic pre-treatment (MP) reactors were operated as first-stages of TPAD systems treating activated sludge for 340 days. The TP system was operated sequentially at 50, 60 and 65 °C, while the MP rector was held at 35 °C for the entire period. The composition of microbial communities associated with the MP and TP pre-treatment reactors was characterised weekly using terminal-restriction fragment length polymorphism (T-RFLP) supported by clone library sequencing of 16S rRNA gene amplicons. The outcomes of this approach were confirmed using 454 pyrosequencing of gene amplicons and fluorescence in-situ hybridisation (FISH). TP associated bacterial communities were dominated by populations affiliated to the Firmicutes, Thermotogae, Proteobacteria and Chloroflexi. In particular there was a progression from Thermotogae to Lutispora and Coprothermobacter and diversity decreased as temperature and hydrolysis performance increased. While change in the composition of TP associated bacterial communities was attributable to temperature, that of MP associated bacterial communities was related to the composition of the incoming feed. This study determined processes driving the dynamics of key microbial populations that are correlated with an enhanced hydrolytic functionality of the TPAD system. Copyright © 2013 Elsevier Ltd. All rights reserved.

Performance of a catalytic reactor at simulated gas turbine combustor operating conditions

NASA Technical Reports Server (NTRS)

Anderson, D. N.; Tacina, R. R.; Mroz, T. S.

1975-01-01

The performance of a catalytic reactor 12 cm in diameter and 17 cm long was evaluated at simulated gas turbine combustor operating conditions using premixed propane and air. Inlet temperatures of 600 and 800 K, pressures of 3 and 6 atm, and reference velocities of 9 to 30 m/s were tested. Data were taken for equivalence ratios as high as 0.43. The operating range was limited on the low-temperature side by very poor efficiency; the minimum exit temperature for good performance ranged from 1400 to 1600 K depending on inlet conditions. As exit temperatures were raised above this minimum, emissions of unburned hydrocarbons decreased, carbon monoxide emissions became generally less than 1 g CO/kg fuel, and nitrogen oxides were less than about 0.1 g NO2/kg fuel.

Plasma reactor waste management systems

NASA Technical Reports Server (NTRS)

Ness, Robert O., Jr.; Rindt, John R.; Ness, Sumitra R.

1992-01-01

The University of North Dakota is developing a plasma reactor system for use in closed-loop processing that includes biological, materials, manufacturing, and waste processing. Direct-current, high-frequency, or microwave discharges will be used to produce plasmas for the treatment of materials. The plasma reactors offer several advantages over other systems, including low operating temperatures, low operating pressures, mechanical simplicity, and relatively safe operation. Human fecal material, sunflowers, oats, soybeans, and plastic were oxidized in a batch plasma reactor. Over 98 percent of the organic material was converted to gaseous products. The solids were then analyzed and a large amount of water and acid-soluble materials were detected. These materials could possibly be used as nutrients for biological systems.

Dimensionless Numbers Expressed in Terms of Common CVD Process Parameters

NASA Technical Reports Server (NTRS)

Kuczmarski, Maria A.

1999-01-01

A variety of dimensionless numbers related to momentum and heat transfer are useful in Chemical Vapor Deposition (CVD) analysis. These numbers are not traditionally calculated by directly using reactor operating parameters, such as temperature and pressure. In this paper, these numbers have been expressed in a form that explicitly shows their dependence upon the carrier gas, reactor geometry, and reactor operation conditions. These expressions were derived for both monatomic and diatomic gases using estimation techniques for viscosity, thermal conductivity, and heat capacity. Values calculated from these expressions compared well to previously published values. These expressions provide a relatively quick method for predicting changes in the flow patterns resulting from changes in the reactor operating conditions.

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

The effect of catalyst length and downstream reactor distance on catalytic combustor performance

NASA Technical Reports Server (NTRS)

Anderson, D.

1980-01-01

A study was made to determine the effects on catalytic combustor performance which resulted from independently varying the length of a catalytic reactor and the length available for gas-phase reactions downstream of the catalyst. Monolithic combustion catalysts from three manufacturers were tested in a combustion test rig with no. 2 diesel fuel. Catalytic reactor lengths of 2.5 and 5.4 cm, and downstream gas-phase reaction distances of 7.3, 12.4, 17.5, and 22.5 cm were evaluated. Measurements of carbon monoxide, unburned hydrocarbons, nitrogen oxides, and pressure drop were made. The catalytic-reactor pressure drop was less than 1 percent of the upstream total pressure for all test configurations and test conditions. Nitrogen oxides and unburned hydrocarbons emissions were less than 0.25 g NO2/kg fuel and 0.6 g HC/kg fuel, respectively. The minimum operating temperature (defined as the adiabatic combustion temperature required to obtain carbon monoxide emissions below a reference level of 13.6 g CO/kg fuel) ranged from 1230 K to 1500 K for the various conditions and configurations tested. The minimum operating temperature decreased with increasing total (catalytic-reactor-plus-downstream-gas-phase-reactor-zone) residence time but was independent of the relative times spent in each region when the catalytic-reactor residence time was greater than or equal to 1.4 ms.

Design and manufacture of a D-shape coil-based toroid-type HTS DC reactor using 2nd generation HTS wire

NASA Astrophysics Data System (ADS)

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

2014-09-01

This paper describes the design specifications and performance of a real toroid-type high temperature superconducting (HTS) DC reactor. The HTS DC reactor was designed using 2G HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The target inductance of the HTS DC reactor was 400 mH. The expected operating temperature was under 20 K. 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. Performances of the toroid-type HTS DC reactor were analyzed through experiments conducted under the steady-state and charge conditions. The fundamental design specifications and the data obtained from this research will be applied to the design of a commercial-type HTS DC reactor.

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.

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

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.

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.

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.

PROCESS FOR COOLING A NUCLEAR REACTOR

DOEpatents

Borst, L.B.

1962-12-11

This patent relates to the operation of a reactor cooled by liquid sulfur dioxide. According to the invention the pressure on the sulfur dioxide in the reactor is maintained at least at the critical pressure of the sulfur dioxide. Heating the sulfur dioxide to its critical temperature results in vaporization of the sulfur dioxide without boiling. (AEC)

Nuclear breeder reactor fuel element with axial tandem stacking and getter

DOEpatents

Gibby, Ronald L.; Lawrence, Leo A.; Woodley, Robert E.; Wilson, Charles N.; Weber, Edward T.; Johnson, Carl E.

1981-01-01

A breeder reactor fuel element having a tandem arrangement of fissile and fertile fuel with a getter for fission product cesium disposed between the fissile and fertile sections. The getter is effective at reactor operating temperatures to isolate the cesium generated by the fissile material from reacting with the fertile fuel section.

Rapid starting methanol reactor system

DOEpatents

Chludzinski, Paul J.; Dantowitz, Philip; McElroy, James F.

1984-01-01

The invention relates to a methanol-to-hydrogen cracking reactor for use with a fuel cell vehicular power plant. The system is particularly designed for rapid start-up of the catalytic methanol cracking reactor after an extended shut-down period, i.e., after the vehicular fuel cell power plant has been inoperative overnight. Rapid system start-up is accomplished by a combination of direct and indirect heating of the cracking catalyst. Initially, liquid methanol is burned with a stoichiometric or slightly lean air mixture in the combustion chamber of the reactor assembly. The hot combustion gas travels down a flue gas chamber in heat exchange relationship with the catalytic cracking chamber transferring heat across the catalyst chamber wall to heat the catalyst indirectly. The combustion gas is then diverted back through the catalyst bed to heat the catalyst pellets directly. When the cracking reactor temperature reaches operating temperature, methanol combustion is stopped and a hot gas valve is switched to route the flue gas overboard, with methanol being fed directly to the catalytic cracking reactor. Thereafter, the burner operates on excess hydrogen from the fuel cells.

A novel plant protection strategy for transient reactors

NASA Astrophysics Data System (ADS)

Bhattacharyya, Samit K.; Lipinski, Walter C.; Hanan, Nelson A.

The present plant protection system (PPS) has been defined for use in the TREAT-upgrade (TU) reactor for controlled transient operation of reactor-fuel behavior testing under simulated reactor-accident conditions. A PPS with energy-dependent trip set points lowered worst-case clad temperatures by as much as 180 K, relative to the use of conventional fixed-level trip set points. The multilayered multilevel protection strategy represents the state-of-the-art in terrestrial transient reactor protection systems, and should be applicable to multi-MW space reactors.

Flow instability in particle-bed nuclear reactors

NASA Technical Reports Server (NTRS)

Kerrebrock, J. L.; Kalamas, J.

1993-01-01

A three-dimensional model of the stability of the particle-bed reactor is presented, in which the fluid has mobility in three dimensions. The model accurately represents the stability at low Re numbers as well as the effects of the cold and hot frits and of the heat conduction and radiation in the particle bed. The model can be easily extended to apply to the cylindrical geometry of particle-bed reactors. Exemplary calculations are carried out, showing that a particle bed without a cold frit would be subject to instability if operated at the high-temperature ratios used for nuclear rockets and at power densities below about 4 MW/l; since the desired power density for such a reactor is about 40 MW/l, the operation at design exit temperature but at reduced power could be hazardous. Calculations show however that it might be possible to remove the instability problem by appropriate combinations of cold and hot frits.

Low-temperature overpressurization protection system setpoint analysis using RETRAN-02/MOD5 for Salem

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

Dodson, R.J.; Feltus, M.A.

The low-temperature overpressurization protection system (LTOPS) is designed to protect the reactor pressure vessel (RPV) from brittle failure during startup and cooldown maneuvers in Westinghouse pressurized water reactors. For the Salem power plants, the power-operated relief valves (PORVs) mitigate pressure increases above a setpoint where an operational startup transient may put the RPV in the embrittlement fracture zone. The Title 10, Part 50, Code of Federal Regulations Appendix G limit, given by plant technical specifications, conservatively bounds the maximum pressure allowed during those transients where the RPV can suffer brittle fracture (usually below 350{degrees}F). The Appendix G limit is amore » pressure versus temperature curve that is more restrictive at lower RPV temperatures and allows for higher pressures as the temperature approaches the upper bounding fracture temperature.« less

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.

Investigation on discharge characteristics of a coaxial dielectric barrier discharge reactor driven by AC and ns power sources

NASA Astrophysics Data System (ADS)

Qian, WANG; Feng, LIU; Chuanrun, MIAO; Bing, YAN; Zhi, FANG

2018-03-01

A coaxial dielectric barrier discharge (DBD) reactor with double layer dielectric barriers has been developed for exhaust gas treatment and excited either by AC power or nanosecond (ns) pulse to generate atmospheric pressure plasma. The comparative study on the discharge characteristics of the discharge uniformity, power deposition, energy efficiency, and operation temperature between AC and ns pulsed coaxial DBD is carried out in terms of optical and electrical characteristics and operation temperature for optimizing the coaxial DBD reactor performance. The voltages across the air gap and dielectric layer and the conduction and displacement currents are extracted from the applied voltages and measured currents of AC and ns pulsed coaxial DBDs for the calculation of the power depositions and energy efficiencies through an equivalent electrical model. The discharge uniformity and operating temperature of the coaxial DBD reactor are monitored and analyzed by optical images and infrared camera. A heat conduction model is used to calculate the temperature of the internal quartz tube. It is found that the ns pulsed coaxial DBD has a much higher instantaneous power deposition in plasma, a lower total power consumption, and a higher energy efficiency compared with that excited by AC power and is more homogeneous and stable. The temperature of the outside wall of the AC and ns pulse excited coaxial DBD reaches 158 °C and 64.3 °C after 900 s operation, respectively. The experimental results on the comparison of the discharge characteristics of coaxial DBDs excited by different powers are significant for understanding of the mechanism of DBDs, reducing energy loss, and optimizing the performance of coaxial DBD in industrial applications.

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.

Thermodynamic Simulation of Equilibrium Composition of Reaction Products at Dehydration of a Technological Channel in a Uranium-Graphite Reactor

NASA Astrophysics Data System (ADS)

Pavliuk, A. O.; Zagumennov, V. S.; Kotlyarevskiy, S. G.; Bespala, E. V.

2018-01-01

The problems of accumulation of nuclear fuel spills in the graphite stack in the course of operation of uranium-graphite nuclear reactors are considered. The results of thermodynamic analysis of the processes in the graphite stack at dehydration of a technological channel, fuel element shell unsealing and migration of fission products, and activation of stable nuclides in structural elements of the reactor and actinides inside the graphite moderator are given. The main chemical reactions and compounds that are produced in these modes in the reactor channel during its operation and that may be hazardous after its shutdown and decommissioning are presented. Thermodynamic simulation of the equilibrium composition is performed using the specialized code TERRA. The results of thermodynamic simulation of the equilibrium composition in different cases of technological channel dehydration in the course of the reactor operation show that, if the temperature inside the active core of the nuclear reactor increases to the melting temperature of the fuel element, oxides and carbides of nuclear fuel are produced. The mathematical model of the nonstationary heat transfer in a graphite stack of a uranium-graphite reactor in the case of the technological channel dehydration is presented. The results of calculated temperature evolution at the center of the fuel element, the replaceable graphite element, the air gap, and in the surface layer of the block graphite are given. The numerical results show that, in the case of dehydration of the technological channel in the uranium-graphite reactor with metallic uranium, the main reaction product is uranium dioxide UO2 in the condensed phase. Low probability of production of pyrophoric uranium compounds (UH3) in the graphite stack is proven, which allows one to disassemble the graphite stack without the risk of spontaneous graphite ignition in the course of decommissioning of the uranium-graphite nuclear reactor.

Numerical study of radiative heat transfer and effects of thermal boundary conditions on CLC fuel reactor

NASA Astrophysics Data System (ADS)

Ben-Mansour, R.; Li, H.; Habib, M. A.; Hossain, M. M.

2018-02-01

Global warming has become a worldwide concern due to its severe impacts and consequences on the climate system and ecosystem. As a promising technology proving good carbon capture ability with low-efficiency penalty, Chemical Looping Combustion technology has risen much interest. However, the radiative heat transfer was hardly studied, nor its effects were clearly declared. The present work provides a mathematical model for radiative heat transfer within fuel reactor of chemical looping combustion systems and conducts a numerical research on the effects of boundary conditions, solid particles reflectivity, particles size, and the operating temperature. The results indicate that radiative heat transfer has very limited impacts on the flow pattern. Meanwhile, the temperature variations in the static bed region (where solid particles are dense) brought by radiation are also insignificant. However, the effects of radiation on temperature profiles within free bed region (where solid particles are very sparse) are obvious, especially when convective-radiative (mixed) boundary condition is applied on fuel reactor walls. Smaller oxygen carrier particle size results in larger absorption & scattering coefficients. The consideration of radiative heat transfer within fuel reactor increases the temperature gradient within free bed region. On the other hand, the conversion performance of fuel is nearly not affected by radiation heat transfer within fuel reactor. However, the consideration of radiative heat transfer enhances the heat transfer between the gas phase and solid phase, especially when the operating temperature is low.

Liquid metal hydrogen barriers

DOEpatents

Grover, George M.; Frank, Thurman G.; Keddy, Edward S.

1976-01-01

Hydrogen barriers which comprise liquid metals in which the solubility of hydrogen is low and which have good thermal conductivities at operating temperatures of interest. Such barriers are useful in nuclear fuel elements containing a metal hydride moderator which has a substantial hydrogen dissociation pressure at reactor operating temperatures.

Simultaneous saccharification and extractive fermentation of lignocellulosic materials into lactic acid in a two-zone fermentor-extractor system.

PubMed

Iyer, P V; Lee, Y Y

1999-01-01

Simultaneous saccharification and extractive fermentation of lignocellulosic materials into lactic acid was investigated using a two-zone bioreactor. The system is composed of an immobilized cell reactor, a separate column reactor containing the lignocellulosic substrate and a hollow-fiber membrane. It is operated by recirculating the cell free enzyme (cellulase) solution from the immobilized cell reactor to the column reactor through the membrane. The enzyme and microbial reactions thus occur at separate locations, yet simultaneously. This design provides flexibility in reactor operation as it allows easy separation of the solid substrate from the microorganism, in situ removal of the product and, if desired, different temperatures in the two reactor sections. This reactor system was tested using pretreated switchgrass as the substrate. It was operated under a fed-batch mode with continuous removal of lactic acid by solvent extraction. The overall lactic acid yield obtainable from this bioreactor system is 77% of the theoretical.

Emissivity of Candidate Materials for VHTR Applicationbs: Role of Oxidation and Surface Modification Treatments

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

Sridharan, Kumar; Allen, Todd; Anderson, Mark

The Generation IV (GEN IV) Nuclear Energy Systems Initiative was instituted by the Department of Energy (DOE) with the goal of researching and developing technologies and materials necessary for various types of future reactors. These GEN IV reactors will employ advanced fuel cycles, passive safety systems, and other innovative systems, leading to significant differences between these future reactors and current water-cooled reactors. The leading candidate for the Next Generation Nuclear Plant (NGNP) to be built at Idaho National Lab (INL) in the United States is the Very High Temperature Reactor (VHTR). Due to the high operating temperatures of the VHTR,more » the Reactor Pressure Vessel (RPV) will partially rely on heat transfer by radiation for cooling. Heat expulsion by radiation will become all the more important during high temperature excursions during off-normal accident scenarios. Radiant power is dictated by emissivity, a material property. The NGNP Materials Research and Development Program Plan [1] has identified emissivity and the effects of high temperature oxide formation on emissivity as an area of research towards the development of the VHTR.« less

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W. Y.

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc.; a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR); NERVA-derivative; and other concepts are discussed. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggest that full-scale PBR elements could be tested at an average energy deposition of approximately 60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of approximately 100 MW/L may be achievable.

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.

Small reactor power system for space application

NASA Technical Reports Server (NTRS)

Shirbacheh, M.

1987-01-01

A development history and comparative performance capability evaluation is presented for spacecraft nuclear powerplant Small Reactor Power System alternatives. The choice of power conversion technology depends on the reactor's operating temperature; thermionic, thermoelectric, organic Rankine, and Alkali metal thermoelectric conversion are the primary power conversion subsystem technology alternatives. A tabulation is presented for such spacecraft nuclear reactor test histories as those of SNAP-10A, SP-100, and NERVA.

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.

Evaluation of an Integrated Gas-Cooled Reactor Simulator and Brayton Turbine-Generator

NASA Technical Reports Server (NTRS)

Hissam, David Andy; Stewart, Eric T.

2006-01-01

A closed-loop brayton cycle, powered by a fission reactor, offers an attractive option for generating both planetary and in-space electric power. Non-nuclear testing of this type of system provides the opportunity to safely work out integration and system control challenges for a modest investment. Recognizing this potential, a team at Marshall Space Flight Center has evaluated the viability of integrating and testing an existing gas-cooled reactor simulator and a modified commercially available, off-the-shelf, brayton turbine-generator. Since these two systems were developed independently of one another, this evaluation had to determine if they could operate together at acceptable power levels, temperatures, and pressures. Thermal, fluid, and structural analyses show that this combined system can operate at acceptable power levels and temperatures. In addition, pressure drops across the reactor simulator, although higher than desired, are also viewed as acceptable. Three potential working fluids for the system were evaluated: N2, He/Ar, and He/Xe. Other potential issues, such as electrical breakdown in the generator and the operation of the brayton foil bearings using various gas mixtures, were also investigated.

Technical note on results of production test IP-628-AI: Crosstie flow test and flushing calibration, and the earlier production tests IP-573-AI and IP-499-AI

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

Jones, S.S.

1964-06-15

A series of three experimental tests was conducted on the operation and adequacy of the K-Reactors` third, or last-ditch, cooling system. The first test showed considerable line corrosion the second test was performed directly after line cleaning, and the third test showed a significant amount of additional line corrosion after only nine months of service. The present cooling adequacy of this last-ditch system at the KE and KW reactors is summarized these show the power levels for which we have adequate last-ditch cooling as a function of the crosstie coolant temperature. These figures include the effects of increasing the numbermore » of pumps that remain in operation at the other K-Reactor, and various other operating or emergency conditions. These curves are for diesel pump speeds of 750 rpm which are planned for June of this year. In these figures the crosstie temperature is assumed to be a conservative 5{degrees}C above the process water inlet temperature.« less

Cluster formation in in-service thermally aged pressurizer welds

NASA Astrophysics Data System (ADS)

Lindgren, Kristina; Boåsen, Magnus; Stiller, Krystyna; Efsing, Pål; Thuvander, Mattias

2018-06-01

Thermal aging of reactor pressure vessel steel welds at elevated temperatures may affect the ductile-to-brittle transition temperature. In this study, unique weld material from a pressurizer, with a composition similar to that of the reactor pressure vessel, that has been in operation for 28 years at 345 °C is examined. Despite the relatively low temperature, the weld becomes hardened during operation. This is attributed to nanometre sized Cu-rich clusters, mainly located at Mo- and C-enriched dislocation lines and on boundaries. The welds have been characterized using atom probe tomography, and the characteristics of the precipitates/clusters is related to the hardness increase, giving the best agreement for the Russell-Brown model.

Outdoor cultivation of temperature-tolerant Chlorella sorokiniana in a column photobioreactor under low power-input.

PubMed

Béchet, Quentin; Muñoz, Raul; Shilton, Andy; Guieysse, Benoit

2013-01-01

Temperature-tolerant Chlorella sorokiniana was cultivated in a 51-L column photobioreactor with a 1.1 m(2) illuminated area. The reactor was operated outdoors under tropical meteorological conditions (Singapore) without controlling temperature and the culture was mixed at a power input of 7.5 W/m(3) by sparging CO(2) -enriched air at 1.2 L/min (gas hold-up of 0.02). Biomass productivity averaged 10 ± 2.2 g/m(2) -day over six batch studies, yielding an average photosynthetic efficiency (PE) of 4.8 ± 0.5% of the total solar radiation (P = 0.05, N = 6). This demonstrates that temperature-tolerant microalgae can be cultivated at high PE under a mixing input sevenfold to ninefold lower than current operational guidelines (50-70 W/m(3)) and without the need for temperature control (the culture broth temperature reached 41 °C during operation). In this study, the PE value was determined based on the amount of solar radiation actually reaching the algae and this amount was estimated using a mathematical model fed with onsite solar irradiance data. This determination was found to be particularly sensitive to the value of the atmospheric diffusion coefficient, which generated a significant uncertainty in the PE calculation. The use of the mathematical model, however, confirmed that the vertical reactor geometry supported efficient photosynthesis by reducing the duration and intensity of photoinhibition events. The model also revealed that all three components of direct, diffuse, and reflected solar radiation were quantitatively important for the vertical column photobioreactor, accounting for 14%, 65%, and 21% of the total solar radiation reaching the culture, respectively. The accurate prediction of the discrete components of solar radiation reaching the algae as a function of climatic, geographic, and design parameters is therefore crucial to optimize the individual reactor geometry and the layout/spacing between the individual reactors in a reactor farm. Copyright © 2012 Wiley Periodicals, Inc.

Optimization of a horizontal-flow biofilm reactor for the removal of methane at low temperatures.

PubMed

Clifford, E; Kennelly, C; Walsh, R; Gerrity, S; Reilly, E O; Collins, G

2012-10-01

Three pilot-scale, horizontal-flow biofilm reactors (HFBRs 1-3) were used to treat methane (CH4)-contaminated air to assess the potential of this technology to manage emissions from agricultural activities, waste and wastewater treatment facilities, and landfills. The study was conducted over two phases (Phase 1, lasting 90 days and Phase 2, lasting 45 days). The reactors were operated at 10 degrees C (typical of ambient air and wastewater temperatures in northern Europe), and were simultaneously dosed with CH4-contaminated air and a synthetic wastewater (SWW). The influent loading rates to the reactors were 8.6 g CH4/m3/hr (4.3 g CH4/m2 TPSA/hr; where TPSA is top plan surface area). Despite the low operating temperatures, an overall average removal of 4.63 g CH4/m3/day was observed during Phase 2. The maximum removal efficiency (RE) for the trial was 88%. Potential (maximum) rates of methane oxidation were measured and indicated that biofilm samples taken from various regions in the HFBRs had mostly equal CH4 removal potential. In situ activity rates were dependent on which part of the reactor samples were obtained. The results indicate the potential of the HFBR, a simple and robust technology, to biologically treat CH4 emissions. The results of this study indicate that the HFBR technology could be effectively applied to the reduction of greenhouse gas emissions from wastewater treatment plants and agricultural facilities at lower temperatures common to northern Europe. This could reduce the carbon footprint of waste treatment and agricultural livestock facilities. Activity tests indicate that methanotrophic communities can be supported at these temperatures. Furthermore, these data can lead to improved reactor design and optimization by allowing conditions to be engineered to allow for improved removal rates, particularly at lower temperatures. The technology is simple to construct and operate, and with some optimization of the liquid phase to improve mass transfer, the HFBR represents a viable, cost-effective solution for these emissions.

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

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.

METHOD AND APPARATUS FOR EARTH PENETRATION

DOEpatents

Adams, W.M.

1963-12-24

A nuclear reactor apparatus for penetrating into the earth's crust is described. The apparatus comprises a cylindrical nuclear core operating at a temperature that is higher than the melting temperature of rock. A high-density ballast member is coupled to the nuclear core such that the overall density of the core-ballast assembly is greater than the density of molten rock. The nuclear core is thermally insulated so that its heat output is constrained to flow axially, with radial heat flow being minimized. In operation, the apparatus is placed in contact with the earth's crust at the point desired to be penetrated. The heat output of the reactor melts the underlying rock, and the apparatus sinks through the resulting magma. The fuel loading of the reactor core determines the ultimate depth of crust penetration. (AEC)

Etherification process

DOEpatents

Smith, Jr., Lawrence A.; Hearn, Dennis; Jones, Jr., Edward M.

1990-01-01

A liquid phase process for oligomerization of C.sub.4 and C.sub.5 isoolefins or the etherification thereof with C.sub.1 to C.sub.6 alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120.degree. to 300.degree. F. wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

Oligomerization process

DOEpatents

Smith, Jr., Lawrence A.; hearn, Dennis; Jones, Jr., Edward M.

1991-01-01

A liquid phase process for oligomerization of C.sub.4 and C.sub.5 isoolefins or the etherification thereof with C.sub.1 to C.sub.6 alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300.degree. F. wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

Effect of temperature and temperature fluctuation on thermophilic anaerobic digestion of cattle manure.

PubMed

El-Mashad, Hamed M; Zeeman, Grietje; van Loon, Wilko K P; Bot, Gerard P A; Lettinga, Gatze

2004-11-01

The influence of temperature, 50 and 60 degrees C, at hydraulic retention times (HRTs) of 20 and 10 days, on the performance of anaerobic digestion of cow manure has been investigated in completely stirred tank reactors (CSTRs). Furthermore, the effect of both daily downward and daily upward temperature fluctuations has been studied. In the daily downward temperature fluctuation regime the temperatures of each reactor was reduced by 10 degrees C for 10 h while in the daily upward fluctuation regime the temperature of each reactor was increased 10 degrees C for 5 h. The results show that the methane production rate at 60 degrees C is lower than that at 50 degrees C at all experimental conditions of imposed HRT except when downward temperature fluctuations were applied at an HRT of 10 days. It also was found that the free ammonia concentration not only affects the acetate-utilising bacteria but also the hydrolysis and acidification process. The upward temperature fluctuation affects the maximum specific methanogenesis activity more severely as compared to imposed downward temperature fluctuations. The results clearly reveal the possibility of using available solar energy at daytime to heat up the reactor(s) without the need of heat storage during nights, especially at an operational temperature of 50 degrees C and at a 20 days HRT, and without the jeopardising of the overheating.

Non-equilibrium radiation nuclear reactor

NASA Technical Reports Server (NTRS)

Thom, K.; Schneider, R. T. (Inventor)

1978-01-01

An externally moderated thermal nuclear reactor is disclosed which is designed to provide output power in the form of electromagnetic radiation. The reactor is a gaseous fueled nuclear cavity reactor device which can operate over wide ranges of temperature and pressure, and which includes the capability of processing and recycling waste products such as long-lived transuranium actinides. The primary output of the device may be in the form of coherent radiation, so that the reactor may be utilized as a self-critical nuclear pumped laser.

Fission-powered in-core thermoacoustic sensor

DOE PAGES

Garrett, Steven L.; Smith, James A.; Smith, Robert W. M.; ...

2016-04-07

A thermoacoustic engine is operated within the core of a nuclear reactor to acoustically telemeter coolant temperature (frequency-encoded) and reactor power level (amplitude-encoded) outside the reactor, thus providing the values of these important parameters without external electrical power or wiring. We present data from two hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. Furthermore, these signals have been detected even in the presence of substantial background noise generated by the reactor's fluid pumps.

Fission-powered in-core thermoacoustic sensor

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

Garrett, Steven L.; Smith, James A.; Smith, Robert W. M.

2016-04-04

A thermoacoustic engine is operated within the core of a nuclear reactor to acoustically telemeter coolant temperature (frequency-encoded) and reactor power level (amplitude-encoded) outside the reactor, thus providing the values of these important parameters without external electrical power or wiring. We present data from two hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. These signals have been detected even in the presence of substantial background noise generated by the reactor's fluid pumps.

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.

Next generation fuel irradiation capability in the High Flux Reactor Petten

NASA Astrophysics Data System (ADS)

Fütterer, Michael A.; D'Agata, Elio; Laurie, Mathias; Marmier, Alain; Scaffidi-Argentina, Francesco; Raison, Philippe; Bakker, Klaas; de Groot, Sander; Klaassen, Frodo

2009-07-01

This paper describes selected equipment and expertise on fuel irradiation testing at the High Flux Reactor (HFR) in Petten, The Netherlands. The reactor went critical in 1961 and holds an operating license up to at least 2015. While HFR has initially focused on Light Water Reactor fuel and materials, it also played a decisive role since the 1970s in the German High Temperature Reactor (HTR) development program. A variety of tests related to fast reactor development in Europe were carried out for next generation fuel and materials, in particular for Very High Temperature Reactor (V/HTR) fuel, fuel for closed fuel cycles (U-Pu and Th-U fuel cycle) and transmutation, as well as for other innovative fuel types. The HFR constitutes a significant European infrastructure tool for the development of next generation reactors. Experimental facilities addressed include V/HTR fuel tests, a coated particle irradiation rig, and tests on fast reactor, transmutation and thorium fuel. The rationales for these tests are given, results are provided and further work is outlined.

Passive heat transfer means for nuclear reactors

DOEpatents

Burelbach, James P.

1984-01-01

An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. Means such as shrouding normally isolated the secondary condensing section from effective heat transfer with the heat sink, but a sensor responds to overheat conditions of the reactor to open the shrouding, which thereby increases the cooling capacity of the heat pipe. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

Space reactor fuel element testing in upgraded TREAT

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

Todosow, M.; Bezler, P.; Ludewig, H.

1993-01-14

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. Ifmore » the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.« less

Space reactor fuel element testing in upgraded TREAT

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

Todosow, M.; Bezler, P.; Ludewig, H.

1993-05-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. Ifmore » the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.« less

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.

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.

Real-time LMR control parameter generation using advanced adaptive synthesis

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

King, R.W.; Mott, J.E.

1990-01-01

The reactor delta T'', the difference between the average core inlet and outlet temperatures, for the liquid-sodium-cooled Experimental Breeder Reactor 2 is empirically synthesized in real time from, a multitude of examples of past reactor operation. The real-time empirical synthesis is based on reactor operation. The real-time empirical synthesis is based on system state analysis (SSA) technology embodied in software on the EBR 2 data acquisition computer. Before the real-time system is put into operation, a selection of reactor plant measurements is made which is predictable over long periods encompassing plant shutdowns, core reconfigurations, core load changes, and plant startups.more » A serial data link to a personal computer containing SSA software allows the rapid verification of the predictability of these plant measurements via graphical means. After the selection is made, the real-time synthesis provides a fault-tolerant estimate of the reactor delta T accurate to {plus}/{minus}1{percent}. 5 refs., 7 figs.« less

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

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.

BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

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

Sweet, Ryan; George, Nathan M.; Terrani, Kurt A.

2016-08-30

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling themore » integral thermo-mechanical performance of FeCrAl-cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl, and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and operating conditions used are based off the Peach Bottom BWR and design consideration was given to minimize the neutronic penalty of the FeCrAl cladding by changing fuel enrichment and cladding thickness. As this study progressed, systematic parametric analysis of the fuel and cladding creep responses were also performed.« less

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.

Dynamics of heat-pipe reactors

NASA Technical Reports Server (NTRS)

Niederauer, G. F.

1971-01-01

A split-core heat pipe reactor, fueled with either U(233)C or U(235)C in a tungsten cermet and cooled by 7-Li-W heat pipes, was examined for the effects of the heat pipes on reactor while trying to safely absorb large reactivity inputs through inherent shutdown mechanisms. Limits on ramp reactivity inputs due to fuel melting temperature and heat pipe wall heat flux were mapped for the reactor in both startup and at-power operating modes.

Low temperature MBBR nitrification: Microbiome analysis.

PubMed

Young, Bradley; Delatolla, Robert; Kennedy, Kevin; Laflamme, Edith; Stintzi, Alain

2017-03-15

This study aims to investigate post carbon removal moving bed biofilm reactor (MBBR) nitrification through the transition from 20 °C to 1 °C and during through long term operation at 1 °C. Four pilot nitrifying MBBR reactors were operated at various ammonia loading rates to elucidate the temperature effects on ammonia removal rates, cell viability and bacterial communities. The transition from 20 °C to 1 °C and during long term operation at 1 °C were modeled using Arrhenius temperature correction coefficients. Specifically, the steady state removal rates at 1 °C on average were 22.8% of the maximum ammonia removal rate at 20 °C, which corresponds to an Arrhenius temperature correction of 1.086 during steady operation at 1 °C. The microbial communities of the nitrifying MBBR biofilm were shown to be significantly more diverse at 20 °C as compared to 1 °C operation. Although less diverse at 1 °C, 2000 species of bacteria were identified in the nitrifying biofilm during operation at this low temperature. Nitrosomonads were shown to be the dominant ammonia oxidizing bacteria (AOB) and Nitrospira was shown to be the dominant nitrite oxidizing bacteria (NOB) in all the pilot MBBR reactors at all temperatures. The performance of the post carbon removal nitrifying MBBR systems were shown to be enhanced at 1 °C by an increase in the viable embedded biomass as well as thicker biofilm. This effectively increases the number of viable cell present during low temperature operation, which partially compensates for the significant decrease in rate of ammonia removal per nitrifying cell. Operation at the highest loading conditions tested in this study at 1 °C were shown to reduce the ammonia removal rate compared to lower loading conditions at 1 °C. The lower performance at higher loading conditions at 1 °C demonstrated an enrichment in the stress response metagenomics pathways of the system. Copyright © 2017 Elsevier Ltd. All rights reserved.

Analysis on the Role of RSG-GAS Pool Cooling System during Partial Loss of Heat Sink Accident

NASA Astrophysics Data System (ADS)

Susyadi; Endiah, P. H.; Sukmanto, D.; Andi, S. E.; Syaiful, B.; Hendro, T.; Geni, R. S.

2018-02-01

RSG-GAS is a 30 MW reactor that is mostly used for radioisotope production and experimental activities. Recently, it is regularly operated at half of its capacity for efficiency reason. During an accident, especially loss of heat sink, the role of its pool cooling system is very important to dump decay heat. An analysis using single failure approach and partial modeling of RELAP5 performed by S. Dibyo, 2010 shows that there is no significant increase in the coolant temperature if this system is properly functioned. However lessons learned from the Fukushima accident revealed that an accident can happen due to multiple failures. Considering ageing of the reactor, in this research the role of pool cooling system is to be investigated for a partial loss of heat sink accident which is at the same time the protection system fails to scram the reactor when being operated at 15 MW. The purpose is to clarify the transient characteristics and the final state of the coolant temperature. The method used is by simulating the system in RELAP5 code. Calculation results shows the pool cooling systems reduce coolant temperature for about 1 K as compared without activating them. The result alsoreveals that when the reactor is being operated at half of its rated power, it is still in safe condition for a partial loss of heat sink accident without scram.

Evaluation of Bosch-Based Systems Using Non-Traditional Catalysts at Reduced Temperatures

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, J. Matthew

2011-01-01

Oxygen and water resupply make open loop atmosphere revitalization (AR) systems unfavorable for long-term missions beyond low Earth orbit. Crucial to closing the AR loop are carbon dioxide reduction systems with low mass and volume, minimal power requirements, and minimal consumables. For this purpose, NASA is exploring using Bosch-based systems. The Bosch process is favorable over state-of-the-art Sabatier-based processes due to complete loop closure. However, traditional operation of the Bosch required high reaction temperatures, high recycle rates, and significant consumables in the form of catalyst resupply due to carbon fouling. A number of configurations have been proposed for next-generation Bosch systems. First, alternative catalysts (catalysts other than steel wool) can be used in a traditional single-stage Bosch reactor to improve reaction kinetics and increase carbon packing density. Second, the Bosch reactor may be split into separate stages wherein the first reactor stage is dedicated to carbon monoxide and water formation via the reverse water-gas shift reaction and the second reactor stage is dedicated to carbon formation. A series system will enable maximum efficiency of both steps of the Bosch reaction, resulting in optimized operation and maximum carbon formation rate. This paper details the results of testing of both single-stage and two-stage Bosch systems with alternative catalysts at reduced temperatures. These results are compared to a traditional Bosch system operated with a steel wool catalyst.

Gasification reactor engineering approach to understanding the formation of biochar properties

PubMed Central

2016-01-01

The correlation between thermochemical provenance and biochar functionality is poorly understood. To this end, operational reactor temperatures (spanning the reduction zone), pressure and product gas composition measurements were obtained from a downdraft gasifier and compared against elemental composition, surface morphology and polyaromatic hydrocarbon content (PAH) of the char produced. Pine feedstock moisture with values of 7% and 17% was the experimental variable. Moderately high steady-state temperatures were observed inside the reactor, with a ca 50°C difference in how the gasifier operated between the two feedstock types. Both chars exhibited surface properties comparable to activated carbon, but the relatively small differences in temperature caused significant variations in biochar surface area and morphology: micropore area 584 against 360 m2 g−1, and micropore volume 0.287 against 0.172 cm3 g−1. Differences in char extractable PAH content were also observed, with higher concentrations (187 µg g−1 ± 18 compared with 89 ± 19 µg g−1 Σ16EPA PAH) when the gasifier was operated with higher moisture content feedstock. It is recommended that greater detail on operational conditions during biochar production should be incorporated to future biochar characterization research as a consequence of these results. PMID:27616911

Research and Engineering Operation, Irradiation Processing Department monthly record report, May 1965

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

Ambrose, T.W.

1965-06-04

Process and development activities reported include: depleted uranium irradiations, thoria irradiation, and hot die sizing. Reactor engineering activities include: brittle fracture of 190-C tanks, increased graphite temperature limits for the F reactor, VSR channel caulking, K reactor downcomer flow, zircaloy hydriding, and ribbed zircaloy process tubes. Reactor physics activities include: thoria irradiations, E-D irradiations, boiling protection with the high speed scanner, and in-core flux monitoring. Radiological engineering activities include: radiation control, classification, radiation occurrences, effluent activity data, and well car shielding. Process standards are listed, along with audits, and fuel failure experience. Operational physics and process physics studies are presented.more » Lastly, testing activities are detailed.« less

Etherification process

DOEpatents

Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

1990-08-21

A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled. 2 figs.

Method for conducting exothermic reactions

DOEpatents

Smith, L. Jr.; Hearn, D.; Jones, E.M. Jr.

1993-01-05

A liquid phase process for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

Method for conducting exothermic reactions

DOEpatents

Smith, Jr., Lawrence; Hearn, Dennis; Jones, Jr., Edward M.

1993-01-01

A liquid phase process for oligomerization of C.sub.4 and C.sub.5 isoolefins or the etherification thereof with C.sub.1 to C.sub.6 alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120.degree. to 300.degree. F. wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

Oligomerization process

DOEpatents

Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

1991-03-26

A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled. 2 figures.

SAFEGUARDS REPORT FOR THE NORTHROP PULSE RADIATION FACILITY

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

Feinauer, E.; Thomas, R.D.

1961-03-22

Ae description is given of the Northrop pulse Radiation Facility, (NPRF), which consists of a TRlGA Mark-F reactor and associated supporting equipment. The NPRF was designed to operate in the following modes: Mode 1-100 kw steady-state operation; Mode II--Pulsed operation up to a maximum transient giving a maximum measured fuel element temperature of 470 deg C, which corresponds to an energy release of about 18 Mw-sec (approximately 1.9% sigma K/ K). The movable reactor will be operated in three general areas in the pool: adjacent to the exposure room; adjacent to the beam ponts; or at intermediate positions. Based onmore » the analyses presented and operating experience with the prototype TRIGA Mark F and other TRlGA reactors, it is concluded that operation of the NPRF does not present any undue hazard to the health and safety of the operating personnel or the public. (auth)« less

Reactor Operations Monitoring System

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

Hart, M.M.

1989-01-01

The Reactor Operations Monitoring System (ROMS) is a VME based, parallel processor data acquisition and safety action system designed by the Equipment Engineering Section and Reactor Engineering Department of the Savannah River Site. The ROMS will be analyzing over 8 million signal samples per minute. Sixty-eight microprocessors are used in the ROMS in order to achieve a real-time data analysis. The ROMS is composed of multiple computer subsystems. Four redundant computer subsystems monitor 600 temperatures with 2400 thermocouples. Two computer subsystems share the monitoring of 600 reactor coolant flows. Additional computer subsystems are dedicated to monitoring 400 signals from assortedmore » process sensors. Data from these computer subsystems are transferred to two redundant process display computer subsystems which present process information to reactor operators and to reactor control computers. The ROMS is also designed to carry out safety functions based on its analysis of process data. The safety functions include initiating a reactor scram (shutdown), the injection of neutron poison, and the loadshed of selected equipment. A complete development Reactor Operations Monitoring System has been built. It is located in the Program Development Center at the Savannah River Site and is currently being used by the Reactor Engineering Department in software development. The Equipment Engineering Section is designing and fabricating the process interface hardware. Upon proof of hardware and design concept, orders will be placed for the final five systems located in the three reactor areas, the reactor training simulator, and the hardware maintenance center.« 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.

Operation of staged membrane oxidation reactor systems

DOEpatents

Repasky, John Michael

2012-10-16

A method of operating a multi-stage ion transport membrane oxidation system. The method comprises providing a multi-stage ion transport membrane oxidation system with at least a first membrane oxidation stage and a second membrane oxidation stage, operating the ion transport membrane oxidation system at operating conditions including a characteristic temperature of the first membrane oxidation stage and a characteristic temperature of the second membrane oxidation stage; and controlling the production capacity and/or the product quality by changing the characteristic temperature of the first membrane oxidation stage and/or changing the characteristic temperature of the second membrane oxidation stage.

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.

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

Pyrolysis of softwood carbohydrates in a fluidized bed reactor.

PubMed

Aho, Atte; Kumar, Narendra; Eränen, Kari; Holmbom, Bjarne; Hupa, Mikko; Salmi, Tapio; Murzin, Dmitry Yu

2008-09-01

In the present work pyrolysis of pure pine wood and softwood carbohydrates, namely cellulose and galactoglucomannan (the major hemicellulose in coniferous wood), was conducted in a batch mode operated fluidized bed reactor. Temperature ramping (5 degrees C/min) was applied to the heating until a reactor temperature of 460 degrees C was reached. Thereafter the temperature was kept until the release of non-condensable gases stopped. The different raw materials gave significantly different bio-oils. Levoglucosan was the dominant product in the cellulose pyrolysis oil. Acetic acid was found in the highest concentrations in both the galactoglucomannan and in the pine wood pyrolysis oils. Acetic acid is most likely formed by removal of O-acetyl groups from mannose units present in GGM structure.

Pyrolysis of Softwood Carbohydrates in a Fluidized Bed Reactor

PubMed Central

Aho, Atte; Kumar, Narendra; Eränen, Kari; Holmbom, Bjarne; Hupa, Mikko; Salmi, Tapio; Murzin, Dmitry Yu.

2008-01-01

In the present work pyrolysis of pure pine wood and softwood carbohydrates, namely cellulose and galactoglucomannan (the major hemicellulose in coniferous wood), was conducted in a batch mode operated fluidized bed reactor. Temperature ramping (5 °C/min) was applied to the heating until a reactor temperature of 460 °C was reached. Thereafter the temperature was kept until the release of non-condensable gases stopped. The different raw materials gave significantly different bio-oils. Levoglucosan was the dominant product in the cellulose pyrolysis oil. Acetic acid was found in the highest concentrations in both the galactoglucomannan and in the pine wood pyrolysis oils. Acetic acid is most likely formed by removal of O-acetyl groups from mannose units present in GGM structure. PMID:19325824

Analysis of Water Volume Changes and Temperature Measurement Location Effect to the Accuracy of RTP Power Calibration

NASA Astrophysics Data System (ADS)

Lanyau, T.; Hamzah, N. S.; Jalal Bayar, A. M.; Karim, J. Abdul; Phongsakorn, P. K.; Suhaimi, K. Mohammad; Hashim, Z.; Razi, H. Md; Fazli, Z. Mohd; Ligam, A. S.; Mustafa, M. K. A.

2018-01-01

Power calibration is one of the important aspect for safe operation of the reactor. In RTP, the calorimetric method has been applied in reactor power calibration. This method involves measurement of water temperature in the RTP tank. Water volume and location of the temperature measurement may play an important role to the accuracy of the measurement. In this study, the analysis of water volume changes and thermocouple location effect to the power calibration accuracy has been done. The changes of the water volume are controlled by the variation of water level in reactor tank. The water level is measured by the ultrasonic measurement device. Temperature measurement has been done by thermocouple placed at three different locations. The accuracy of the temperature trend from various condition of measurement has been determined and discussed in this paper.

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

Impact of Reactor Operating Parameters on Cask Reactivity in BWR Burnup Credit

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

Ilas, Germina; Betzler, Benjamin R; Ade, Brian J

This paper discusses the effect of reactor operating parameters used in fuel depletion calculations on spent fuel cask reactivity, with relevance for boiling-water reactor (BWR) burnup credit (BUC) applications. Assessments that used generic BWR fuel assembly and spent fuel cask configurations are presented. The considered operating parameters, which were independently varied in the depletion simulations for the assembly, included fuel temperature, bypass water density, specific power, and operating history. Different operating history scenarios were considered for the assembly depletion to determine the effect of relative power distribution during the irradiation cycles, as well as the downtime between cycles. Depletion, decay,more » and criticality simulations were performed using computer codes and associated nuclear data within the SCALE code system. Results quantifying the dependence of cask reactivity on the assembly depletion parameters are presented herein.« less

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.

Numerical investigation of flow and heat transfer in a novel configuration multi-tubular fixed bed reactor for propylene to acrolein process

NASA Astrophysics Data System (ADS)

Jiang, Bin; Hao, Li; Zhang, Luhong; Sun, Yongli; Xiao, Xiaoming

2015-01-01

In the present contribution, a numerical study of fluid flow and heat transfer performance in a pilot-scale multi-tubular fixed bed reactor for propylene to acrolein oxidation reaction is presented using computational fluid dynamics (CFD) method. Firstly, a two-dimensional CFD model is developed to simulate flow behaviors, catalytic oxidation reaction, heat and mass transfer adopting porous medium model on tube side to achieve the temperature distribution and investigate the effect of operation parameters on hot spot temperature. Secondly, based on the conclusions of tube-side, a novel configuration multi-tubular fixed-bed reactor comprising 790 tubes design with disk-and-doughnut baffles is proposed by comparing with segmental baffles reactor and their performance of fluid flow and heat transfer is analyzed to ensure the uniformity condition using molten salt as heat carrier medium on shell-side by three-dimensional CFD method. The results reveal that comprehensive performance of the reactor with disk-and-doughnut baffles is better than that of with segmental baffles. Finally, the effects of operating conditions to control the hot spots are investigated. The results show that the flow velocity range about 0.65 m/s is applicable and the co-current cooling system flow direction is better than counter-current flow to control the hottest temperature.

Multi-physics design and analyses of long life reactors for lunar outposts

NASA Astrophysics Data System (ADS)

Schriener, Timothy M.

Future human exploration of the solar system is likely to include establishing permanent outposts on the surface of the Moon. These outposts will require reliable sources of electrical power in the range of 10's to 100's of kWe to support exploration and resource utilization activities. This need is best met using nuclear reactor power systems which can operate steadily throughout the long ˜27.3 day lunar rotational period, irrespective of location. Nuclear power systems can potentially open up the entire lunar surface for future exploration and development. Desirable features of nuclear power systems for the lunar surface include passive operation, the avoidance of single point failures in reactor cooling and the integrated power system, moderate operating temperatures to enable the use of conventional materials with proven irradiation experience, utilization of the lunar regolith for radiation shielding and as a supplemental neutron reflector, and safe post-operation decay heat removal and storage for potential retrieval. In addition, it is desirable for the reactor to have a long operational life. Only a limited number of space nuclear reactor concepts have previously been developed for the lunar environment, and these designs possess only a few of these desirable design and operation features. The objective of this research is therefore to perform design and analyses of long operational life lunar reactors and power systems which incorporate the desirable features listed above. A long reactor operational life could be achieved either by increasing the amount of highly enriched uranium (HEU) fuel in the core or by improving the neutron economy in the reactor through reducing neutron leakage and parasitic absorption. The amount of fuel in surface power reactors is constrained by the launch safety requirements. These include ensuring that the bare reactor core remains safely subcritical when submerged in water or wet sand and flooded with seawater in the unlikely event of a launch abort accident. Increasing the amount of fuel in the reactor core, and hence its operational life, would be possible by launching the reactor unfueled and fueling it on the Moon. Such a reactor would, thus, not be subject to launch criticality safety requirements. However, loading the reactor with fuel on the Moon presents a challenge, requiring special designs of the core and the fuel elements, which lend themselves to fueling on the lunar surface. This research investigates examples of both a solid core reactor that would be fueled at launch as well as an advanced concept which could be fueled on the Moon. Increasing the operational life of a reactor fueled at launch is exercised for the NaK-78 cooled Sectored Compact Reactor (SCoRe). A multi-physics design and analyses methodology is developed which iteratively couples together detailed Monte Carlo neutronics simulations with 3-D Computational Fluid Dynamics (CFD) and thermal-hydraulics analyses. Using this methodology the operational life of this compact, fast spectrum reactor is increased by reconfiguring the core geometry to reduce neutron leakage and parasitic absorption, for the same amount of HEU in the core, and meeting launch safety requirements. The multi-physics analyses determine the impacts of the various design changes on the reactor's neutronics and thermal-hydraulics performance. The option of increasing the operational life of a reactor by loading it on the Moon is exercised for the Pellet Bed Reactor (PeBR). The PeBR uses spherical fuel pellets and is cooled by He-Xe gas, allowing the reactor core to be loaded with fuel pellets and charged with working fluid on the lunar surface. The performed neutronics analyses ensure the PeBR design achieves a long operational life, and develops safe launch canister designs to transport the spherical fuel pellets to the lunar surface. The research also investigates loading the PeBR core with fuel pellets on the Moon using a transient Discrete Element Method (DEM) analysis in lunar gravity. In addition, this research addresses the post-operation storage of the SCoRe and PeBR concepts, below the lunar surface, to determine the time required for the radioactivity in the used fuel to decrease to a low level to allow for its safe recovery. The SCoRe and PeBR concepts are designed to operate at coolant temperatures ≤ 900 K and use conventional stainless steels and superalloys for the structure in the reactor core and power system. They are emplaced below grade on the Moon to take advantage of the regolith as a supplemental neutron reflector and as shielding of the lunar outpost from the reactors' neutron and gamma radiation.

Nuclear reactor removable radial shielding assembly having a self-bowing feature

DOEpatents

Pennell, William E.; Kalinowski, Joseph E.; Waldby, Robert N.; Rylatt, John A.; Swenson, Daniel V.

1978-01-01

A removable radial shielding assembly for use in the periphery of the core of a liquid-metal-cooled fast-breeder reactor, for closing interassembly gaps in the reactor core assembly load plane prior to reactor criticality and power operation to prevent positive reactivity insertion. The assembly has a lower nozzle portion for inserting into the core support and a flexible heat-sensitive bimetallic central spine surrounded by blocks of shielding material. At refueling temperature and below the spine is relaxed and in a vertical position so that the tolerances permitted by the interassembly gaps allow removal and replacement of the various reactor core assemblies. During an increase in reactor temperature from refueling to hot standby, the bimetallic spine expands, bowing the assembly toward the core center line, exerting a radially inward gap-closing-force on the above core load plane of the reactor core assembly, closing load plane interassembly gaps throughout the core prior to startup and preventing positive reactivity insertion.

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.

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.

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

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.

Optimization of the anaerobic treatment of a waste stream from an enhanced oil recovery process.

PubMed

Alimahmoodi, Mahmood; Mulligan, Catherine N

2011-01-01

The aim of this work was to optimize the anaerobic treatment of a waste stream from an enhanced oil recovery (EOR) process. The treatment of a simulated waste water containing about 150 mg chemical oxygen demand (COD)/L of total petroleum hydrocarbons (TPH) and the saturation level of CO2 was evaluated. A two-step anaerobic system was undertaken in the mesophilic temperature range (30-40°C). The method of evolutionary operation EVOP factorial design was used to optimize pH, temperature and organic loading rate with the target parameters of CO2 reduction and CH4 production in the first reactor and TPH removal in the second reactor. The results showed 98% methanogenic removal of CO2 and CH4 yield of 0.38 L/gCOD in the first reactor and 83% TPH removal in the second reactor. In addition to enhancing CO2 and TPH removal and CH4 production, application of this method showed the degree of importance of the operational variables and their interactive effects for the two reactors in series. Copyright © 2010 Elsevier Ltd. All rights reserved.

STEAM FORMING NEUTRONIC REACTOR AND METHOD OF OPERATING IT

DOEpatents

Untermyer, S.

1960-05-10

The heterogeneous reactor is liquid moderated and cooled by a steam forming coolant and is designed to produce steam from the coolant directly within the active portion of the reactor while avoiding the formation of bubbles in the liquid moderator. This reactor achieves inherent stability as a result of increased neutron leakage and increased neutron resonance absorption in the U/sup 238/ fuel with the formation of bubbles. The invention produces certain conditions under which the formation of vapor bubbles as a result of a neutron flux excursion from the injection of a reactivity increment into the reactor will operate to nullify the reactivity increment within a sufficiently short period of time to prevent unsafe reactor operating conditions from developing. This is obtained by disposing a plurality of fuel elements within a mass of steam forming coolant in the core with the ratio of the volume of steam forming coolant to the volume of fissionable isotopes being within the range yielding a multiplication factor greater than unity and a negative reactivity to core void coefficient at the boiling temperature of the coolant.

Inherently Safe and Long-Life Fission Power System for Lunar Outposts

NASA Astrophysics Data System (ADS)

Schriener, T. M.; El-Genk, Mohamed S.

Power requirements for future lunar outposts, of 10's to 100's kWe, can be fulfilled using nuclear reactor power systems. In addition to the long life and operation reliability, safety is paramount in all phases, including fabrication and assembly, launch, emplacement below grade on the lunar surface, operation, post-operation decay heat removal and long-term storage and eventual retrieval. This paper introduces the Solid Core-Sectored Compact Reactor (SC-SCoRe) and power system with static components and no single point failures. They ensure reliable continuous operation for ~21 years and fulfill the safety requirements. The SC-SCoRe nominally generates 1.0 MWth at liquid NaK-56 coolant inlet and exit temperatures of 850 K and 900 K and the power system provides 38 kWe at high DC voltage using SiGe thermoelectric (TE) conversion assemblies. In case of a loss of coolant or cooling in a reactor core sector, the power system continues to operate; generating ~4 kWe to the outpost for emergency life support needs. The post-operation storage of the reactor below grade on the lunar surface is a safe and practical choice. The total radioactivity in the reactor drops from ~1 million Ci, immediately at shutdown, to below 164 Ci after 300 years of storage. At such time, the reactor is retrieved safely with no contamination or environmental concerns.

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

Experimentation on the anaerobic filter reactor for biogas production using rural domestic wastewater

NASA Astrophysics Data System (ADS)

Leju Celestino Ladu, John; Lü, Xi-wu; Zhong, Zhaoping

2017-08-01

The biogas production from anaerobic filter (AF) reactor was experimented in Taihu Lake Environmental Engineering Research Center of Southeast University, Wuxi, China. Two rounds of experimental operations were conducted in a laboratory scale at different Hydraulic retention time (HRT) and wastewater temperature. The biogas production rate during the experimentation was in the range of 4.63 to 11.78 L/d. In the first experimentation, the average gas production rate was 10.08 L/d, and in the second experimentation, the average gas production rate was 4.97 L/d. The experimentation observed the favorable Hydraulic Retention Time and wastewater temperature in AF was three days and 30.95°C which produced the gas concentration of 11.78 L/d. The HRT and wastewater temperature affected the efficiency of the AF process on the organic matter removal and nutrients removal as well. It can be deduced from the obtained results that HRT and wastewater temperature directly affects the efficiency of the AF reactor in biogas production. In conclusion, anaerobic filter treatment of organic matter substrates from the rural domestic wastewater increases the efficiency of the AF reactor on biogas production and gives a number of benefits for the management of organic wastes as well as reduction in water pollution. Hence, the operation of the AF reactor in rural domestic wastewater treatment can play an important element for corporate economy of the biogas plant, socio-economic aspects and in the development of effective and feasible concepts for wastewater management, especially for people in rural low-income areas.

The effect of temperature and retention time on methane production and microbial community composition in staged anaerobic digesters fed with food waste.

PubMed

Gaby, John Christian; Zamanzadeh, Mirzaman; Horn, Svein Jarle

2017-01-01

Food waste is a large bio-resource that may be converted to biogas that can be used for heat and power production, or as transport fuel. We studied the anaerobic digestion of food waste in a staged digestion system consisting of separate acidogenic and methanogenic reactor vessels. Two anaerobic digestion parameters were investigated. First, we tested the effect of 55 vs. 65 °C acidogenic reactor temperature, and second, we examined the effect of reducing the hydraulic retention time (HRT) from 17 to 10 days in the methanogenic reactor. Process parameters including biogas production were monitored, and the microbial community composition was characterized by 16S amplicon sequencing. Neither organic matter removal nor methane production were significantly different for the 55 and 65 °C systems, despite the higher acetate and butyrate concentrations observed in the 65 °C acidogenic reactor. Ammonium levels in the methanogenic reactors were about 950 mg/L NH 4 + when HRT was 17 days but were reduced to 550 mg/L NH 4 + at 10 days HRT. Methane production increased from ~ 3600 mL/day to ~ 7800 when the HRT was decreased. Each reactor had unique environmental parameters and a correspondingly unique microbial community. In fact, the distinct values in each reactor for just two parameters, pH and ammonium concentration, recapitulate the separation seen in microbial community composition. The thermophilic and mesophilic digesters were particularly distinct from one another. The 55 °C acidogenic reactor was mainly dominated by Thermoanaerobacterium and Ruminococcus , whereas the 65 °C acidogenic reactor was initially dominated by Thermoanaerobacterium but later was overtaken by Coprothermobacter . The acidogenic reactors were lower in diversity (34-101 observed OTU 0.97 , 1.3-2.5 Shannon) compared to the methanogenic reactors (472-513 observed OTU 0.97 , 5.1-5.6 Shannon). The microbial communities in the acidogenic reactors were > 90% Firmicutes, and the Euryarchaeota were higher in relative abundance in the methanogenic reactors. The digestion systems had similar biogas production and COD removal rates, and hence differences in temperature, NH 4 + concentration, and pH in the reactors resulted in distinct but similarly functioning microbial communities over this range of operating parameters. Consequently, one could reduce operational costs by lowering both the hydrolysis temperature from 65 to 55 °C and the HRT from 17 to 10 days.

IN-PILE CORROSION TEST LOOPS FOR AQUEOUS HOMOGENEOUS REACTOR SOLUTIONS

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

Savage, H.C.; Jenks, G.H.; Bohlmann, E.G.

1960-12-21

An in-pile corrosion test loop is described which is used to study the effect of reactor radiation on the corrosion of materials of construction and the chemical stability of fuel solutions of interest to the Aqueous Homogeneous Reactor Program at ORNL. Aqueous solutions of uranyl sulfate are circulated in the loop by means of a 5-gpm canned-rotor pump, and the pump loop is designed for operation at temperatures to 300 ts C and pressures to 2000 psia while exposed to reactor radiation in beam-hole facilities of the LITR and ORR. Operation of the first loop in-pile was begun in Octobermore » 1954, and since that time 17 other in-pile loop experiments were completed. Design criteria of the pump loop and its associated auxiliary equipment and instrumentation are described. In-pile operating procedures, safety features, and operating experience are presented. A cost summary of the design, fabrication, and installation of the loop and experimental facillties is also included. (auth)« less

Tungsten Deposition on Graphite using Plasma Enhanced Chemical Vapour Deposition.

NASA Astrophysics Data System (ADS)

Sharma, Uttam; Chauhan, Sachin S.; Sharma, Jayshree; Sanyasi, A. K.; Ghosh, J.; Choudhary, K. K.; Ghosh, S. K.

2016-10-01

The tokamak concept is the frontrunner for achieving controlled thermonuclear reaction on earth, an environment friendly way to solve future energy crisis. Although much progress has been made in controlling the heated fusion plasmas (temperature ∼ 150 million degrees) in tokamaks, technological issues related to plasma wall interaction topic still need focused attention. In future, reactor grade tokamak operational scenarios, the reactor wall and target plates are expected to experience a heat load of 10 MW/m2 and even more during the unfortunate events of ELM's and disruptions. Tungsten remains a suitable choice for the wall and target plates. It can withstand high temperatures, its ductile to brittle temperature is fairly low and it has low sputtering yield and low fuel retention capabilities. However, it is difficult to machine tungsten and hence usages of tungsten coated surfaces are mostly desirable. To produce tungsten coated graphite tiles for the above-mentioned purpose, a coating reactor has been designed, developed and made operational at the SVITS, Indore. Tungsten coating on graphite has been attempted and successfully carried out by using radio frequency induced plasma enhanced chemical vapour deposition (rf -PECVD) for the first time in India. Tungsten hexa-fluoride has been used as a pre-cursor gas. Energy Dispersive X-ray spectroscopy (EDS) clearly showed the presence of tungsten coating on the graphite samples. This paper presents the details of successful operation and achievement of tungsten coating in the reactor at SVITS.

Analysis of boron dilution in a four-loop PWR

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

Sun, J.G.; Sha, W.T.

1995-12-31

Thermal mixing and boron dilution in a pressurized water reactor were analyzed with COMMIX codes. The reactor system was the four loop Zion reactor. Two boron dilution scenarios were analyzed. In the first scenario, the plant is in cold shutdown and the reactor coolant system has just been filled after maintenance on the steam generators. To flush the air out of the steam generator tubes, a reactor coolant pump (RCP) is started, with the water in the pump suction line devoid of boron and at the same temperature as the coolant in the system. In the second scenario, the plantmore » is at hot standby and the reactor coolant system has been heated up to operating temperature after a long outage. It is assumed that an RCP is started, with the pump suction line filled with cold unborated water, forcing a slug of diluted coolant down the downcomer and subsequently through the reactor core. The subsequent transient thermal mixing and boron dilution that would occur in the reactor system is simulated for these two scenarios. The reactivity insertion rate and the total reactivity are evaluated.« less

Combined Effects of Temperature and Irradiation on Concrete Damage

DOE PAGES

Le Pape, Yann; Giorla, Alain; Sanahuja, Julien

2016-01-01

Aggregate radiation-induced volumetric expansion (RIVE) is a predominant mechanism in the formation of mechanical damage in the hardened cement paste (hcp) of irradiated concrete under fast-neutron flux (Giorla et al. 2015). Among the operating conditions difference between test reactors and light water reactors (LWRs), the difference of irradiation flux and temperature is significant. While a temperature increase is quite generally associated with a direct, or indirect (e.g., by dehydration) loss of mechanical properties (Maruyama et al. 2014), we found that it causes a partial annealing of irradiation amorphization of α-quartz, hence, reducing RIVE rate. Based on data collected by Bykovmore » et al. (1981), an incremental RIVE model coupling neutron fluence and temperature is developed. The elastic properties and coefficient of thermal expansion (CTE) of irradiated polycrystalline quartz are interpreted through analytical homogenization of experimental data on irradiated α-quartz published by Mayer and Lecomte (1960). Moreover, the proposed model, implemented in the meso-scale simulation code AMIE, is compared to experimental data obtained on ordinary concrete made of quartz/quartzite aggregate (Dubrovskii et al. 1967). Substantial discrepancy, in terms of damage and volumetric expansion developments, is found when comparing irradiation scenarios assuming constant flux and temperature, as opposed to more realistic test reactor operation conditions.« less

Behavior of toxic metals and radionuclides during molten salt oxidation of chlorinated plastics.

PubMed

Yang, Hee-Chul; Cho, Yong-Jun; Eun, Hee-Chul; Yoo, Jae-Hyung; Kim, Joon-Hyung

2004-01-01

Molten salt oxidation is one of the promising alternatives to incineration for chlorinated organics without the emission of chlorinated organic pollutants. This study investigated the behavior of three hazardous metals (Cd, Pb, and Cr) and four radioactive metal surrogates (Cs, Ce, Gd, and Sm) in the molten Na2CO3 oxidation reactor during the destruction of PVC plastics. In the tested temperature ranges (1143 1223K) and NaCl content (0-10%), the impact of temperature on the retention of cadmium and lead in the molten salt reactor was very small, but that of the NaCl content for their retention was relatively higher. The influence of NaCl accumulation was, however, proven to be practically negligible due to the low-temperature operating characteristics of the molten salt oxidation system. Neither temperature increase nor chlorine accumulation in the MSO reactor reduced the retention of Cr, Ce, Gd, and Sm. Over 99.98% of these metals remained in the reactor. The influence of the temperature on the cesium behavior is relatively large for a chlorine addition, however, over 99.7% of cesium remained in the reactor throughout the entire test. The experimental metal entrainment rate and the entrained metal particle size distribution agree well with the theoretical equilibrium metal distributions.

Physics Features of TRU-Fueled VHTRs

DOE PAGES

Lewis, Tom G.; Tsvetkov, Pavel V.

2009-01-01

The current waste management strategy for spent nuclear fuel (SNF) mandated by the US Congress is the disposal of high-level waste (HLW) in a geological repository at Yucca Mountain. Ongoing efforts on closed-fuel cycle options and difficulties in opening and safeguarding such a repository have led to investigations of alternative waste management strategies. One potential strategy for the US fuel cycle would be to make use of fuel loadings containing high concentrations of transuranic (TRU) nuclides in the next-generation reactors. The use of such fuels would not only increase fuel supply but could also potentially facilitate prolonged operation modes (viamore » fertile additives) on a single fuel loading. The idea is to approach autonomous operation on a single fuel loading that would allow marketing power units as nuclear batteries for worldwide deployment. Studies have already shown that high-temperature gas-cooled reactors (HTGRs) and their Generation IV (GEN IV) extensions, very-high-temperature reactors (VHTRs), have encouraging performance characteristics. This paper is focused on possible physics features of TRU-fueled VHTRs. One of the objectives of a 3-year U.S. DOE NERI project was to show that TRU-fueled VHTRs have the possibility of prolonged operation on a single fuel loading. A 3D temperature distribution was developed based on conceivable operation conditions of the 600 MWth VHTR design. Results of extensive criticality and depletion calculations with varying fuel loadings showed that VHTRs are capable for autonomous operation and HLW waste reduction when loaded with TRU fuel.« 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

76 FR 48184 - Exelon Nuclear, Peach Bottom Atomic Power Station, Unit 1; Exemption From Certain Security...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-08-08

... nuclear reactor facility. PBAPS Unit 1 was a high-temperature, gas-cooled reactor that was operated from... the safeguards contingency plan.'' Part 73 of 10 CFR, ``Physical Protection of Plant and Materials... physical protection system which will have capabilities for the protection of special nuclear material at...

Nuclear core positioning system

DOEpatents

Garkisch, Hans D.; Yant, Howard W.; Patterson, John F.

1979-01-01

A structural support system for the core of a nuclear reactor which achieves relatively restricted clearances at operating conditions and yet allows sufficient clearance between fuel assemblies at refueling temperatures. Axially displaced spacer pads having variable between pad spacing and a temperature compensated radial restraint system are utilized to maintain clearances between the fuel elements. The core support plates are constructed of metals specially chosen such that differential thermal expansion produces positive restraint at operating temperatures.

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.)

A study of increasing radical density and etch rate using remote plasma generator system

NASA Astrophysics Data System (ADS)

Lee, Jaewon; Kim, Kyunghyun; Cho, Sung-Won; Chung, Chin-Wook

2013-09-01

To improve radical density without changing electron temperature, remote plasma generator (RPG) is applied. Multistep dissociation of the polyatomic molecule was performed using RPG system. RPG is installed to inductively coupled type processing reactor; electrons, positive ions, radicals and polyatomic molecule generated in RPG and they diffused to processing reactor. The processing reactor dissociates the polyatomic molecules with inductively coupled power. The polyatomic molecules are dissociated by the processing reactor that is operated by inductively coupled power. Therefore, the multistep dissociation system generates more radicals than single-step system. The RPG was composed with two cylinder type inductively coupled plasma (ICP) using 400 kHz RF power and nitrogen gas. The processing reactor composed with two turn antenna with 13.56 MHz RF power. Plasma density, electron temperature and radical density were measured with electrical probe and optical methods.

Feasibility Study of a Nuclear-Stirling Power Plant for the Jupiter Icy Moons Orbiter

NASA Astrophysics Data System (ADS)

Schmitz, Paul C.; Schreiber, Jeffrey G.; Penswick, L. Barry

2005-02-01

NASA is undertaking the design of a new spacecraft to explore the planet Jupiter and its three moons Calisto, Ganymede and Europa. This proposed mission, known as Jupiter Icy Moons Orbiter (JIMO) would use a nuclear reactor and an associated electrical generation system (Reactor Power Plant - RPP) to provide power to the spacecraft. The JIMO spacecraft is envisioned to use this power for science and communications as well as Electric Propulsion (EP). Among other potential power-generating concepts, previous studies have considered Thermoelectric and Brayton power conversion systems, coupled to a liquid metal reactor for the JIMO mission. This paper will explore trades in system mass and radiator area for a nuclear reactor power conversion system, however this study will focus on Stirling power conversion. Stirling convertors have a long heritage operating in both power generation and the cooler industry, and are currently in use in a wide variety of applications. The Stirling convertor modeled in this study is based upon the Component Test Power Convertor design that was designed and operated successfully under the Civil Space Technology Initiative for use with the SP-100 nuclear reactor in the 1980's and early 1990's. The baseline RPP considered in this study consists of four dual-opposed Stirling convertors connected to the reactor by a liquid lithium loop. The study design is such that two of the four convertors would operate at any time to generate the 100 kWe while the others are held in reserve. For this study the Stirling convertors hot-side temperature is 1050 K, would operate at a temperature ratio of 2.4 for a minimum mass system and would have a system efficiency of 29%. The Stirling convertor would generate high voltage (400 volt), 100 Hz single phase AC that is supplied to the Power Management and Distribution system. The waste heat is removed from the Stirling convertors by a flowing liquid sodium-potassium eutectic and then rejected by a shared radiator. The radiator consists of two coplanar wings, which would be deployed after the reactor is in space. For this study design, the radiators would be located behind the conical radiation shield of the reactor and fan out as the radiator's distance from the reactor increases. System trades were performed to vary cycle state point temperatures and convertor design as well as power output. Other redundancy combinations were considered to understand the affects of convertor size and number of spares to the system mass.

Feasibility Study of a Nuclear-Stirling Power Plant for the Jupiter Icy Moons Orbiter

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

Schmitz, Paul C.; Schreiber, Jeffrey G.; Penswick, L. Barry

2005-02-06

NASA is undertaking the design of a new spacecraft to explore the planet Jupiter and its three moons Calisto, Ganymede and Europa. This proposed mission, known as Jupiter Icy Moons Orbiter (JIMO) would use a nuclear reactor and an associated electrical generation system (Reactor Power Plant - RPP) to provide power to the spacecraft. The JIMO spacecraft is envisioned to use this power for science and communications as well as Electric Propulsion (EP). Among other potential power-generating concepts, previous studies have considered Thermoelectric and Brayton power conversion systems, coupled to a liquid metal reactor for the JIMO mission. This papermore » will explore trades in system mass and radiator area for a nuclear reactor power conversion system, however this study will focus on Stirling power conversion. Stirling convertors have a long heritage operating in both power generation and the cooler industry, and are currently in use in a wide variety of applications. The Stirling convertor modeled in this study is based upon the Component Test Power Convertor design that was designed and operated successfully under the Civil Space Technology Initiative for use with the SP-100 nuclear reactor in the 1980's and early 1990's. The baseline RPP considered in this study consists of four dual-opposed Stirling convertors connected to the reactor by a liquid lithium loop. The study design is such that two of the four convertors would operate at any time to generate the 100 kWe while the others are held in reserve. For this study the Stirling convertors hot-side temperature is 1050 K, would operate at a temperature ratio of 2.4 for a minimum mass system and would have a system efficiency of 29%. The Stirling convertor would generate high voltage (400 volt), 100 Hz single phase AC that is supplied to the Power Management and Distribution system. The waste heat is removed from the Stirling convertors by a flowing liquid sodium-potassium eutectic and then rejected by a shared radiator. The radiator consists of two coplanar wings, which would be deployed after the reactor is in space. For this study design, the radiators would be located behind the conical radiation shield of the reactor and fan out as the radiator's distance from the reactor increases. System trades were performed to vary cycle state point temperatures and convertor design as well as power output. Other redundancy combinations were considered to understand the affects of convertor size and number of spares to the system mass.« less

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.

Baseline Concept Description of a Small Modular High Temperature Reactor

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

Hans Gougar

2014-05-01

The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNPmore » were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the Generation IV program and its specific R&D needs will be included in this report when appropriate for comparison. The distinguishing features of the HTGR are the refractory (TRISO) coated particle fuel, the low-power density, graphite-moderated core, and the high outlet temperature of the inert helium coolant. The low power density and fuel form effectively eliminate the possibility of core melt, even upon a complete loss of coolant pressure and flow. The graphite, which constitutes the bulk of the core volume and mass, provides a large thermal buffer that absorbs fission heat such that thermal transients occur over a timespan of hours or even days. As chemically-inert helium is already a gas, there is no coolant temperature or void feedback on the neutronics and no phase change or corrosion product that could degrade heat transfer. Furthermore, the particle coatings and interstitial graphite retain fission products such that the source terms at the plant boundary remain well below actionable levels under all anticipated nominal and off-normal operating conditions. These attributes enable the reactor to supply process heat to a collocated industrial plant with negligible risk of contamination and minimal dynamic coupling of the facilities (Figure 1). The exceptional retentive properties of coated particle fuel in a graphite matrix were first demonstrated in the DRAGON reactor, a European research facility that began operation in 1964.« less

Baseline Concept Description of a Small Modular High Temperature Reactor

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

Gougar, Hans D.

2014-10-01

The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNPmore » were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the Generation IV program and its specific R&D needs will be included in this report when appropriate for comparison. The distinguishing features of the HTGR are the refractory (TRISO) coated particle fuel, the low-power density, graphite-moderated core, and the high outlet temperature of the inert helium coolant. The low power density and fuel form effectively eliminate the possibility of core melt, even upon a complete loss of coolant pressure and flow. The graphite, which constitutes the bulk of the core volume and mass, provides a large thermal buffer that absorbs fission heat such that thermal transients occur over a timespan of hours or even days. As chemically-inert helium is already a gas, there is no coolant temperature or void feedback on the neutronics and no phase change or corrosion product that could degrade heat transfer. Furthermore, the particle coatings and interstitial graphite retain fission products such that the source terms at the plant boundary remain well below actionable levels under all anticipated nominal and off-normal operating conditions. These attributes enable the reactor to supply process heat to a collocated industrial plant with negligible risk of contamination and minimal dynamic coupling of the facilities (Figure 1). The exceptional retentive properties of coated particle fuel in a graphite matrix were first demonstrated in the DRAGON reactor, a European research facility that began operation in 1964.« less

High Temperature Water Heat Pipes Radiator for a Brayton Space Reactor Power System

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.; Tournier, Jean-Michel

2006-01-01

A high temperature water heat pipes radiator design is developed for a space power system with a sectored gas-cooled reactor and three Closed Brayton Cycle (CBC) engines, for avoidance of single point failures in reactor cooling and energy conversion and rejection. The CBC engines operate at turbine inlet and exit temperatures of 1144 K and 952 K. They have a net efficiency of 19.4% and each provides 30.5 kWe of net electrical power to the load. A He-Xe gas mixture serves as the turbine working fluid and cools the reactor core, entering at 904 K and exiting at 1149 K. Each CBC loop is coupled to a reactor sector, which is neutronically and thermally coupled, but hydraulically decoupled to the other two sectors, and to a NaK-78 secondary loop with two water heat pipes radiator panels. The segmented panels each consist of a forward fixed segment and two rear deployable segments, operating hydraulically in parallel. The deployed radiator has an effective surface area of 203 m2, and when the rear segments are folded, the stowed power system fits in the launch bay of the DELTA-IV Heavy launch vehicle. For enhanced reliability, the water heat pipes operate below 50% of their wicking limit; the sonic limit is not a concern because of the water, high vapor pressure at the temperatures of interest (384 - 491 K). The rejected power by the radiator peaks when the ratio of the lengths of evaporator sections of the longest and shortest heat pipes is the same as that of the major and minor widths of the segments. The shortest and hottest heat pipes in the rear segments operate at 491 K and 2.24 MPa, and each rejects 154 W. The longest heat pipes operate cooler (427 K and 0.52 MPa) and because they are 69% longer, reject more power (200 W each). The longest and hottest heat pipes in the forward segments reject the largest power (320 W each) while operating at ~ 46% of capillary limit. The vapor temperature and pressure in these heat pipes are 485 K and 1.97 MPa. By contrast, the shortest water heat pipes in the forward segments operate much cooler (427 K and 0.52 MPa), and reject a much lower power of 45 W each. The radiator with six fixed and 12 rear deployable segments rejects a total of 324 kWth, weights 994 kg and has an average specific power of 326 Wth/kg and a specific mass of 5.88 kg/m2.

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

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?

Proposed Design and Operation of a Heat Pipe Reactor using the Sandia National Laboratories Annular Core Test Facility and Existing UZrH Fuel Pins

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

Wright, Steven A.; Lipinski, Ronald J.; Pandya, Tara

2005-02-06

Heat Pipe Reactors (HPR) for space power conversion systems offer a number of advantages not easily provided by other systems. They require no pumping, their design easily deals with freezing and thawing of the liquid metal, and they can provide substantial levels of redundancy. Nevertheless, no reactor has ever been operated and cooled with heat pipes, and the startup and other operational characteristics of these systems remain largely unknown. Signification deviations from normal reactor heat removal mechanisms exist, because the heat pipes have fundamental heat removal limits due to sonic flow issues at low temperatures. This paper proposes an earlymore » prototypic test of a Heat Pipe Reactor (using existing 20% enriched nuclear fuel pins) to determine the operational characteristics of the HPR. The proposed design is similar in design to the HOMER and SAFE-300 HPR designs (Elliot, Lipinski, and Poston, 2003; Houts, et. al, 2003). However, this reactor uses existing UZrH fuel pins that are coupled to potassium heat pipes modules. The prototype reactor would be located in the Sandia Annular Core Research Reactor Facility where the fuel pins currently reside. The proposed reactor would use the heat pipes to transport the heat from the UZrH fuel pins to a water pool above the core, and the heat transport to the water pool would be controlled by adjusting the pressure and gas type within a small annulus around each heat pipe. The reactor would operate as a self-critical assembly at power levels up to 200 kWth. Because the nuclear heated HPR test uses existing fuel and because it would be performed in an existing facility with the appropriate safety authorization basis, the test could be performed rapidly and inexpensively. This approach makes it possible to validate the operation of a HPR and also measure the feedback mechanisms for a typical HPR design. A test of this nature would be the world's first operating Heat Pipe Reactor. This reactor is therefore called 'HPR-1'.« less

Key role of temperature monitoring in interpretation of microwave effect on transesterification and esterification reactions for biodiesel production.

PubMed

Mazubert, Alex; Taylor, Cameron; Aubin, Joelle; Poux, Martine

2014-06-01

Microwave effects have been quantified, comparing activation energies and pre-exponential factors to those obtained in a conventionally-heated reactor for biodiesel production from waste cooking oils via transesterification and esterification reactions. Several publications report an enhancement of biodiesel production using microwaves, however recent reviews highlight poor temperature measurements in microwave reactors give misleading reaction performances. Operating conditions have therefore been carefully chosen to investigate non-thermal microwave effects alone. Temperature is monitored by an optical fiber sensor, which is more accurate than infrared sensors. For the transesterification reaction, the activation energy is 37.1kJ/mol (20.1-54.2kJ/mol) in the microwave-heated reactor compared with 31.6kJ/mol (14.6-48.7kJ/mol) in the conventionally-heated reactor. For the esterification reaction, the activation energy is 45.4kJ/mol (31.8-58.9kJ/mol) for the microwave-heated reactor compared with 56.1kJ/mol (55.7-56.4kJ/mol) for conventionally-heated reactor. The results confirm the absence of non-thermal microwave effects for homogenous-catalyzed reactions. Copyright © 2014 Elsevier Ltd. All rights reserved.

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

Dr. Mohit Jain; Dr. Ganesh Skandan; Dr. Gordon E. Khose

Generation IV Very High Temperature power generating nuclear reactors will operate at temperatures greater than 900 oC. At these temperatures, the components operating in these reactors need to be fabricated from materials with excellent thermo-mechanical properties. Conventional pure or composite materials have fallen short in delivering the desired performance. New materials, or conventional materials with new microstructures, and associated processing technologies are needed to meet these materials challenges. Using the concept of functionally graded materials, we have fabricated a composite material which has taken advantages of the mechanical and thermal properties of ceramic and metals. Functionally-graded composite samples with variousmore » microstructures were fabricated. It was demonstrated that the composition and spatial variation in the composition of the composite can be controlled. Some of the samples were tested for irradiation resistance to neutrons. The samples did not degrade during initial neutron irradiation testing.« less

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 electrical characteristics of the dielectric barrier discharges

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

Yehia, Ashraf, E-mail: yehia30161@yahoo.com; Department of Physics, Faculty of Science, Assiut University, Assiut 71516

2016-06-15

The electrical characteristics of the dielectric barrier discharges have been studied in this paper under different operating conditions. The dielectric barrier discharges were formed inside two reactors composed of electrodes in the shape of two parallel plates. The dielectric layers inside these reactors were pasted on the surface of one electrode only in the first reactor and on the surfaces of the two electrodes in the second reactor. The reactor under study has been fed by atmospheric air that flowed inside it with a constant rate at the normal temperature and pressure, in parallel with applying a sinusoidal ac voltagemore » between the electrodes of the reactor. The amount of the electric charge that flows from the reactors to the external circuit has been studied experimentally versus the ac peak voltage applied to them. An analytical model has been obtained for calculating the electrical characteristics of the dielectric barrier discharges that were formed inside the reactors during a complete cycle of the ac voltage. The results that were calculated by using this model have agreed well with the experimental results under the different operating conditions.« less

Production of aluminum-silicon alloy and ferrosilicon and commercial purity aluminum by the direct reduction process. Second interim technical report, Phase B, December 1, 1978-February 28, 1979

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

Bruno, M.J.

1979-03-01

Experimental runs were made to determine the effect of a cooler product reservoir on metal alloy yield and recovery. The reservoir temperature had no significant effect. Difficulties were experienced with operation of an oxygen injected bench scale reactor. Many tests were terminated by burden bridging or flooding of the oxygen tuyeres with metal and slag. Runs were made in which refluxing vapors were condensed in a liquid slag. The addition of CaO decreased the tendency for formation of thick, strong burden bridges but did not completely eliminate bridging. Reduction of flame temperatures did not affect the volatilization rate in themore » bench reactor. Operation of VSR-1 pilot reactor with O injection was achieved after resolving reactor shell leakage problems, by replacing the permeable ceramic shell with impermeable fused silica. Various combustion parameters were investigated, including coke size, burden height and oxygen flow rate. Steady state operation of the oxygen-coke system was attained with smooth burden movement and a 2000/sup 0/C bed temperature in the raceway vicinity. To further reduce heat losses from the raceway area. VSR-1 was redesigned to facilitate locating an induction coil below the oxygen inlets. Further evaluation of effects of impurities on alloy purification in the bench scale unit indicated a 50% decrease in product yield for starting charges containing Fe greater than 5%. Site installation for the entire alloy purification complex was completed. Operations were continued in the bench scale units to obtain design information for the pilot commercial grade Al purification unit. Procurement of construction material was established.« less

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

Tan, Lizhen; Yang, Ying; Tyburska-Puschel, Beata

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 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 computational tools)more » is an important path to more efficient alloy development and process optimization. Ferritic-martensitic (FM) steels are important structural materials for nuclear reactors due to their advantages over other applicable materials like austenitic stainless steels, notably their resistance to void swelling, low thermal expansion coefficients, and higher thermal conductivity. However, traditional FM steels exhibit a noticeable yield strength reduction at elevated temperatures above ~500°C, which limits their applications in advanced nuclear reactors which target operating temperatures at 650°C or higher. Although oxide-dispersion-strengthened (ODS) ferritic steels have shown excellent high-temperature performance, their extremely high cost, limited size and fabricability of products, as well as the great difficulty with welding and joining, have limited or precluded their commercial applications. Zirconium has shown many benefits to Fe-base alloys such as grain refinement, improved phase stability, and reduced radiation-induced segregation. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of a new generation of Zr-bearing ferritic alloys to be fabricated using conventional steelmaking practices, which have excellent radiation resistance and enhanced high-temperature creep performance greater than Grade 91.« less

Metal halides vapor lasers with inner reactor and small active volume.

NASA Astrophysics Data System (ADS)

Shiyanov, D. V.; Sukhanov, V. B.; Evtushenko, G. S.

2018-04-01

Investigation of the energy characteristics of copper, manganese, lead halide vapor lasers with inner reactor and small active volume 90 cm3 was made. The optimal operating pulse repetition rates, temperatures, and buffer gas pressure for gas discharge tubes with internal and external electrodes are determined. Under identical pump conditions, such systems are not inferior in their characteristics to standard metal halide vapor lasers. It is shown that the use of a zeolite halogen generator provides lifetime laser operation.

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

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.

Reactor Simulator Testing Overview

NASA Technical Reports Server (NTRS)

Schoenfeld, Michael P.

2013-01-01

Test Objectives Summary: a) Verify operation of the core simulator, the instrumentation & control system, and the ground support gas and vacuum test equipment. b) Examine cooling & heat regeneration performance of the cold trap purification. c) Test the ALIP pump at voltages beyond 120V to see if the targeted mass flow rate of 1.75 kg/s can be obtained in the RxSim. Testing Highlights: a) Gas and vacuum ground support test equipment performed effectively for operations (NaK fill, loop pressurization, and NaK drain). b) Instrumentation & Control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings and ramped within prescribed constraints. It effectively interacted with reactor simulator control model and defaulted back to temperature control mode if the transient fluctuations didn't dampen. c) Cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the minimum temperature indicating the design provided some heat regeneration. d) ALIP produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

Magnetic switch for reactor control rod. [LMFBR

DOEpatents

Germer, J.H.

1982-09-30

A magnetic reed switch assembly is described for activating an electromagnetic grapple utilized to hold a control rod in position above a reactor core. In normal operation the magnetic field of a permanent magnet is short-circuited by a magnetic shunt, diverting the magnetic field away from the reed switch. The magnetic shunt is made of a material having a Curie-point at the desired release temperature. Above that temperature the material loses its ferromagnetic properties, and the magnetic path is diverted to the reed switch which closes and short-circuits the control circuit for the control rod electro-magnetic grapple which allows the control rod to drop into the reactor core for controlling the reactivity of the core.

Magnetic switch for reactor control rod

DOEpatents

Germer, John H.

1986-01-01

A magnetic reed switch assembly for activating an electromagnetic grapple utilized to hold a control rod in position above a reactor core. In normal operation the magnetic field of a permanent magnet is short-circuited by a magnetic shunt, diverting the magnetic field away from the reed switch. The magnetic shunt is made of a material having a Curie-point at the desired release temperature. Above that temperature the material loses its ferromagnetic properties, and the magnetic path is diverted to the reed switch which closes and short-circuits the control circuit for the control rod electromagnetic grapple which allows the control rod to drop into the reactor core for controlling the reactivity of the core.

Gravity Scaling of a Power Reactor Water Shield

NASA Technical Reports Server (NTRS)

Reid, Robert S.; Pearson, J. Boise

2007-01-01

A similarity analysis on a water-based reactor shield examined the effect of gravity on free convection between a reactor shield inner and outer vessel boundaries. Two approaches established similarity between operation on the Earth and the Moon: 1) direct scaling of Rayleigh number equating gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant. Nusselt number for natural convection (laminar and turbulent) is assumed of form Nu = CRa(sup n).

Using the sound of nuclear energy

DOE PAGES

Garrett, Steven; Smith, James; Smith, Robert; ...

2016-08-01

The generation of sound by heat has been documented as an “acoustical curiosity” since a Buddhist monk reported the loud tone generated by a ceremonial rice-cooker in his diary, in 1568. Over the last four decades, significant progress has been made in understanding “thermoacoustic processes,” enabling the design of thermoacoustic engines and refrigerators. Motivated by the Fukushima nuclear reactor disaster, we have developed and tested a thermoacoustic engine that exploits the energy-rich conditions in the core of a nuclear reactor to provide core condition information to the operators without a need for external electrical power. The heat engine is self-poweredmore » and can wirelessly transmit the temperature and reactor power level by generation of a pure tone which can be detected outside the reactor. We report here the first use of a fission-powered thermoacoustic engine capable of serving as a performance and safety sensor in the core of a research reactor and present data from the hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. These measurements confirmed that the frequency of the sound produced indicates the reactor’s coolant temperature and that the amplitude (above an onset threshold) is related to the reactor’s operating power level. Furthermore, these signals can be detected even in the presence of substantial background noise generated by the reactor’s fluid pumps.« less

Using the sound of nuclear energy

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

Garrett, Steven; Smith, James; Smith, Robert

The generation of sound by heat has been documented as an “acoustical curiosity” since a Buddhist monk reported the loud tone generated by a ceremonial rice-cooker in his diary, in 1568. Over the last four decades, significant progress has been made in understanding “thermoacoustic processes,” enabling the design of thermoacoustic engines and refrigerators. Motivated by the Fukushima nuclear reactor disaster, we have developed and tested a thermoacoustic engine that exploits the energy-rich conditions in the core of a nuclear reactor to provide core condition information to the operators without a need for external electrical power. The heat engine is self-poweredmore » and can wirelessly transmit the temperature and reactor power level by generation of a pure tone which can be detected outside the reactor. We report here the first use of a fission-powered thermoacoustic engine capable of serving as a performance and safety sensor in the core of a research reactor and present data from the hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. These measurements confirmed that the frequency of the sound produced indicates the reactor’s coolant temperature and that the amplitude (above an onset threshold) is related to the reactor’s operating power level. Furthermore, these signals can be detected even in the presence of substantial background noise generated by the reactor’s fluid pumps.« less

SCW Pressure-Channel Nuclear Reactor Some Design Features

NASA Astrophysics Data System (ADS)

Pioro, Igor L.; Khan, Mosin; Hopps, Victory; Jacobs, Chris; Patkunam, Ruban; Gopaul, Sandeep; Bakan, Kurtulus

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30-35% to about 45-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (˜1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Fast Salt Reactor (MFSR)

NASA Astrophysics Data System (ADS)

Laureau, A.; Rubiolo, P. R.; Heuer, D.; Merle-Lucotte, E.; Brovchenko, M.

2014-06-01

Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated.

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

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.

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.

Design and testing of a self-actuated shut down system for the protection of liquid metal fast breeder reactors (LMFBRs)

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

Josephson, J.; Sowa, E.S.

1977-04-01

The design and testing of a simple and reliable Self-Actuated Shutdown System (SASS) for the protection of Liquid Metal Fast Breeder Reactors (LMFBRs) is described. A ferromagnetic Curie temperature permanent magnet holding device has been selected for the design of the Self-Actuated Shutdown System in order to enhance the safety of liquid metal cooled fast reactors (LMFBRs). The self-actuated, self-contained device operates such that accident conditions, resulting in increased coolant temperature or neutron flux reduce the magnetic holding force suspending a neutron absorber above the core by raising the temperature of the trigger mechanism above the Curie point. Neutron absorbermore » material is then inserted into the core, under gravity, terminating the accident. Two possible design variations of the selected concept are presented.« less

Development of small-size tubular-flow continuous reactors for the analysis of operational stability of enzymes in low-water systems.

PubMed

Pirozzi, D; Halling, P J

2001-01-20

A very small-scale continuous flow reactor has been designed for use with enzymes in organic media, particularly for operational stability studies. It is constructed from fairly inexpensive components, and typically uses 5 mg of catalyst and flow rates of 1 to 5 mL/h, so only small quantities of feedstock need to be handled. The design allows control of the thermodynamic water activity of the feed, and works with temperatures up to at least 80 degrees C. The reactor has been operated with both nonpolar (octane) and polar (4-methyl-pentan-2-one) solvents, and with the more viscous solvent-free reactant mixture. It has been applied to studies of the operational stability of lipases from Chromobacterium viscosum (lyophilized powder or polypropylene-adsorbed) and Rhizomucor miehei (Lipozyme) in different experimental conditions. Transesterification of geraniol and ethylcaproate has been adopted as a model transformation.

In-Pile Instrumentation Multi- Parameter System Utilizing Photonic Fibers and Nanovision

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

Burgett, Eric

2015-10-13

An advanced in-pile multi-parameter reactor monitoring system is being proposed in this funding opportunity. The proposed effort brings cutting edge, high fidelity optical measurement systems into the reactor environment in an unprecedented fashion, including in-core, in-cladding and in-fuel pellet itself. Unlike instrumented leads, the proposed system provides a unique solution to a multi-parameter monitoring need in core while being minimally intrusive in the reactor core. Detector designs proposed herein can monitor fuel compression and expansion in both the radial and axial dimensions as well as monitor linear power profiles and fission rates during the operation of the reactor. In additionmore » to pressure, stress, strain, compression, neutron flux, neutron spectra, and temperature can be observed inside the fuel bundle and fuel rod using the proposed system. The proposed research aims at developing radiation-hard, harsh-environment multi-parameter systems for insertion into the reactor environment. The proposed research holds the potential to drastically increase the fidelity and precision of in-core instrumentation with little or no impact in the neutron economy in the reactor environment while providing a measurement system capable of operation for entire operating cycles.« less

Performance evaluation and phylogenetic characterization of anaerobic fluidized bed reactors using ground tire and pet as support materials for biohydrogen production.

PubMed

Barros, Aruana Rocha; Adorno, Maria Angela Tallarico; Sakamoto, Isabel Kimiko; Maintinguer, Sandra Imaculada; Varesche, Maria Bernadete Amâncio; Silva, Edson Luiz

2011-02-01

This study evaluated two different support materials (ground tire and polyethylene terephthalate [PET]) for biohydrogen production in an anaerobic fluidized bed reactor (AFBR) treating synthetic wastewater containing glucose (4000 mg L(-1)). The AFBR, which contained either ground tire (R1) or PET (R2) as support materials, were inoculated with thermally pretreated anaerobic sludge and operated at a temperature of 30°C. The AFBR were operated with a range of hydraulic retention times (HRT) between 1 and 8h. The reactor R1 operating with a HRT of 2h showed better performance than reactor R2, reaching a maximum hydrogen yield of 2.25 mol H(2)mol(-1) glucose with 1.3mg of biomass (as the total volatile solids) attached to each gram of ground tire. Subsequent 16S rRNA gene sequencing and phylogenetic analysis of particle samples revealed that reactor R1 favored the presence of hydrogen-producing bacteria such as Clostridium, Bacillus, and Enterobacter. Copyright © 2010 Elsevier Ltd. All rights reserved.

Automatic reactor control system for transient operation

NASA Astrophysics Data System (ADS)

Lipinski, Walter C.; Bhattacharyya, Samit K.; Hanan, Nelson A.

Various programmatic considerations have delayed the upgrading of the TREAT reactor and the performance of the control system is not yet experimentally verified. The current schedule calls for the upgrading activities to occur last in the calendar year 1987. Detailed simulation results, coupled with earlier validation of individual components of the control strategy in TREAT, verify the performance of the algorithms. The control system operates within the safety envelope provided by a protection system designed to ensure reactor safety under conditions of spurious reactivity additions. The approach should be directly applicable to MMW systems, with appropriate accounting of temperature rate limitations of key components and of the inertia of the secondary system components.

Performance and microbial community composition dynamics of aerobic granular sludge from sequencing batch bubble column reactors operated at 20 degrees C, 30 degrees C, and 35 degrees C.

PubMed

Ebrahimi, Sirous; Gabus, Sébastien; Rohrbach-Brandt, Emmanuelle; Hosseini, Maryam; Rossi, Pierre; Maillard, Julien; Holliger, Christof

2010-07-01

Two bubble column sequencing batch reactors fed with an artificial wastewater were operated at 20 degrees C, 30 degrees C, and 35 degrees C. In a first stage, stable granules were obtained at 20 degrees C, whereas fluffy structures were observed at 30 degrees C. Molecular analysis revealed high abundance of the operational taxonomic unit 208 (OTU 208) affiliating with filamentous bacteria Leptothrix spp. at 30 degrees C, an OTU much less abundant at 20 degrees C. The granular sludge obtained at 20 degrees C was used for the second stage during which one reactor was maintained at 20 degrees C and the second operated at 30 degrees C and 35 degrees C after prior gradual increase of temperature. Aerobic granular sludge with similar physical properties developed in both reactors but it had different nutrient elimination performances and microbial communities. At 20 degrees C, acetate was consumed during anaerobic feeding, and biological phosphorous removal was observed when Rhodocyclaceae-affiliating OTU 214 was present. At 30 degrees C and 35 degrees C, acetate was mainly consumed during aeration and phosphorous removal was insignificant. OTU 214 was almost absent but the Gammaproteobacteria-affiliating OTU 239 was more abundant than at 20 degrees C. Aerobic granular sludge at all temperatures contained abundantly the OTUs 224 and 289 affiliating with Sphingomonadaceae indicating that this bacterial family played an important role in maintaining stable granular structures.

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

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.

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.

Online Oxide Contamination Measurement and Purification Demonstration

NASA Technical Reports Server (NTRS)

Bradley, D. E.; Godfroy, T. J.; Webster, K. L.; Garber, A. E.; Polzin, K. A.; Childers, D. J.

2011-01-01

Liquid metal sodium-potassium (NaK) has advantageous thermodynamic properties indicating its use as a fission reactor coolant for a surface (lunar, martian) power system. A major area of concern for fission reactor cooling systems is system corrosion due to oxygen contaminants at the high operating temperatures experienced. A small-scale, approximately 4-L capacity, simulated fission reactor cooling system employing NaK as a coolant was fabricated and tested with the goal of demonstrating a noninvasive oxygen detection and purification system. In order to generate prototypical conditions in the simulated cooling system, several system components were designed, fabricated, and tested. These major components were a fully-sealed, magnetically-coupled mechanical NaK pump, a graphite element heated reservoir, a plugging indicator system, and a cold trap. All system components were successfully demonstrated at a maximum system flow rate of approximately 150 cc/s at temperatures up to 550 C. Coolant purification was accomplished using a cold trap before and after plugging operations which showed a relative reduction in oxygen content.

Degradation of pharmaceuticals in UV (LP)/H₂O₂ reactors simulated by means of kinetic modeling and computational fluid dynamics (CFD).

PubMed

Wols, B A; Harmsen, D J H; Wanders-Dijk, J; Beerendonk, E F; Hofman-Caris, C H M

2015-05-15

UV/H2O2 treatment is a well-established technique to degrade organic micropollutants. A CFD model in combination with an advanced kinetic model is presented to predict the degradation of organic micropollutants in UV (LP)/H2O2 reactors, accounting for the hydraulics, fluence rate, complex (photo)chemical reactions in the water matrix and the interactions between these processes. The model incorporates compound degradation by means of direct UV photolysis, OH radical and carbonate radical reactions. Measurements of pharmaceutical degradations in pilot-scale UV/H2O2 reactors are presented under different operating conditions. A comparison between measured and modeled degradation for a group of 35 pharmaceuticals resulted in good model predictions for most of the compounds. The research also shows that the degradation of organic micropollutants can be dependent on temperature, which is relevant for full-scale installations that are operated at different temperatures over the year. Copyright © 2015 Elsevier Ltd. All rights reserved.

Preliminary analysis of hot spot factors in an advanced reactor for space electric power systems

NASA Technical Reports Server (NTRS)

Lustig, P. H.; Holms, A. G.; Davison, H. W.

1973-01-01

The maximum fuel pin temperature for nominal operation in an advanced power reactor is 1370 K. Because of possible nitrogen embrittlement of the clad, the fuel temperature was limited to 1622 K. Assuming simultaneous occurrence of the most adverse conditions a deterministic analysis gave a maximum fuel temperature of 1610 K. A statistical analysis, using a synthesized estimate of the standard deviation for the highest fuel pin temperature, showed probabilities of 0.015 of that pin exceeding the temperature limit by the distribution free Chebyshev inequality and virtually nil assuming a normal distribution. The latter assumption gives a 1463 K maximum temperature at 3 standard deviations, the usually assumed cutoff. Further, the distribution and standard deviation of the fuel-clad gap are the most significant contributions to the uncertainty in the fuel temperature.

High temperature ceramic composition for hydrogen retention

DOEpatents

Webb, R.W.

1974-01-01

A ceramic coating for H retention in fuel elements is described. The coating has relatively low thermal neutron cross section, is not readily reduced by H at 1500 deg F, is adherent to the fuel element base metal, and is stable at reactor operating temperatures. (JRD)

Effect of process parameters on greenhouse gas generation by wastewater treatment plants.

PubMed

Yerushalmi, L; Shahabadi, M Bani; Haghighat, F

2011-05-01

The effect of key process parameters on greenhouse gas (GHG) emission by wastewater treatment plants was evaluated, and the governing parameters that exhibited major effects on the overall on- and off-site GHG emissions were identified. This evaluation used aerobic, anaerobic, and hybrid anaerobic/aerobic treatment systems with food processing industry wastewater. The operating temperature of anaerobic sludge digester was identified to have the highest effect on GHG generation in the aerobic treatment system. The total GHG emissions of 2694 kg CO2e/d were increased by 72.5% with the increase of anaerobic sludge digester temperature from 20 to 40 degrees C. The operating temperature of the anaerobic reactor was the dominant controlling parameter in the anaerobic and hybrid treatment systems. Raising the anaerobic reactor's temperature from 25 to 40 degrees C increased the total GHG emissions from 5822 and 6617 kg CO2e/d by 105.6 and 96.5% in the anaerobic and hybrid treatment systems, respectively.

Investigation of the effects of radiolytic-gas bubbles on the long-term operation of solution reactors for medical-isotope production

NASA Astrophysics Data System (ADS)

Souto Mantecon, Francisco Javier

One of the most common and important medical radioisotopes is 99Mo, which is currently produced using the target irradiation technology in heterogeneous nuclear reactors. The medical isotope 99Mo can also be produced from uranium fission using aqueous homogeneous solution reactors. In solution reactors, 99Mo is generated directly in the fuel solution, resulting in potential advantages when compared with the target irradiation process in heterogeneous reactors, such as lower reactor power, less waste heat, and reduction by a factor of about 100 in the generation of spent fuel. The commercial production of medical isotopes in solution reactors requires steady-state operation at about 200 kW. At this power regime, the formation of radiolytic-gas bubbles creates a void volume in the fuel solution that introduces a negative coefficient of reactivity, resulting in power reduction and instabilities that may impede reactor operation for medical-isotope production. A model has been developed considering that reactivity effects are due to the increase in the fuel-solution temperature and the formation of radiolytic-gas bubbles. The model has been validated against experimental results from the Los Alamos National Laboratory uranyl fluoride Solution High-Energy Burst Assembly (SHEBA), and the SILENE uranyl nitrate solution reactor, commissioned at the Commissariat a l'Energie Atomique, in Valduc, France. The model shows the feasibility of solution reactors for the commercial production of medical isotopes and reveals some of the important parameters to consider in their design, including the fuel-solution type, 235U enrichment, uranium concentration, reactor vessel geometry, and neutron reflectors surrounding the reactor vessel. The work presented herein indicates that steady-state operation at 200 kW can be achieved with a solution reactor consisting of 120 L of uranyl nitrate solution enriched up to 20% with 235U and a uranium concentration of 145 kg/m3 in a graphite-reflected cylindrical geometry.

Performance characteristics of anaerobic downflow stationary fixed film reactors

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

van den Berg, L.; Kennedy, K.J.

1982-01-01

Stationary fixed film reactors operated to ensure a net downflow of substrate have several characteristics different from other retained biomass reactors. The active biomass attaches itself to stationary surface and hence is difficult to wash out. Performance is related to the surface-to-volume of the film support as well as to the composition of the support. Methane production rates of up to 8 cym day at loading rates of up to 30 kg COD/m cym day, are possible. Severe hydraulic and organic overloadings can be tolerated with operation back to normal 24 hours following cessation of mistreatment. Reactors can operate withmore » dilute and concentrated wastes (4000-130,000 mg COD/L) and can change readily over from one waste to another. Intermittent loading at high loading rates are possible. Methane production rates and loading rates decreased linearly with temperature (35) to 10); at 10 C they were about 20% of those at 35 C.« less

Phenol removal from hypersaline wastewaters in a Membrane Biological Reactor (MBR): operation and microbiological characterisation.

PubMed

Dosta, J; Nieto, J M; Vila, J; Grifoll, M; Mata-Álvarez, J

2011-03-01

In this study, two Membrane Biological Reactors (MBR) with submerged flat membranes, one at lab-scale conditions and the other at pilot-plant conditions, were operated at environmental temperature to treat an industrial wastewater characterised by low phenol concentrations (8-16 mg L(-1)) and high salinity (∼ 150-160 mS cm(-1)). During the operation of both reactors, the phenol loading rate was progressively increased and less than 1mg phenol L(-1) was detected even at very low HRTs (0.5-0.7 days). Membrane fouling was minimized by the cross flow aeration rate inside the MBRs and by intermittent permeation. Microbial community analysis of both reactors revealed that members of the genera Halomonas and Marinobacter (gammaproteobacteria) were major components. Growth-linked phenol degradation by pure cultures of Marinobacter isolates demonstrated that this bacterium played a major role in the removal of phenol from the bioreactors. Copyright © 2010 Elsevier Ltd. All rights reserved.

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

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

VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS

DOEpatents

Furgerson, W.T.

1963-12-17

A hollow, porous-walled fuel element filled with fissionable fuel and provided with an outlet port through its wall is described. In operation in a gas-cooled reactor, the element is connected, through its outlet port, to the vacuum side of a pump that causes a portion of the coolant gas flowing over the exterior surface of the element to be drawn through the porous walls thereof and out through the outlet port. This continuous purging gas flow sweeps away gaseous fission products as they are released by the fissioning fuel. (AEC) A fuel element for a nuclear reactor incorporating a body of metal of melting point lower than the temperature of operation of the reactor and a nuclear fuel in finely divided form dispersed in the body of metal as a settled slurry is presented. (AEC)

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.

Thermal Analysis of ZPPR High Pu Content Stored Fuel

DOE PAGES

Solbrig, Charles W.; Pope, Chad L.; Andrus, Jason P.

2014-09-17

The Zero Power Physics Reactor (ZPPR) operated from April 18, 1969, until 1990. ZPPR operated at low power for testing nuclear reactor designs. This paper examines the temperature of Pu content ZPPR fuel while it is in storage. Heat is generated in the fuel due to Pu and Am decay and is a concern for possible cladding damage. Damage to the cladding could lead to fuel hydriding and oxidizing. A series of computer simulations were made to determine the range of temperatures potentially occuring in the ZPPR fuel. The maximum calculated fuel temperature is 292°C (558°F). Conservative assumptions in themore » model intentionally overestimate temperatures. The stored fuel temperatures are dependent on the distribution of fuel in the surrounding storage compartments, the heat generation rate of the fuel, and the orientation of fuel. Direct fuel temperatures could not be measured but storage bin doors, storage sleeve doors, and storage canister temperatures were measured. Comparison of these three temperatures to the calculations indicates that the temperatures calculated with conservative assumptions are, as expected, higher than the actual temperatures. The maximum calculated fuel temperature with the most conservative assumptions is significantly below the fuel failure criterion of 600°C (1,112°F).« less

U.S./CIS eye joint nuclear rocket venture

NASA Technical Reports Server (NTRS)

Clark, John S.; Mcilwain, Melvin C.; Smetanikov, Vladimir; D'Yakov, Evgenij K.; Pavshuk, Vladimir A.

1993-01-01

An account is given of the significance for U.S. spacecraft development of a nuclear thermal rocket (NTR) reactor concept that has been developed in the (formerly Soviet) Commonwealth of Independent States (CIS). The CIS NTR reactor employs a hydrogen-cooled zirconium hydride moderator and ternary carbide fuels; the comparatively cool operating temperatures associated with this design promise overall robustness.

Automatic coolant flow control device for a nuclear reactor assembly

DOEpatents

Hutter, E.

1984-01-27

A device which controls coolant flow through a nuclear reactor assembly comprises a baffle means at the exit end of said assembly having a plurality of orifices, and a bimetallic member in operative relation to the baffle means such that at increased temperatures said bimetallic member deforms to unblock some of said orifices and allow increased coolant flow therethrough.

Automatic coolant flow control device for a nuclear reactor assembly

DOEpatents

Hutter, Ernest

1986-01-01

A device which controls coolant flow through a nuclear reactor assembly comprises a baffle means at the exit end of said assembly having a plurality of orifices, and a bimetallic member in operative relation to the baffle means such that at increased temperatures said bimetallic member deforms to unblock some of said orifices and allow increased coolant flow therethrough.

Isotopic composition and neutronics of the Okelobondo natural reactor

NASA Astrophysics Data System (ADS)

Palenik, Christopher Samuel

The Oklo-Okelobondo and Bangombe uranium deposits, in Gabon, Africa host Earth's only known natural nuclear fission reactors. These 2 billion year old reactors represent a unique opportunity to study used nuclear fuel over geologic periods of time. The reactors in these deposits have been studied as a means by which to constrain the source term of fission product concentrations produced during reactor operation. The source term depends on the neutronic parameters, which include reactor operation duration, neutron flux and the neutron energy spectrum. Reactor operation has been modeled using a point-source computer simulation (Oak Ridge Isotope Generation and Depletion, ORIGEN, code) for a light water reactor. Model results have been constrained using secondary ionization mass spectroscopy (SIMS) isotopic measurements of the fission products Nd and Te, as well as U in uraninite from samples collected in the Okelobondo reactor zone. Based upon the constraints on the operating conditions, the pre-reactor concentrations of Nd (150 ppm +/- 75 ppm) and Te (<1 ppm) in uraninite were estimated. Related to the burnup measured in Okelobondo samples (0.7 to 13.8 GWd/MTU), the final fission product inventories of Nd (90 to 1200 ppm) and Te (10 to 110 ppm) were calculated. By the same means, the ranges of all other fission products and actinides produced during reactor operation were calculated as a function of burnup. These results provide a source term against which the present elemental and decay abundances at the fission reactor can be compared. Furthermore, they provide new insights into the extent to which a "fossil" nuclear reactor can be characterized on the basis of its isotopic signatures. In addition, results from the study of two other natural systems related to the radionuclide and fission product transport are included. A detailed mineralogical characterization of the uranyl mineralogy at the Bangombe uranium deposit in Gabon, Africa was completed to improve geochemical models of the solubility-limiting phase. A study of the competing effects of radiation damage and annealing in a U-bearing crystal of zircon shows that low temperature annealing in actinide-bearing phases is significant in the annealing of radiation damage.

Temperature, inocula and substrate: Contrasting electroactive consortia, diversity and performance in microbial fuel cells.

PubMed

Heidrich, E S; Dolfing, J; Wade, M J; Sloan, W T; Quince, C; Curtis, T P

2018-02-01

The factors that affect microbial community assembly and its effects on the performance of bioelectrochemical systems are poorly understood. Sixteen microbial fuel cell (MFC) reactors were set up to test the importance of inoculum, temperature and substrate: Arctic soil versus wastewater as inoculum; warm (26.5°C) versus cold (7.5°C) temperature; and acetate versus wastewater as substrate. Substrate was the dominant factor in determining performance and diversity: unexpectedly the simple electrogenic substrate delivered a higher diversity than a complex wastewater. Furthermore, in acetate fed reactors, diversity did not correlate with performance, yet in wastewater fed ones it did, with greater diversity sustaining higher power densities and coulombic efficiencies. Temperature had only a minor effect on power density, (Q 10 : 2 and 1.2 for acetate and wastewater respectively): this is surprising given the well-known temperature sensitivity of anaerobic bioreactors. Reactors were able to operate at low temperature with real wastewater without the need for specialised inocula; it is speculated that MFC biofilms may have a self-heating effect. Importantly, the warm acetate fed reactors in this study did not act as direct model for cold wastewater fed systems. Application of this technology will encompass use of real wastewater at ambient temperatures. Copyright © 2017 The Authors. Published by Elsevier B.V. All rights reserved.

Applications of plasma core reactors to terrestrial energy systems

NASA Technical Reports Server (NTRS)

Latham, T. S.; Biancardi, F. R.; Rodgers, R. J.

1974-01-01

Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrial applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times.-

Autoheated thermophilic aerobic digestion

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

Deeny, K.; Hahn, H.; Leonhard, D.

1991-10-01

Autothermal thermophilic aerobic digestion (ATAD) is first and foremost a digestion process, the primary purpose of which is to decompose a portion of the waste organic solids generated from wastewater treatment. As a result of the high operating temperature, digestion is expected to occur within a short time period (6 days) and accomplish a high degree of pathogen reduction. ATAD systems are two-stage aerobic digestion processes that operate under thermophilic temperature conditions (40 to 80C) without supplemental heat. Like composting, the systems rely on the conservation of heat released during digestion itself to attain and sustain the desired operating temperature.more » Typical ATAD systems operate at 55C and may reach temperatures of 60 to 65C in the second-stage reactor. Perhaps because of the high operating temperature, this process has been referred to as Liquid Composting.' Major advantages associated with thermophilic operation include high biological reaction rates and a substantial degree of pathogen reduction.« less

Distributed optical fibre temperature measurements in a low dose rate radiation environment based on Rayleigh backscattering

NASA Astrophysics Data System (ADS)

Faustov, A.; Gussarov, A.; Wuilpart, M.; Fotiadi, A. A.; Liokumovich, L. B.; Kotov, O. I.; Zolotovskiy, I. O.; Tomashuk, A. L.; Deschoutheete, T.; Mégret, P.

2012-04-01

On-line monitoring of environmental conditions in nuclear facilities is becoming a more and more important problem. Standard electronic sensors are not the ideal solution due to radiation sensitivity and difficulties in installation of multiple sensors. In contrast, radiation-hard optical fibres can sustain very high radiation doses and also naturally offer multi-point or distributed monitoring of external perturbations. Multiple local electro-mechanical sensors can be replaced by just one measuring fibre. At present, there are over four hundred operational nuclear power plants (NPPs) in the world 1. Operating experience has shown that ineffective control of the ageing degradation of major NPP components can threaten plant safety and also plant life. Among those elements, cables are vital components of I&C systems in NPPs. To ensure their safe operation and predict remaining life, environmental monitoring is necessary. In particular, temperature and radiation dose are considered to be the two most important parameters. The aim of this paper is to assess experimentally the feasibility of optical fibre temperature measurements in a low doserate radiation environment, using a commercially available reflectometer based on Rayleigh backscattering. Four different fibres were installed in the Sub-Pile Room of the BR2 Material testing nuclear reactor in Mol, Belgium. This place is man-accessible during the reactor shut-down, allowing easy fibre installation. When the reactor operates, the dose-rates in the room are in a range 0.005-5 Gy/h with temperatures of 40-60 °C, depending on the location. Such a surrounding is not much different to some "hot" environments in NPPs, where I&C cables are located.

Ultra-High Temperature ContinuousReactors based on Electro-thermal FluidizedBed Concept

DOE PAGES

Fedorov, Sergiy S.; Rohatgi, Upendra Singh; Barsukov, Igor V.; ...

2015-12-08

This paper presents the results of research and development in high-temperature (i.e. 2,000- 3,000ºС) continuous furnaces operating on the principle of electro-thermal fluidized bed for the purification of recycled, finely sized carbon materials. The basis of this fluidized bed furnace is specific electrical resistance and a new correlation has been developed to predict specific electrical resistance for the natural graphite-based precursors entering the fluidized bed reactor This correlation has been validated with the data from a fully functional pilot furnace whose throughput capacity is 10 kg per hour built as part of this work. Data collected in the course ofmore » graphite refining experiments demonstrated that difference between the calculated and measured values of specific electrical resistance of fluidized bed does not exceed 25%. It was concluded that due to chaotic nature of electro-thermal fluidized bed reactors this discrepancy is acceptable. The fluid mechanics of the three types of operating regimes, have been described. The numerical relationships obtained as part of this work allowed proposing an algorithm for selection of technological operational modes with large- scale high-temperature furnaces rated for throughputs of several tons of product per hour. Optimizations proposed now allow producing natural graphite-based end product with the purity level of 99.98+ wt%C which is the key passing criteria for applications in the advanced battery markets.« less

Remediation of 1,2,3-trichlorobenzene contaminated soil using a combined thermal desorption-molten salt oxidation reactor system.

PubMed

Li, Jin-hui; Sun, Xiao-fei; Yao, Zhi-tong; Zhao, Xiang-yang

2014-02-01

A combined thermal desorption (TD)-molten salt oxidation (MSO) reactor system was applied to remediate the 1,2,3-trichlorobenzene (1,2,3-TCB) contaminated soil. The TD reactor was used to enrich the contaminant from soil, and its dechlorination of the contaminant was achieved in the MSO reactor. The optimum operating conditions of TD, and the effects of MSO reactor temperatures, additive amounts of the TCB on destruction and removal efficiency (DRE) of TCB and chlorine retention efficiency (CRE) were investigated. The reaction mechanism and pathway were proposed as well. The combined system could remediate the contaminated soil at a large scale of concentration from 5 to 25gkg(-1), and the DRE and CRE reached more than 99% and 95%, respectively, at temperatures above 850°C. The reaction emissions included C6H6, CH4, CO and CO2, and chlorinated species were not detected. It was found that a little increase in the temperature can considerably reduce the emission of C6H6, CH4, and CO, while the CO2 level increased. Copyright © 2014. Published by Elsevier Ltd.

Commercial-Scale Demonstration of the Liquid Phase Methanol (LPMEOH) Process

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

None

he Liquid Phase Methanol (LPMEOW) Demonstration Project at Kingsport Tennessee, is a $213.7 million cooperative agreement between the U.S. Department of Energy (DOE) and Air Products Liquid Phase Conversion Company, L.P. (the Partnership) to produce methanol from coal-derived synthesis gas (syngas). Air Products and Chemicals, Inc. (Air Products) and Eastman Chemical Company (Eastman) formed the Partnership to execute the Demonstration Project. The LPMEOEP Process Demonstration Unit was built at a site located at the Eastman coal-to-chemicals complex in Kingsport. The LPMEOHW Demonstration Facility completed its first year of operation on 02 April 1998. The LPMEOW Demonstration Facility also completed themore » longest continuous operating run (65 days) on 21 April 1998. Catalyst activity, as defined by the ratio of the rate constant at any point in time to the rate constant for freshly reduced catalyst (as determined in the laboratory autoclave), was monitored throughout the reporting period. During a six-week test at a reactor temperature of 225oC and Balanced Gas flowrate of 700 KSCFH, the rate of decline in catalyst activity was steady at 0.29-0.36% per day. During a second one-month test at a reactor temperature of 220oC and a Balanced Gas flowrate of 550-600 KSCFH, the rate of decline in catalyst activity was 0.4% per day, which matched the pefiorrnance at 225"C, as well as the 4-month proof-of-concept run at the LaPorte AFDU in 1988/89. Beginning on 08 May 1998, the LPMEOW Reactor temperature was increased to 235oC, which was the operating temperature tier the December 1997 restart with the fresh charge of catalyst (50'Yo of design loading). The flowrate of the primary syngas feed stream (Balanced Gas) was also increased to 700-750 KSCFH. During two stable operating periods between 08 May and 09 June 1998, the average catalyst deactivation rate was 0.8% per day. Due to the scatter of the statistical analysis of the results, this test was extended to better quanti& the catalyst aging behavior. During the reporting perio~ two batches of fresh catalyst were activated and transferred to the reactor (on 02 April and 20 June 1998). The weight of catalyst in the LPMEOW Reactor has reached 80% of the design value. At the end of the reporting period, a step-change in the pressure-drop profile within the LPMEOW Reactor and an increase in the pressure of the steam system which provides cooling to the LPMEOW Reactor were observed. No change in the calculated activity of the catalyst was detected during either of these transients. These parameters will be monitored closely for any additional changes.« less

Heat and fuel coupled operation of a high temperature polymer electrolyte fuel cell with a heat exchanger methanol steam reformer

NASA Astrophysics Data System (ADS)

Schuller, G.; Vázquez, F. Vidal; Waiblinger, W.; Auvinen, S.; Ribeirinha, P.

2017-04-01

In this work a methanol steam reforming (MSR) reactor has been operated thermally coupled to a high temperature polymer electrolyte fuel cell stack (HT-PEMFC) utilizing its waste heat. The operating temperature of the coupled system was 180 °C which is significantly lower than the conventional operating temperature of the MSR process which is around 250 °C. A newly designed heat exchanger reformer has been developed by VTT (Technical Research Center of Finland LTD) and was equipped with commercially available CuO/ZnO/Al2O3 (BASF RP-60) catalyst. The liquid cooled, 165 cm2, 12-cell stack used for the measurements was supplied by Serenergy A/S. The off-heat from the electrochemical fuel cell reaction was transferred to the reforming reactor using triethylene glycol (TEG) as heat transfer fluid. The system was operated up to 0.4 A cm-2 generating an electrical power output of 427 Wel. A total stack waste heat utilization of 86.4% was achieved. It has been shown that it is possible to transfer sufficient heat from the fuel cell stack to the liquid circuit in order to provide the needed amount for vaporizing and reforming of the methanol-water-mixture. Furthermore a set of recommendations is given for future system design considerations.

Catalytic combustion of residual fuels

NASA Technical Reports Server (NTRS)

Bulzan, D. L.; Tacina, R. R.

1981-01-01

A noble metal catalytic reactor was tested using two grades of petroleum derived residual fuels at specified inlet air temperatures, pressures, and reference velocities. Combustion efficiencies greater than 99.5 percent were obtained. Steady state operation of the catalytic reactor required inlet air temperatures of at least 800 K. At lower inlet air temperatures, upstream burning in the premixing zone occurred which was probably caused by fuel deposition and accumulation on the premixing zone walls. Increasing the inlet air temperature prevented this occurrence. Both residual fuels contained about 0.5 percent nitrogen by weight. NO sub x emissions ranged from 50 to 110 ppm by volume at 15 percent excess O2. Conversion of fuel-bound nitrogen to NO sub x ranged from 25 to 50 percent.

Analysis of boron dilution in a four-loop PWR

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

Sun, J.G.; Sha, W.T.

1995-03-01

Thermal mixing and boron dilution in a pressurized water reactor were analyzed with COMMIX codes. The reactor system was the four-loop Zion reactor. Two boron dilution scenarios were analyzed. In the first scenario, the plant is in cold shutdown and the reactor coolant system has just been filled after maintenance on the steam generators. To flush the air out of the steam generator tubes, a reactor coolant pump (RCP) is started, with the water in the pump suction line devoid of boron and at the same temperature as the coolant in the system. In the second scenario, the plant ismore » at hot standby and the reactor coolant system has been heated to operating temperature after a long outage. It is assumed that an RCP is started, with the pump suction line filled with cold unborated water, forcing a slug of diluted coolant down the downcomer and subsequently through the reactor core. The subsequent transient thermal mixing and boron dilution that would occur in the reactor system is simulated for these two scenarios. The reactivity insertion rate and the total reactivity are evaluated and a sensitivity study is performed to assess the accuracy of the numerical modeling of the geometry of the reactor coolant system.« less

Low temperature conversion of plastic waste into light hydrocarbons.

PubMed

Shah, Sajid Hussain; Khan, Zahid Mahmood; Raja, Iftikhar Ahmad; Mahmood, Qaisar; Bhatti, Zulfiqar Ahmad; Khan, Jamil; Farooq, Ather; Rashid, Naim; Wu, Donglei

2010-07-15

Advance recycling through pyrolytic technology has the potential of being applied to the management of plastic waste (PW). For this purpose 1 l volume, energy efficient batch reactor was manufactured locally and tested for pyrolysis of waste plastic. The feedstock for reactor was 50 g waste polyethylene. The average yield of the pyrolytic oil, wax, pyrogas and char from pyrolysis of PW were 48.6, 40.7, 10.1 and 0.6%, respectively, at 275 degrees C with non-catalytic process. Using catalyst the average yields of pyrolytic oil, pyrogas, wax and residue (char) of 50 g of PW was 47.98, 35.43, 16.09 and 0.50%, respectively, at operating temperature of 250 degrees C. The designed reactor could work at low temperature in the absence of a catalyst to obtain similar products as for a catalytic process. 2010 Elsevier B.V. All rights reserved.

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.

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

Control of autothermal reforming reactor of diesel fuel

NASA Astrophysics Data System (ADS)

Dolanc, Gregor; Pregelj, Boštjan; Petrovčič, Janko; Pasel, Joachim; Kolb, Gunther

2016-05-01

In this paper a control system for autothermal reforming reactor for diesel fuel is presented. Autothermal reforming reactors and the pertaining purification reactors are used to convert diesel fuel into hydrogen-rich reformate gas, which is then converted into electricity by the fuel cell. The purpose of the presented control system is to control the hydrogen production rate and the temperature of the autothermal reforming reactor. The system is designed in such a way that the two control loops do not interact, which is required for stable operation of the fuel cell. The presented control system is a part of the complete control system of the diesel fuel cell auxiliary power unit (APU).

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

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

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

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

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

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 use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 LiF-BeF 2) 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. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologiesmore » include TRISO particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Several preconceptual and conceptual design efforts that have been conducted on FHR concepts bear a significant influence on the FHR DR design. Specific designs include the Oak Ridge National Laboratory (ORNL) advanced high-temperature reactor (AHTR) with 3400/1500 MWt/megawatts of electric output (MWe), as well as a 125 MWt small modular AHTR (SmAHTR) from ORNL. Other important examples are the Mk1 pebble bed FHR (PB-FHR) concept from the University of California, Berkeley (UCB), and an FHR test reactor design developed at the Massachusetts Institute of Technology (MIT). The MIT FHR test reactor is based on a prismatic fuel platform and is directly relevant to the present FHR DR design effort. These FHR concepts are based on reasonable assumptions for credible commercial prototypes. The FHR DR concept also directly benefits from the operating experience of the Molten Salt Reactor Experiment (MSRE), as well as the detailed design efforts for a large molten salt reactor concept and its breeder variant, the Molten Salt Breeder Reactor. The FHR DR technology is most representative of the 3400 MWt AHTR concept, and it will demonstrate key operational features of that design. The FHR DR will be closely scaled to the SmAHTR concept in power and flows, so any technologies demonstrated will be directly applicable to a reactor concept of that size. The FHR DR is not a commercial prototype design, but rather a DR that serves a cost and risk mitigation function for a later commercial prototype. It is expected to have a limited operational lifetime compared to a commercial plant. It is designed to be a low-cost reactor compared to more mature advanced prototype DRs. A primary reason to build the FHR DR is to learn about salt reactor technologies and demonstrate solutions to remaining technical gaps.« less

Preliminary Analysis of SiC BWR Channel Box Performance under Normal Operation

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

Wirth, Brian; Singh, Gyanender P.; Gorton, Jacob

SiC-SiC composites are being considered for applications in the core components, including BWR channel box and fuel rod cladding, of light water reactors to improve accident tolerance. In the extreme nuclear reactor environment, core components like the BWR channel box will be exposed to neutron damage and a corrosive environment. To ensure reliable and safe operation of a SiC channel box, it is important to assess its deformation behavior under in-reactor conditions including the expected neutron flux and temperature distributions. In particular, this work has evaluated the effect of non-uniform dimensional changes caused by spatially varying neutron flux and temperaturesmore » on the deformation behavior of the channel box over the course of one cycle of irradiation. These analyses have been performed using the fuel performance modeling code BISON and the commercial finite element analysis code Abaqus, based on fast flux and temperature boundary conditions have been calculated using the neutronics and thermal-hydraulics codes Serpent2 and COBRA-TF, respectively. The dependence of dimensions and thermophysical properties on fast flux and temperature has been incorporated into the material models. These initial results indicate significant bowing of the channel box with a lateral displacement greater than 6.5mm. The channel box bowing behavior is time dependent, and driven by the temperature dependence of the SiC irradiation-induced swelling and the neutron flux/fluence gradients. The bowing behavior gradually recovers during the course of the operating cycle as the swelling of the SiC-SiC material saturates. However, the bending relaxation due to temperature gradients does not fully recover and residual bending remains after the swelling saturates in the entire channel box.« less

Emulation of reactor irradiation damage using ion beams

DOE PAGES

Was, G. S.; Jiao, Z.; Getto, E.; ...

2014-06-14

The continued operation of existing light water nuclear reactors and the development of advanced nuclear reactor depend heavily on understanding how damage by radiation to levels degrades materials that serve as the structural components in reactor cores. The first high dose ion irradiation experiments on a ferritic-martensitic steel showing that ion irradiation closely emulates the full radiation damage microstructure created in-reactor are described. Ferritic-martensitic alloy HT9 (heat 84425) in the form of a hexagonal fuel bundle duct (ACO-3) accumulated 155 dpa at an average temperature of 443°C in the Fast Flux Test Facility (FFTF). Using invariance theory as a guide,more » irradiation of the same heat was conducted using self-ions (Fe++) at 5 MeV at a temperature of 460°C and to a dose of 188 displacements per atom. The void swelling was nearly identical between the two irradiation and the size and density of precipitates and loops following ion irradiation are within a factor of two of those for neutron irradiation. The level of agreement across all of the principal microstructure changes between ion and reactor irradiation establishes the capability of tailoring ion irradiation to emulate the reactor-irradiated microstructure.« less

The prototype fast reactor at Dounreay, Scotland. Process and engineering development for sodium removal

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

Mann, A.; Herrick, R.; Gunn, J.

2007-07-01

Dounreay was home to commercial fast reactor development in the UK. Following the construction and operation of the Dounreay Fast Reactor, a sodium-cooled Prototype Fast Reactor (PFR), was constructed. PFR started operating in 1974, closed in 1994 and is presently being decommissioned. To date the bulk of the sodium has been removed and treated. Due to the design of the existing extraction system however, a sodium pool will remain in the heel of the reactor. To remove this sodium, a pump/camera system was developed, tested and deployed. The Water Vapour Nitrogen (WVN) process has been selected to allow removal ofmore » the final sodium residues from the reactor. Due to the design of the reactor and potential for structural damage should Normal WVN (which produces hydrated sodium hydroxide) be used, Low Concentration WVN (LC WVN) has been developed. Pilot scale testing has shown that it is possible treat the reactor within 18 months at a WVN concentration of up to 4% v/v and temperature of 120 deg. C. At present the equipment that will be used to apply LC WVN to the reactor is being developed at the detail design stage. and is expected to be deployed within the next few years. (authors)« less

Evaluation of decentralized treatment of sewage employing Upflow Septic Tank/Baffled Reactor (USBR) in developing countries.

PubMed

Sabry, Tarek

2010-02-15

A new concept for a low-cost modified septic tank, named Upflow Septic Tank/Baffled Reactor (USBR), was constructed and tested in a small village in Egypt. During almost one year of continuous operation and monitoring, this system was found to have very satisfactory removal results, where the average results of COD, BOD, and TSS removal efficiencies were 84%, 81%, and 89%, respectively, and the results of the experiment proved that the second compartment (Anaerobic Baffled Reactor) was the main treatment unit in removing the pollutants during the start-up period and at the very early steady-state stage. However, after this period and during the steady-state operation conditions, the second compartment served as a polishing step. Also, it was observed that the USBR system was not affected by the imposed shock loads at the peak flow and organic periods. The results showed that the system is slightly influenced by the drop in the temperature. Decreasing in BOD and COD removal by factor of 9% was observed, when temperature decreases from the average of 35 degrees C in summer time (for the first 127 days) to the average of 22 degrees C in winter time (between day 252 and day 280). Whereas, the TSS removals were not affected by the drop in temperature. The results of the sewage flow variations during one year of operation were compared with Goodrich Formula to see the applicability of this equation in rural developing countries. MAIN FINDING OF THE WORK: The Upflow Septic Tank/Baffled Reactor system could become a promising alternative to the conventional treatment plants in rural developing countries.

Ozone generation by negative direct current corona discharges in dry air fed coaxial wire-cylinder reactors

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

Yehia, Ashraf; Mizuno, Akira

An analytical study was made in this paper for calculating the ozone generation by negative dc corona discharges. The corona discharges were formed in a coaxial wire-cylinder reactor. The reactor was fed by dry air flowing with constant rates at atmospheric pressure and room temperature, and stressed by a negative dc voltage. The current-voltage characteristics of the negative dc corona discharges formed inside the reactor were measured in parallel with concentration of the generated ozone under different operating conditions. An empirical equation was derived from the experimental results for calculating the ozone concentration generated inside the reactor. The results, thatmore » have been recalculated by using the derived equation, have agreed with the experimental results over the whole range of the investigated parameters, except in the saturation range for the ozone concentration. Therefore, the derived equation represents a suitable criterion for expecting the ozone concentration generated by negative dc corona discharges in dry air fed coaxial wire-cylinder reactors under any operating conditions in range of the investigated parameters.« less

Control rod calibration and reactivity effects at the IPEN/MB-01 reactor

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

Pinto, Letícia Negrão; Gonnelli, Eduardo; Santos, Adimir dos

2014-11-11

Researches that aim to improve the performance of neutron transport codes and quality of nuclear cross section databases are very important to increase the accuracy of simulations and the quality of the analysis and prediction of phenomena in the nuclear field. In this context, relevant experimental data such as reactivity worth measurements are needed. Control rods may be made of several neutron absorbing materials that are used to adjust the reactivity of the core. For the reactor operation, these experimental data are also extremely important: with them it is possible to estimate the reactivity worth by the movement of themore » control rod, understand the reactor response at each rod position and to operate the reactor safely. This work presents a temperature correction approach for the control rod calibration problem. It is shown the control rod calibration data of the IPEN/MB-01 reactor, the integral and differential reactivity curves and a theoretical analysis, performed by the MCNP-5 reactor physics code, developed and maintained by Los Alamos National Laboratory, using the ENDF/B-VII.0 nuclear data library.« 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

SIMULTANEOUS DIFFERENTIAL EQUATION COMPUTER

DOEpatents

Collier, D.M.; Meeks, L.A.; Palmer, J.P.

1960-05-10

A description is given for an electronic simulator for a system of simultaneous differential equations, including nonlinear equations. As a specific example, a homogeneous nuclear reactor system including a reactor fluid, heat exchanger, and a steam boiler may be simulated, with the nonlinearity resulting from a consideration of temperature effects taken into account. The simulator includes three operational amplifiers, a multiplier, appropriate potential sources, and interconnecting R-C networks.

Initial conceptual design study of self-critical nuclear pumped laser systems

NASA Technical Reports Server (NTRS)

Rodgers, R. J.

1979-01-01

An analytical study of self-critical nuclear pumped laser system concepts was performed. Primary emphasis was placed on reactor concepts employing gaseous uranium hexafluoride (UF6) as the fissionable material. Relationships were developed between the key reactor design parameters including reactor power level, critical mass, neutron flux level, reactor size, operating pressure, and UF6 optical properties. The results were used to select a reference conceptual laser system configuration. In the reference configuration, the 3.2 m cubed lasing volume is surrounded by a graphite internal moderator and a region of heavy water. Results of neutronics calculations yield a critical mass of 4.9 U(235) in the form (235)UF6. The configuration appears capable of operating in a continuous steady-state mode. The average gas temperature in the core is 600 K and the UF6 partial pressure within the lasing volume is 0.34 atm.

Safety considerations in testing a fuel-rich aeropropulsion gas generator

NASA Technical Reports Server (NTRS)

Rollbuhler, R. James; Hulligan, David D.

1991-01-01

A catalyst containing reactor is being tested using a fuel-rich mixture of Jet A fuel and hot input air. The reactor product is a gaseous fuel that can be utilized in aeropropulsion gas turbine engines. Because the catalyst material is susceptible to damage from high temperature conditions, fuel-rich operating conditions are attained by introducing the fuel first into an inert gas stream in the reactor and then displacing the inert gas with reaction air. Once a desired fuel-to-air ratio is attained, only limited time is allowed for a catalyst induced reaction to occur; otherwise the inert gas is substituted for the air and the fuel flow is terminated. Because there presently is not a gas turbine combustor in which to burn the reactor product gas, the gas is combusted at the outlet of the test facility flare stack. This technique in operations has worked successfully in over 200 tests.

Low cost silicon solar array project silicon materials task: Establishment of the feasibility of a process capable of low-cost, high volume production of silane (step 1) and the pyrolysis of silane to semiconductor-grade silicon (step 2)

NASA Technical Reports Server (NTRS)

Breneman, W. C.; Cheung, H.; Farrier, E. G.; Morihara, H.

1977-01-01

A quartz fluid bed reactor capable of operating at temperatures of up to 1000 C was designed, constructed, and successfully operated. During a 30 minute experiment, silane was decomposed within the reactor with no pyrolysis occurring on the reactor wall or on the gas injection system. A hammer mill/roller-crusher system appeared to be the most practical method for producing seed material from bulk silicon. No measurable impurities were detected in the silicon powder produced by the free space reactor, using the cathode layer emission spectroscopic technique. Impurity concentration followed by emission spectroscopic examination of the residue indicated a total impurity level of 2 micrograms/gram. A pellet cast from this powder had an electrical resistivity of 35 to 45 ohm-cm and P-type conductivity.

Feasibility Study of a Nuclear-Stirling Plant for the Jupiter Icy Moons Orbiter

NASA Technical Reports Server (NTRS)

Schmitz, Paul C.; Schreiber, Jeffrey G.; Penswick, L. Barry

2005-01-01

NASA is undertaking the design of a new spacecraft to explore the planet Jupiter and its three moons Calisto, Ganymede and Europa. This proposed mission, known as Jupiter Icy Moons Orbiter (JIMO) would use a nuclear reactor and an associated electrical generation system (Reactor Power Plant-RPP) to provide power to the spacecraft. The JIMO spacecraft is envisioned to use this power for science and communications as well as Electric Propulsion (EP). Among other potential power-generating concepts, previous studies have considered Thermoelectric and Brayton Power conversion systems, coupled to a liquid metal reactor for the JIMO mission. This paper will explore trades in system mass and radiator area for a nuclear reactor power conversion system, however this study will focus on Stirling power conversion. The Stirling convertor modeled in this study is based upon the Component Test Power Convertor design that was designed and operated successfully under the Civil Space Technology Initiative for use with the SP-100 nuclear reactor i the 1980's and early 1990's. The study design is such that two of the four convertors would operate at any time to generate the 100 kWe while the others are held in reserve. For this study the Stirling convertors hot-side temperature is 1050 K, would operate at a temperature ratio of 2.4 for a minimum mass system and would have a system efficiency of 29%. The Stirling convertor would generate high voltage (400 volt), 100 Hz single phase AC that is supplied to the Power Management and Distribution system. The waste hear is removed from the Stirling convertors by a flowing liquid sodium-potassium eutectic and then rejected by a shared radiator. The radiator consists of two coplanar wings, which would be deployed after the reactor is in space. System trades were performed to vary cycle state point temperatures and convertor design as well as power output. Other redundancy combinations were considered to understand the affects of convertor size and number of spares to the system mass.

Flow instability in particle-bed nuclear reactors

NASA Astrophysics Data System (ADS)

Kerrebrock, Jack L.

The particle-bed core offers mitigation of some of the problems of solid-core nuclear rocket reactors. Dividing the fuel elements into small spherical particles contained in a cylindrical bed through which the propellant flows radially, may reduce the thermal stress in the fuel elements, allowing higher propellant temperatures to be reached. The high temperature regions of the reactor are confined to the interior of cylindrical fuel assemblies, so most of the reactor can be relatively cool. This enables the use of structural and moderating materials which reduce the minimum critical size and mass of the reactor. One of the unresolved questions about this concept is whether the flow through the particle-bed will be well behaved, or will be subject to destructive flow instabilities. Most of the recent analyses of the stability of the particle-bed reactor have been extensions of the approach of Bussard and Delauer, where the bed is essentially treated as an array of parallel passages, so that the mass flow is continuous from inlet to outlet through any one passage. A more general three dimensional model of the bed is adopted, in which the fluid has mobility in three dimensions. Comparison of results of the earlier approach to the present one shows that the former does not accurately represent the stability at low Re. The more complete model presented should be capable of meeting this deficiency while accurately representing the effects of the cold and hot frits, and of heat conduction and radiation in the particle-bed. It can be extended to apply to the cylindrical geometry of particle-bed reactors without difficulty. From the exemplary calculations which were carried out, it can be concluded that a particle-bed without a cold frit would be subject to instability if operated at the high temperatures desired for nuclear rockets, and at power densities below about 4 megawatts per liter. Since the desired power density is about 40 megawatts per liter, it can be concluded that operation at design exit temperature but at reduced power could be hazardous for such a reactor. But the calculations also show that an appropriate cold frit could very likely cure the instability. More definite conclusions must await calculations for specific designs.

Flow instability in particle-bed nuclear reactors

NASA Technical Reports Server (NTRS)

Kerrebrock, Jack L.

1993-01-01

The particle-bed core offers mitigation of some of the problems of solid-core nuclear rocket reactors. Dividing the fuel elements into small spherical particles contained in a cylindrical bed through which the propellant flows radially, may reduce the thermal stress in the fuel elements, allowing higher propellant temperatures to be reached. The high temperature regions of the reactor are confined to the interior of cylindrical fuel assemblies, so most of the reactor can be relatively cool. This enables the use of structural and moderating materials which reduce the minimum critical size and mass of the reactor. One of the unresolved questions about this concept is whether the flow through the particle-bed will be well behaved, or will be subject to destructive flow instabilities. Most of the recent analyses of the stability of the particle-bed reactor have been extensions of the approach of Bussard and Delauer, where the bed is essentially treated as an array of parallel passages, so that the mass flow is continuous from inlet to outlet through any one passage. A more general three dimensional model of the bed is adopted, in which the fluid has mobility in three dimensions. Comparison of results of the earlier approach to the present one shows that the former does not accurately represent the stability at low Re. The more complete model presented should be capable of meeting this deficiency while accurately representing the effects of the cold and hot frits, and of heat conduction and radiation in the particle-bed. It can be extended to apply to the cylindrical geometry of particle-bed reactors without difficulty. From the exemplary calculations which were carried out, it can be concluded that a particle-bed without a cold frit would be subject to instability if operated at the high temperatures desired for nuclear rockets, and at power densities below about 4 megawatts per liter. Since the desired power density is about 40 megawatts per liter, it can be concluded that operation at design exit temperature but at reduced power could be hazardous for such a reactor. But the calculations also show that an appropriate cold frit could very likely cure the instability. More definite conclusions must await calculations for specific designs.

THE EXPERIENCE IN THE UNITED STATES WITH REACTOR OPERATION AND REACTOR SAFEGUARDS

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

McCullough, C.R.

1958-10-31

Reactors are operating or planned at locations in the United States in cities, near cities, and at remote locations. There is a general pattern that the higher power reactors are not in, but fairly uear cities, and the testing reactors for more hazardous experiments are at remote locations. A great deal has been done on the theoretical and experimental study of importunt features of reactor design. The metal-water reaction is still a theoretical possibility but tests of fuel element burnout under conditions approaching reactor operation gave no reaction. It appears that nucleate boiling does not necessarily result in steam blanketingmore » and fuel melting. Much attention is being given to the calculation of core kinetics but it is being found that temperature, power, and void coefficients cannot be calculated with accuracy and experiments are required. Some surprises are found giving positive localized void coefficients. Possible oscillatory behavior of reactors is being given careful study. No dangerous oscillations have been found in operating reactors but osciliations hare appeared in experimeats. The design of control and safety systems varies wvith different constructors. The relation of control to the kinetic behavior of the reactor is being studied. The importance of sensing element locations in order to know actual local reactor power level is being recognized. The time constants of instrumentation as related to reactor kinetics are being studied. Pressure vessels for reactors are being designed and manufactured. Many of these are beyond any previous experience. The stress problem is being given careful study. The effect of radiation is being studied experimentally. The stress problems of piping and pressure vessels is a difficult design problem being met successfully in reactor plants. The proper organization and procedure for operation of reactors is being evolved for resourch, testing, and power reactors. The importance of written standards and instructions for both normal and abnormal operating conditions is recogmized. Corfinement of radioactive materials either by tight steel shells, tight buildings, or semi-tight structures vented through filters is considered necessary in the United States. A discussion will be given of specifications, construction, and testing of these structures. The need for emergency plans has been stressed by recent experiences in radioactive releases. The problems of such plans to cover all grades of accidents will be discussed. The theoretical consequences of releases of radioactive materials have been studied and these results will be compared with actual experience. The problem of exposures from normal and abnormal operetion of reactors is a problem of desiga and operation on one hand and the amount of damage to be expected on the other. The safeguard problem is closely related to the acceptable doses of radiouctivity which the ICRP recommend. The future of atomic energy depends upon adequate safeguards and economical design and operation. Accepted criteria are required to guide designers as to the proper balance of caution and boldness. (auth)« less

Cold adaptation and replicable microbial community development during long-term low-temperature anaerobic digestion treatment of synthetic sewage

PubMed Central

Hughes, D; Mahony, T; Cysneiros, D; Ijaz, U Z; Smith, C J; O'Flaherty, V

2018-01-01

ABSTRACT The development and activity of a cold-adapting microbial community was monitored during low-temperature anaerobic digestion (LtAD) treatment of wastewater. Two replicate hybrid anaerobic sludge bed-fixed-film reactors treated a synthetic sewage wastewater at 12°C, at organic loading rates of 0.25–1.0 kg chemical oxygen demand (COD) m−3 d−1, over 889 days. The inoculum was obtained from a full-scale anaerobic digestion reactor, which was operated at 37°C. Both LtAD reactors readily degraded the influent with COD removal efficiencies regularly exceeding 78% for both the total and soluble COD fractions. The biomass from both reactors was sampled temporally and tested for activity against hydrolytic and methanogenic substrates at 12°C and 37°C. Data indicated that significantly enhanced low-temperature hydrolytic and methanogenic activity developed in both systems. For example, the hydrolysis rate constant (k) at 12°C had increased 20–30-fold by comparison to the inoculum by day 500. Substrate affinity also increased for hydrolytic substrates at low temperature. Next generation sequencing demonstrated that a shift in a community structure occurred over the trial, involving a 1-log-fold change in 25 SEQS (OTU-free approach) from the inoculum. Microbial community structure changes and process performance were replicable in the LtAD reactors. PMID:29846574

Cold adaptation and replicable microbial community development during long-term low-temperature anaerobic digestion treatment of synthetic sewage.

PubMed

Keating, C; Hughes, D; Mahony, T; Cysneiros, D; Ijaz, U Z; Smith, C J; O'Flaherty, V

2018-07-01

The development and activity of a cold-adapting microbial community was monitored during low-temperature anaerobic digestion (LtAD) treatment of wastewater. Two replicate hybrid anaerobic sludge bed-fixed-film reactors treated a synthetic sewage wastewater at 12°C, at organic loading rates of 0.25-1.0 kg chemical oxygen demand (COD) m-3 d-1, over 889 days. The inoculum was obtained from a full-scale anaerobic digestion reactor, which was operated at 37°C. Both LtAD reactors readily degraded the influent with COD removal efficiencies regularly exceeding 78% for both the total and soluble COD fractions. The biomass from both reactors was sampled temporally and tested for activity against hydrolytic and methanogenic substrates at 12°C and 37°C. Data indicated that significantly enhanced low-temperature hydrolytic and methanogenic activity developed in both systems. For example, the hydrolysis rate constant (k) at 12°C had increased 20-30-fold by comparison to the inoculum by day 500. Substrate affinity also increased for hydrolytic substrates at low temperature. Next generation sequencing demonstrated that a shift in a community structure occurred over the trial, involving a 1-log-fold change in 25 SEQS (OTU-free approach) from the inoculum. Microbial community structure changes and process performance were replicable in the LtAD reactors.

Grey water treatment in UASB reactor at ambient temperature.

PubMed

Elmitwalli, T A; Shalabi, M; Wendland, C; Otterpohl, R

2007-01-01

In this paper, the feasibility of grey water treatment in a UASB reactor was investigated. The batch recirculation experiments showed that a maximum total-COD removal of 79% can be obtained in grey-water treatment in the UASB reactor. The continuous operational results of a UASB reactor treating grey water at different hydraulic retention time (HRT) of 20, 12 and 8 hours at ambient temperature (14-24 degrees C) showed that 31-41% of total COD was removed. These results were significantly higher than that achieved by a septic tank (11-14%), the most common system for grey water pre-treatment, at HRT of 2-3 days. The relatively lower removal of total COD in the UASB reactor was mainly due to a higher amount of colloidal COD in the grey water, as compared to that reported in domestic wastewater. The grey water had a limited amount of nitrogen, which was mainly in particulate form (80-90%). The UASB reactor removed 24-36% and 10-24% of total nitrogen and total phosphorus, respectively, in the grey water, due to particulate nutrients removal by physical entrapment and sedimentation. The sludge characteristics of the UASB reactor showed that the system had stable performance and the recommended HRT for the reactor is 12 hours.

THE ARMOUR DUST FUELED REACTOR (ADFR). Quarterly Progress Report No. 1 for the Period February 21, 1958 to May 21, 1958

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

Loewe, W.E.; Krucoff, D.

1958-10-31

Work has begun on the ADFR, a reactor using a new fuel form -- fissionable dust carried in an inent gas. Temperatures in the range 2,000 to 3,000 deg F appear feasible in an all-ceramic system. Experimental study of the fuel form was initiated, and a loop to circulate the fuel dust was constructed. Initial operation is encouraging. Theoretical studies were carried on in the areas of reactor physics, heat transfer, and safety. (auth)

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

1993-01-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ˜60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ˜100 MW/L may be achievable.

[Kinetics of catalytic wet air oxidation of phenol in trickle bed reactor].

PubMed

Li, Guang-ming; Zhao, Jian-fu; Wang, Hua; Zhao, Xiu-hua; Zhou, Yang-yuan

2004-05-01

By using a trickle bed reactor which was designed by the authors, the catalytic wet air oxidation reaction of phenol on CuO/gamma-Al2O3 catalyst was studied. The results showed that in mild operation conditions (at temperature of 180 degrees C, pressure of 3 MPa, liquid feed rate of 1.668 L x h(-1) and oxygen feed rate of 160 L x h(-1)), the removal of phenol can be over 90%. The curve of phenol conversion is similar to "S" like autocatalytic reaction, and is accordance with chain reaction of free radical. The kinetic model of pseudo homogenous reactor fits the catalytic wet air oxidation reaction of phenol. The effects of initial concentration of phenol, liquid feed rate and temperature for reaction also were investigated.

Alpha-ray spectrometry at high temperature by using a compound semiconductor detector.

PubMed

Ha, Jang Ho; Kim, Han Soo

2013-11-01

The use of conventional radiation detectors in harsh environments is limited by radiation damage to detector materials and by temperature constraints. We fabricated a wide-band gap semiconductor radiation detector based on silicon carbide. All the detector components were considered for an application in a high temperature environment like a nuclear reactor core. The radiation response, especially to alpha particles, was measured using an (241)Am source at variable operating voltages at room temperature in the air. The temperature on detector was controlled from 30°C to 250°C. The alpha-particle spectra were measured at zero bias operation. Even though the detector is operated at high temperature, the energy resolution as a function of temperature is almost constant within 3.5% deviation. Copyright © 2013 Elsevier Ltd. All rights reserved.

Stable partial nitritation for low-strength wastewater at low temperature in an aerobic granular reactor.

PubMed

Isanta, Eduardo; Reino, Clara; Carrera, Julián; Pérez, Julio

2015-09-01

Partial nitritation for a low-strength wastewater at low temperature was stably achieved in an aerobic granular reactor. A bench-scale granular sludge bioreactor was operated in continuous mode treating an influent of 70 mg N-NH4(+) L(-1) to mimic pretreated municipal nitrogenous wastewater and the temperature was progressively decreased from 30 to 12.5 °C. A suitable effluent nitrite to ammonium concentrations ratio to a subsequent anammox reactor was maintained stable during 300 days at 12.5 °C. The average applied nitrogen loading rate at 12.5 °C was 0.7 ± 0.3 g N L(-1) d(-1), with an effluent nitrate concentration of only 2.5 ± 0.7 mg N-NO3(-) L(-1). The biomass fraction of nitrite-oxidizing bacteria (NOB) in the granular sludge decreased from 19% to only 1% in 6 months of reactor operation at 12.5 °C. Nitrobacter spp. where found as the dominant NOB population, whereas Nitrospira spp. were not detected. Simulations indicated that: (i) NOB would only be effectively repressed when their oxygen half-saturation coefficient was higher than that of ammonia-oxidizing bacteria; and (ii) a lower specific growth rate of NOB was maintained at any point in the biofilm (even at 12.5 °C) due to the bulk ammonium concentration imposed through the control strategy. Copyright © 2015 Elsevier Ltd. All rights reserved.

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

Adaption of a microwave plasma source for low temperature diamond deposition

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

Ulczynski, M.; Reinhard, D.K.; Asmussen, J.

1996-12-31

This report describes the adaption of a microwave plasma reactor for low temperature diamond deposition. The reactor is of a resonant cavity design. Three approaches have been taken to establish plasma conditions for diamond deposition on substrates which are in the range of 450 C to 550 C. In the first, the substrate is heated only by the plasma and the source is operated at pressures on the order of 10 torr, such that the volumetric power density is sufficiently low to achieve these temperatures. In the second, the plasma pressure and microwave input power were reduced and a substratemore » heater was used to maintain the desired deposition temperatures. In the third approach, the plasma pressure and microwave power were increased and a substrate cooler was used to keep the substrate temperature in the desired range. Reactor performance and deposition results will be described for the three configurations. For the plasma heated substrate assembly, substrate dimensions were up to 10 cm diameter. For the heated and cooled substrate assemblies, substrate dimensions were up to 7.5 cm diameter. Deposition results on a variety of substrates will be reported including low-temperature substrates such as borosilicate glass.« less

Bimodal Nuclear Thermal Rocket Analysis Developments

NASA Technical Reports Server (NTRS)

Belair, Michael; Lavelle, Thomas; Saimento, Charles; Juhasz, Albert; Stewart, Mark

2014-01-01

Nuclear thermal propulsion has long been considered an enabling technology for human missions to Mars and beyond. One concept of operations for these missions utilizes the nuclear reactor to generate electrical power during coast phases, known as bimodal operation. This presentation focuses on the systems modeling and analysis efforts for a NERVA derived concept. The NERVA bimodal operation derives the thermal energy from the core tie tube elements. Recent analysis has shown potential temperature distributions in the tie tube elements that may limit the thermodynamic efficiency of the closed Brayton cycle used to generate electricity with the current design. The results of this analysis are discussed as well as the potential implications to a bimodal NERVA type reactor.

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

Nygaard, E. T.; Williams, M. M. R.; Angelo, P. L.

Babcock and Wilcox Technical Services Group (B and W) has identified aqueous homogeneous reactors (AHRs) as a technology well suited to produce the medical isotope molybdenum 99 (Mo-99). AHRs have never been specifically designed or built for this specialized purpose. However, AHRs have a proven history of being safe research reactors. In fact, in 1958, AHRs had 'a longer history of operation than any other type of research reactor using enriched fuel' and had 'experimentally demonstrated to be among the safest of all various type of research reactor now in use [1].' A 'Level 1' model representing B and W'smore » proposed Medical Isotope Production System (MIPS) reactor has been developed. The Level 1 model couples a series of differential equations representing neutronics, temperature, and voiding. Neutronics are represented by point reactor kinetics while temperature and voiding terms are axially varying (one-dimensional). While this model was developed specifically for the MIPS reactor, its applicability to the Japanese TRACY reactor was assessed. The results from the Level 1 model were in good agreement with TRACY experimental data and found to be conservative over most of the time domains considered. The Level 1 model was used to study the MIPS reactor. An analysis showed the Level 1 model agreed well with a more complex computational model of the MIPS reactor (a FETCH model). Finally, a significant reactivity insertion was simulated with the Level 1 model to study the MIPS reactor's time-dependent response. (authors)« less

Entropy Generation Minimization in Dimethyl Ether Synthesis: A Case Study

NASA Astrophysics Data System (ADS)

Kingston, Diego; Razzitte, Adrián César

2018-04-01

Entropy generation minimization is a method that helps improve the efficiency of real processes and devices. In this article, we study the entropy production (due to chemical reactions, heat exchange and friction) in a conventional reactor that synthesizes dimethyl ether and minimize it by modifying different operating variables of the reactor, such as composition, temperature and pressure, while aiming at a fixed production of dimethyl ether. Our results indicate that it is possible to reduce the entropy production rate by nearly 70 % and that, by changing only the inlet composition, it is possible to cut it by nearly 40 %, though this comes at the expense of greater dissipation due to heat transfer. We also study the alternative of coupling the reactor with another, where dehydrogenation of methylcyclohexane takes place. In that case, entropy generation can be reduced by 54 %, when pressure, temperature and inlet molar flows are varied. These examples show that entropy generation analysis can be a valuable tool in engineering design and applications aiming at process intensification and efficient operation of plant equipment.

Evaluation of thermodynamically favourable operating conditions for production of hydrogen in three different reforming technologies

NASA Astrophysics Data System (ADS)

Seo, Y.-S.; Shirley, A.; Kolaczkowski, S. T.

With the aid of thermodynamic analysis using AspenPlus™, the characteristics of three different types of reforming process are investigated. These include: steam-methane reforming (SMR), partial oxidation (POX) and autothermal reforming (ATR). Thereby, favourable operating conditions are identified for each process. The optimum steam-to-carbon (S:C) ratio of the SMR reactor is found to be 1.9. The optimum air ratio of the POX reactor is 0.3 at a preheat temperature of 312 °C. The optimum air ratio and S:C ratio of the ATR reactor are 0.29 and 0.35, respectively at a preheat temperature of 400 °C. Simulated material and energy balances show that the CH 4 flow rates required to generate 1 mol s -1 of hydrogen are 0.364 mol s -1 for POX, 0.367 mol s -1 for ATR and 0.385 mol s -1 for the SMR. These results demonstrate that the POX reforming system has the lowest energy cost to produce the same amount of hydrogen from CH 4.

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).

Ultrapyrolytic upgrading of plastic wastes and plastics/heavy oil mixtures to valuable light gas products

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

Lovett, S.; Berruti, F.; Behie, L.A.

1997-11-01

Viable operating conditions were identified experimentally for maximizing the production of high-value products such as ethylene, propylene, styrene, and benzene, from the ultrapyrolysis of waste plastics. Using both a batch microreactor and a pilot-plant-sized reactor, the key operating variables considered were pyrolysis temperature, product reaction time, and quench time. In the microreactor experiments, polystyrene (PS), a significant component of waste plastics, was pyrolyzed at temperatures ranging from 800 to 965 C, with total reaction times ranging from 500 to 1,000 ms. At a temperature of 965 C and 500 ms, the yields of styrene plus benzene were greater than 95more » wt %. In the pilot-plant experiments, the recently patented internally circulating fluidized bed (ICFB) reactor (Milne et al., US Patent Number 5,370,789, 1994b) was used to ultrapyrolyze low-density polyethylene (LDPE) in addition to LDPE (5% by weight)/heavy oil mixtures at a residence time of 600 ms. Both experiments produced light olefin yields greater than 55 wt % at temperatures above 830 C.« less

Numerical Study of Pyrolysis of Biomass in Fluidized Beds

NASA Technical Reports Server (NTRS)

Bellan, Josette; Lathouwers, Danny

2003-01-01

A report presents a numerical-simulation study of pyrolysis of biomass in fluidized-bed reactors, performed by use of the mathematical model described in Model of Fluidized Bed Containing Reacting Solids and Gases (NPO-30163), which appears elsewhere in this issue of NASA Tech Briefs. The purpose of the study was to investigate the effect of various operating conditions on the efficiency of production of condensable tar from biomass. The numerical results indicate that for a fixed particle size, the fluidizing-gas temperature is the foremost parameter that affects the tar yield. For the range of fluidizing-gas temperatures investigated, and under the assumption that the pyrolysis rate exceeds the feed rate, the optimum steady-state tar collection was found to occur at 750 K. In cases in which the assumption was not valid, the optimum temperature for tar collection was found to be only slightly higher. Scaling up of the reactor was found to exert a small negative effect on tar collection at the optimal operating temperature. It is also found that slightly better scaling is obtained by use of shallower fluidized beds with greater fluidization velocities.

Demonstration of a 100-kWth high-temperature solar thermochemical reactor pilot plant for ZnO dissociation

NASA Astrophysics Data System (ADS)

Koepf, E.; Villasmil, W.; Meier, A.

2016-05-01

Solar thermochemical H2O and CO2 splitting is a viable pathway towards sustainable and large-scale production of synthetic fuels. A reactor pilot plant for the solar-driven thermal dissociation of ZnO into metallic Zn has been successfully developed at the Paul Scherrer Institute (PSI). Promising experimental results from the 100-kWth ZnO pilot plant were obtained in 2014 during two prolonged experimental campaigns in a high flux solar simulator at PSI and a 1-MW solar furnace in Odeillo, France. Between March and June the pilot plant was mounted in the solar simulator and in-situ flow-visualization experiments were conducted in order to prevent particle-laden fluid flows near the window from attenuating transparency by blocking incoming radiation. Window flow patterns were successfully characterized, and it was demonstrated that particle transport could be controlled and suppressed completely. These results enabled the successful operation of the reactor between August and October when on-sun experiments were conducted in the solar furnace in order to demonstrate the pilot plant technology and characterize its performance. The reactor was operated for over 97 hours at temperatures as high as 2064 K; over 28 kg of ZnO was dissociated at reaction rates as high as 28 g/min.

Electrocoagulation using a rotated anode: A novel reactor design for textile wastewater treatment.

PubMed

Naje, Ahmed Samir; Chelliapan, Shreeshivadasan; Zakaria, Zuriati; Abbas, Saad A

2016-07-01

This paper investigates the optimum operational conditions of a novel rotated bed electrocoagulation (EC) reactor for the treatment of textile wastewater. The effect of various operational parameters such as rotational speed, current density (CD), operational time (RT), pH, temperature, and inter-electrode distance (IED) on the pollutant removal efficiency were examined. In addition, the consumption of aluminum (Al) and electrical energy, as well as operating costs at optimum conditions were also calculated. The results indicated that the optimum conditions for the treatment of textile wastewater were achieved at CD = 4 mA/cm(2), RT = 10 min, rotational speed = 150 rpm, pH = 4.57, temperature = 25 °C, and IED = 1 cm. The electrode consumption, energy consumption, and operating costs were 0.038 kg/m(3), 4.66 kWh/m(3) and 0.44 US$/m(3), respectively. The removal efficiencies of chemical oxygen demand (COD), biological oxygen demand (BOD), total suspended solid (TSS), turbidity and color were 97.10%, 95.55%, 98%, 96% and 98.50%, respectively, at the first 10 min of reaction time, while the phenol compound of the wastewater was almost entirely removed (99.99%). The experimental results confirm that the new reactor design with rotated anode impellers and cathode rings provided high treatment efficiency at a reduced reaction time and with lower energy consumption. Copyright © 2016 Elsevier Ltd. All rights reserved.

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.

Transient Response to Rapid Cooling of a Stainless Steel Sodium Heat Pipe

NASA Technical Reports Server (NTRS)

Mireles, Omar R.; Houts, Michael G.

2011-01-01

Compact fission power systems are under consideration for use in long duration space exploration missions. Power demands on the order of 500 W, to 5 kW, will be required for up to 15 years of continuous service. One such small reactor design consists of a fast spectrum reactor cooled with an array of in-core alkali metal heat pipes coupled to thermoelectric or Stirling power conversion systems. Heat pipes advantageous attributes include a simplistic design, lack of moving parts, and well understood behavior. Concerns over reactor transients induced by heat pipe instability as a function of extreme thermal transients require experimental investigations. One particular concern is rapid cooling of the heat pipe condenser that would propagate to cool the evaporator. Rapid cooling of the reactor core beyond acceptable design limits could possibly induce unintended reactor control issues. This paper discusses a series of experimental demonstrations where a heat pipe operating at near prototypic conditions experienced rapid cooling of the condenser. The condenser section of a stainless steel sodium heat pipe was enclosed within a heat exchanger. The heat pipe - heat exchanger assembly was housed within a vacuum chamber held at a pressure of 50 Torr of helium. The heat pipe was brought to steady state operating conditions using graphite resistance heaters then cooled by a high flow of gaseous nitrogen through the heat exchanger. Subsequent thermal transient behavior was characterized by performing an energy balance using temperature, pressure and flow rate data obtained throughout the tests. Results indicate the degree of temperature change that results from a rapid cooling scenario will not significantly influence thermal stability of an operating heat pipe, even under extreme condenser cooling conditions.

Improving online risk assessment with equipment prognostics and health monitoring

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

Coble, Jamie B.; Liu, Xiaotong; Briere, Chris

The current approach to evaluating the risk of nuclear power plant (NPP) operation relies on static probabilities of component failure, which are based on industry experience with the existing fleet of nominally similar light water reactors (LWRs). As the nuclear industry looks to advanced reactor designs that feature non-light water coolants (e.g., liquid metal, high temperature gas, molten salt), this operating history is not available. Many advanced reactor designs use advanced components, such as electromagnetic pumps, that have not been used in the US commercial nuclear fleet. Given the lack of rich operating experience, we cannot accurately estimate the evolvingmore » probability of failure for basic components to populate the fault trees and event trees that typically comprise probabilistic risk assessment (PRA) models. Online equipment prognostics and health management (PHM) technologies can bridge this gap to estimate the failure probabilities for components under operation. The enhanced risk monitor (ERM) incorporates equipment condition assessment into the existing PRA and risk monitor framework to provide accurate and timely estimates of operational risk.« less

Process for removing copper in a recoverable form from solid scrap metal

DOEpatents

Hartman, Alan D.; Oden, Laurance L.; White, Jack C.

1995-01-01

A process for removing copper in a recoverable form from a copper/solid ferrous scrap metal mix is disclosed. The process begins by placing a copper/solid ferrous scrap metal mix into a reactor vessel. The atmosphere within the reactor vessel is purged with an inert gas or oxidizing while the reactor vessel is heated in the area of the copper/solid ferrous scrap metal mix to raise the temperature within the reactor vessel to a selected elevated temperature. Air is introduced into the reactor vessel and thereafter hydrogen chloride is introduced into the reactor vessel to obtain a desired air-hydrogen chloride mix. The air-hydrogen chloride mix is operable to form an oxidizing and chloridizing atmosphere which provides a protective oxide coating on the surface of the solid ferrous scrap metal in the mix and simultaneously oxidizes/chloridizes the copper in the mix to convert the copper to a copper monochloride gas for transport away from the solid ferrous scrap metal. After the copper is completely removed from the copper/solid ferrous scrap metal mix, the flows of air and hydrogen chloride are stopped and the copper monochloride gas is collected for conversion to a recoverable copper species.

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.

Transient thermohydraulic heat pipe modeling

NASA Astrophysics Data System (ADS)

Hall, Michael L.; Doster, Joseph M.

Many space based reactor designs employ heat pipes as a means of conveying heat. In these designs, thermal radiation is the principle means for rejecting waste heat from the reactor system, making it desirable to operate at high temperatures. Lithium is generally the working fluid of choice as it undergoes a liquid-vapor transformation at the preferred operating temperature. The nature of remote startup, restart, and reaction to threats necessitates an accurate, detailed transient model of the heat pipe operation. A model is outlined of the vapor core region of the heat pipe which is part of a large model of the entire heat pipe thermal response. The vapor core is modeled using the area averaged Navier-Stokes equations in one dimension, which take into account the effects of mass, energy and momentum transfer. The core model is single phase (gaseous), but contains two components: lithium gas and a noncondensible vapor. The vapor core model consists of the continuity equations for the mixture and noncondensible, as well as mixture equations for internal energy and momentum.

Thermo-kinetic instabilities in model reactors. Examples in experimental tests

NASA Astrophysics Data System (ADS)

Lavadera, Marco Lubrano; Sorrentino, Giancarlo; Sabia, Pino; de Joannon, Mara; Cavaliere, Antonio; Ragucci, Raffaele

2017-11-01

The use of advanced combustion technologies (such as MILD, LTC, etc.) is among the most promising methods to reduce emission of pollutants. For such technologies, working temperatures are enough low to boost the formation of several classes of pollutants, such as NOx and soot. To access this temperature range, a significant dilution as well as preheating of reactants is required. Such conditions are usually achieved by a strong recirculation of exhaust gases that simultaneously dilute and pre-heat the fresh reactants. These peculiar operative conditions also imply strong fuel flexibility, thus allowing the use of low calorific value (LCV) energy carriers with high efficiency. However, the intersection of low combustion temperatures and highly diluted mixtures with intense pre-heating alters the evolution of the combustion process with respect to traditional flames, leading to features such as the susceptibility to oscillations, which are undesirable during combustion. Therefore, an effective use of advanced combustion technologies requires a thorough analysis of the combustion kinetic characteristics in order to identify optimal operating conditions and control strategies with high efficiency and low pollutant emissions. The present work experimentally and numerically characterized the ignition and oxidation processes of methane and propane, highly diluted in nitrogen, at atmospheric pressure, in a Plug Flow Reactor and a Perfectly Stirred Reactor under a wide range of operating conditions involving temperatures, mixture compositions and dilution levels. The attention was focused particularly on the chemistry of oscillatory phenomena and multistage ignitions. The global behavior of these systems can be qualitatively and partially quantitatively modeled using the detailed kinetic models available in the literature. Results suggested that, for diluted conditions and lower adiabatic flame temperatures, the competition among several pathways, i.e. intermediate- and high-temperature branching, branching and recombination channels, oxidation and recombination/pyrolysis pathways, is enhanced, thus permitting the onset of phenomena that are generally hidden during conventional combustion processes.

Development of tungsten fibre-reinforced tungsten composites towards their use in DEMO—potassium doped tungsten wire

NASA Astrophysics Data System (ADS)

Riesch, J.; Han, Y.; Almanstötter, J.; Coenen, J. W.; Höschen, T.; Jasper, B.; Zhao, P.; Linsmeier, Ch; Neu, R.

2016-02-01

For the next step fusion reactor the use of tungsten is inevitable to suppress erosion and allow operation at elevated temperature and high heat loads. Tungsten fibre-reinforced composites overcome the intrinsic brittleness of tungsten and its susceptibility to operation embrittlement and thus allow its use as a structural as well as an armour material. That this concept works in principle has been shown in recent years. In this contribution we present a development approach towards its use in a future fusion reactor. A multilayer approach is needed addressing all composite constituents and manufacturing steps. A huge potential lies in the optimization of the tungsten wire used as fibre. We discuss this aspect and present studies on potassium doped tungsten wire in detail. This wire, utilized in the illumination industry, could be a replacement for the so far used pure tungsten wire due to its superior high temperature properties. In tensile tests the wire showed high strength and ductility up to an annealing temperature of 2200 K. The results show that the use of doped tungsten wire could increase the allowed fabrication temperature and the overall working temperature of the composite itself.

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.

Effects of water vapor pretreatment time and reaction temperature on CO(2) capture characteristics of a sodium-based solid sorbent in a bubbling fluidized-bed reactor.

PubMed

Seo, Yongwon; Jo, Sung-Ho; Ryu, Chong Kul; Yi, Chang-Keun

2007-10-01

CO(2) capture from flue gas using a sodium-based solid sorbent was investigated in a bubbling fluidized-bed reactor. Carbonation and regeneration temperature on CO(2) removal was determined. The extent of the chemical reactivity after carbonation or regeneration was characterized via (13)C NMR. In addition, the physical properties of the sorbent such as pore size, pore volume, and surface area after carbonation or regeneration were measured by gas adsorption method (BET). With water vapor pretreatment, near complete CO(2) removal was initially achieved and maintained for about 1-2min at 50 degrees C with 2s gas residence time, while without proper water vapor pretreatment CO(2) removal abruptly decreased from the beginning. Carbonation was effective at the lower temperature over the 50-70 degrees C temperature range, while regeneration more effective at the higher temperature over the 135-300 degrees C temperature range. To maintain the initial 90% CO(2) removal, it would be necessary to keep the regeneration temperature higher than about 135 degrees C. The results obtained in this study can be used as basic data for designing and operating a large scale CO(2) capture process with two fluidized-bed reactors.

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

Principles and Applications of Solid Polymer Electrolyte Reactors for Electrochemical Hydrodehalogenation of Organic Pollutants

NASA Astrophysics Data System (ADS)

Cheng, Hua; Scott, Keith

The ability to re-cycle halogenated liquid wastes, based on electrochemical hydrodehalogenation (EHDH), will provide a significant economic advantage and will reduce the environmental burden in a number of processes. The use of a solid polymer electrolyte (SPE) reactor is very attractive for this purpose. Principles and features of electrochemical HDH technology and SPE EHDH reactors are described. The SPE reactor enables selective dehalogenation of halogenated organic compounds in both aqueous and non-aqueous media with high current efficiency and low energy consumption. The influence of operating conditions, including cathode material, current density, reactant concentration and temperature on the HDH process and its stability are examined.

Jet pump-drive system for heat removal

NASA Technical Reports Server (NTRS)

French, J. R. (Inventor)

1985-01-01

A jet pump, in combination with a TEMP, is employed to assure safe cooling of a nuclear reactor after shutdown. A TEMP, responsive to the heat from the coolant in the secondary flow path, automatically pumps the 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. At this lower temperature, the TEMP/jet jump combination ceases its circulation boosting operation. The TEMP/jet pump combination is automatic, self-regulating and provides an emergency pumping system free of moving parts.

Thermal and hydraulic analysis of a cylindrical blanket module design for a tokamak reactor

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

Lee, A.Y.

1978-10-01

Various existing blanket design concepts for a tokamak fusion reactor were evaluated and assessed. These included the demonstration power reactors of ORNL, GA and others. As a result of this study, a cylindrical, modularized blanket design concept was developed. The module is a double-walled, stainless steel 316 cylinder containing liquid lithium for tritium breeding and is cooled by pressurized helium. Steady state and transient thermal conditions under normal and some off-design conditions were analyzed and presented. At the steady state reference operating point the maximum structure temperature is 452/sup 0/C at the maximum stressed location and is 495/sup 0/C atmore » the less stressed location. The coolant inlet pressure is 54.4 atm, the inlet temperature is 200/sup 0/C and the exit temperature is 435/sup 0/C. The coolant could be utilized with a helium/steam turbine power conversion system with a cycle thermal efficiency of 30.8%.« less

The performance and microbial diversity of temperature-phased hyperthermophilic and thermophilic anaerobic digestion system fed with organic waste.

PubMed

Lee, M Y; Cheon, J H; Hidaka, T; Tsuno, H

2008-01-01

The objective of this study was to evaluate the performances and microbial diversities for development of the effective hyperthermophilic digester system that consists of a hyperthermophilic reactor and hyperthermophilic or thermophilic reactor in series. Lab-scale reactors were operated continuously fed with artificial kitchen garbage. The effect of temperature on the acidification step was firstly investigated. Results indicated that 20.8% of COD solubilization was achieved at 70 degrees C, with 12.6% at 80 degrees C. The average protein solubilization reached 31% at 80 degrees C. Methane conversion efficiency following the acidification was around 85% on average at 55 degrees C, but decreased with increasing temperature and methane gas was not produced over 73 degrees C. As well, bacteria affiliated with the methanogens dominated the population below 65 degrees C, while those affiliated with acidogens were predominant over 73 degrees C. These results indicated that the hyperthermophilic process has considerable benefits to treat wastewater or waste containing high concentration of protein.

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

Production of aluminum--silicon alloy and ferrosilicon and commercial purity aluminum by the direct reduction process. First interim technical report, Phase B, September 1--November 30, 1978

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

Bruno, M.J.

1978-12-01

The parameters of charge content, reaction temperatures and residence time were studied in a bench reactor concerning the production of Al--Si and Fe--Si alloys. Results confirmed that minimum final stage reaction temperature is 1950 to 2000/sup 0/C. Residence time varied with initial charge concentration. Fe/sub 2/O/sub 3/ additions to the charge produced a significant increase in metallic yield. A burden preparation procedure was developed for making acceptable agglomerates containing Fe/sub 2/O/sub 3/, bauxite, clay and coke. Particle size distribution of starting materials was correlated with agglomerate strength. A new bench scale reactor was designed and built to facilitate semi-continuous operation,more » using O/sub 2/ injection to burn coke supporting the burden, resulting in burden movement. In a number of runs bridging of the burden material occurred due to condensation of volatilized sub-oxides in the cooler zones of the reactor. The reactor operated smoothly as an iron blast furnace at 1500/sup 0/C, demonstrating the validity of the equipment and test procedures. Initial construction of pilot reactor VSR-1 was completed. Bench scale fractional crystallizer runs were continued to determine the impurity effects of Fe up to 6.9% and Ti up to 1.25% on alloy product concentration and yield. High levels of impurities formed intermetallic complexes with Al and reduced product yield.« 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.

A coupled nuclear reactor thermal energy storage system for enhanced load following operation

NASA Astrophysics Data System (ADS)

Alameri, Saeed A.

Nuclear power plants usually provide base-load electric power and operate most economically at a constant power level. In an energy grid with a high fraction of renewable energy sources, future nuclear reactors may be subject to significantly variable power demands. These variable power demands can negatively impact the effective capacity factor of the reactor and result in severe economic penalties. Coupling the reactor to a large Thermal Energy Storage (TES) block will allow the reactor to better respond to variable power demands. In the system described in this thesis, a Prismatic-core Advanced High Temperature Reactor (PAHTR) operates at constant power with heat provided to a TES block that supplies power as needed to a secondary energy conversion system. The PAHTR is designed to have a power rating of 300 MW th, with 19.75 wt% enriched Tri-Structural-Isotropic UO 2 fuel and a five year operating cycle. The passive molten salt TES system will operate in the latent heat region with an energy storage capacity of 150 MWd. Multiple smaller TES blocks are used instead of one large block to enhance the efficiency and maintenance complexity of the system. A transient model of the coupled reactor/TES system is developed to study the behavior of the system in response to varying load demands. The model uses six-delayed group point kinetics and decay heat models coupled to thermal-hydraulic and heat transfer models of the reactor and TES system. Based on the transient results, the preferred TES design consists of 1000 blocks, each containing 11000 LiCl phase change material tubes. A safety assessment of major reactor events demonstrates the inherent safety of the coupled system. The loss of forced circulation study determined the minimum required air convection heat removal rate from the reactor core and the lowest possible reduced primary flow rate that can maintain the reactor in a safe condition. The loss of ultimate heat sink study demonstrated the ability of the TES to absorb the decay heat of the reactor fuel while cooling the PAHTR after an emergency shutdown. The simulated reactivity insertion accident assessment determined the maximum allowable reactivity insertion to the PAHTR as a function of shutdown response times.

Silicon Chemical Vapor Deposition Process Using a Half-Inch Silicon Wafer for Minimal Manufacturing System

NASA Astrophysics Data System (ADS)

Li, Ning; Habuka, Hitoshi; Ikeda, Shin-ichi; Hara, Shiro

A chemical vapor deposition reactor for producing thin silicon films was designed and developed for achieving a new electronic device production system, the Minimal Manufacturing, using a half-inch wafer. This system requires a rapid process by a small footprint reactor. This was designed and verified by employing the technical issues, such as (i) vertical gas flow, (ii) thermal operation using a highly concentrated infrared flux, and (iii) reactor cleaning by chlorine trifluoride gas. The combination of (i) and (ii) could achieve a low heating power and a fast cooling designed by the heat balance of the small wafer placed at a position outside of the reflector. The cleaning process could be rapid by (iii). The heating step could be skipped because chlorine trifluoride gas was reactive at any temperature higher than room temperature.

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

Gerczak, Tyler J.; Smith, Kurt R.; Petrie, Christian M.

Tristructural-isotropic (TRISO)–coated particle fuel is a promising advanced fuel concept consisting of a spherical fuel kernel made of uranium oxide and uranium carbide, surrounded by a porous carbonaceous buffer layer and successive layers of dense inner pyrolytic carbon (IPyC), silicon carbide (SiC) deposited by chemical vapor , and dense outer pyrolytic carbon (OPyC). This fuel concept is being considered for advanced reactor applications such as high temperature gas-cooled reactors (HTGRs) and molten salt reactors (MSRs), as well as for accident-tolerant fuel for light water reactors (LWRs). Development and implementation of TRISO fuel for these reactor concepts support the US Departmentmore » of Energy (DOE) Office of Nuclear Energy mission to promote safe, reliable nuclear energy that is sustainable and environmentally friendly. During operation, the SiC layer serves as the primary barrier to metallic fission products and actinides not retained in the kernel. It has been observed that certain fission products are released from TRISO fuel during operation, notably, Ag, Eu, and Sr [1]. Release of these radioisotopes causes safety and maintenance concerns.« less

Reactor performance and microbial community of an EGSB reactor operated at 20 and 15 degrees C.

PubMed

Xing, W; Zuo, J-E; Dai, N; Cheng, J; Li, J

2009-09-01

To investigate the effects of low temperatures on the performance and microbial community of anaerobic wastewater treatment. An expanded granular sludge bed (EGSB) reactor was employed to treat synthetic brewery wastewater at 20 and 15 degrees C. Reactor performance was represented by chemical oxygen demand (COD) removal efficiency, while the microbial community was analysed using denaturing gradient gel electrophoresis (DGGE) and clone technology. When the hydraulic retention time (HRT) was maintained at 18 h, COD removal efficiencies above 85% were obtained at both 20 and 15 degrees C, with influent COD concentrations up to 7300 and 4100 mg l(-1), respectively. At 15 degrees C, the COD removal efficiency was more easily manipulated by increasing the influent COD concentration. DGGE and clone results for both temperatures revealed that Methanosaeta and Methanobacterium were two dominant methanogens, and that the majority of the eubacterial clones were represented by Firmicutes. When the temperature decreased from 20 to 15 degrees C, both archaeal and eubacterial communities had higher diversity, and the proportion of Methanosaeta (acetate-utilizing methanogens) decreased markedly from 60.0% to 49.3%, together with an increase in proportions of hydrogen-utilizing methanogens (especially Methanospirillum). The feasibility of psychrophilic anaerobic treatment of low and medium strength organic wastewaters was demonstrated, although lower temperature could significantly affect both reactor performance and the anaerobic microbial community. The findings enrich the theory involving the microbial community and the application of anaerobic treatment in a psychrophilic environment.

Effect of inlet temperature on the performance of a catalytic reactor. [air pollution control

NASA Technical Reports Server (NTRS)

Anderson, D. N.

1978-01-01

A 12 cm diameter by 15 cm long catalytic reactor was tested with No. 2 diesel fuel in a combustion test rig at inlet temperatures of 700, 800, 900, and 1000 K. Other test conditions included pressures of 3 and 6 x 10 to the 5th power Pa, reference velocities of 10, 15, and 20 m/s, and adiabatic combustion temperatures in the range 1100 to 1400 K. The combustion efficiency was calculated from measurements of carbon monoxide and unburned hydrocarbon emissions. Nitrogen oxide emissions and reactor pressure drop were also measured. At a reference velocity of 10 m/s, the CO and unburned hydrocarbons emissions, and, therefore, the combustion efficiency, were independent of inlet temperature. At an inlet temperature of 1000 K, they were independent of reference velocity. Nitrogen oxides emissions resulted from conversion of the small amount (135 ppm) of fuel-bound nitrogen in the fuel. Up to 90 percent conversion was observed with no apparent effect of any of the test variables. For typical gas turbine operating conditions, all three pollutants were below levels which would permit the most stringent proposed automotive emissions standards to be met.

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.

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)

Progress in the development of the neutron flux monitoring system of the French GEN-IV SFR: simulations and experimental validations [ANIMMA--2015-IO-98

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

Jammes, C.; Filliatre, P.; De Izarra, G.

The neutron flux monitoring system of the French GEN-IV sodium-cooled fast reactor will rely on high temperature fission chambers installed in the reactor vessel and capable of operating over a wide-range neutron flux. The definition of such a system is presented and the technological solutions are justified with the use of simulation and experimental results. (authors)

Array of planar membrane modules for producing hydrogen

DOEpatents

Vencill, Thomas R [Albuquerque, NM; Chellappa, Anand S [Albuquerque, NM; Rathod, Shailendra B [Hillsboro, OR

2012-05-08

A shared or common environment membrane reactor containing a plurality of planar membrane modules with top and bottom thin foil membranes supported by both an intermediary porous support plate and a central base which has both solid extended members and hollow regions or a hollow region whereby the two sides of the base are in fluid communication. The membrane reactor operates at elevate temperatures for generating hydrogen from hydrogen rich feed fuels.

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

A Gas-Cooled-Reactor Closed-Brayton-Cycle Demonstration with Nuclear Heating

NASA Astrophysics Data System (ADS)

Lipinski, Ronald J.; Wright, Steven A.; Dorsey, Daniel J.; Peters, Curtis D.; Brown, Nicholas; Williamson, Joshua; Jablonski, Jennifer

2005-02-01

A gas-cooled reactor may be coupled directly to turbomachinery to form a closed-Brayton-cycle (CBC) system in which the CBC working fluid serves as the reactor coolant. Such a system has the potential to be a very simple and robust space-reactor power system. Gas-cooled reactors have been built and operated in the past, but very few have been coupled directly to the turbomachinery in this fashion. In this paper we describe the option for testing such a system with a small reactor and turbomachinery at Sandia National Laboratories. Sandia currently operates the Annular Core Research Reactor (ACRR) at steady-state powers up to 4 MW and has an adjacent facility with heavy shielding in which another reactor recently operated. Sandia also has a closed-Brayton-Cycle test bed with a converted commercial turbomachinery unit that is rated for up to 30 kWe of power. It is proposed to construct a small experimental gas-cooled reactor core and attach this via ducting to the CBC turbomachinery for cooling and electricity production. Calculations suggest that such a unit could produce about 20 kWe, which would be a good power level for initial surface power units on the Moon or Mars. The intent of this experiment is to demonstrate the stable start-up and operation of such a system. Of particular interest is the effect of a negative temperature power coefficient as the initially cold Brayton gas passes through the core during startup or power changes. Sandia's dynamic model for such a system would be compared with the performance data. This paper describes the neutronics, heat transfer, and cycle dynamics of this proposed system. Safety and radiation issues are presented. The views expressed in this document are those of the author and do not necessarily reflect agreement by the government.

Optimization of operating parameters in polysilicon chemical vapor deposition reactor with response surface methodology

NASA Astrophysics Data System (ADS)

An, Li-sha; Liu, Chun-jiao; Liu, Ying-wen

2018-05-01

In the polysilicon chemical vapor deposition reactor, the operating parameters are complex to affect the polysilicon's output. Therefore, it is very important to address the coupling problem of multiple parameters and solve the optimization in a computationally efficient manner. Here, we adopted Response Surface Methodology (RSM) to analyze the complex coupling effects of different operating parameters on silicon deposition rate (R) and further achieve effective optimization of the silicon CVD system. Based on finite numerical experiments, an accurate RSM regression model is obtained and applied to predict the R with different operating parameters, including temperature (T), pressure (P), inlet velocity (V), and inlet mole fraction of H2 (M). The analysis of variance is conducted to describe the rationality of regression model and examine the statistical significance of each factor. Consequently, the optimum combination of operating parameters for the silicon CVD reactor is: T = 1400 K, P = 3.82 atm, V = 3.41 m/s, M = 0.91. The validation tests and optimum solution show that the results are in good agreement with those from CFD model and the deviations of the predicted values are less than 4.19%. This work provides a theoretical guidance to operate the polysilicon CVD process.

Design and analysis of a nuclear reactor core for innovative small light water reactors

NASA Astrophysics Data System (ADS)

Soldatov, Alexey I.

In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design issues related with increased fuel enrichment (8.0%) and specific MASLWR operational conditions (such as lower operational pressure and temperature, and increased leakage due to small core). Neutron physics calculations are performed with the commercial nuclear industry tools CASMO-4 and SIMULATE-3, developed by Studsvik Scandpower Inc. The first set of results are generated from infinite lattice level calculations with CASMO-4, and focus on evaluation of the principal differences between standard PWR fuel and MASLWR fuel. Chapter 4-1 covers aspects of fuel isotopic composition changes with burnup, evaluation of kinetic parameters and reactivity coefficients. Chapter 4-2 discusses gadolinium self-shielding and shadowing effects, and subsequent impacts on power generation peaking and Reactor Control System shadowing. The second aspect of the research is dedicated to core design issues, such as reflector design (chapter 4-3), burnable absorber distribution and programmed fuel burnup and fuel use strategy (chapter 4-4). This section also includes discussion of the parameters important for safety and evaluation of Reactor Control System options for the proposed core design. An evaluation of the sensitivity of the proposed design to uncertainty in calculated parameters is presented in chapter 4-5. The results presented in this dissertation cover a new area of reactor design and operational parameters, and may be applicable to other small and large pressurized water reactor designs.

EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR

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

REID, ROBERT S.; PEARSON, J. BOSIE; STEWART, ERIC T.

2007-01-16

Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WSTmore » is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.« less

TRACE/PARCS Analysis of ATWS with Instability for a MELLLA+BWR/5

DOE PAGES

L. Y. Cheng; Baek, J. S.; Cuadra, A.; ...

2016-06-06

A TRACE/PARCS model has been developed to analyze anticipated transient without SCRAM (ATWS) events for a boiling water reactor (BWR) operating in the maximum extended load line limit analysis-plus (MELLLA+) expanded operating domain. The MELLLA+ domain expands allowable operation in the power/flow map of a BWR to low flow rates at high power conditions. Such operation exacerbates the likelihood of large amplitude power/flow oscillations during certain ATWS scenarios. The analysis shows that large amplitude power/flow oscillations, both core-wide and out-of-phase, arise following the establishment of natural circulation flow in the reactor pressure vessel (RPV) after the trip of the recirculationmore » pumps and an increase in core inlet subcooling. The analysis also indicates a mechanism by which the fuel may experience heat-up that could result in localized fuel damage. TRACE predicts the heat-up to occur when the cladding surface temperature exceeds the minimum stable film boiling temperature after periodic cycles of dryout and rewet; and the fuel becomes “locked” into a film boiling regime. Further, the analysis demonstrates the effectiveness of the simulated manual operator actions to suppress the instability.« less

Study of parameters affecting the conversion in a plug flow reactor for reactions of the type 2A→B

NASA Astrophysics Data System (ADS)

Beltran-Prieto, Juan Carlos; Long, Nguyen Huynh Bach Son

2018-04-01

Modeling of chemical reactors is an important tool to quantify reagent conversion, product yield and selectivity towards a specific compound and to describe the behavior of the system. Proposal of differential equations describing the mass and energy balance are among the most important steps required during the modeling process as they play a special role in the design and operation of the reactor. Parameters governing transfer of heat and mass have a strong relevance in the rate of the reaction. Understanding this information is important for the selection of reactor and operating regime. In this paper we studied the irreversible gas-phase reaction 2A→B. We model the conversion that can be achieved as function of the reactor volume and feeding temperature. Additionally, we discuss the effect of activation energy and the heat of reaction on the conversion achieved in the tubular reactor. Furthermore, we considered that dimerization occurs instantaneously in the catalytic surface to develop equations for the determination of rate of reaction per unit area of three different catalytic surface shapes. This data can be combined with information about the global rate of conversion in the reactor to improve regent conversion and yield of product.

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.

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

Apparatus, components and operating methods for circulating fluidized bed transport gasifiers and reactors

DOEpatents

Vimalchand, Pannalal; Liu, Guohai; Peng, Wan Wang

2015-02-24

The improvements proposed in this invention provide a reliable apparatus and method to gasify low rank coals in a class of pressurized circulating fluidized bed reactors termed "transport gasifier." The embodiments overcome a number of operability and reliability problems with existing gasifiers. The systems and methods address issues related to distribution of gasification agent without the use of internals, management of heat release to avoid any agglomeration and clinker formation, specific design of bends to withstand the highly erosive environment due to high solid particles circulation rates, design of a standpipe cyclone to withstand high temperature gasification environment, compact design of seal-leg that can handle high mass solids flux, design of nozzles that eliminate plugging, uniform aeration of large diameter Standpipe, oxidant injection at the cyclone exits to effectively modulate gasifier exit temperature and reduction in overall height of the gasifier with a modified non-mechanical valve.

High temperature ultrasonic immersion measurements using a BS-PT based piezoelectric transducer without a delay line

NASA Astrophysics Data System (ADS)

Bilgunde, Prathamesh N.; Bond, Leonard J.

2018-04-01

Ultrasonic imaging is a key enabling technology required for in-service inspection of advanced sodium fast reactors at the hot stand-by operating mode (˜250C). Current work presents development of a single element, 2.4MHz, planar, ultrasonic immersion transducer for a potential application in ranging, inspection and imaging of the reactor components. The prototype immersion transducer is first tested in water for three thermal cycles up to 92C. The transducer is further evaluated for four thermal cycles in silicone oil, with total seven thermal cycles that exceeded operation period of 21 hours. Moreover, the preliminary data acquired for speed of sound in silicone oil indicates 24% reduction from 22C to 142C. Sensitivity of the ultrasonic transducer is also measured as a function of temperature and demonstrates the effect of multiple thermal cycles on the transducer components.

Pm-1 Reactor Core Final Design Report

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

Bagley, R. O.; Cox, F. H.; Carnasale, A.

1962-01-01

The PM-1 water cooled and moderated core contains 741 highly enriched stainless steel cermet tubular fuel elements and 90 lumped B stainless steel burnable poison elements, and it is controlled by 6 Y-shaped europium titanate movable control rods. The core has a lifetime of 1.95 years when operated at its design power level of 9.37 mw of thermal energy. The control of the core is designed so that there is a positive shutdown margin at all times with either one rod stuck completely out or the core or with two rods stuck in the operating condition. The core power ismore » removed by 2125 gpm of pressurized water at an average temperature of 463 deg F and pressure of 1300 psia. In reactors of this type, the core is stable with a negative temperature coefficient of approximately 2.5 x 10/sup -4/ DELTA K/K/ deg F.« less

Anaerobic treatment for C and S removal in "zero-discharge" paper mills: effects of process design on S removal efficiencies.

PubMed

van Lier, J B; Lens, P N; Pol, L W

2001-01-01

Stringent environmental laws in Europe and Northern America lead to the development towards closure of the process water streams in pulp and paper mills. Application of a "zero-discharge" process is already a feasible option for the board and packaging paper industry, provided in-line treatment is applied. Concomitant energy conservation inside the mill results in process water temperatures of 50-60 degrees C. Thermophilic anaerobic treatment complemented with appropriate post-treatment is considered as the most cost-effective solution to meet re-use criteria of the process water and to keep its temperature. In the proposed closed-cycle, the anaerobic treatment step removes the largest fraction of the biodegradable COD and eliminates "S" as H2S from the process stream, without the use of additional chemicals. The anaerobic step is regarded as the only possible location to bleed "S" from the process water cycle. In laboratory experiments, the effect of upward liquid velocity (Vupw) and the specific gas loading rate (Vgas) on the S removal capacity of thermophilic anaerobic bio-reactors was investigated. Acidifying, sulphate reducing sludge bed reactors were fed with partly acidified synthetic paper mill wastewater and were operated at 55 degrees C and pH 6. The reactors were operated at organic loading rates up to 50 g COD.l-1.day-1 at COD/SO4(2-) ratios of 10. The effect of Vupw was researched by comparing the performance of a UASB reactor operated at 1.0 m.h-1 and an EGSB reactor, operated at 6.8 m.h-1. The Vupw had a strong effect on the fermentation patterns. In the UASB reactor, acidification yielded H2, acetate and propionate, leading to an accumulation of reducing equivalents. These were partly disposed of by the production of n-butyrate and n-valerate from propionate. In the EGSB reactor net acetate consumption was observed as well as high volumetric gas (CO2 and CH4) production rates. The higher gas production rates in the EGSB reactor resulted in higher S-stripping efficiencies. The effect of Vgas was further researched by comparing 2 UASB reactors which were sparged with N2 gas at a specific gas loading rate of 30 m3.m-2.day-1. In contrast to the regular UASB reactors, the gas-supplied UASB showed a more stable performance when the organic loading rates were increased. Also, the H2S stripping efficiency was 3-4 times higher in the gas-supplied UASB, reaching values of 67%. Higher values were not obtained owing to the relatively poor sulphate reduction efficiencies.

Recovering hydrogen production performance of upflow anaerobic sludge blanket reactor (UASBR) fed with galactose via repeated heat treatment strategy.

PubMed

Sivagurunathan, Periyasamy; Anburajan, Parthiban; Kumar, Gopalakrishnan; Park, Jong-Hun; Kim, Sang-Hyoun

2017-09-01

This study evaluated the effect of repeated heat treatment towards the enhancement of hydrogen fermentation from galactose in an upflow anaerobic sludge blanket reactor with the hydraulic retention time of 6h and the operation temperature of 37°C. The hydrogen production rate (HPR) and hydrogen yield (HY) gradually increased up to 9.1L/L/d and 1.1mol/mol galactose, respectively, until the 33rd day of operation. When heat treatment at 80°C for 30min was applied, hydrogen production performance was enhanced by 37% with the enrichment of hydrogen producing bacteria population. The HPR and HY were achieved at 12.5L/L/d and 1.5mol/mol hexose, respectively, during further 30 cycles of reactor operation. The repeated heat treatment would be a viable strategy to warrant reliable continuous hydrogen production using mixed culture. Copyright © 2017 Elsevier Ltd. All rights reserved.

Compatibility of refractory materials for nuclear reactor poison control systems

NASA Technical Reports Server (NTRS)

Sinclair, J. H.

1974-01-01

Metal-clad poison rods have been considered for the control system of an advanced space power reactor concept studied at the NASA Lewis Research Center. Such control rods may be required to operate at temperatures of about 140O C. Selected poison materials (including boron carbide and the diborides of zirconium, hafnium, and tantalum) were subjected to 1000-hour screening tests in contact with candidate refractory metal cladding materials (including tungsten and alloys of tantalum, niobium, and molybdenum) to assess the compatibility of these materials combinations at the temperatures of interest. Zirconium and hafnium diborides were compatible with refractory metals at 1400 C, but boron carbide and tantalum diboride reacted with the refractory metals at this temperature. Zirconium diboride also showed promise as a reaction barrier between boron carbide and tungsten.

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

Microstructure and Property Evolution in Advanced Cladding and Duct Materials Under Long-Term and Elevated Temperature Irradiation: Modeling and Experimental Investigation

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

Wirth, Brian; Morgan, Dane; Kaoumi, Djamel

2013-12-01

The in-service degradation of reactor core materials is related to underlying changes in the irradiated microstructure. During reactor operation, structural components and cladding experience displacement of atoms by collisions with neutrons at temperatures at which the radiation-induced defects are mobile, leading to microstructure evolution under irradiation that can degrade material properties. At the doses and temperatures relevant to fast reactor operation, the microstructure evolves by dislocation loop formation and growth, microchemistry changes due to radiation-induced segregation, radiation-induced precipitation, destabilization of the existing precipitate structure, and in some cases, void formation and growth. These processes do not occur independently; rather, theirmore » evolution is highly interlinked. Radiationinduced segregation of Cr and existing chromium carbide coverage in irradiated alloy T91 track each other closely. The radiation-induced precipitation of Ni-Si precipitates and RIS of Ni and Si in alloys T91 and HCM12A are likely related. Neither the evolution of these processes nor their coupling is understood under the conditions required for materials performance in fast reactors (temperature range 300-600°C and doses beyond 200 dpa). Further, predictive modeling is not yet possible as models for microstructure evolution must be developed along with experiments to characterize these key processes and provide tools for extrapolation. To extend the range of operation of nuclear fuel cladding and structural materials in advanced nuclear energy and transmutation systems to that required for the fast reactor, the irradiation-induced evolution of the microstructure, microchemistry, and the associated mechanical properties at relevant temperatures and doses must be understood. Predictive modeling relies on an understanding of the physical processes and also on the development of microstructure and microchemical models to describe their evolution under irradiation. This project will focus on modeling microstructural and microchemical evolution of irradiated alloys by performing detailed modeling of such microstructure evolution processes coupled with well-designed in situ experiments that can provide validation and benchmarking to the computer codes. The broad scientific and technical objectives of this proposal are to evaluate the microstructure and microchemical evolution in advanced ferritic/martensitic and oxide dispersion strengthened (ODS) alloys for cladding and duct reactor materials under long-term and elevated temperature irradiation, leading to improved ability to model structural materials performance and lifetime. Specifically, we propose four research thrusts, namely Thrust 1: Identify the formation mechanism and evolution for dislocation loops with Burgers vector of a<100> and determine whether the defect microstructure (predominately dislocation loop/dislocation density) saturates at high dose. Thrust 2: Identify whether a threshold irradiation temperature or dose exists for the nucleation of growing voids that mark the beginning of irradiation-induced swelling, and begin to probe the limits of thermal stability of the tempered Martensitic structure under irradiation. Thrust 3: Evaluate the stability of nanometer sized Y- Ti-O based oxide dispersion strengthened (ODS) particles at high fluence/temperature. Thrust 4: Evaluate the extent to which precipitates form and/or dissolve as a function of irradiation temperature and dose, and how these changes are driven by radiation induced segregation and microchemical evolutions and determined by the initial microstructure.« less

A CFD model for biomass fast pyrolysis in fluidized-bed reactors

NASA Astrophysics Data System (ADS)

Xue, Qingluan; Heindel, T. J.; Fox, R. O.

2010-11-01

A numerical study is conducted to evaluate the performance and optimal operating conditions of fluidized-bed reactors for fast pyrolysis of biomass to bio-oil. A comprehensive CFD model, coupling a pyrolysis kinetic model with a detailed hydrodynamics model, is developed. A lumped kinetic model is applied to describe the pyrolysis of biomass particles. Variable particle porosity is used to account for the evolution of particle physical properties. The kinetic scheme includes primary decomposition and secondary cracking of tar. Biomass is composed of reference components: cellulose, hemicellulose, and lignin. Products are categorized into groups: gaseous, tar vapor, and solid char. The particle kinetic processes and their interaction with the reactive gas phase are modeled with a multi-fluid model derived from the kinetic theory of granular flow. The gas, sand and biomass constitute three continuum phases coupled by the interphase source terms. The model is applied to investigate the effect of operating conditions on the tar yield in a fluidized-bed reactor. The influence of various parameters on tar yield, including operating temperature and others are investigated. Predicted optimal conditions for tar yield and scale-up of the reactor are discussed.

An Integrated Chemical Reactor-Heat Exchanger Based on Ammonium Carbamate (POSTPRINT)

DTIC Science & Technology

2012-10-01

With the scrubber and exhaust operating, the test cell ammonia concentration remains below 5 ppm. To further reduce NH3 release into the test cell...material has a high decomposition enthalpy and exhibits decomposition over a wide range of temperatures. AC decomposition produces ammonia and carbon...installation due to toxic gas ( ammonia ) generation during operation. Therefore, the experiment is intended to be remotely operated. A secondary control

Anaerobic treatability of wastewater contaminated with propylene glycol.

PubMed

Sezgin, Naim; Tonuk, Gulseven Ubay

2013-09-01

The purpose of this study was to investigate the biodegradability of propylene glycol in anaerobic conditions by using methanogenic culture. A master reactor was set up to develop a culture that would be acclimated to propylene glycol. After reaching steady-state, culture was transferred to serum bottles. Three reactors with same initial conditions were run for consistency. Propylene glycol was completely biodegradable under anaerobic methanogenic conditions. Semi-continuous reactors operated at a temperature of 35°C had consistently achieved a propylene glycol removal of higher than 95 % based on chemical oxygen demand (COD). It was found that in semi-continuous reactors, anaerobic treatment of propylene glycol at concentrations higher than 1,500 mg COD m(-3) day(-1) was not convenient due to instable effluent COD.

CFD Analysis of Upper Plenum Flow for a Sodium-Cooled Small Modular Reactor

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

Kraus, A.; Hu, R.

2015-01-01

Upper plenum flow behavior is important for many operational and safety issues in sodium fast reactors. The Prototype Gen-IV Sodium Fast Reactor (PGSFR), a pool-type, 150 MWe output power design, was used as a reference case for a detailed characterization of upper plenum flow for normal operating conditions. Computational Fluid Dynamics (CFD) simulation was utilized with detailed geometric modeling of major structures. Core outlet conditions based on prior system-level calculations were mapped to approximate the outlet temperatures and flow rates for each core assembly. Core outlet flow was found to largely bypass the Upper Internal Structures (UIS). Flow curves overmore » the shield and circulates within the pool before exiting the plenum. Cross-flows and temperatures were evaluated near the core outlet, leading to a proposed height for the core outlet thermocouples to ensure accurate assembly-specific temperature readings. A passive scalar was used to evaluate fluid residence time from core outlet to IHX inlet, which can be used to assess the applicability of various methods for monitoring fuel failure. Additionally, the gas entrainment likelihood was assessed based on the CFD simulation results. Based on the evaluation of velocity gradients and turbulent kinetic energies and the available gas entrainment criteria in the literature, it was concluded that significant gas entrainment is unlikely for the current PGSFR design.« less

Modeling the UO2 ex-AUC pellet process and predicting the fuel rod temperature distribution under steady-state operating condition

NASA Astrophysics Data System (ADS)

Hung, Nguyen Trong; Thuan, Le Ba; Thanh, Tran Chi; Nhuan, Hoang; Khoai, Do Van; Tung, Nguyen Van; Lee, Jin-Young; Jyothi, Rajesh Kumar

2018-06-01

Modeling uranium dioxide pellet process from ammonium uranyl carbonate - derived uranium dioxide powder (UO2 ex-AUC powder) and predicting fuel rod temperature distribution were reported in the paper. Response surface methodology (RSM) and FRAPCON-4.0 code were used to model the process and to predict the fuel rod temperature under steady-state operating condition. Fuel rod design of AP-1000 designed by Westinghouse Electric Corporation, in these the pellet fabrication parameters are from the study, were input data for the code. The predictive data were suggested the relationship between the fabrication parameters of UO2 pellets and their temperature image in nuclear reactor.

Quantification of syntrophic acetate-oxidizing microbial communities in biogas processes

PubMed Central

Westerholm, Maria; Dolfing, Jan; Sherry, Angela; Gray, Neil D; Head, Ian M; Schnürer, Anna

2011-01-01

Changes in communities of syntrophic acetate-oxidizing bacteria (SAOB) and methanogens caused by elevated ammonia levels were quantified in laboratory-scale methanogenic biogas reactors operating at moderate temperature (37°C) using quantitative polymerase chain reaction (qPCR). The experimental reactor was subjected to gradually increasing ammonia levels (0.8–6.9 g NH4+-N l−1), whereas the level of ammonia in the control reactor was kept low (0.65–0.90 g NH4+-N l−1) during the entire period of operation (660 days). Acetate oxidation in the experimental reactor, indicated by increased production of 14CO2 from acetate labelled in the methyl carbon, occurred when ammonia levels reached 5.5 and 6.9 g NH4+-N l−1. Syntrophic acetate oxidizers targeted by newly designed qPCR primers were Thermacetogenium phaeum, Clostridium ultunense, Syntrophaceticus schinkii and Tepidanaerobacter acetatoxydans. The results showed a significant increase in abundance of all these bacteria except T. phaeum in the ammonia-stressed reactor, coincident with the shift to syntrophic acetate oxidation. As the abundance of the bacteria increased, a simultaneous decrease was observed in the abundance of aceticlastic methanogens from the families Methanosaetaceae and Methanosarcinaceae. qPCR analyses of sludge from two additional high ammonia processes, in which methane production from acetate proceeded through syntrophic acetate oxidation (reactor SB) or through aceticlastic degradation (reactor DVX), demonstrated that SAOB were significantly more abundant in the SB reactor than in the DVX reactor. PMID:23761313

Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

NASA Astrophysics Data System (ADS)

Hashim, Zaredah; Lanyau, Tonny Anak; Farid, Mohamad Fairus Abdul; Kassim, Mohammad Suhaimi; Azhar, Noraishah Syahirah

2016-01-01

The 1-MW PUSPATI Research Reactor (RTP) is the one and only nuclear pool type research reactor developed by General Atomic (GA) in Malaysia. It was installed at Malaysian Nuclear Agency and has reached the first criticality on 8 June 1982. Based on the initial core which comprised of 80 standard TRIGA fuel elements, the very fundamental thermal hydraulic model was investigated during steady state operation using the PARET-code. The main objective of this paper is to determine the variation of temperature profiles and Departure of Nucleate Boiling Ratio (DNBR) of RTP at full power operation. The second objective is to confirm that the values obtained from PARET-code are in agreement with Safety Analysis Report (SAR) for RTP. The code was employed for the hot and average channels in the core in order to calculate of fuel's center and surface, cladding, coolant temperatures as well as DNBR's values. In this study, it was found that the results obtained from the PARET-code showed that the thermal hydraulic parameters related to safety for initial core which was cooled by natural convection was in agreement with the designed values and safety limit in SAR.

Synthesis of biodiesel from waste cooking oil using sonochemical reactors.

PubMed

Hingu, Shishir M; Gogate, Parag R; Rathod, Virendra K

2010-06-01

Investigation into newer routes of biodiesel synthesis is a key research area especially due to the fluctuations in the conventional fuel prices and the environmental advantages of biodiesel. The present work illustrates the use of sonochemical reactors for the synthesis of biodiesel from waste cooking oil. Transesterification of used frying oil with methanol, in the presence of potassium hydroxide as a catalyst has been investigated using low frequency ultrasonic reactor (20 kHz). Effect of different operating parameters such as alcohol-oil molar ratio, catalyst concentration, temperature, power, pulse and horn position on the extent of conversion of oil have been investigated. The optimum conditions for the transesterification process have been obtained as molar ratio of alcohol to oil as 6:1, catalyst concentration of 1 wt.%, temperature as 45 degrees C and ultrasound power as 200 W with an irradiation time of 40 min. The efficacy of using ultrasound has been compared with the conventional stirring approach based on the use of a six blade turbine with diameter of 1.5 cm operating at 1000 rpm. Also the purification aspects of the final product have been investigated. (c) 2010 Elsevier B.V. All rights reserved.

Flat-plate solar array project. Task 1: Silicon material. Investigation of the hydrochlorination of SiC14

NASA Technical Reports Server (NTRS)

Mui, J. Y. P.

1982-01-01

A two inch diameter stainless steel reactor was designed and built to operate at pressures up to 500 psig for the experimental studies on the hydrochlorination of SiCl4 and metallurgical grade (m.g.) silicon metal to SiHCl3. In order to clearly see the effect of pressure, the experiments were carried out at low reactor pressures of 73 psig and 150 psig, respectively. A large pressure effect on the hydrochlorination reaction was observed between the results of the low pressure experiments and the results of the high pressure experiments. In general, higher pressure produces a higher conversion of SiHCl3, but at a lower reaction rate. The effect of temperature on the reaction rate was studied at 73 psig. Higher reaction temperature gave a higher conversion and a higher reaction rate. Samples of the materials used to construct the hydrochlorination reactor were prepared for corrosion tests.

Coal desulfurization by low temperature chlorinolysis, phase 3

NASA Technical Reports Server (NTRS)

Kalvinskas, J. J.; Rohatgi, N. K.; Ernest, J.

1981-01-01

Laboratory scale, bench scale batch reactor, and minipilot plant tests were conducted on 22 bituminous, subbituminous, and lignite coals. Chemical pretreatment and post treatment of coals relative to the chlorination were tried as a means of enhancing desulfurization by the chlorinolysis process. Elevated temperature (500-700 C) hydrogen treatment of chlorinolysis-processed coal at atmospheric pressure was found to substantially increase coal desulfurization up to 90 percent. Sulfur forms, proximate and ultimate analyses of the processed coal are included. Minipilot plant operation indicates that the continuous flow reactor provides coal desulfurization results comparable to those obtained in the batch reactor. Seven runs were conducted at coal feed rates of 1.5 to 8.8 kg per hour using water and methylchloroform solvents, gaseous chlorine feed of 3 to 31.4 SCFH at 21 to 70 C, and atmospheric pressure for retention times of 20 to 120 minutes.

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.

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

Recovery and biological oxidation of dissolved methane in effluent from UASB treatment of municipal sewage using a two-stage closed downflow hanging sponge system.

PubMed

Matsuura, Norihisa; Hatamoto, Masashi; Sumino, Haruhiko; Syutsubo, Kazuaki; Yamaguchi, Takashi; Ohashi, Akiyoshi

2015-03-15

A two-stage closed downflow hanging sponge (DHS) reactor was used as a post-treatment to prevent methane being emitted from upflow anaerobic sludge blanket (UASB) effluents containing unrecovered dissolved methane. The performance of the closed DHS reactor was evaluated using real municipal sewage at ambient temperatures (10-28 °C) for one year. The first stage of the closed DHS reactor was intended to recover dissolved methane from the UASB effluent and produce a burnable gas with a methane concentration greater than 30%, and its recovery efficiency was 57-88%, although the amount of dissolved methane in the UASB effluent fluctuated in the range of 46-68 % of methane production greatly depending on the temperature. The residual methane was oxidized and the remaining organic carbon was removed in the second closed DHS reactor, and this reactor performed very well, removing more than 99% of the dissolved methane during the experimental period. The rate at which air was supplied to the DHS reactor was found to be one of the most important operating parameters. Microbial community analysis revealed that seasonal changes in the methane-oxidizing bacteria were key to preventing methane emissions. Copyright © 2014 Elsevier Ltd. All rights reserved.

Effect of operational conditions on the degradation of organic matter and development of microalgae-bacteria consortia when treating swine slurry.

PubMed

González-Fernández, Cristina; Riaño-Irazábal, Berta; Molinuevo-Salces, Beatriz; Blanco, Saúl; García-González, Maria Cruz

2011-05-01

There is great controversy regarding the best substrate (fresh or anaerobically digested swine slurry) for the development of microalgae-bacteria consortia. This study aims to elucidate the best substrate by assessing biomass productivity, microorganism predominance, and their ability for organic matter removal. In addition to the different substrates, different operational conditions and influent strengths were evaluated. Increasing organic matter content when favourable temperature and illumination conditions were present improved biomass production. However, these conditions were not favourable for microalgal growth, but they were favourable for bacteria. Regardless of the operational conditions, reactors fed with fresh slurry not only resulted in the highest biomass productivity, but also the greatest removal of total and soluble chemical oxygen demand (COD). On the other hand, reactors fed with digested slurry showed biomass productivity and COD removal values lower than those obtained for reactors fed with fresh slurry, most probably due to the recalcitrant nature of the former. Nevertheless, digested slurry was the substrate more appropriate for microalgae growth under harsh operational conditions (16 °C and 9-h illumination) at low influent strength and optimum operational conditions (30 °C and 24-h illumination) at higher influent strength.

The effects of temperatures on the pebble flow in a pebble bed high temperature reactor

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

Sen, R. S.; Cogliati, J. J.; Gougar, H. D.

2012-07-01

The core of a pebble bed high temperature reactor (PBHTR) moves during operation, a feature which leads to better fuel economy (online refueling with no burnable poisons) and lower fuel stress. The pebbles are loaded at the top and trickle to the bottom of the core after which the burnup of each is measured. The pebbles that are not fully burned are recirculated through the core until the target burnup is achieved. The flow pattern of the pebbles through the core is of importance for core simulations because it couples the burnup distribution to the core temperature and power profiles,more » especially in cores with two or more radial burnup 'zones '. The pebble velocity profile is a strong function of the core geometry and the friction between the pebbles and the surrounding structures (other pebbles or graphite reflector blocks). The friction coefficient for graphite in a helium environment is inversely related to the temperature. The Thorium High Temperature Reactor (THTR) operated in Germany between 1983 and 1989. It featured a two-zone core, an inner core (IC) and outer core (OC), with different fuel mixtures loaded in each zone. The rate at which the IC was refueled relative to the OC in THTR was designed to be 0.56. During its operation, however, this ratio was measured to be 0.76, suggesting the pebbles in the inner core traveled faster than expected. It has been postulated that the positive feedback effect between inner core temperature, burnup, and pebble flow was underestimated in THTR. Because of the power shape, the center of the core in a typical cylindrical PBHTR operates at a higher temperature than the region next to the side reflector. The friction between pebbles in the IC is lower than that in the OC, perhaps causing a higher relative flow rate and lower average burnup, which in turn yield a higher local power density. Furthermore, the pebbles in the center region have higher velocities than the pebbles next to the side reflector due to the interaction between the pebbles and the immobile graphite reflector as well as the geometry of the discharge conus near the bottom of the core. In this paper, the coupling between the temperature profile and the pebble flow dynamics was analyzed by using PEBBED/THERMIX and PEBBLES codes by modeling the HTR-10 reactor in China. Two extreme and opposing velocity profiles are used as a starting point for the iterations. The PEBBED/THERMIX code is used to calculate the burnup, power and temperature profiles with one of the velocity profiles as input. The resulting temperature profile is then passed to PEBBLES code to calculate the updated pebble velocity profile taking the new temperature profile into account. If the aforementioned hypothesis is correct, the strong temperature effect upon the friction coefficients would cause the two cases to converge to different final velocity and temperature profiles. The results of this analysis indicates that a single zone pebble bed core is self-stabilizing in terms of the pebble velocity profile and the effect of the temperature profile on the pebble flow is insignificant. (authors)« less

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

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

Flat-plate collector research area: Silicon material task

NASA Technical Reports Server (NTRS)

Lutwack, R.

1982-01-01

Silane decomposition in a fluidized-bed reactor (FBR) process development unit (PDU) to make semiconductor-grade Si is reviewed. The PDU was modified by installation of a new heating system to provide the required temperature profile and better control, and testing was resumed. A process for making trichlorosilane by the hydrochlorination of metallurgical-grade Si and silicon tetrachloride is reported. Fabrication and installation of the test system employing a new 2-in.-dia reactor was completed. A process that converts trichlorosilane to dichlorosilane (DCS), which is reduced by hydrogen to make Si by a chemical vapor deposition step in a Siemens-type reactor is described. Testing of the DCS PDU integraled with Si deposition reactors continued. Experiments in a 2-in.-dia reactor to define the operating window and to investigate the Si deposition kinetics were completed.

Effect of small-scale biomass gasification at the state of refractory lining the fixed bed reactor

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

Janša, Jan, E-mail: jan.jansa@vsb.cz; Peer, Vaclav, E-mail: vaclav.peer@vsb.cz; Pavloková, Petra, E-mail: petra.pavlokova@vsb.cz

The article deals with the influence of biomass gasification on the condition of the refractory lining of a fixed bed reactor. The refractory lining of the gasifier is one part of the device, which significantly affects the operational reliability and durability. After removing the refractory lining of the gasifier from the experimental reactor, there was done an assessment how gasification of different kinds of biomass reflected on its condition in terms of the main factors affecting its life. Gasification of biomass is reflected on the lining, especially through sticking at the bottom of the reactor. Measures for prolonging the lifemore » of lining consist in the reduction of temperature in the reactor, in this case, in order to avoid ash fusion biomass which it is difficult for this type of gasifier.« less

Heat pipe nuclear reactor for space power

NASA Technical Reports Server (NTRS)

Koening, D. R.

1976-01-01

A heat-pipe-cooled nuclear reactor has been designed to provide 3.2 MWth to an out-of-core thermionic conversion system. The reactor is a fast reactor designed to operate at a nominal heat-pipe temperature of 1675 K. Each reactor fuel element consists of a hexagonal molybdenum block which is bonded along its axis to one end of a molybdenum/lithium-vapor heat pipe. The block is perforated with an array of longitudinal holes which are loaded with UO2 pellets. The heat pipe transfers heat directly to a string of six thermionic converters which are bonded along the other end of the heat pipe. An assembly of 90 such fuel elements forms a hexagonal core. The core is surrounded by a thermal radiation shield, a thin thermal neutron absorber, and a BeO reflector containing boron-loaded control drums.

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.

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

Burgett, Eric; Al-Sheikhly, Mohamad; Summers, Christopher

An advanced in-pile multi-parameter reactor monitoring system is being proposed in this funding opportunity. The proposed effort brings cutting edge, high-fidelity optical measurement systems into the reactor environment in an unprecedented fashion, including in-core, in-cladding and in-fuel pellet itself. Unlike instrumented leads, the proposed system provides a unique solution to a multi-parameter monitoring need in core while being minimally intrusive in the reactor core. Detector designs proposed herein can monitor fuel compression and expansion in both the radial and axial dimensions as well as monitor linear power profiles and fission rates during the operation of the reactor. In addition tomore » pressure, stress, strain, compression, neutron flux, neutron spectra, and temperature can be observed inside the fuel bundle and fuel rod using the proposed system. The proposed research aims at developing radiation-hard, harsh-environment multi-parameter systems for insertion into the reactor environment. The proposed research holds the potential to drastically increase the fidelity and precision of in-core instrumentation with little or no impact in the neutron economy in the reactor environment while providing a measurement system capable of operation for entire operating cycles. Significant work has been done over the last few years on the use of nanoparticle-based scintillators. Through the use of metamaterials, the PIs aim to develop planar neutron detectors and large-volume neutron detectors. These detectors will have high efficiencies for neutron detection and will have a high gamma discrimination capability.« less

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

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.

Catalysis in high-temperature fuel cells.

PubMed

Föger, K; Ahmed, K

2005-02-17

Catalysis plays a critical role in solid oxide fuel cell systems. The electrochemical reactions within the cell--oxygen dissociation on the cathode and electrochemical fuel combustion on the anode--are catalytic reactions. The fuels used in high-temperature fuel cells, for example, natural gas, propane, or liquid hydrocarbons, need to be preprocessed to a form suitable for conversion on the anode-sulfur removal and pre-reforming. The unconverted fuel (economic fuel utilization around 85%) is commonly combusted using a catalytic burner. Ceramic Fuel Cells Ltd. has developed anodes that in addition to having electrochemical activity also are reactive for internal steam reforming of methane. This can simplify fuel preprocessing, but its main advantage is thermal management of the fuel cell stack by endothermic heat removal. Using this approach, the objective of fuel preprocessing is to produce a methane-rich fuel stream but with all higher hydrocarbons removed. Sulfur removal can be achieved by absorption or hydro-desulfurization (HDS). Depending on the system configuration, hydrogen is also required for start-up and shutdown. Reactor operating parameters are strongly tied to fuel cell operational regimes, thus often limiting optimization of the catalytic reactors. In this paper we discuss operation of an authothermal reforming reactor for hydrogen generation for HDS and start-up/shutdown, and development of a pre-reformer for converting propane to a methane-rich fuel stream.

Reactor Simulator Testing

NASA Technical Reports Server (NTRS)

Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise J.

2013-01-01

As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

Reactor Simulator Integration and Testing

NASA Technical Reports Server (NTRS)

Schoenfield, M. P.; Webster, K. L.; Pearson, J. B.

2013-01-01

As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator (RxSim) test loop was designed and built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing were to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V because the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This Technical Memorandum summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained, which was lower than the predicted 750 K but 156 K higher than the cold temperature, indicating the design provided some heat regeneration. The annular linear induction pump tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

Feasibility study of full-reactor gas core demonstration test

NASA Technical Reports Server (NTRS)

Kunze, J. F.; Lofthouse, J. H.; Shaffer, C. J.; Macbeth, P. J.

1973-01-01

Separate studies of nuclear criticality, flow patterns, and thermodynamics for the gas core reactor concept have all given positive indications of its feasibility. However, before serious design for a full scale gas core application can be made, feasibility must be shown for operation with full interaction of the nuclear, thermal, and hydraulic effects. A minimum sized, and hence minimum expense, test arrangement is considered for a full gas core configuration. It is shown that the hydrogen coolant scattering effects dominate the nuclear considerations at elevated temperatures. A cavity diameter of somewhat larger than 4 ft (122 cm) will be needed if temperatures high enough to vaporize uranium are to be achieved.

Engineering kinetics of short residence time coal liquefaction processes

NASA Astrophysics Data System (ADS)

Traeger, R. K.

1980-06-01

Conversion of coal to liquid products occurs rapidly at temperatures over 350 C and can be significant in preheaters or short residence time reactors. The extent of conversion can have an effect on the operation of preheaters or effectiveness of subsequent reactors. To obtain process information, Illinois No. 6 coal in SRC II heavy distillate was reacted at 13.8 MPa, temperatures of 400, 425, and 450 C, and at slurry space velocities of 3200-96,000 kg/h-cu m. Product compositions and viscosities were measured. High concentrations of preasphaltenes occur in early reactions resulting in a high viscosity product, but subsequent reactions to asphaltenes and oils are less rapid.

Bayesian network representing system dynamics in risk analysis of nuclear systems

NASA Astrophysics Data System (ADS)

Varuttamaseni, Athi

2011-12-01

A dynamic Bayesian network (DBN) model is used in conjunction with the alternating conditional expectation (ACE) regression method to analyze the risk associated with the loss of feedwater accident coupled with a subsequent initiation of the feed and bleed operation in the Zion-1 nuclear power plant. The use of the DBN allows the joint probability distribution to be factorized, enabling the analysis to be done on many simpler network structures rather than on one complicated structure. The construction of the DBN model assumes conditional independence relations among certain key reactor parameters. The choice of parameter to model is based on considerations of the macroscopic balance statements governing the behavior of the reactor under a quasi-static assumption. The DBN is used to relate the peak clad temperature to a set of independent variables that are known to be important in determining the success of the feed and bleed operation. A simple linear relationship is then used to relate the clad temperature to the core damage probability. To obtain a quantitative relationship among different nodes in the DBN, surrogates of the RELAP5 reactor transient analysis code are used. These surrogates are generated by applying the ACE algorithm to output data obtained from about 50 RELAP5 cases covering a wide range of the selected independent variables. These surrogates allow important safety parameters such as the fuel clad temperature to be expressed as a function of key reactor parameters such as the coolant temperature and pressure together with important independent variables such as the scram delay time. The time-dependent core damage probability is calculated by sampling the independent variables from their probability distributions and propagate the information up through the Bayesian network to give the clad temperature. With the knowledge of the clad temperature and the assumption that the core damage probability has a one-to-one relationship to it, we have calculated the core damage probably as a function of transient time. The use of the DBN model in combination with ACE allows risk analysis to be performed with much less effort than if the analysis were done using the standard techniques.

Some methods for achieving more efficient performance of fuel assemblies

NASA Astrophysics Data System (ADS)

Boltenko, E. A.

2014-07-01

More efficient operation of reactor plant fuel assemblies can be achieved through the use of new technical solutions aimed at obtaining more uniform distribution of coolant over the fuel assembly section, more intense heat removal on convex heat-transfer surfaces, and higher values of departure from nucleate boiling ratio (DNBR). Technical solutions using which it is possible to obtain more intense heat removal on convex heat-transfer surfaces and higher DNBR values in reactor plant fuel assemblies are considered. An alternative heat removal arrangement is described using which it is possible to obtain a significantly higher power density in a reactor plant and essentially lower maximal fuel rod temperature.

HIGH TEMPERATURE, HIGH POWER HETEROGENEOUS NUCLEAR REACTOR

DOEpatents

Hammond, R.P.; Wykoff, W.R.; Busey, H.M.

1960-06-14

A heterogeneous nuclear reactor is designed comprising a stationary housing and a rotatable annular core being supported for rotation about a vertical axis in the housing, the core containing a plurality of radial fuel- element supporting channels, the cylindrical empty space along the axis of the core providing a central plenum for the disposal of spent fuel elements, the core cross section outer periphery being vertically gradated in radius one end from the other to provide a coolant duct between the core and the housing, and means for inserting fresh fuel elements in the supporting channels under pressure and while the reactor is in operation.

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.

A comparative study of thermophilic and mesophilic anaerobic co-digestion of food waste and wheat straw: Process stability and microbial community structure shifts.

PubMed

Shi, Xuchuan; Guo, Xianglin; Zuo, Jiane; Wang, Yajiao; Zhang, Mengyu

2018-05-01

Renewable energy recovery from organic solid waste via anaerobic digestion is a promising way to provide sustainable energy supply and eliminate environmental pollution. However, poor efficiency and operational problems hinder its wide application of anaerobic digestion. The effects of two key parameters, i.e. temperature and substrate characteristics on process stability and microbial community structure were studied using two lab-scale anaerobic reactors under thermophilic and mesophilic conditions. Both the reactors were fed with food waste (FW) and wheat straw (WS). The organic loading rates (OLRs) were maintained at a constant level of 3 kg VS/(m 3 ·d). Five different FW:WS substrate ratios were utilized in different operational phases. The synergetic effects of co-digestion improved the stability and performance of the reactors. When FW was mono-digested, both reactors were unstable. The mesophilic reactor eventually failed due to volatile fatty acid accumulation. The thermophilic reactor had better performance compared to mesophilic one. The biogas production rate of the thermophilic reactor was 4.9-14.8% higher than that of mesophilic reactor throughout the experiment. The shifts in microbial community structures throughout the experiment in both thermophilic and mesophilic reactors were investigated. With increasing FW proportions, bacteria belonging to the phylum Thermotogae became predominant in the thermophilic reactor, while the phylum Bacteroidetes was predominant in the mesophilic reactor. The genus Methanosarcina was the predominant methanogen in the thermophilic reactor, while the genus Methanothrix remained predominant in the mesophilic reactor. The methanogenesis pathway shifted from acetoclastic to hydrogenotrophic when the mesophilic reactor experienced perturbations. Moreover, the population of lignocellulose-degrading microorganisms in the thermophilic reactor was higher than those in mesophilic reactor, which explained the better performance of the thermophilic reactor. Copyright © 2018. Published by Elsevier Ltd.

Pauson-Khand reactions in a photochemical flow microreactor.

PubMed

Asano, Keisuke; Uesugi, Yuki; Yoshida, Jun-ichi

2013-05-17

Pauson-Khand reactions were achieved at ambient temperature without any additive using a photochemical flow microreactor. The efficiency of the reaction was better than that in a conventional batch reactor, and the reaction could be operated continuously for 1 h.

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

Pyrolysis of furan in a microreactor

NASA Astrophysics Data System (ADS)

Urness, Kimberly N.; Guan, Qi; Golan, Amir; Daily, John W.; Nimlos, Mark R.; Stanton, John F.; Ahmed, Musahid; Ellison, G. Barney

2013-09-01

A silicon carbide microtubular reactor has been used to measure branching ratios in the thermal decomposition of furan, C4H4O. The pyrolysis experiments are carried out by passing a dilute mixture of furan (approximately 0.01%) entrained in a stream of helium through the heated reactor. The SiC reactor (0.66 mm i.d., 2 mm o.d., 2.5 cm long) operates with continuous flow. Experiments were performed with a reactor inlet pressure of 100-300 Torr and a wall temperature between 1200 and 1600 K; characteristic residence times in the reactor are 60-150 μs. The unimolecular decomposition pathway of furan is confirmed to be: furan (+ M) rightleftharpoons α-carbene or β-carbene. The α-carbene fragments to CH2=C=O + HC≡CH while the β-carbene isomerizes to CH2=C=CHCHO. The formyl allene can isomerize to CO + CH3C≡CH or it can fragment to H + CO + HCCCH2. Tunable synchrotron radiation photoionization mass spectrometry is used to monitor the products and to measure the branching ratio of the two carbenes as well as the ratio of [HCCCH2]/[CH3C≡CH]. The results of these pyrolysis experiments demonstrate a preference for 80%-90% of furan decomposition to occur via the β-carbene. For reactor temperatures of 1200-1400 K, no propargyl radicals are formed. As the temperature rises to 1500-1600 K, at most 10% of the decomposition of CH2=C=CHCHO produces H + CO + HCCCH2 radicals. Thermodynamic conditions in the reactor have been modeled by computational fluid dynamics and the experimental results are compared to the predictions of three furan pyrolysis mechanisms. Uncertainty in the pressure-dependency of the initiation reaction rates is a possible a source of discrepancy between experimental results and theoretical predictions.

Poisoning of Heat Pipes

NASA Technical Reports Server (NTRS)

Gillies, Donald; Lehoczky, Sandor; Palosz, Witold; Carpenter, Paul; Salvail, Pat

2007-01-01

Thermal management is critical to space exploration efforts. In particular, efficient transfer and control of heat flow is essential when operating high energy sources such as nuclear reactors. Thermal energy must be transferred to various energy conversion devices, and to radiators for safe and efficient rejection of excess thermal energy. Applications for space power demand exceptionally long periods of time with equipment that is accessible for limited maintenance only. Equally critical is the hostile and alien environment which includes high radiation from the reactor and from space (galactic) radiation. In space or lunar applications high vacuum is an issue, while in Martian operations the systems will encounter a CO2 atmosphere. The effect of contact at high temperature with local soil (regolith) in surface operations on the moon or other terrestrial bodies (Mars, asteroids) must be considered.

A review of carbide fuel corrosion for nuclear thermal propulsion applications

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; El-Genk, Mohamed S.; Butt, Darryl P.

1993-10-01

At the operation conditions of interest in nuclear thermal propulsion reactors, carbide materials have been known to exhibit a number of life limiting phenomena. These include the formation of liquid, loss by vaporization, creep and corresponding gas flow restrictions, and local corrosion and fuel structure degradation due to excessive mechanical and/or thermal loading. In addition, the radiation environment in the reactor core can produce a substantial change in its local physical properties, which can produce high thermal stresses and corresponding stress fractures (cracking). Time-temperature history and cyclic operation of the nuclear reactor can also accelerate some of these processes. The University of New Mexico's Institute for Space Nuclear Power Studies, under NASA sponsorship has recently initiated a study to model the complicated hydrogen corrosion process. In support of this effort, an extensive review of the open literature was performed, and a technical expert workshop was conducted. This paper summarizes the results of this review.

A Review of Carbide Fuel Corrosion for Nuclear Thermal Propulsion Applications

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; El-Genk, Mohamed S.; Butt, Darryl P.

1994-07-01

At the operation conditions of interest in nuclear thermal propulsion reactors, carbide materials have been known to exhibit a number of life limiting phenomena. These include the formation of liquid, loss by vaporization, creep and corresponding gas flow restrictions, and local corrosion and fuel structure degradation due to excessive mechanical and/or thermal loading. In addition, the radiation environment in the reactor core can produce a substantial change in its local physical properties, which can produce high thermal stresses and corresponding stress fractures (cracking). Time-temperature history and cyclic operation of the nuclear reactor can also accelerate some of these processes. The University of New Mexico's Institute for Space Nuclear Power Studies, under NASA sponsorship has recently initiated a study to model the complicated hydrogen corrosion process. In support of this effort, an extensive review of the open literature was performed, and a technical expert workshop was conducted. This paper summarizes the results of this review.

Army gas-cooled reactor systems program. Preliminary design report off-normal scram system

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

Bushnell, W.H.; Malmstrom, S.A.

1965-06-01

The maximum allowable ML-1 fuel element cladding (hot spot) temperature is established by ANTS 201 at 1750/sup 0/F. The existing ML-1 design makes no provision for automatic scram when this limit is reached. Operating experience has indicated a requirement for such an automatic system during plant startup and a revised hot spot envelope (generated during conceptual design of the scram system) established the desirability of extending this protection to operation at full power conditions. It was also determined that the scram system should include circuitry to initiate an automatic scram if reactor ..delta..T exceeded 450/sup 0/F (the limit established inmore » ANTS 201) and if reactor power exceeded 6 kw(t) without coolant flow in the main loop. The preliminary design of the scram system (designated off-normal scram system) which will provide the required protection is described.« less

Design modification for the modular helium reactor for higher temperature operation and reliability studies for nuclear hydrogen production processes

NASA Astrophysics Data System (ADS)

Reza, S. M. Mohsin

Design options have been evaluated for the Modular Helium Reactor (MHR) for higher temperature operation. An alternative configuration for the MHR coolant inlet flow path is developed to reduce the peak vessel temperature (PVT). The coolant inlet path is shifted from the annular path between reactor core barrel and vessel wall through the permanent side reflector (PSR). The number and dimensions of coolant holes are varied to optimize the pressure drop, the inlet velocity, and the percentage of graphite removed from the PSR to create this inlet path. With the removal of ˜10% of the graphite from PSR the PVT is reduced from 541°C to 421°C. A new design for the graphite block core has been evaluated and optimized to reduce the inlet coolant temperature with the aim of further reduction of PVT. The dimensions and number of fuel rods and coolant holes, and the triangular pitch have been changed and optimized. Different packing fractions for the new core design have been used to conserve the number of fuel particles. Thermal properties for the fuel elements are calculated and incorporated into these analyses. The inlet temperature, mass flow and bypass flow are optimized to limit the peak fuel temperature (PFT) within an acceptable range. Using both of these modifications together, the PVT is reduced to ˜350°C while keeping the outlet temperature at 950°C and maintaining the PFT within acceptable limits. The vessel and fuel temperatures during low pressure conduction cooldown and high pressure conduction cooldown transients are found to be well below the design limits. The reliability and availability studies for coupled nuclear hydrogen production processes based on the sulfur iodine thermochemical process and high temperature electrolysis process have been accomplished. The fault tree models for both these processes are developed. Using information obtained on system configuration, component failure probability, component repair time and system operating modes and conditions, the system reliability and availability are assessed. Required redundancies are made to improve system reliability and to optimize the plant design for economic performance. The failure rates and outage factors of both processes are found to be well below the maximum acceptable range.

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.

Strategic need for a multi-purpose thermal hydraulic loop for support of advanced reactor technologies

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

O'Brien, James E.; Sabharwall, Piyush; Yoon, Su -Jong

2014-09-01

This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs)more » at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation needs. The experimental database will guide development of appropriate predictive methods and be available for code verification and validation (V&V) related to these systems.« less

Rapid formation and pollutant removal ability of aerobic granules in a sequencing batch airlift reactor at low temperature.

PubMed

Jiang, Yu; Shang, Yu; Wang, Hongyu; Yang, Kai

2016-12-01

The start-up of an aerobic granular sludge (AGS) reactor at low temperature was more difficult than at ambient temperature.The rapid formation and characteristics of AGS in a sequencing batch airlift reactor at low temperature were investigated. The nutrient removal ability of the system was also evaluated. It was found that compact granules with clear boundary were formed within 10 days and steady state was achieved within 25 days. The settling properties of sludge were improved with the increasing secretion of extracellular polymeric substances and removal performances of pollutants were enhanced along with granulation. The average removal efficiencies of COD, NH4(+)-N, total nitrogen (TN), total phosphorus (TP) after aerobic granules maturing were over 90.9%, 94.7%, 75.4%, 80.2%, respectively. The rise of temperature had little impact on pollutant biodegradation while the variation of dissolved oxygen caused obvious changes in TN and TP removal rates. COD concentrations of effluents were below 30 mg l(-1) in most cycles of operation with a wide range of organic loading rates (0.6-3.0 kg COD m(-3) d(-1)). The rapid granulation and good performance of pollutant reduction by the system might provide an alternate for wastewater treatment in cold regions.

Direct Estimation of Power Distribution in Reactors for Nuclear Thermal Space Propulsion

NASA Astrophysics Data System (ADS)

Aldemir, Tunc; Miller, Don W.; Burghelea, Andrei

2004-02-01

A recently proposed constant temperature power sensor (CTPS) has the capability to directly measure the local power deposition rate in nuclear reactor cores proposed for space thermal propulsion. Such a capability reduces the uncertainties in the estimated power peaking factors and hence increases the reliability of the nuclear engine. The CTPS operation is sensitive to the changes in the local thermal conditions. A procedure is described for the automatic on-line calibration of the sensor through estimation of changes in thermal .conditions.

Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

NASA Astrophysics Data System (ADS)

Meriyanti, Su'ud, Zaki; Rijal, K.; Zuhair, Ferhat, A.; Sekimoto, H.

2010-06-01

In this study a fesibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850° C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticallity was obtained for this reactor.

A diesel fuel processor for fuel-cell-based auxiliary power unit applications

NASA Astrophysics Data System (ADS)

Samsun, Remzi Can; Krekel, Daniel; Pasel, Joachim; Prawitz, Matthias; Peters, Ralf; Stolten, Detlef

2017-07-01

Producing a hydrogen-rich gas from diesel fuel enables the efficient generation of electricity in a fuel-cell-based auxiliary power unit. In recent years, significant progress has been achieved in diesel reforming. One issue encountered is the stable operation of water-gas shift reactors with real reformates. A new fuel processor is developed using a commercial shift catalyst. The system is operated using optimized start-up and shut-down strategies. Experiments with diesel and kerosene fuels show slight performance drops in the shift reactor during continuous operation for 100 h. CO concentrations much lower than the target value are achieved during system operation in auxiliary power unit mode at partial loads of up to 60%. The regeneration leads to full recovery of the shift activity. Finally, a new operation strategy is developed whereby the gas hourly space velocity of the shift stages is re-designed. This strategy is validated using different diesel and kerosene fuels, showing a maximum CO concentration of 1.5% at the fuel processor outlet under extreme conditions, which can be tolerated by a high-temperature PEFC. The proposed operation strategy solves the issue of strong performance drop in the shift reactor and makes this technology available for reducing emissions in the transportation sector.

COMPARTMENTED REACTOR FUEL ELEMENT

DOEpatents

Cain, F.M. Jr.

1962-09-11

A method of making a nuclear reactor fuel element of the elongated red type is given wherein the fissionable fuel material is enclosed within a tubular metal cladding. The method comprises coating the metal cladding tube on its inside wall with a brazing alloy, inserting groups of cylindrical pellets of fissionable fuel material into the tube with spacing members between adjacent groups of pellets, sealing the ends of the tubes to leave a void space therewithin, heating the tube and its contents to an elevated temperature to melt the brazing alloy and to expand the pellets to their maximum dimensions under predetermined operating conditions thereby automatically positioning the spacing members along the tube, and finally cooling the tube to room temperature whereby the spacing disks become permanently fixed at their edges in the brazing alloy and define a hermetically sealed compartment for each fl group of fuel pellets. Upon cooling, the pellets contract thus leaving a space to accommodate thermal expansion of the pellets when in use in a reactor. The spacing members also provide lateral support for the tubular cladding to prevent collapse thereof when subjected to a reactor environment. (AEC)

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.

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

Anaerobic on-site treatment of black water and dairy parlour wastewater in UASB-septic tanks at low temperatures.

PubMed

Luostarinen, Sari A; Rintala, Jukka A

2005-01-01

Anaerobic on-site treatment of synthetic black water (BW) and dairy parlour wastewater (DPWW) was studied in two-phased upflow anaerobic sludge blanket (UASB)-septic tanks at low temperatures (10-20 degrees C). At all temperatures, total chemical oxygen demand (COD(t)) removal was above 90% with BW and above 80% with DPWW and removal of total suspended solids (TSS) above 90% with both wastewaters. Moreover, dissolved COD (COD(dis)) removal was approx. 70% with both wastewaters indicating good biological activity of the sludges. With BW, a single-phased reactor was found sufficient for good COD removals, while with DPWW, a two-phased process was required. Temperature optimum of reactor sludges was still 35 degrees C after long (398d) operation. Most of the nutrients from BW were removed with TSS, while with DPWW nutrient removal was low. In conclusion, UASB-septic tank was found feasible for (pre)treatment of BW and DPWW at low temperatures.

Development of next generation tempered and ODS reduced activation ferritic/martensitic steels for fusion energy applications

DOE PAGES

Zinkle, S. J.; Boutard, J. L.; Hoelzer, D. T.; ...

2017-06-09

Reduced activation ferritic/martensitic steels are currently the most technologically mature option for the structural material of proposed fusion energy reactors. Advanced next-generation higher performance steels offer the opportunity for improvements in fusion reactor operational lifetime and reliability, superior neutron radiation damage resistance, higher thermodynamic efficiency, and reduced construction costs. The two main strategies for developing improved steels for fusion energy applications are based on (1) an evolutionary pathway using computational thermodynamics modelling and modified thermomechanical treatments (TMT) to produce higher performance reduced activation ferritic/martensitic (RAFM) steels and (2) a higher risk, potentially higher payoff approach based on powder metallurgy techniquesmore » to produce very high strength oxide dispersion strengthened (ODS) steels capable of operation to very high temperatures and with potentially very high resistance to fusion neutron-induced property degradation. The current development status of these next-generation high performance steels is summarized, and research and development challenges for the successful development of these materials are outlined. In conclusion, material properties including temperature-dependent uniaxial yield strengths, tensile elongations, high-temperature thermal creep, Charpy impact ductile to brittle transient temperature (DBTT) and fracture toughness behaviour, and neutron irradiation-induced low-temperature hardening and embrittlement and intermediate-temperature volumetric void swelling (including effects associated with fusion-relevant helium and hydrogen generation) are described for research heats of the new steels.« less

Development of next generation tempered and ODS reduced activation ferritic/martensitic steels for fusion energy applications

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

Zinkle, S. J.; Boutard, J. L.; Hoelzer, D. T.

Reduced activation ferritic/martensitic steels are currently the most technologically mature option for the structural material of proposed fusion energy reactors. Advanced next-generation higher performance steels offer the opportunity for improvements in fusion reactor operational lifetime and reliability, superior neutron radiation damage resistance, higher thermodynamic efficiency, and reduced construction costs. The two main strategies for developing improved steels for fusion energy applications are based on (1) an evolutionary pathway using computational thermodynamics modelling and modified thermomechanical treatments (TMT) to produce higher performance reduced activation ferritic/martensitic (RAFM) steels and (2) a higher risk, potentially higher payoff approach based on powder metallurgy techniquesmore » to produce very high strength oxide dispersion strengthened (ODS) steels capable of operation to very high temperatures and with potentially very high resistance to fusion neutron-induced property degradation. The current development status of these next-generation high performance steels is summarized, and research and development challenges for the successful development of these materials are outlined. In conclusion, material properties including temperature-dependent uniaxial yield strengths, tensile elongations, high-temperature thermal creep, Charpy impact ductile to brittle transient temperature (DBTT) and fracture toughness behaviour, and neutron irradiation-induced low-temperature hardening and embrittlement and intermediate-temperature volumetric void swelling (including effects associated with fusion-relevant helium and hydrogen generation) are described for research heats of the new steels.« less

Development of next generation tempered and ODS reduced activation ferritic/martensitic steels for fusion energy applications

NASA Astrophysics Data System (ADS)

Zinkle, S. J.; Boutard, J. L.; Hoelzer, D. T.; Kimura, A.; Lindau, R.; Odette, G. R.; Rieth, M.; Tan, L.; Tanigawa, H.

2017-09-01

Reduced activation ferritic/martensitic steels are currently the most technologically mature option for the structural material of proposed fusion energy reactors. Advanced next-generation higher performance steels offer the opportunity for improvements in fusion reactor operational lifetime and reliability, superior neutron radiation damage resistance, higher thermodynamic efficiency, and reduced construction costs. The two main strategies for developing improved steels for fusion energy applications are based on (1) an evolutionary pathway using computational thermodynamics modelling and modified thermomechanical treatments (TMT) to produce higher performance reduced activation ferritic/martensitic (RAFM) steels and (2) a higher risk, potentially higher payoff approach based on powder metallurgy techniques to produce very high strength oxide dispersion strengthened (ODS) steels capable of operation to very high temperatures and with potentially very high resistance to fusion neutron-induced property degradation. The current development status of these next-generation high performance steels is summarized, and research and development challenges for the successful development of these materials are outlined. Material properties including temperature-dependent uniaxial yield strengths, tensile elongations, high-temperature thermal creep, Charpy impact ductile to brittle transient temperature (DBTT) and fracture toughness behaviour, and neutron irradiation-induced low-temperature hardening and embrittlement and intermediate-temperature volumetric void swelling (including effects associated with fusion-relevant helium and hydrogen generation) are described for research heats of the new steels.

Aerobic granules formation and nutrients removal characteristics in sequencing batch airlift reactor (SBAR) at low temperature.

PubMed

Bao, Ruiling; Yu, Shuili; Shi, Wenxin; Zhang, Xuedong; Wang, Yulan

2009-09-15

To understand the effect of low temperature on the formation of aerobic granules and their nutrient removal characteristics, an aerobic granular sequencing batch airlift reactor (SBAR) has been operated at 10 degrees C using a mixed carbon source of glucose and sodium acetate. The results showed that aerobic granules were obtained and that the reactor performed in stable manner under the applied conditions. The granules had a compact structure and a clear out-surface. The average parameters of the granules were: diameter 3.4mm, wet density 1.036 g mL(-1), sludge volume index 37 mL g(-1), and settling velocity 18.6-65.1 cm min(-1). Nitrite accumulation was observed, with a nitrite accumulation rate (NO(2)(-)-N/NO(x)(-)-N) between 35% and 43% at the beginning of the start-up stage. During the stable stage, NO(x) was present at a level below the detection limit. However, when the influent COD concentration was halved (resulting in COD/N a reduction of the COD/N from 20:1 to 10:1) nitrite accumulation was observed once more with an effluent nitrite accumulation rate of 94.8%. Phosphorus release was observed in the static feeding phase and also during the initial 20-30 min of the aerobic phase. Neither the low temperature nor adjustment of the COD/P ratio from 100:1 to 25:1 had any influence on the phosphorus removal efficiency under the operating conditions. In the granular reactor with the influent load rates for COD, NH(4)(+)-N, and PO(4)(3-)-P of 1.2-2.4, 0.112 and 0.012-0.024 kg m(-3)d(-1), the respective removal efficiencies at low temperature were 90.6-95.4%, 72.8-82.1% and 95.8-97.9%.

Enrichment of acetogenic bacteria in high rate anaerobic reactors under mesophilic and thermophilic conditions.

PubMed

Ryan, P; Forbes, C; McHugh, S; O'Reilly, C; Fleming, G T A; Colleran, E

2010-07-01

The objective of the current study was to expand the knowledge of the role of acetogenic Bacteria in high rate anaerobic digesters. To this end, acetogens were enriched by supplying a variety of acetogenic growth supportive substrates to two laboratory scale high rate upflow anaerobic sludge bed (UASB) reactors operated at 37 degrees C (R1) and 55 degrees C (R2). The reactors were initially fed a glucose/acetate influent. Having achieved high operational performance and granular sludge development and activity, both reactors were changed to homoacetogenic bacterial substrates on day 373 of the trial. The reactors were initially fed with sodium vanillate as a sole substrate. Although % COD removal indicated that the 55 degrees C reactor out performed the 37 degrees C reactor, effluent acetate levels from R2 were generally higher than from R1, reaching values as high as 5023 mg l(-1). Homoacetogenic activity in both reactors was confirmed on day 419 by specific acetogenic activity (SAA) measurement, with higher values obtained for R2 than R1. Sodium formate was introduced as sole substrate to both reactors on day 464. It was found that formate supported acetogenic activity at both temperatures. By the end of the trial, no specific methanogenic activity (SMA) was observed against acetate and propionate indicating that the methane produced was solely by hydrogenotrophic Archaea. Higher SMA and SAA values against H(2)/CO(2) suggested development of a formate utilising acetogenic population growing in syntrophy with hydrogenotrophic methanogens. Throughout the formate trial, the mesophilic reactor performed better overall than the thermophilic reactor. Copyright 2010 Elsevier Ltd. All rights reserved.

Start-up of a sequential dry anaerobic digestion of paunch under psychrophilic and mesophilic temperatures.

PubMed

Nkemka, Valentine Nkongndem; Hao, Xiying

2018-04-01

The present laboratory study evaluated the sequential leach bed dry anaerobic digestion (DAD) of paunch under psychrophilic (22°C) and mesophilic (40°C) temperatures. Three leach bed reactors were operated under the mesophilic temperature in sequence at a solid retention time (SRT) of 40d with a new batch started 27d into the run of the previous one. A total of six batches were operated for 135d. The results showed that the mesophilic DAD of paunch was efficient, reaching methane yields of 126.9-212.1mLg -1 volatile solid (VS) and a VS reduction of 32.9-55.5%. The average daily methane production rate increased from 334.3mLd -1 to 571.4mLd -1 and 825.7mLd -1 when one, two and three leach bed reactors were in operation, respectively. The psychrophilic DAD of paunch was operated under a SRT of 100d and a total of three batches were performed in sequence for 300d with each batch starting after completion of the previous one. Improvements in the methane yield from 93.9 to 107.3 and 148.3mLg -1 VS and VS reductions of 24.8, 30.2 and 38.6% were obtained in the consecutive runs, indicating the adaptation of anaerobic microbes from mesophilic to psychrophilic temperatures. In addition, it took three runs for anaerobic microbes to reduce the volatile fatty acid accumulation observed in the first and second trials. This study demonstrates the potential of renewable energy recovery from paunch under psychrophilic and mesophilic temperatures. Crown Copyright © 2016. Published by Elsevier Ltd. All rights reserved.

Investigation on the influence of electrode geometry on characteristics of coaxial dielectric barrier discharge reactor driven by an oscillating microsecond pulsed power supply

NASA Astrophysics Data System (ADS)

Miao, Chuanrun; Liu, Feng; Wang, Qian; Cai, Meiling; Fang, Zhi

2018-03-01

In this paper, an oscillating microsecond pulsed power supply with rise time of several tens of nanosecond (ns) is used to excite a coaxial DBD with double layer dielectric barriers. The effects of various electrode geometries by changing the size of inner quartz tube (different electrode gaps) on the discharge uniformity, power deposition, energy efficiency, and operation temperature are investigated by electrical, optical, and temperature diagnostics. The electrical parameters of the coaxial DBD are obtained from the measured applied voltage and current using an equivalent electrical model. The energy efficiency and the power deposition in air gap of coaxial DBD with various electrode geometries are also obtained with the obtained electrical parameters, and the heat loss and operation temperature are analyzed by a heat conduction model. It is found that at the same applied voltage, with the increasing of the air gap, the discharge uniformity becomes worse and the discharge power deposition and the energy efficiency decrease. At 2.5 mm air gap and 24 kV applied voltage, the energy efficiency of the coaxial DBD reaches the maximum value of 68.4%, and the power deposition in air gap is 23.6 W and the discharge uniformity is the best at this case. The corresponding operation temperature of the coaxial DBD reaches 64.3 °C after 900 s operation and the temperature of the inner dielectric barrier is 114.4 °C under thermal balance. The experimental results provide important experimental references and are important to optimize the design and the performance of coaxial DBD reactor.

Performance evaluation of a completely stirred anaerobic reactor treating pig manure at a low range of mesophilic conditions

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

Guo, Jianbin, E-mail: jianbinguo@gmail.com; Dong, Renjie; Clemens, Joachim

Highlights: • The biogas process can run stably at 20 °C at extremely low OLR after long-term acclimation of bacteria. • A biogas plant running at 28 °C seems as efficient as that operated at 38 °C at low OLR of 1.3 g ODM L{sup −1} d{sup −1}. • Lower temperature operation is inadvisable for the commercial biogas plant running at rather high OLR. • The estimated sludge yield at 28 °C is higher than that at 38 °C. - Abstract: Many Chinese biogas plants run in the lower range of mesophilic conditions. This study evaluated the performance of amore » completely stirred anaerobic reactor treating pig manure at different temperatures (20, 28 and 38 °C). The start-up phase of the reactor at 20 °C was very long and extremely poor performance was observed with increasing organic loading rate (OLR). At an OLR of 4.3 g ODM L{sup −1} d{sup −1}, methane production at 28 °C was comparable (3% less) with that at 38 °C, but the risk of acidification was high at 28 °C. At low OLR (1.3 g ODM L{sup −1} d{sup −1}), the biogas process appeared stable at 28 °C and gave same methane yields as compared to the reactor operating at 38 °C. The estimated sludge yield at 28 °C was 0.065 g VSS g{sup −1} COD{sub removed,} which was higher than that at 38 °C (0.016 g VSS g{sup −1} COD{sub removed})« less

Complete Non-Radioactive Operability Tests for Cladding Hull Chlorination

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

Collins, Emory D; Johnson, Jared A.; Hylton, Tom D.

2016-04-01

Non-radioactive operability tests were made to test the metal chlorination reactor and condenser and their accessories using batch chlorinations of non-radioactive cladding samples and to identify optimum operating practices and components that need further modifications prior to installation of the equipment into the hot cell for tests on actual used nuclear fuel (UNF) cladding. The operability tests included (1) modifications to provide the desired heating and reactor temperature profile; and (2) three batch chlorination tests using, respectively, 100, 250, and 500 g of cladding. During the batch chlorinations, metal corrosion of the equipment was assessed, pressurization of the gas inletmore » was examined and the best method for maintaining solid salt product transfer through the condenser was determined. Also, additional accessing equipment for collection of residual ash and positioning of the unit within the hot cell were identified, designed, and are being fabricated.« less

Low cost silicon solar array project silicon materials task: Establishment of the feasibility of a process capable of low-cost, high volume production of silane (step 1) and the pyrolysis of silane to semiconductor-grade silicon (step 2)

NASA Technical Reports Server (NTRS)

Breneman, W. C.; Farrier, E. G.; Rexer, J.

1977-01-01

Extended operation of a small process-development unit routinely produced high quality silane in 97+% yield from dichlorosilane. The production rate was consistent with design loadings for the fractionating column and for the redistribution reactor. A glass fluid-bed reactor was constructed for room temperature operation. The behavior of a bed of silcon particles was observed as a function of various feedstocks, component configurations, and operating conditions. For operating modes other than spouting, the bed behaved in an erratic and unstable manner. A method was developed for casting molten silicon powder into crack-free solid pellets for process evaluation. The silicon powder was melted and cast into thin walled quartz tubes that sacrificially broke on cooling.

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

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

Joule-Heated Molten Regolith Electrolysis Reactor Concepts for Oxygen and Metals Production on the Moon and Mars

NASA Technical Reports Server (NTRS)

Sibille, Laurent; Dominguez, Jesus A.

2012-01-01

The technology of direct electrolysis of molten lunar regolith to produce oxygen and molten metal alloys has progressed greatly in the last few years. The development of long-lasting inert anodes and cathode designs as well as techniques for the removal of molten products from the reactor has been demonstrated. The containment of chemically aggressive oxide and metal melts is very difficult at the operating temperatures ca. 1600 C. Containing the molten oxides in a regolith shell can solve this technical issue and can be achieved by designing a Joule-heated (sometimes called 'self-heating') reactor in which the electrolytic currents generate enough Joule heat to create a molten bath. Solutions obtained by multiphysics modeling allow the identification of the critical dimensions of concept reactors.

Combustion engine system

NASA Technical Reports Server (NTRS)

Houseman, John (Inventor); Voecks, Gerald E. (Inventor)

1986-01-01

A flow through catalytic reactor which selectively catalytically decomposes methanol into a soot free hydrogen rich product gas utilizing engine exhaust at temperatures of 200 to 650 C to provide the heat for vaporizing and decomposing the methanol is described. The reactor is combined with either a spark ignited or compression ignited internal combustion engine or a gas turbine to provide a combustion engine system. The system may be fueled entirely by the hydrogen rich gas produced in the methanol decomposition reactor or the system may be operated on mixed fuels for transient power gain and for cold start of the engine system. The reactor includes a decomposition zone formed by a plurality of elongated cylinders which contain a body of vapor permeable, methanol decomposition catalyst preferably a shift catalyst such as copper-zinc.

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.

Experimental and Computational Study of the Hydrodynamics of Trickle Bed Flow Reactor Operating Under Different Pressure Conditions

NASA Astrophysics Data System (ADS)

Rabbani, S.; Ben Salem, I.; Nadeem, H.; Kurnia, J. C.; Shamim, T.; Sassi, M.

2014-12-01

Pressure drop estimation and prediction of liquid holdup play a crucial role in design and operation of trickle bed reactors. Experiments are performed for Light Gas Oil (LGO)-nitrogen system in ambient temperature conditions in an industrial pilot plant with reactor height 0.79 m and diameter of 0.0183 m and pressure ranging from atmospheric to 10 bars. It was found that pressure drop increased with increase in system pressure, superficial gas velocity and superficial liquid velocity. It was demonstrated in the experiments that liquid holdup of the system increases with the increase in superficial liquid velocity and tends to decrease with increase in superficial gas velocity which is in good agreement with existing literature. Similar conditions were also simulated using CFD-software FLUENT. The Volume of Fluid (VoF) technique was employed in combination with "discrete particle approach" and results were compared with that of experiments. The overall pressure drop results were compared with the different available models and a new comprehensive model was proposed to predict the pressure drop in Trickle Bed Flow Reactor.

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

Flat-plate solar array project. Task 1: Silicon material: Investigation of the hydrochlorination of SiC1sub4

NASA Technical Reports Server (NTRS)

Mui, J. Y. P.

1981-01-01

A two inch-diameter stainless steel reactor was designed to operate at pressure up to 500 psig and at temperature up to 600 C in order to study the hydrochlorination of silicon tetrachloride and metallurgical grade (m.g.) silicon metal to trichlorosilane. The hydrochlorination apparatus is described and operation safety and pollution control are discussed.

Closed Brayton cycle power conversion systems for nuclear reactors :

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

Wright, Steven A.; Lipinski, Ronald J.; Vernon, Milton E.

2006-04-01

This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors,more » reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at the manufacturers site (Barber-Nichols Inc.) and installed and operated at Sandia. A sufficiently detailed description of the loop is provided in this report along with the design characteristics of the turbo-alternator-compressor set to allow other researchers to compare their results with those measured in the Sandia test-loop. The third task consisted of a validation effort. In this task the test loop was operated and compared with the modeled results to develop a more complete understanding of this electrically heated closed power generation system and to validate the model. The measured and predicted system temperatures and pressures are in good agreement, indicating that the model is a reasonable representation of the test loop. Typical deviations between the model and the hardware results are less than 10%. Additional tests were performed to assess the capability of the Brayton engine to continue to remove decay heat after the reactor/heater is shutdown, to develop safe and effective control strategies, and to access the effectiveness of gas inventory control as an alternative means to provide load following. In one test the heater power was turned off to simulate a rapid reactor shutdown, and the turbomachinery was driven solely by the sensible heat stored in the heater for over 71 minutes without external power input. This is an important safety feature for CBC systems as it means that the closed Brayton loop will keep cooling the reactor without the need for auxiliary power (other than that needed to circulate the waste heat rejection coolant) provided the heat sink is available.« less

Advanced Reactors-Intermediate Heat Exchanger (IHX) Coupling: Theoretical Modeling and Experimental Validation

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

Utgikar, Vivek; Sun, Xiaodong; Christensen, Richard

2016-12-29

The overall goal of the research project was to model the behavior of the advanced reactorintermediate heat exchange system and to develop advanced control techniques for off-normal conditions. The specific objectives defined for the project were: 1. To develop the steady-state thermal hydraulic design of the intermediate heat exchanger (IHX); 2. To develop mathematical models to describe the advanced nuclear reactor-IHX-chemical process/power generation coupling during normal and off-normal operations, and to simulate models using multiphysics software; 3. To develop control strategies using genetic algorithm or neural network techniques and couple these techniques with the multiphysics software; 4. To validate themore » models experimentally The project objectives were accomplished by defining and executing four different tasks corresponding to these specific objectives. The first task involved selection of IHX candidates and developing steady state designs for those. The second task involved modeling of the transient and offnormal operation of the reactor-IHX system. The subsequent task dealt with the development of control strategies and involved algorithm development and simulation. The last task involved experimental validation of the thermal hydraulic performances of the two prototype heat exchangers designed and fabricated for the project at steady state and transient conditions to simulate the coupling of the reactor- IHX-process plant system. The experimental work utilized the two test facilities at The Ohio State University (OSU) including one existing High-Temperature Helium Test Facility (HTHF) and the newly developed high-temperature molten salt facility.« less

Pre-Conceptual Design for Northstar ⁹⁹Mo Process Tritium Removal System

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

Nobile, Arthur; Reichert, Heidi; Hollis, William Kirk

2016-01-12

In this report we describe a preliminary concept for a Tritium Removal System (TRS) to remove tritium that is generated in the ⁹⁹Mo production process. Preliminary calculations have been performed to evaluate an approximate size for the system. The concept described utilizes well-established detritiation technology based on catalytic oxidation of tritium and tritiated hydrocarbons to water in a high temperature (400 °C) reactor and capture of water in a molecular sieve bed. The TRS concept involves use of a single system that would cycle through each of the seven online target systems and remove tritium that has been accumulated aftermore » one week’s run time. The TRS would perform cleanup operations on each target system for a period of approximately 24 hours. This would occur while the system is still online and just prior to target replacement, so tritium levels would at their minimum values for target replacement. In the concept, during normal operation a small fraction (1%) of the helium recirculating in the system would be diverted through the TRS and returned to the flow loop. With this approach sufficient levels of detritiation can be accomplished in a 24 hour period. In the study it was found that because of the need to maintain low oxygen levels in the system (<100 ppm) this increases the size of the catalytic reactor. As a result of this finding, consideration should be given to other methods for removing tritium from the system. Other methods such as catalytic exchange of tritium with an unsaturated organic compound and subsequent trapping on activated carbon or molecular sieve could offer advantages of reducing reactor size and operation at lower reactor temperature. However the most significant advantage of such an approach would be the ability to operate in very low oxygen environments, which would eliminate any concerns for oxidation of the target.« less

A laser scanning system for metrology and viewing in ITER

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

Spampinato, P.T.; Barry, R.E.; Menon, M.M.

1996-05-01

The construction and operation of a next-generation fusion reactor will require metrology to achieve and verify precise alignment of plasma-facing components and inspection in the reactor vessel. The system must be compatible with the vessel environment of high gamma radiation (10{sup 4} Gy/h), ultra-high-vacuum (10{sup {minus}8} torr), and elevated temperature (200 C). The high radiation requires that the system be remotely deployed. A coherent frequency modulated laser radar-based system will be integrated with a remotely operated deployment mechanism to meet these requirements. The metrology/viewing system consists of a compact laser transceiver optics module which is linked through fiber optics tomore » the laser source and imaging units that are located outside of a biological shield. The deployment mechanism will be a mast-like positioning system. Radiation-damage tests will be conducted on critical sensor components at Oak Ridge National Laboratory to determine threshold damage levels and effects on data transmission. This paper identifies the requirements for International Thermonuclear Experimental Reactor metrology and viewing and describes a remotely operated precision ranging and surface mapping system.« less

Fermentative hydrogen production from soybean protein processing wastewater in an anaerobic baffled reactor (ABR) using anaerobic mixed consortia.

PubMed

Zhu, Ge-fu; Li, Jian-zheng; Liu, Chao-xiang

2012-09-01

Fermentative H(2) production from soybean protein processing wastewater (SPPW) was investigated in a four-compartment anaerobic baffled reactor (ABR) using anaerobic mixed cultures under continuous flow condition in the present study. After being inoculated with aerobic activated sludge and operated at the inoculants of 5.98 gVSS L(-1), COD of 5000 mg L(-1), HRT of 16 h and temperature of (35 ± 1) °C for 22 days, the ABR achieved stable ethanol-type fermentation. The specific hydrogen production rate of anaerobic activated sludge was 165 LH(2)kg MLVSS(-1) day(-1), the substrate conversion rate was 600.83 LH(2)kg COD(-1)and the COD removal efficiency was 44.73% at the stable operation status. The ABR system exhibited a better stability and higher hydrogen yields than continuous stirring tank reactor under the same operational condition. The experimental data documented the feasibility of substrate degradation along with molecular H(2) generation utilizing SPPW as primary carbon source in the ABR system.

Assessment of segregation kinetics in water-moderated reactors pressure vessel steels under long-term operation

NASA Astrophysics Data System (ADS)

Kuleshova, E. A.; Gurovich, B. A.; Lavrukhina, Z. V.; Saltykov, M. A.; Fedotova, S. V.; Khodan, A. N.

2016-08-01

In reactor pressure vessel (RPV) bcc-lattice steels temper embrittlement is developed under the influence of both operating temperature of ∼300 °C and neutron irradiation. Segregation processes in the grain boundaries (GB) begin to play a special role in the assessment of the safe operation of the RPV in case of its lifetime extension up to 60 years or more. The most reliable information on the RPV material condition can be obtained by investigating the surveillance specimens (SS) that are exposed to operational factors simultaneously with the RPV itself. In this paper the GB composition in the specimens with different thermal exposure time at the RPV operating temperature as well as irradiated by fast neutrons (E ≥ 0.5 MeV) to different fluences (20-71)·1022 m-2 was studied by means of Auger electron spectroscopy (AES) including both impurity and main alloying elements content. The data obtained allowed to trace the trend of the operating temperature and radiation-stimulated diffusion influence on the overall segregants level in GB. The revealed differences in the concentration levels of GB segregants in different steels, are due to the different chemical composition of the steels and also due to different grain boundary segregation levels in initial (unexposed) state. The data were used to estimate the RPV steels working capacity for 60 years. The estimation was carried out using both the well-known Langmuir-McLean model and the one specially developed for RPV steels, which takes into account the structure and phase composition of VVER-1000 RPV steels, as well as the long-term influence of operational factors.

Interim Safe Storage of Plutonium Production Reactors at the US DOE Hanford Site - 13438

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

Schilperoort, Daryl L.; Faulk, Darrin

2013-07-01

Nine plutonium production reactors located on DOE's Hanford Site are being placed into an Interim Safe Storage (ISS) period that extends to 2068. The Environmental Impact Statement (EIS) for ISS [1] was completed in 1993 and proposed a 75-year storage period that began when the EIS was finalized. Remote electronic monitoring of the temperature and water level alarms inside the safe storage enclosure (SSE) with visual inspection inside the SSE every 5 years are the only planned operational activities during this ISS period. At the end of the ISS period, the reactor cores will be removed intact and buried inmore » a landfill on the Hanford Site. The ISS period allows for radioactive decay of isotopes, primarily Co-60 and Cs-137, to reduce the dose exposure during disposal of the reactor cores. Six of the nine reactors have been placed into ISS by having an SSE constructed around the reactor core. (authors)« 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

Membrane reactor for water detritiation: a parametric study on operating parameters

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

Mascarade, J.; Liger, K.; Troulay, M.

2015-03-15

This paper presents the results of a parametric study done on a single stage finger-type packed-bed membrane reactor (PBMR) used for heavy water vapor de-deuteration. Parametric studies have been done on 3 operating parameters which are: the membrane temperature, the total feed flow rate and the feed composition through D{sub 2}O content variations. Thanks to mass spectrometer analysis of streams leaving the PBMR, speciation of deuterated species was achieved. Measurement of the amounts of each molecular component allowed the calculation of reaction quotient at the packed-bed outlet. While temperature variation mainly influences permeation efficiency, feed flow rate perturbation reveals dependencemore » of conversion and permeation properties to contact time between catalyst and reacting mixture. The study shows that isotopic exchange reactions occurring on the catalyst particles surface are not thermodynamically balanced. Moreover, the variation of the heavy water content in the feed exhibits competition between permeation and conversion kinetics.« less

Development and Analysis of Cold Trap for Use in Fission Surface Power-Primary Test Circuit

NASA Technical Reports Server (NTRS)

Wolfe, T. M.; Dervan, C. A.; Pearson, J. B.; Godfroy, T. J.

2012-01-01

The design and analysis of a cold trap proposed for use in the purification of circulated eutectic sodium potassium (NaK-78) loops is presented. The cold trap is designed to be incorporated into the Fission Surface Power-Primary Test Circuit (FSP-PTC), which incorporates a pumped NaK loop to simulate in-space nuclear reactor-based technology using non-nuclear test methodology as developed by the Early Flight Fission-Test Facility. The FSP-PTC provides a test circuit for the development of fission surface power technology. This system operates at temperatures that would be similar to those found in a reactor (500-800 K). By dropping the operating temperature of a specified percentage of NaK flow through a bypass containing a forced circulation cold trap, the NaK purity level can be increased by precipitating oxides from the NaK and capturing them within the cold trap. This would prevent recirculation of these oxides back through the system, which may help prevent corrosion.

The preliminary design of bearings for the control system of a high-temperature lithium-cooled nuclear reactor

NASA Technical Reports Server (NTRS)

Yacobucci, H. G.; Waldron, W. D.; Walowit, J. A.

1973-01-01

The design of bearings for the control system of a fast reactor concept is presented. The bearings are required to operate at temperatures up to 2200 F in one of two fluids, lithium or argon. Basic bearing types are the same regardless of the fluid. Crowned cylindrical journals were selected for radially loaded bearings and modified spherical bearings were selected for bearings under combined thrust and radial loads. Graphite and aluminum oxide are the materials selected for the argon atmosphere bearings while cermet compositions (carbides or nitrides bonded with refractory metals) were selected for the lithium lubricated bearings. Mounting of components is by shrink fit or by axial clamping utilizing differential thermal expansion.

STUDIES OF FAST REACTOR FUEL ELEMENT BEHAVIOR UNDER TRANSIENT HEATING TO FAILURE. I. INITIAL EXPERIMENTS ON METALLIC SAMPLES IN THE ABSENCE OF COOLANT

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

Dickerman, C. E.; Sowa, E. S.; Okrent, D.

1961-08-01

Meltdown tests on single metallic unirradiated fuel elements in TREAT are described. The fuel elements (EBRII Mark I fuel pins, EBR-II fuel pins with retractory Nb or Ta cladding, and Fermi-I fuel pins) are tested in an inert atmosphere, with no coolant. The fuel elements are exposed to reactor power bursts of 200 msec to 25 sec duration, under conditions simulating fast reactor operations. For these tests, the type of power burst, the integrated power, the fuel enrichment, the maximum cladding temperature, and the effects of the test on the fuel element are recorded. ( T.F.H.)

Application of high rate, high temperature anaerobic digestion to fungal thermozyme hydrolysates from carbohydrate wastes.

PubMed

Forbes, C; O'Reilly, C; McLaughlin, L; Gilleran, G; Tuohy, M; Colleran, E

2009-05-01

The objective of this study was to examine the feasibility of using a two-step, fully biological and sustainable strategy for the treatment of carbohydrate rich wastes. The primary step in this strategy involves the application of thermostable enzymes produced by the thermophilic, aerobic fungus, Talaromyces emersonii, to carbohydrate wastes producing a liquid hydrolysate discharged at elevated temperatures. To assess the potential of thermophilic treatment of this hydrolysate, a comparative study of thermophilic and mesophilic digestion of four sugar rich thermozyme hydrolysate waste streams was conducted by operating two high rate upflow anaerobic hybrid reactors (UAHR) at 37 degrees C (R1) and 55 degrees C (R2). The operational performance of both reactors was monitored from start-up by assessing COD removal efficiencies, volatile fatty acid (VFA) discharge and % methane of the biogas produced. Rapid start-up of both R1 and R2 was achieved on an influent composed of the typical sugar components of the organic fraction of municipal solid waste (OFMSW). Both reactors were subsequently challenged in terms of volumetric loading rate (VLR) and it was found that a VLR of 9 gCOD l(-1)d(-1) at a hydraulic retention time (HRT) of 1 day severely affected the thermophilic reactor with instability characterised by a build up of volatile fatty acid (VFA) intermediates in the effluent. The influent to both reactors was changed to a simple glucose and sucrose-based influent supplied at a VLR of 4.5 gCOD l(-1)d(-1) and HRT of 2 days prior to the introduction of thermozyme hydrolysates. Four unique thermozyme hydrolysates were subsequently supplied to the reactors, each for a period of 10 HRTs. The applied hydrolysates were derived from apple pulp, bread, carob powder and cardboard, all of which were successfully and comparably converted by both reactors. The % total carbohydrate removal by both reactors was monitored during the application of the sugar rich thermozyme hydrolysates. This approach offers a sustainable technology for the treatment of carbohydrate rich wastes and highlights the potential of these wastes as substrates for the generation of second-generation biofuels.

Use of aluminum nitride to obtain temperature measurements in a high temperature and high radiation environment

DOEpatents

Wernsman, Bernard R.; Blasi, Raymond J.; Tittman, Bernhard R.; Parks, David A.

2016-04-26

An aluminum nitride piezoelectric ultrasonic transducer successfully operates at temperatures of up to 1000.degree. C. and fast (>1 MeV) neutron fluencies of more than 10.sup.18 n/cm.sup.2. The transducer comprises a transparent, nitrogen rich aluminum nitride (AlN) crystal wafer that is coupled to an aluminum cylinder for pulse-echo measurements. The transducer has the capability to measure in situ gamma heating within the core of a nuclear reactor.

Reactor engineering support of operations at the Davis-Besse nuclear power station

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

Kelley, D.B.

1995-12-31

Reactor engineering functions differ greatly from unit to unit; however, direct support of the reactor operators during reactor startups and operational transients is common to all units. This paper summarizes the support the reactor engineers provide the reactor operators during reactor startups and power changes through the use of automated computer programs at the Davis-Besse nuclear power station.

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.

Deuterium-tritium experiments on the Tokamak Fusion Test reactor

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

Hosea, J.; Adler, J.H.; Alling, P.

The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to {approx}9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning;more » possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the PFUS {approx}6 MW level. Instability in the TAE mode frequency range has been observed at PFUS > 7 MW and its effect on performance in under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored.« less

Temperature and solids retention time control microbial population dynamics and volatile fatty acid production in replicated anaerobic digesters

PubMed Central

Vanwonterghem, Inka; Jensen, Paul D.; Rabaey, Korneel; Tyson, Gene W.

2015-01-01

Anaerobic digestion is a widely used technology for waste stabilization and generation of biogas, and has recently emerged as a potentially important process for the production of high value volatile fatty acids (VFAs) and alcohols. Here, three reactors were seeded with inoculum from a stably performing methanogenic digester, and selective operating conditions (37°C and 55°C; 12 day and 4 day solids retention time) were applied to restrict methanogenesis while maintaining hydrolysis and fermentation. Replicated experiments performed at each set of operating conditions led to reproducible VFA production profiles which could be correlated with specific changes in microbial community composition. The mesophilic reactor at short solids retention time showed accumulation of propionate and acetate (42 ± 2% and 15 ± 6% of CODhydrolyzed, respectively), and dominance of Fibrobacter and Bacteroidales. Acetate accumulation (>50% of CODhydrolyzed) was also observed in the thermophilic reactors, which were dominated by Clostridium. Under all tested conditions, there was a shift from acetoclastic to hydrogenotrophic methanogenesis, and a reduction in methane production by >50% of CODhydrolyzed. Our results demonstrate that shortening the SRT and increasing the temperature are effective strategies for driving microbial communities towards controlled production of high levels of specific volatile fatty acids. PMID:25683239

Temperature and solids retention time control microbial population dynamics and volatile fatty acid production in replicated anaerobic digesters

NASA Astrophysics Data System (ADS)

Vanwonterghem, Inka; Jensen, Paul D.; Rabaey, Korneel; Tyson, Gene W.

2015-02-01

Anaerobic digestion is a widely used technology for waste stabilization and generation of biogas, and has recently emerged as a potentially important process for the production of high value volatile fatty acids (VFAs) and alcohols. Here, three reactors were seeded with inoculum from a stably performing methanogenic digester, and selective operating conditions (37°C and 55°C 12 day and 4 day solids retention time) were applied to restrict methanogenesis while maintaining hydrolysis and fermentation. Replicated experiments performed at each set of operating conditions led to reproducible VFA production profiles which could be correlated with specific changes in microbial community composition. The mesophilic reactor at short solids retention time showed accumulation of propionate and acetate (42 +/- 2% and 15 +/- 6% of CODhydrolyzed, respectively), and dominance of Fibrobacter and Bacteroidales. Acetate accumulation (>50% of CODhydrolyzed) was also observed in the thermophilic reactors, which were dominated by Clostridium. Under all tested conditions, there was a shift from acetoclastic to hydrogenotrophic methanogenesis, and a reduction in methane production by >50% of CODhydrolyzed. Our results demonstrate that shortening the SRT and increasing the temperature are effective strategies for driving microbial communities towards controlled production of high levels of specific volatile fatty acids.

LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

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

Carmack, William Jonathan; Barrett, Kristine Eloise; Chichester, Heather Jean MacLean

2015-09-01

The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirementsmore » for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.« less

Continuation of studies on thermoregulation of fish and turtles in thermally stressed habitats. Annual progress report, 1 October 1979-30 September 1980

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

Spotila, J.R.

1980-05-01

Fundamental and realized climate spaces were calculated for the turtle Chrysemys scripta. These allow predictions about the effect of microclimate and thermal effluents on the behavior of these animals to be made. A conceptual model to define the biophysical-behavioral thermoregulatory mechanisms employed by this turtle is being finalized. Operative environmental temperature (T/sub e/) is a good predictor of the basking behavior of turtles. T/sub e/ is positively related to visible and thermal radiation and air temperature. Turtles generally do not bask until T/sub e/ exceeds 28/sup 0/C, thus implicating thermoregulation as a major factor in determining the basking behavior ofmore » C. scripta. Water temperature was very important in determining the distribution of largemouth bass, Micropterus salmoides, in a South Carolina reservoir receiving thermal effluent from a nuclear reactor. Bass were restricted in movement by lethal water temperatures, selecting temperatures close to 30/sup 0/C and avoiding temperatures above 31/sup 0/C. Under normal, unheated conditions, bass dispersed throughout the reservoir. During reactor operation, hot water at temperatures lethal to fish (approx. 55/sup 0/C), forced bass to retreat to refuges in two coves and a deep spring. Distribution of bass varied seasonally. Multichannel radio transmitters were surgically implanted in free ranging fish, permitting the telemetry of temperatures from five parts of the body and from surrounding water. In general, body temperatures followed water temperatures closely, but rapidly changing temperatures produced lags between body temperatures and water of as much as 3.5/sup 0/C. (ERB)« less

A Kinetic Model for GaAs Growth by Hydride Vapor Phase Epitaxy

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

Schulte, Kevin L.; Simon, John; Jain, Nikhil

2016-11-21

Precise control of the growth of III-V materials by hydride vapor phase epitaxy (HVPE) is complicated by the fact that the growth rate depends on the concentrations of nearly all inputs to the reactor and also the reaction temperature. This behavior is in contrast to metalorganic vapor phase epitaxy (MOVPE), which in common practice operates in a mass transport limited regime where growth rate and alloy composition are controlled almost exclusively by flow of the Group III precursor. In HVPE, the growth rate and alloy compositions are very sensitive to temperature and reactant concentrations, which are strong functions of themore » reactor geometry. HVPE growth, particularly the growth of large area materials and devices, will benefit from the development of a growth model that can eventually be coupled with a computational fluid dynamics (CFD) model of a specific reactor geometry. In this work, we develop a growth rate law using a Langmuir-Hinshelwood (L-H) analysis, fitting unknown parameters to growth rate data from the literature that captures the relevant kinetic and thermodynamic phenomena of the HVPE process. We compare the L-H rate law to growth rate data from our custom HVPE reactor, and develop quantitative insight into reactor performance, demonstrating the utility of the growth model.« less

Parametric study of irradiation effects on the ductile damage and flow stress behavior in ferritic-martensitic steels

NASA Astrophysics Data System (ADS)

Chakraborty, Pritam; Biner, S. Bulent

2015-10-01

Ferritic-martensitic steels are currently being considered as structural materials in fusion and Gen-IV nuclear reactors. These materials are expected to experience high dose radiation, which can increase their ductile to brittle transition temperature and susceptibility to failure during operation. Hence, to estimate the safe operational life of the reactors, precise evaluation of the ductile to brittle transition temperatures of ferritic-martensitic steels is necessary. Owing to the scarcity of irradiated samples, particularly at high dose levels, micro-mechanistic models are being employed to predict the shifts in the ductile to brittle transition temperatures. These models consider the ductile damage evolution, in the form of nucleation, growth and coalescence of voids; and the brittle fracture, in the form of probabilistic cleavage initiation, to estimate the influence of irradiation on the ductile to brittle transition temperature. However, the assessment of irradiation dependent material parameters is challenging and influences the accuracy of these models. In the present study, the effects of irradiation on the overall flow stress and ductile damage behavior of two ferritic-martensitic steels is parametrically investigated. The results indicate that the ductile damage model parameters are mostly insensitive to irradiation levels at higher dose levels though the resulting flow stress behavior varies significantly.

Oxide Morphology of a FeCrAl Alloy, Kanthal APMT, following Extended Aging at 300-600C

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

Li, Nan; Parker, Stephen Scott; Wood, Elizabeth Sooby

Iron-chromium-aluminum (FeCrAl) alloys are of interest to the nuclear materials community due to their resistance to high temperature steam oxidation under accident conditions. The present work investigates oxide formation at temperatures relevant to light water reactor cladding operation following extended aging to assess growth kinetics, chemical composition, and microstructure of oxide formation on a commercial FeCrAl alloy, Fe-21wt.%Cr-5wt.%Al-3wt.%Mo (Kanthal APMT). Aging treatments were performed for 100-1000 hours in stagnant air at 300, 400, 500, and 600 °C, respectively. Oxide growth behavior under the investigated conditions follows a logarithmic time dependence. When the oxidization temperature is 400 °C or below, themore » oxide is amorphous. At 500 °C, isolated crystalline regions start to appear during short period aging time and expand with extended exposures. Crystalline α-Al2O3 oxide film develops at 600 °C and the correlated logarithmic rate constant decreases significantly, indicating enhanced oxidation resistance of the formed oxide film. In addition, Mo segregation at grain boundaries has been observed when the aging temperature exceeds 500 °C. The results of this study can be viewed as an upper bounding result for potential oxide coarsening during reactor operation.« less

Reactor Simulator Testing

NASA Technical Reports Server (NTRS)

Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise Jon

2013-01-01

As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz. Keywords: fission, space power, nuclear, liquid metal, NaK.

Reduction of dioxin emission by a multi-layer reactor with bead-shaped activated carbon in simulated gas stream and real flue gas of a sinter plant.

PubMed

Hung, Pao Chen; Lo, Wei Chiao; Chi, Kai Hsien; Chang, Shu Hao; Chang, Moo Been

2011-01-01

A laboratory-scale multi-layer system was developed for the adsorption of PCDD/Fs from gas streams at various operating conditions, including gas flow rate, operating temperature and water vapor content. Excellent PCDD/F removal efficiency (>99.99%) was achieved with the multi-layer design with bead-shaped activated carbons (BACs). The PCDD/F removal efficiency achieved with the first layer adsorption bed decreased as the gas flow rate was increased due to the decrease of the gas retention time. The PCDD/F concentrations measured at the outlet of the third layer adsorption bed were all lower than 0.1 ng I-TEQ Nm⁻³. The PCDD/Fs desorbed from BAC were mainly lowly chlorinated congeners and the PCDD/F outlet concentrations increased as the operating temperature was increased. In addition, the results of pilot-scale experiment (real flue gases of an iron ore sintering plant) indicated that as the gas flow rate was controlled at 15 slpm, the removal efficiencies of PCDD/F congeners achieved with the multi-layer reactor with BAC were better than that in higher gas flow rate condition (20 slpm). Overall, the lab-scale and pilot-scale experiments indicated that PCDD/F removal achieved by multi-layer reactor with BAC strongly depended on the flow rate of the gas stream to be treated. Copyright © 2010 Elsevier Ltd. All rights reserved.

High temperature oxidation behavior of ODS steels

NASA Astrophysics Data System (ADS)

Kaito, T.; Narita, T.; Ukai, S.; Matsuda, Y.

2004-08-01

Oxide dispersion strengthened (ODS) steels are being developing for application as advanced fast reactor cladding and fusion blanket materials, in order to allow increased operation temperature. Oxidation testing of ODS steel was conducted under a controlled dry air atmosphere to evaluate the high temperature oxidation behavior. This showed that 9Cr-ODS martensitic steels and 12Cr-ODS ferritic steels have superior high temperature oxidation resistance compared to 11 mass% Cr PNC-FMS and 17 mass% Cr ferritic stainless steel. This high temperature resistance is attributed to earlier formation of the protective α-Cr 2O 3 on the outer surface of ODS steels.

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

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

Thermal Characterization of a Simulated Fission Engine via Distributed Fiber Bragg Gratings

NASA Astrophysics Data System (ADS)

Duncan, Roger G.; Fielder, Robert S.; Seeley, Ryan J.; Kozikowski, Carrie L.; Raum, Matthew T.

2005-02-01

We report the use of distributed fiber Bragg gratings to monitor thermal conditions within a simulated nuclear reactor core located at the Early Flight Fission Test Facility of the NASA Marshall Space Flight Center. Distributed fiber-optic temperature measurements promise to add significant capability and advance the state-of-the-art in high-temperature sensing. For the work reported herein, seven probes were constructed with ten sensors each for a total of 70 sensor locations throughout the core. These discrete temperature sensors were monitored over a nine hour period while the test article was heated to over 700 °C and cooled to ambient through two operational cycles. The sensor density available permits a significantly elevated understanding of thermal effects within the simulated reactor. Fiber-optic sensor performance is shown to compare very favorably with co-located thermocouples where such co-location was feasible.

Evaluation of the finite element fuel rod analysis code (FRANCO)

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

Lee, K.; Feltus, M.A.

1994-12-31

Knowledge of temperature distribution in a nuclear fuel rod is required to predict the behavior of fuel elements during operating conditions. The thermal and mechanical properties and performance characteristics are strongly dependent on the temperature, which can vary greatly inside the fuel rod. A detailed model of fuel rod behavior can be described by various numerical methods, including the finite element approach. The finite element method has been successfully used in many engineering applications, including nuclear piping and reactor component analysis. However, fuel pin analysis has traditionally been carried out with finite difference codes, with the exception of Electric Powermore » Research Institute`s FREY code, which was developed for mainframe execution. This report describes FRANCO, a finite element fuel rod analysis code capable of computing temperature disrtibution and mechanical deformation of a single light water reactor fuel rod.« less

Superconducting RF Linacs Driving Subcritical Reactors for Profitable Disposition of Surplus Weapons-grade Plutonium

NASA Astrophysics Data System (ADS)

Cummings, Mary Anne; Johnson, Rolland

Acceptable capital and operating costs of high-power proton accelerators suitable for profitable commercial electric-power and process-heat applications have been demonstrated. However, studies have pointed out that even a few hundred trips of an accelerator lasting a few seconds would lead to unacceptable thermal stresses as each trip causes fission to be turned off in solid fuel structures found in conventional reactors. The newest designs based on the GEM*STAR concept take such trips in stride by using molten-salt fuel, where fuel pin fatigue is not an issue. Other aspects of the GEM*STAR concept which address all historical reactor failures include an internal spallation neutron target and high temperature molten salt fuel with continuous purging of volatile radioactive fission products such that the reactor contains less than a critical mass and almost a million times fewer volatile radioactive fission products than conventional reactors. GEM*STAR is a reactor that without redesign will burn spent nuclear fuel, natural uranium, thorium, or surplus weapons material. It will operate without the need for a critical core, fuel enrichment, or reprocessing making it an excellent candidate for export. As a first application, the design for a pilot plant is described for the profitable disposition of surplus weapons-grade plutonium by using process heat to produce green diesel fuel for the Department of Defense (DOD) from natural gas and renewable carbon.

Performance and stability of an expanded granular sludge bed reactor modified with zeolite addition subjected to step increases of organic loading rate (OLR) and to organic shock load (OSL).

PubMed

Pérez-Pérez, T; Pereda-Reyes, I; Pozzi, E; Oliva-Merencio, D; Zaiat, M

2018-01-01

This paper shows the effect of organic shock loads (OSLs) on the anaerobic digestion (AD) of synthetic swine wastewater using an expanded granular sludge bed (EGSB) reactor modified with zeolite. Two reactors (R1 and R2), each with an effective volume of 3.04 L, were operated for 180 days at a controlled temperature of 30 °C and hydraulic retention time of 12 h. In the case of R2, 120 g of zeolite was added. The reactors were operated with an up-flow velocity of 6 m/h. The evolution of pH, total Kjeldahl nitrogen, chemical oxygen demand (COD) and volatile fatty acids (VFAs) was monitored during the AD process with OSL and increases in the organic loading rate (OLR). In addition, the microbial composition and changes in the structure of the bacterial and archaeal communities were assessed. The principal results demonstrate that the presence of zeolite in an EGSB reactor provides a more stable process at higher OLRs and after applying OSL, based on both COD and VFA accumulation, which presented with significant differences compared to the control. Denaturing gradient gel electrophoresis band profiles indicated differences in the populations of Bacteria and Archaea between the R1 and R2 reactors, attributed to the presence of zeolite.

Catalytic multi-stage process for hydroconversion and refining hydrocarbon feeds

DOEpatents

Comolli, Alfred G.; Lee, Lap-Keung

2001-01-01

A multi-stage catalytic hydrogenation and hydroconversion process for heavy hydrocarbon feed materials such as coal, heavy petroleum fractions, and plastic waste materials. In the process, the feedstock is reacted in a first-stage, back-mixed catalytic reactor with a highly dispersed iron-based catalyst having a powder, gel or liquid form. The reactor effluent is pressure-reduced, vapors and light distillate fractions are removed overhead, and the heavier liquid fraction is fed to a second stage back-mixed catalytic reactor. The first and second stage catalytic reactors are operated at 700-850.degree. F. temperature, 1000-3500 psig hydrogen partial pressure and 20-80 lb./hr per ft.sup.3 reactor space velocity. The vapor and light distillates liquid fractions removed from both the first and second stage reactor effluent streams are combined and passed to an in-line, fixed-bed catalytic hydrotreater for heteroatom removal and for producing high quality naphtha and mid-distillate or a full-range distillate product. The remaining separator bottoms liquid fractions are distilled at successive atmospheric and vacuum pressures, low and intermediate-boiling hydrocarbon liquid products are withdrawn, and heavier distillate fractions are recycled and further upgraded to provide additional low-boiling hydrocarbon liquid products. This catalytic multistage hydrogenation process provides improved flexibility for hydroprocessing the various carbonaceous feedstocks and adjusting to desired product structures and for improved economy of operations.

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.

DELTA PHASE PLUTONIUM ALLOYS

DOEpatents

Cramer, E.M.; Ellinger, F.H.; Land. C.C.

1960-03-22

Delta-phase plutonium alloys were developed suitable for use as reactor fuels. The alloys consist of from 1 to 4 at.% zinc and the balance plutonium. The alloys have good neutronic, corrosion, and fabrication characteristics snd possess good dimensional characteristics throughout an operating temperature range from 300 to 490 deg C.

Introduction to the special issue on the technical status of materials for a fusion reactor

NASA Astrophysics Data System (ADS)

Stork, D.; Zinkle, S. J.

2017-09-01

Materials determine in a fundamental way the performance and environmental attractiveness of a fusion reactor: through the size (power fluxes to the divertor, neutron fluxes to the first wall); economics (replacement lifetime of critical in-vessel components, thermodynamic efficiency through operating temperature etc); plasma performance (erosion by plasma fluxes to the divertor surfaces); robustness against off-normal accidents (safety); and the effects of post-operation radioactivity on waste disposal and maintenance. The major philosophies and methodologies used to formulate programmes for the development of fusion materials are outlined, as the basis for other articles in this special issue, which deal with the fundamental understanding of the issues regarding these materials and their technical status and prospects for development.

Onset conditions for gas phase reaction and nucleation in the CVD of transition metal oxides

NASA Technical Reports Server (NTRS)

Collins, J.; Rosner, D. E.; Castillo, J.

1992-01-01

A combined experimental/theoretical study is presented of the onset conditions for gas phase reaction and particle nucleation in hot substrate/cold gas CVD of transition metal oxides. Homogeneous reaction onset conditions are predicted using a simple high activation energy reacting gas film theory. Experimental tests of the basic theory are underway using an axisymmetric impinging jet CVD reactor. No vapor phase ignition has yet been observed in the TiCl4/O2 system under accessible operating conditions (below substrate temperature Tw = 1700 K). The goal of this research is to provide CVD reactor design and operation guidelines for achieving acceptable deposit microstructures at the maximum deposition rate while simultaneously avoiding homogeneous reaction/nucleation and diffusional limitations.

Advanced propulsion engine assessment based on a cermet reactor

NASA Technical Reports Server (NTRS)

Parsley, Randy C.

1993-01-01

A preferred Pratt & Whitney conceptual Nuclear Thermal Rocket Engine (NTRE) has been designed based on the fundamental NASA priorities of safety, reliability, cost, and performance. The basic philosophy underlying the design of the XNR2000 is the utilization of the most reliable form of ultrahigh temperature nuclear fuel and development of a core configuration which is optimized for uniform power distribution, operational flexibility, power maneuverability, weight, and robustness. The P&W NTRE system employs a fast spectrum, cermet fueled reactor configured in an expander cycle to ensure maximum operational safety. The cermet fuel form provides retention of fuel and fission products as well as high strength. A high level of confidence is provided by benchmark analysis and independent evaluations.

Atmosphere Processing Module Automation and Catalyst Durability Analysis for Mars ISRU Pathfinder

NASA Technical Reports Server (NTRS)

Petersen, Elspeth M.

2016-01-01

The Mars In-Situ Resource Utilization Pathfinder was designed to create fuel using components found in the planet’s atmosphere and regolith for an ascension vehicle to return a potential sample return or crew return vehicle from Mars. The Atmosphere Processing Module (APM), a subunit of the pathfinder, uses cryocoolers to isolate and collect carbon dioxide from Mars simulant gas. The carbon dioxide is fed with hydrogen into a Sabatier reactor where methane is produced. The APM is currently undergoing the final stages of testing at Kennedy Space Center prior to process integration testing with the other subunits of the pathfinder. The automation software for the APM cryocoolers was tested and found to perform nominally. The catalyst used for the Sabatier reactor was investigated to determine the factors contributing to catalyst failure. The results from the catalyst testing require further analysis, but it appears that the rapid change in temperature during reactor start up or the elevated operating temperature is responsible for the changes observed in the catalyst.

Catalytic two-stage coal hydrogenation and hydroconversion process

DOEpatents

MacArthur, James B.; McLean, Joseph B.; Comolli, Alfred G.

1989-01-01

A process for two-stage catalytic hydrogenation and liquefaction of coal to produce increased yields of low-boiling hydrocarbon liquid and gas products. In the process, the particulate coal is slurried with a process-derived liquid solvent and fed at temperature below about 650.degree. F. into a first stage catalytic reaction zone operated at conditions which promote controlled rate liquefaction of the coal, while simultaneously hydrogenating the hydrocarbon recycle oils at conditions favoring hydrogenation reactions. The first stage reactor is maintained at 650.degree.-800.degree. F. temperature, 1000-4000 psig hydrogen partial pressure, and 10-60 lb coal/hr/ft.sup.3 reactor space velocity. The partially hydrogenated material from the first stage reaction zone is passed directly to the close-coupled second stage catalytic reaction zone maintained at a temperature at least about 25.degree. F. higher than for the first stage reactor and within a range of 750.degree.-875.degree. F. temperature for further hydrogenation and thermal hydroconversion reactions. By this process, the coal feed is successively catalytically hydrogenated and hydroconverted at selected conditions, which results in significantly increased yields of desirable low-boiling hydrocarbon liquid products and minimal production of undesirable residuum and unconverted coal and hydrocarbon gases, with use of less energy to obtain the low molecular weight products, while catalyst life is substantially increased.

Composting on Mars or the Moon: II. Temperature feedback control with top-wise introduction of waste material and air

NASA Technical Reports Server (NTRS)

Finstein, M. S.; Hogan, J. A.; Sager, J. C.; Cowan, R. M.; Strom, P. F.; Janes, H. W. (Principal Investigator)

1999-01-01

Whereas Earth-based composting reactors that effectively control the process are batch operations with bottom-to-top airflow, in extraterrestrial application both the fresh waste and the air need to be introduced from above. Stabilized compost and used air would exit below. This materials flow pattern permits the addition of waste whenever generated, obviating the need for multiple reactors, and the incorporation of a commode in the lid. Top loading in turn dictates top-down aeration, so that the most actively decomposing material (greatest need for heat removal and O2 replenishment) is first encountered. This novel material and aeration pattern was tested in conjunction with temperature feedback process control. Reactor characteristics were: working, volume, 0.15 m3; charge, 2 kg dry biomass per day (comparable to a 3-4 person self-sufficient bioregenerative habitat); retention time, 7 days. Judging from temperature profile, O2 level, air usage, pressure head loss, moisture, and odor, the system was effectively controlled over a 35-day period. Dry matter disappearance averaged 25% (10-42%). The compost product was substantially, though not completely, stabilized. This demonstrates the compatibility of top-wise introduction of waste and air with temperature feedback process control.

Composting on Mars or the Moon: II. Temperature feedback control with top-wise introduction of waste material and air.

PubMed

Finstein, M S; Hogan, J A; Sager, J C; Cowan, R M; Strom, P F

1999-01-01

Whereas Earth-based composting reactors that effectively control the process are batch operations with bottom-to-top airflow, in extraterrestrial application both the fresh waste and the air need to be introduced from above. Stabilized compost and used air would exit below. This materials flow pattern permits the addition of waste whenever generated, obviating the need for multiple reactors, and the incorporation of a commode in the lid. Top loading in turn dictates top-down aeration, so that the most actively decomposing material (greatest need for heat removal and O2 replenishment) is first encountered. This novel material and aeration pattern was tested in conjunction with temperature feedback process control. Reactor characteristics were: working, volume, 0.15 m3; charge, 2 kg dry biomass per day (comparable to a 3-4 person self-sufficient bioregenerative habitat); retention time, 7 days. Judging from temperature profile, O2 level, air usage, pressure head loss, moisture, and odor, the system was effectively controlled over a 35-day period. Dry matter disappearance averaged 25% (10-42%). The compost product was substantially, though not completely, stabilized. This demonstrates the compatibility of top-wise introduction of waste and air with temperature feedback process control.

Chemistry in microstructured reactors.

PubMed

Jähnisch, Klaus; Hessel, Volker; Löwe, Holger; Baerns, Manfred

2004-01-16

The application of microstructured reactors in the chemical process industry has gained significant importance in recent years. Companies that offer not only microstructured reactors, but also entire chemical process plants and services relating to them, are already in existence. In addition, many institutes and universities are active within this field, and process-engineering-oriented reviews and a specialized book are available. Microstructured systems can be applied with particular success in the investigation of highly exothermic and fast reactions. Often the presence of temperature-induced side reactions can be significantly reduced through isothermal operations. Although microstructured reaction techniques have been shown to optimize many synthetic procedures, they have not yet received the attention they deserve in organic chemistry. For this reason, this Review aims to address this by providing an overview of the chemistry in microstructured reactors, grouped into liquid-phase, gas-phase, and gas-liquid reactions.

Non-equilibrium plasma reactors for organic solvent destruction

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

Yang, C.L.; Beltran, M.R.; Kravets, Z.

1997-12-31

Two non-equilibrium plasma reactors were evaluated for their ability to destroy three widely used organic solvents, i.e., 2-butanone, toluene and ethyl acetate. The catalyzed plasma reactor (CPR) with 6 mm glass beads destroys 98% of 50 ppm toluene in air at 24 kV/cm and space velocities of 1,400 v/v/hr. Eighty-five percent of ethyl acetate and 2-butanone are destroyed under the same conditions. The tubular plasma reactor (TPR) has an efficiency of 10% to 20% lower than that of a CPR under the same conditions. The 1,400 v/v/hr in a CPR is equal to a residence time of 2.6 seconds inmore » a TPR. The operating temperatures, corona characteristics, as well as the kinetics of VOC destruction in both TPR and CPR were studied.« less

A new high temperature reactor for operando XAS: Application for the dry reforming of methane over Ni/ZrO2 catalyst.

PubMed

Aguilar-Tapia, Antonio; Ould-Chikh, Samy; Lahera, Eric; Prat, Alain; Delnet, William; Proux, Olivier; Kieffer, Isabelle; Basset, Jean-Marie; Takanabe, Kazuhiro; Hazemann, Jean-Louis

2018-03-01

The construction of a high-temperature reaction cell for operando X-ray absorption spectroscopy characterization is reported. A dedicated cell was designed to operate as a plug-flow reactor using powder samples requiring gas flow and thermal treatment at high temperatures. The cell was successfully used in the reaction of dry reforming of methane (DRM). We present X-ray absorption results in the fluorescence detection mode on a 0.4 wt. % Ni/ZrO 2 catalyst under realistic conditions at 750 °C, reproducing the conditions used for a conventional dynamic microreactor for the DRM reaction. The setup includes a gas distribution system that can be fully remotely operated. The reaction cell offers the possibility of transmission and fluorescence detection modes. The complete setup dedicated to the study of catalysts is permanently installed on the Collaborating Research Groups French Absorption spectroscopy beamline in Material and Environmental sciences (CRG-FAME) and French Absorption spectroscopy beamline in Material and Environmental sciences at Ultra-High Dilution (FAME-UHD) beamlines (BM30B and BM16) at the European Synchrotron Radiation Facility in Grenoble, France.

A new high temperature reactor for operando XAS: Application for the dry reforming of methane over Ni/ZrO2 catalyst

NASA Astrophysics Data System (ADS)

Aguilar-Tapia, Antonio; Ould-Chikh, Samy; Lahera, Eric; Prat, Alain; Delnet, William; Proux, Olivier; Kieffer, Isabelle; Basset, Jean-Marie; Takanabe, Kazuhiro; Hazemann, Jean-Louis

2018-03-01

The construction of a high-temperature reaction cell for operando X-ray absorption spectroscopy characterization is reported. A dedicated cell was designed to operate as a plug-flow reactor using powder samples requiring gas flow and thermal treatment at high temperatures. The cell was successfully used in the reaction of dry reforming of methane (DRM). We present X-ray absorption results in the fluorescence detection mode on a 0.4 wt. % Ni/ZrO2 catalyst under realistic conditions at 750 °C, reproducing the conditions used for a conventional dynamic microreactor for the DRM reaction. The setup includes a gas distribution system that can be fully remotely operated. The reaction cell offers the possibility of transmission and fluorescence detection modes. The complete setup dedicated to the study of catalysts is permanently installed on the Collaborating Research Groups French Absorption spectroscopy beamline in Material and Environmental sciences (CRG-FAME) and French Absorption spectroscopy beamline in Material and Environmental sciences at Ultra-High Dilution (FAME-UHD) beamlines (BM30B and BM16) at the European Synchrotron Radiation Facility in Grenoble, France.

Dilute Acid and Autohydrolysis Pretreatment

NASA Astrophysics Data System (ADS)

Yang, Bin; Wyman, Charles E.

Exposure of cellulosic biomass to temperatures of about 120-210°C can remove most of the hemicellulose and produce cellulose-rich solids from which high glucose yields are possible with cellulase enzymes. Furthermore, the use of dilute sulfuric acid in this pretreatment operation can increase recovery of hemicellulose sugars substantially to about 85-95% of the maximum possible versus only about 65% if no acid is employed. The use of small-diameter tubes makes it possible to employ high solids concentrations similar to those preferred for commercial operations, with rapid heat-up, good temperature control, and accurate closure of material balances. Mixed reactors can be employed to pretreat larger amounts of biomass than possible in such small-diameter tubes, but solids concentrations are limited to about 15% or less to provide uniform temperatures. Pretreatment of large amounts of biomass at high solids concentrations is best carried out using direct steam injection and rapid pressure release, but closure of material balances in such “steam gun” devices is more difficult. Although flow of water alone or containing dilute acid is not practical commercially, such flow-through configurations provide valuable insight into biomass deconstruction kinetics not possible in the batch tubes, mixed reactors, or steam gun systems.

Heat transfer system

DOEpatents

Not Available

1980-03-07

A heat transfer system for a nuclear reactor is described. Heat transfer is accomplished within a sealed vapor chamber which is substantially evacuated prior to use. A heat transfer medium, which is liquid at the design operating temperatures, transfers heat from tubes interposed in the reactor primary loop to spaced tubes connected to a steam line for power generation purposes. Heat transfer is accomplished by a two-phase liquid-vapor-liquid process as used in heat pipes. Condensible gases are removed from the vapor chamber through a vertical extension in open communication with the chamber interior.

Heat transfer system

DOEpatents

McGuire, Joseph C.

1982-01-01

A heat transfer system for a nuclear reactor. Heat transfer is accomplished within a sealed vapor chamber which is substantially evacuated prior to use. A heat transfer medium, which is liquid at the design operating temperatures, transfers heat from tubes interposed in the reactor primary loop to spaced tubes connected to a steam line for power generation purposes. Heat transfer is accomplished by a two-phase liquid-vapor-liquid process as used in heat pipes. Condensible gases are removed from the vapor chamber through a vertical extension in open communication with the chamber interior.

Materials interactions between the thermoelectric converter and the 5kwe reactor system

NASA Technical Reports Server (NTRS)

Ferry, P. B.

1973-01-01

The integration of a compact thermoelectric converter with a 5-kwe reactor system is described. Material interaction uncertainties study is also presented. This includes degradation of the required austenitic - refractory metal transition joint during operation at high temperatures; loss of corrosion resistance; embrittlement by the presence of hydrogen; and loss of design margin by transport of interstitial elements. Analysis and limited experimental evidence indicate that these potential materials interactions can be adequately controlled. Group 5-2 refractory metals can be utilized without unacceptable adverse effect on system reliability.

NaK Plugging Meter Design for the Feasibility Test Loops

NASA Technical Reports Server (NTRS)

Pearson, J. Boise; Godfroy, Thomas J.; Reid, Robert S.; Polzin, Kurt A.

2008-01-01

The design and predicted performance of a plugging meter for use in the measurement of NaK impurity levels are presented. The plugging meter is incorporated into a Feasibility Test Loop (FTL), which is a small pumped-NaK loop designed to enable the rapid, small-scale evaluation of techniques such as in situ purification methods and to permit the measurement of bulk material transport effects (not mechanisms) under flow conditions that are representative of a fission surface power reactor. The FTL operates at temperatures similar to those found in a reactor, with a maximum hot side temperature of 900 K and a corresponding cold side temperature of 860 K. In the plugging meter a low flow rate bypass loop is cooled until various impurities (primarily oxides) precipitate out of solution. The temperatures at which these impurities precipitate are indicative of the level of impurities in the NaK. The precipitates incrementally plug a small orifice in the bypass loop, which is detected by monitoring changes in the liquid metal flow rate.

Impact of temperature, microwave radiation and organic loading rate on methanogenic community and biogas production during fermentation of dairy wastewater.

PubMed

Zielińska, Magdalena; Cydzik-Kwiatkowska, Agnieszka; Zieliński, Marcin; Dębowski, Marcin

2013-02-01

This study analyzed dairy wastewater fermentation in convection- and microwave-heated hybrid reactors at loadings of 1 and 2 kg COD/(m3 d) and temperatures of 35 and 55 °C. The biomass was investigated at a molecular level to determine the links between the operational parameters of anaerobic digestion and methanogenic Archaea structure. The highest production of biogas with methane content of ca. 67% was noted in the mesophilic microwave-heated reactors. The production of methane-rich biogas and the overall diversity of Archaea was determined by Methanosarcinaceae presence. The temperature and the application of microwaves were the main factors explaining the variations in the methanogen community. At 35 °C, the microwave heating stimulated the growth of highly diverse methanogen assemblages, promoting Methanosarcina barkeri presence and excluding Methanosarcina harudinacea from the biomass. A temperature increase to 55 °C lowered Methanosarcinaceae abundance and induced a replacement of Methanoculleus palmolei by Methanosarcina thermophila. Copyright © 2012 Elsevier Ltd. All rights reserved.

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

Aerosol reactor production of uniform submicron powders

NASA Technical Reports Server (NTRS)

Flagan, Richard C. (Inventor); Wu, Jin J. (Inventor)

1991-01-01

A method of producing submicron nonagglomerated particles in a single stage reactor includes introducing a reactant or mixture of reactants at one end while varying the temperature along the reactor to initiate reactions at a low rate. As homogeneously small numbers of seed particles generated in the initial section of the reactor progress through the reactor, the reaction is gradually accelerated through programmed increases in temperature along the length of the reactor to promote particle growth by chemical vapor deposition while minimizing agglomerate formation by maintaining a sufficiently low number concentration of particles in the reactor such that coagulation is inhibited within the residence time of particles in the reactor. The maximum temperature and minimum residence time is defined by a combination of temperature and residence time that is necessary to bring the reaction to completion. In one embodiment, electronic grade silane and high purity nitrogen are introduced into the reactor and temperatures of approximately 770.degree. K. to 1550.degree. K. are employed. In another embodiment silane and ammonia are employed at temperatures from 750.degree. K. to 1800.degree. K.

Aerosol reactor production of uniform submicron powders

DOEpatents

Flagan, Richard C.; Wu, Jin J.

1991-02-19

A method of producing submicron nonagglomerated particles in a single stage reactor includes introducing a reactant or mixture of reactants at one end while varying the temperature along the reactor to initiate reactions at a low rate. As homogeneously small numbers of seed particles generated in the initial section of the reactor progress through the reactor, the reaction is gradually accelerated through programmed increases in temperature along the length of the reactor to promote particle growth by chemical vapor deposition while minimizing agglomerate formation by maintaining a sufficiently low number concentration of particles in the reactor such that coagulation is inhibited within the residence time of particles in the reactor. The maximum temperature and minimum residence time is defined by a combination of temperature and residence time that is necessary to bring the reaction to completion. In one embodiment, electronic grade silane and high purity nitrogen are introduced into the reactor and temperatures of approximately 770.degree. K. to 1550.degree. K. are employed. In another embodiment silane and ammonia are employed at temperatures from 750.degree. K. to 1800.degree. K.

ICP-MS analysis of fission product diffusion in graphite for High-Temperature Gas-Cooled Reactors

NASA Astrophysics Data System (ADS)

Carter, Lukas M.

Release of radioactive fission products from nuclear fuel during normal reactor operation or in accident scenarios is a fundamental safety concern. Of paramount importance are the understanding and elucidation of mechanisms of chemical interaction, nuclear interaction, and transport phenomena involving fission products. Worldwide efforts to reduce fossil fuel dependence coupled with an increasing overall energy demand have generated renewed enthusiasm toward nuclear power technologies, and as such, these mechanisms continue to be the subjects of vigorous research. High-Temperature Gas-Cooled Reactors (HTGRs or VHTRs) remain one of the most promising candidates for the next generation of nuclear power reactors. An extant knowledge gap specific to HTGR technology derives from an incomplete understanding of fission product transport in major core materials under HTGR operational conditions. Our specific interest in the current work is diffusion in reactor graphite. Development of methods for analysis of diffusion of multiple fission products is key to providing accurate models for fission product release from HTGR core components and the reactor as a whole. In the present work, a specialized diffusion cell has been developed and constructed to facilitate real-time diffusion measurements via ICP-MS. The cell utilizes a helium gas-jet system which transports diffusing fission products to the mass spectrometer using carbon nanoparticles. The setup was designed to replicate conditions present in a functioning HTGR, and can be configured for real-time release or permeation measurements of single or multiple fission products from graphite or other core materials. In the present work, we have analyzed release rates of cesium in graphite grades IG-110, NBG-18, and a commercial grade of graphite, as well as release of iodine in IG-110. Additionally we have investigated infusion of graphite samples with Cs, I, Sr, Ag, and other surrogate fission products for use in release or profile measurements of diffusion coefficients.

Pressure-accelerated azide-alkyne cycloaddition: micro capillary versus autoclave reactor performance.

PubMed

Borukhova, Svetlana; Seeger, Andreas D; Noël, Timothy; Wang, Qi; Busch, Markus; Hessel, Volker

2015-02-01

Pressure effects on regioselectivity and yield of cycloaddition reactions have been shown to exist. Nevertheless, high pressure synthetic applications with subsequent benefits in the production of natural products are limited by the general availability of the equipment. In addition, the virtues and limitations of microflow equipment under standard conditions are well established. Herein, we apply novel-process-window (NPWs) principles, such as intensification of intrinsic kinetics of a reaction using high temperature, pressure, and concentration, on azide-alkyne cycloaddition towards synthesis of Rufinamide precursor. We applied three main activation methods (i.e., uncatalyzed batch, uncatalyzed flow, and catalyzed flow) on uncatalyzed and catalyzed azide-alkyne cycloaddition. We compare the performance of two reactors, a specialized autoclave batch reactor for high-pressure operation up to 1800 bar and a capillary flow reactor (up to 400 bar). A differentiated and comprehensive picture is given for the two reactors and the three methods of activation. Reaction speedup and consequent increases in space-time yields is achieved, while the process window for favorable operation to selectively produce Rufinamide precursor in good yields is widened. The best conditions thus determined are applied to several azide-alkyne cycloadditions to widen the scope of the presented methodology. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Sustainable nitrogen removal by denitrifying anammox applied for anaerobic pre-treated potato wastewater.

PubMed

Mulder, A; Versprille, A I; van Braak, D

2012-01-01

The feasibility of sustainable nitrogen removal was investigated in a two stage biofilm configuration consisting of a MBBR (Moving Bed Biofilm Reactor) and a Deamox reactor (Biobed-EGSB). The MBBR is used for nitrification and the denitrifying ammonium oxidation (Deamox) is aimed at a nitrogen removal process in which part of the required nitrite for the typical anammox reaction originated from nitrate. Anaerobic pre-treated potato wastewater was supplied to a MBBR and Deamox reactor operated in series with a bypass flow of 30%. The MBBR showed stable nitrite production at ammonium-loading rates of 0.9-1.0 kg NH₄-N/m³ d with ammonium conversion rates of 0.80-0.85 kg NH₄-N/m³ d. The nitrogen-loading rate and conversion rate of the Deamox reactor were 1.6-1.8 and 1.6 kg N/m³ d. The maximum ammonium removal capacity in the Deamox reactor was 0.6 kg NH₄-N/m³ d. The removal efficiency of soluble total nitrogen reached 90%. The Deamox process performance was found to be negatively affected during decline of the operating temperature from 33 to 22 °C and by organic loading rates with a chemical oxygen demand (COD)/NO₂-N ratio >1.

Passive load follow analysis of the STAR-LM and STAR-H2 systems

NASA Astrophysics Data System (ADS)

Moisseytsev, Anton

A steady-state model for the calculation of temperature and pressure distributions, and heat and work balance for the STAR-LM and the STAR-H2 systems was developed. The STAR-LM system is designed for electricity production and consists of the lead cooled reactor on natural circulation and the supercritical carbon dioxide Brayton cycle. The STAR-H2 system uses the same reactor which is coupled to the hydrogen production plant, the Brayton cycle, and the water desalination plant. The Brayton cycle produces electricity for the on-site needs. Realistic modules for each system component were developed. The model also performs design calculations for the turbine and compressors for the CO2 Brayton cycle. The model was used to optimize the performance of the entire system as well as every system component. The size of each component was calculated. For the 400 MWt reactor power the STAR-LM produces 174.4 MWe (44% efficiency) and the STAR-H2 system produces 7450 kg H2/hr. The steady state model was used to conduct quasi-static passive load follow analysis. The control strategy was developed for each system; no control action on the reactor is required. As a main safety criterion, the peak cladding temperature is used. It was demonstrated that this temperature remains below the safety limit during both normal operation and load follow.

Treatment of a chocolate industry wastewater in a pilot-scale low-temperature UASB reactor operated at short hydraulic and sludge retention time.

PubMed

Esparza-Soto, M; Arzate-Archundia, O; Solís-Morelos, C; Fall, C

2013-01-01

The aim of this work was to evaluate the performance of a 244-L pilot-scale upflow anaerobic sludge blanket (UASB) reactor during the treatment of chocolate-processing industry wastewater under low-temperature conditions (18 ± 0.6 °C) for approximately 250 d. The applied organic loading rate (OLR) was varied between 4 and 7 kg/m(3)/d by varying the influent soluble chemical oxygen demand (CODsol), while keeping the hydraulic retention time constant (6.4 ± 0.3 h). The CODsol removal efficiency was low (59-78%). The measured biogas production increased from 240 ± 54 to 431 ± 61 L/d during the experiments. A significant linear correlation between the measured biogas production and removed OLR indicated that 81.69 L of biogas were produced per kg/m(3) of CODsol removed. Low average reactor volatile suspended solids (VSS) (2,700-4,800 mg/L) and high effluent VSS (177-313 mg/L) were derived in a short sludge retention time (SRT) (4.9 d). The calculated SRT was shorter than those reported in the literature, but did not affect the reactor's performance. Average sludge yield was 0.20 kg-VSS/kg-CODsol. The low-temperature anaerobic treatment was a good option for the pre-treatment of chocolate-processing industry wastewater.

THERMOCHEMICAL HEAT STORAGE FOR CONCENTRATED SOLAR POWER

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

PROJECT STAFF

2011-10-31

Thermal energy storage (TES) is an integral part of a concentrated solar power (CSP) system. It enables plant operators to generate electricity beyond on sun hours and supply power to the grid to meet peak demand. Current CSP sensible heat storage systems employ molten salts as both the heat transfer fluid and the heat storage media. These systems have an upper operating temperature limit of around 400 C. Future TES systems are expected to operate at temperatures between 600 C to 1000 C for higher thermal efficiencies which should result in lower electricity cost. To meet future operating temperature andmore » electricity cost requirements, a TES concept utilizing thermochemical cycles (TCs) based on multivalent solid oxides was proposed. The system employs a pair of reduction and oxidation (REDOX) reactions to store and release heat. In the storage step, hot air from the solar receiver is used to reduce the oxidation state of an oxide cation, e.g. Fe3+ to Fe2+. Heat energy is thus stored as chemical bonds and the oxide is charged. To discharge the stored energy, the reduced oxide is re-oxidized in air and heat is released. Air is used as both the heat transfer fluid and reactant and no storage of fluid is needed. This project investigated the engineering and economic feasibility of this proposed TES concept. The DOE storage cost and LCOE targets are $15/kWh and $0.09/kWh respectively. Sixteen pure oxide cycles were identified through thermodynamic calculations and literature information. Data showed the kinetics of re-oxidation of the various oxides to be a key barrier to implementing the proposed concept. A down selection was carried out based on operating temperature, materials costs and preliminary laboratory measurements. Cobalt oxide, manganese oxide and barium oxide were selected for developmental studies to improve their REDOX reaction kinetics. A novel approach utilizing mixed oxides to improve the REDOX kinetics of the selected oxides was proposed. It partially replaces some of the primary oxide cations with selected secondary cations. This causes a lattice charge imbalance and increases the anion vacancy density. Such vacancies enhance the ionic mass transport and lead to faster re-oxidation. Reoxidation fractions of Mn3O4 to Mn2O3 and CoO to Co3O4 were improved by up to 16 fold through the addition of a secondary oxide. However, no improvement was obtained in barium based mixed oxides. In addition to enhancing the short term re-oxidation kinetics, it was found that the use of mixed oxides also help to stabilize or even improve the TES properties after long term thermal cycling. Part of this improvement could be attributed to a reduced grain size in the mixed oxides. Based on the measurement results, manganese-iron, cobalt-aluminum and cobalt iron mixed oxides have been proposed for future engineering scale demonstration. Using the cobalt and manganese mixed oxides, we were able to demonstrate charge and discharge of the TES media in both a bench top fixed bed and a rotary kiln-moving bed reactor. Operations of the fixed bed configuration are straight forward but require a large mass flow rate and higher fluid temperature for charging. The rotary kiln makes direct solar irradiation possible and provides significantly better heat transfer, but designs to transport the TES oxide in and out of the reactor will need to be defined. The final reactor and system design will have to be based on the economics of the CSP plant. A materials compatibility study was also conducted and it identified Inconel 625 as a suitable high temperature engineering material to construct a reactor holding either cobalt or manganese mixed oxides. To assess the economics of such a CSP plant, a packed bed reactor model was established as a baseline. Measured cobalt-aluminum oxide reaction kinetics were applied to the model and the influences of bed properties and process parameters on the overall system design were investigated. The optimal TES system design was found to be a network of eight fixed bed reactors at 18.75 MWth each with charge and discharge temperatures between 1200 C and 600 C, which provides a constant output temperature of 900 C. The charge and discharge time are 8 hours each respectively. This design was integrated into a process flowsheet of a CSP plant and the system's economics were determined using AspenPlus and NREL's Solar Advisory Model. Storage cost is very sensitive to materials cost and was calculated to be based around $40/kWh for cobalt based mixed oxide. It can potentially decrease to $10/kWh based on reduced materials cost on a bulk scale. The corresponding calculated LCOE was between $0.22 and 0.30/kW-h. The high LCOE is a result of the high charging temperature required in this first design and the cost of cobalt oxide. It is expected that a moving bed reactor using manganese oxide will significantly improve the economics of the proposed concept.« less

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

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.

Impact of Radiation Hardness and Operating Temperatures of Silicon Carbide Electronics on Space Power System Mass

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.; Tew, Roy C.; Schwarze, Gene E.

1998-01-01

The effect of silicon carbide (SiC) electronics operating temperatures on Power Management and Distribution (PMAD), or Power Conditioning (PC), subsystem radiator size and mass requirements was evaluated for three power output levels (100 kW(e) , 1 MW(e), and 10 MW(e)) for near term technology ( i.e. 1500 K turbine inlet temperature) Closed Cycle Gas Turbine (CCGT) power systems with a High Temperature Gas Reactor (HTGR) heat source. The study was conducted for assumed PC radiator temperatures ranging from 370 to 845 K and for three scenarios of electrical energy to heat conversion levels which needed to be rejected to space by means of the PC radiator. In addition, during part of the study the radiation hardness of the PC electronics was varied at a fixed separation distance to estimate its effect on the mass of the instrument rated reactor shadow shield. With both the PC radiator and the conical shadow shield representing major components of the overall power system the influence of the above on total power system mass was also determined. As expected, results show that the greatest actual mass savings achieved by the use of SiC electronics occur with high capacity power systems. Moreover, raising the PC radiator temperature above 600 K yields only small additional system mass savings. The effect of increased radiation hardness on total system mass is to reduce system mass by virtue of lowering the shield mass.

Neutron radiation embrittlement studies in support of continued operation, and validation by sampling of Magnox reactor steel pressure vessels and components

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

Jones, R.B.; Bolton, C.J.

1997-02-01

Magnox steel reactor pressure vessels differ significantly from US LWR vessels in terms of the type of steel used, as well as their operating environment (dose level, exposure temperature range, and neutron spectra). The large diameter ferritic steel vessels are constructed from C-Mn steel plates and forgings joined together with manual metal and submerged-arc welds which are stress-relieved. All Magnox vessels are now at least thirty years old and their continued operation is being vigorously pursued. Vessel surveillance and other programmes are summarized which support this objective. The current understanding of the roles of matrix irradiation damage, irradiation-enhanced copper impuritymore » precipitation and intergranular embrittlement effects is described in so far as these influence the form of the embrittlement and hardening trend curves for each material. An update is given on the influence of high temperature exposure, and on the role of differing neutron spectra. Finally, the validation offered by the results of an initial vessel sampling exercise is summarized together with the objectives of a more extensive future sampling programme.« less

Draft environmental impact statement siting, construction, and operation of New Production Reactor capacity. Volume 4, Appendices D-R

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

None

1991-04-01

This Environmental Impact Statement (EIS) assesses the potential environmental impacts, both on a broad programmatic level and on a project-specific level, concerning a proposed action to provide new tritium production capacity to meet the nation`s nuclear defense requirements well into the 21st century. A capacity equivalent to that of about a 3,000-megawatt (thermal) heavy-water reactor was assumed as a reference basis for analysis in this EIS; this is the approximate capacity of the existing production reactors at DOE`s Savannah River Site near Aiken, South Carolina. The EIS programmatic alternatives address Departmental decisions to be made on whether to build newmore » production facilities, whether to build one or more complexes, what size production capacity to provide, and when to provide this capacity. Project-specific impacts for siting, constructing, and operating new production reactor capacity are assessed for three alternative sites: the Hanford Site near Richland, Washington; the Idaho National Engineering Laboratory near Idaho Falls, Idaho; and the Savannah River Site. For each site, the impacts of three reactor technologies (and supporting facilities) are assessed: a heavy-water reactor, a light-water reactor, and a modular high-temperature gas-cooled reactor. Impacts of the no-action alternative also are assessed. The EIS evaluates impacts related to air quality; noise levels; surface water, groundwater, and wetlands; land use; recreation; visual environment; biotic resources; historical, archaeological, and cultural resources; socioeconomics; transportation; waste management; and human health and safety. The EIS describes in detail the potential radioactive releases from new production reactors and support facilities and assesses the potential doses to workers and the general public. This volume contains 15 appendices.« less

Heat-transfer analysis of double-pipe heat exchangers for indirect-cycle SCW NPP

NASA Astrophysics Data System (ADS)

Thind, Harwinder

SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. SuperCritical Water (SCW) Nuclear Power Plants (NPPs) are expected to have much higher operating parameters compared to current NPPs, i.e., pressure of about 25 MPa and outlet temperature up to 625 °C. This study presents the heat transfer analysis of an intermediate Heat exchanger (HX) design for indirect-cycle concepts of Pressure-Tube (PT) and Pressure-Vessel (PV) SCWRs. Thermodynamic configurations with an intermediate HX gives a possibility to have a single-reheat option for PT and PV SCWRs without introducing steam-reheat channels into a reactor. Similar to the current CANDU and Pressurized Water Reactor (PWR) NPPs, steam generators separate the primary loop from the secondary loop. In this way, the primary loop can be completely enclosed in a reactor containment building. This study analyzes the heat transfer from a SCW primary (reactor) loop to a SCW and Super-Heated Steam (SHS) secondary (turbine) loop using a double-pipe intermediate HX. The numerical model is developed with MATLAB and NIST REFPROP software. Water from the primary loop flows through the inner pipe, and water from the secondary loop flows through the annulus in the counter direction of the double-pipe HX. The analysis on the double-pipe HX shows temperature and profiles of thermophysical properties along the heated length of the HX. It was found that the pseudocritical region has a significant effect on the temperature profiles and heat-transfer area of the HX. An analysis shows the effect of variation in pressure, temperature, mass flow rate, and pipe size on the pseudocritical region and the heat-transfer area of the HX. The results from the numerical model can be used to optimize the heat-transfer area of the HX. The higher pressure difference on the hot side and higher temperature difference between the hot and cold sides reduces the pseudocritical-region length, thus decreases the heat-transfer surface area of the HX.

Seasonal-related effects on ammonium removal in activated carbon filter biologically enhanced by heterotrophic nitrifying bacteria for drinking water treatment.

PubMed

Qin, Wen; Li, Wei-Guang; Gong, Xu-Jin; Huang, Xiao-Fei; Fan, Wen-Biao; Zhang, Duoying; Yao, Peng; Wang, Xiao-Ju; Song, Yang

2017-08-01

To determine the potential effects of seasonal changes on water temperature and water quality upon removal of ammonium and organic carbon pollutants and to characterize the variations in microbial characteristics, a pilot-scale activated carbon filter biologically enhanced with heterotrophic nitrifying bacteria was investigated for 528 days. The results show that 69.2 ± 28.6% of ammonium and 23.1 ± 11.6% of the dissolved organic carbon were removed by the biologically enhanced activated carbon (BEAC) reactor. It is shown that higher biodegradable dissolved organic carbon enhances ammonium removal, even at low temperatures. The C/N ratio consumed by the BEAC reactor reached a steady value (i.e., 3.3) after 2 months of operation. Despite seasonal fluctuations and competition of the indigenous community, the heterotrophic nitrifying bacteria (Acinetobacter sp. HRBLi 16 and Acinetobacter harbinensis strain HITLi 7) remained relatively stable. The amount of carbon source was the most significant environmental parameter and dramatically affected the microbial community compositions in the BEAC reactor. The present study provides new insights into the application of a BEAC reactor for ammonium removal from drinking water, resisting strong seasonal changes.

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

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.

Production of hydrogen by direct gasification of coal with steam using nuclear heat

NASA Technical Reports Server (NTRS)

1975-01-01

Problems related to: (1) high helium outlet temperature of the reactor, and (2) gas generator design used in hydrogen production are studied. Special attention was given to the use of Oklahoma coal in the gasification process. Plant performance, operation, and environmental considerations are covered.

Pressure polymerization of polyester

DOEpatents

Maurer, Charles J.; Shaw, Gordon; Smith, Vicky S.; Buelow, Steven J.; Tumas, William; Contreras, Veronica; Martinez, Ronald J.

2000-08-29

A process is disclosed for the preparation of a polyester polymer or polyester copolymer under superatmospheric pressure conditions in a pipe or tubular reaction under turbannular flow conditions. Reaction material having a glycol equivalents to carboxylic acid equivalents mole ratio of from 1.0:1 to 1.2:1, together with a superatmospheric dense gaseous medium are fed co-currently to the reactor. Dicarboxylic acid and/or diol raw materials may be injected into any of the reaction zones in the process during operation to achieve the overall desired mole ratio balance. The process operates at temperatures of from about 220.degree. C. to about 320.degree. C., with turbannular flow achieved before the polymer product and gas exit the reactor process. The pressure in the reaction zones can be in the range from 15 psia to 2500 psia. A polymer product having a DP of a greater than 40, more preferably at least about 70, is achieved by the transfer of water from the reacting material polymer melt to the gaseous medium in the reactor.

Carbide fuels for nuclear thermal propulsion

NASA Astrophysics Data System (ADS)

Matthews, R. B.; Blair, H. T.; Chidester, K. M.; Davidson, K. V.; Stark, W. E.; Storms, E. K.

1991-09-01

A renewed interest in manned exploration of space has revitalized interest in the potential for advancing nuclear rocket technology developed during the 1960's. Carbide fuel performance, melting point, stability, fabricability and compatibility are key technology issues for advanced Nuclear Thermal Propulsion reactors. The Rover fuels development ended with proven carbide fuel forms with demonstrated operating temperatures up to 2700 K for over 100 minutes. The next generation of nuclear rockets will start where the Rover technology ended, but with a more rigorous set of operating requirements including operating lifetime to 10 hours, operating temperatures greater that 3000 K, low fission product release, and compatibility. A brief overview of Rover/NERVA carbide fuel development is presented. A new fuel form with the highest potential combination of operating temperature and lifetime is proposed that consists of a coated uranium carbide fuel sphere with built-in porosity to contain fission products. The particles are dispersed in a fiber reinforced ZrC matrix to increase thermal shock resistance.

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

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.

An experimental study of catalytic and non-catalytic reaction in heat recirculating reactors and applications to power generation

NASA Astrophysics Data System (ADS)

Ahn, Jeongmin

An experimental study of the performance of a Swiss roll heat exchanger and reactor was conducted, with emphasis on the extinction limits and comparison of results with and without Pt catalyst. At Re<40, the catalyst was required to sustain reaction; with the catalyst self-sustaining reaction could be obtained at Re less than 1. Both lean and rich extinction limits were extended with the catalyst, though rich limits were extended much further. At low Re, the lean extinction limit was rich of stoichiometric and rich limit had equivalence ratios 80 in some cases. Non-catalytic reaction generally occurred in a flameless mode near the center of the reactor. With or without catalyst, for sufficiently robust conditions, a visible flame would propagate out of the center, but this flame could only be re-centered with catalyst. Gas chromatography indicated that at low Re, CO and non-C3 H8 hydrocarbons did not form. For higher Re, catalytic limits were slightly broader but had much lower limit temperatures. At sufficiently high Re, catalytic and gas-phase limits merged. Experiments with titanium Swiss rolls have demonstrated reducing wall thermal conductivity and thickness leads to lower heat losses and therefore increases operating temperatures and extends flammability limits. By use of Pt catalysts, reaction of propane-air mixtures at temperatures 54°C was sustained. Such low temperatures suggest that polymers may be employed as a reactor material. A polyimide reactor was built and survived prolonged testing at temperatures up to 500°C. Polymer reactors may prove more practical for microscale devices due to their lower thermal conductivity and ease of manufacturing. Since the ultimate goal of current efforts is to develop combustion driven power generation devices at MEMS like scales, a thermally self-sustaining miniature power generation device was developed utilizing a single-chamber solid-oxide-fuel-cell (SOFC) placed in a Swiss roll. With the single-chamber design, fuel/oxygen crossover due to cracking of seals via thermal cycling is irrelevant and coking on the anode is practically eliminated. SOFC power densities up to 420mW/cm2 were observed at low Re. These results suggest that single-chamber SOFC's integrated with heat-recirculating reactors may be a viable approach for small-scale power generation devices.

Autonomous sensor particle for parameter tracking in large vessels

NASA Astrophysics Data System (ADS)

Thiele, Sebastian; Da Silva, Marco Jose; Hampel, Uwe

2010-08-01

A self-powered and neutrally buoyant sensor particle has been developed for the long-term measurement of spatially distributed process parameters in the chemically harsh environments of large vessels. One intended application is the measurement of flow parameters in stirred fermentation biogas reactors. The prototype sensor particle is a robust and neutrally buoyant capsule, which allows free movement with the flow. It contains measurement devices that log the temperature, absolute pressure (immersion depth) and 3D-acceleration data. A careful calibration including an uncertainty analysis has been performed. Furthermore, autonomous operation of the developed prototype was successfully proven in a flow experiment in a stirred reactor model. It showed that the sensor particle is feasible for future application in fermentation reactors and other industrial processes.

Multiphysics Modeling for Dimensional Analysis of a Self-Heated Molten Regolith Electrolysis Reactor for Oxygen and Metals Production on the Moon and Mars

NASA Technical Reports Server (NTRS)

Dominguez, Jesus; Sibille, Laurent

2010-01-01

The technology of direct electrolysis of molten lunar regolith to produce oxygen and molten metal alloys has progressed greatly in the last few years. The development of long-lasting inert anodes and cathode designs as well as techniques for the removal of molten products from the reactor has been demonstrated. The containment of chemically aggressive oxide and metal melts is very difficult at the operating temperatures ca. 1600 C. Containing the molten oxides in a regolith shell can solve this technical issue and can be achieved by designing a self-heating reactor in which the electrolytic currents generate enough Joule heat to create a molten bath.

FUEL ELEMENT FOR NEUTRONIC REACTORS

DOEpatents

Evans, T.C.; Beasley, E.G.

1961-01-17

A fuel element for neutronic reactors, particularly the gas-cooled type of reactor, is described. The element comprises a fuel-bearing plate rolled to form a cylinder having a spiral passageway passing from its periphery to its center. In operation a coolant is admitted to the passageway at the periphery of the element, is passed through the spiral passageway, and emerges into a central channel defined by the inner turn of the rolled plate. The advantage of the element is that the fully heated coolant (i.e., coolant emerging into the central channel) is separated and thus insulated from the periphery of the element, which may be in contact with a low-temperature moderator, by the intermediate turns of the spiral fuel element.

METHOD AND APPARATUS FOR IMPROVING PERFORMANCE OF A FAST REACTOR

DOEpatents

Koch, L.J.

1959-01-20

A specific arrangement of the fertile material and fissionable material in the active portion of a fast reactor to achieve improvement in performance and to effectively lower the operating temperatures in the center of the reactor is described. According to this invention a group of fuel elements containing fissionable material are assembled to form a hollow fuel core. Elements containing a fertile material, such as depleted uranium, are inserted into the interior of the fuel core to form a central blanket. Additional elemenis of fertile material are arranged about the fuel core to form outer blankets which in tunn are surrounded by a reflector. This arrangement of fuel core and blankets results in substantial flattening of the flux pattern.

Deployment history and design considerations for space reactor power systems

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.

2009-05-01

The history of the deployment of nuclear reactors in Earth orbits is reviewed with emphases on lessons learned and the operation and safety experiences. The former Soviet Union's "BUK" power systems, with SiGe thermoelectric conversion and fast neutron energy spectrum reactors, powered a total of 31 Radar Ocean Reconnaissance Satellites (RORSATs) from 1970 to 1988 in 260 km orbit. Two of the former Soviet Union's TOPAZ reactors, with in-core thermionic conversion and epithermal neutron energy spectrum, powered two Cosmos missions launched in 1987 in ˜800 km orbit. The US' SNAP-10A system, with SiGe energy conversion and a thermal neutron energy spectrum reactor, was launched in 1965 in 1300 km orbit. The three reactor systems used liquid NaK-78 coolant, stainless steel structure and highly enriched uranium fuel (90-96 wt%) and operated at a reactor exit temperature of 833-973 K. The BUK reactors used U-Mo fuel rods, TOPAZ used UO 2 fuel rods and four ZrH moderator disks, and the SNAP-10A used moderated U-ZrH fuel rods. These low power space reactor systems were designed for short missions (˜0.5 kW e and ˜1 year for SNAP-10A, <3.0 kW e and <6 months for BUK, and ˜5.5 kW e and up to 1 year for TOPAZ). The deactivated BUK reactors at the end of mission, which varied in duration from a few hours to ˜4.5 months, were boosted into ˜800 km storage orbit with a decay life of more than 600 year. The ejection of the last 16 BUK reactor fuel cores caused significant contamination of Earth orbits with NaK droplets that varied in sizes from a few microns to 5 cm. Power systems to enhance or enable future interplanetary exploration, in-situ resources utilization on Mars and the Moon, and civilian missions in 1000-3000 km orbits would generate significantly more power of 10's to 100's kW e for 5-10 years, or even longer. A number of design options to enhance the operation reliability and safety of these high power space reactor power systems are presented and discussed.

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

Testing of Liquid Metal Components for Nuclear Surface Power Systems

NASA Technical Reports Server (NTRS)

Polzin, K. A.; Pearson, J. B.; Godfroy, T. J.; Schoenfeld, M.; Webster, K.; Briggs, M. H.; Geng, S. M.; Adkins, H. E.; Werner, J. E.

2010-01-01

The capability to perform testing at both the module/component level and in near prototypic reactor configurations using a non-nuclear test methodology allowed for evaluation of two components critical to the development of a potential nuclear fission power system for the lunar surface. A pair of 1 kW Stirling power convertors, similar to the type that would be used in a reactor system to convert heat to electricity, were integrated into a reactor simulator system to determine their performance using pumped NaK as the hot side working fluid. The performance in the pumped-NaK system met or exceed the baseline performance measurements where the converters were electrically heated. At the maximum hot-side temperature of 550 C the maximum output power was 2375 watts. A specially-designed test apparatus was fabricated and used to quantify the performance of an annular linear induction pump that is similar to the type that could be used to circulate liquid metal through the core of a space reactor system. The errors on the measurements were generally much smaller than the magnitude of the measurements, permitting accurate performance evaluation over a wide range of operating conditions. The pump produced flow rates spanning roughly 0.16 to 5.7 l/s (2.5 to 90 GPM), and delta p levels from less than 1 kPa to 90 kPa (greater than 0.145 psi to roughly 13 psi). At the nominal FSP system operating temperature of 525 C the maximum efficiency was just over 4%.

Experimental investigation of a new method for advanced fast reactor shutdown cooling

NASA Astrophysics Data System (ADS)

Pakholkov, V. V.; Kandaurov, A. A.; Potseluev, A. I.; Rogozhkin, S. A.; Sergeev, D. A.; Troitskaya, Yu. I.; Shepelev, S. F.

2017-07-01

We consider a new method for fast reactor shutdown cooling using a decay heat removal system (DHRS) with a check valve. In this method, a coolant from the decay heat exchanger (DHX) immersed into the reactor upper plenum is supplied to the high-pressure plenum and, then, inside the fuel subassemblies (SAs). A check valve installed at the DHX outlet opens by the force of gravity after primary pumps (PP-1) are shut down. Experimental studies of the new and alternative methods of shutdown cooling were performed at the TISEY test facility at OKBM. The velocity fields in the upper plenum of the reactor model were obtained using the optical particle image velocimetry developed at the Institute of Applied Physics (Russian Academy of Sciences). The study considers the process of development of natural circulation in the reactor and the DHRS models and the corresponding evolution of the temperature and velocity fields. A considerable influence of the valve position in the displacer of the primary pump on the natural circulation of water in the reactor through the DHX was discovered (in some modes, circulation reversal through the DHX was obtained). Alternative DHRS designs without a shell at the DHX outlet with open and closed check valve are also studied. For an open check valve, in spite of the absence of a shell, part of the flow is supplied through the DHX pipeline and then inside the SA simulators. When simulating power modes of the reactor operation, temperature stratification of the liquid was observed, which increased in the cooling mode via the DHRS. These data qualitatively agree with the results of tests at BN-600 and BN-800 reactors.

Process for coal liquefaction in staged dissolvers

DOEpatents

Roberts, George W.; Givens, Edwin N.; Skinner, Ronald W.

1983-01-01

There is described an improved liquefaction process by which coal is converted to a low ash and low sulfur carbonaceous material that can be used as a fuel in an environmentally acceptable manner without costly gas scrubbing equipment. In the process, coal is slurried with a pasting oil, passed through a preheater and at least two dissolvers in series in the presence of hydrogen-rich gases at elevated temperatures and pressures. Solids, including mineral ash and unconverted coal macerals, are separated from the condensed reactor effluent. In accordance with the improved process, the first dissolver is operated at a higher temperature than the second dissolver. This temperature sequence produces improved product selectivity and permits the incorporation of sufficient hydrogen in the solvent for adequate recycle operations.

New WC-Cu thermal barriers for fusion applications: High temperature mechanical behaviour

NASA Astrophysics Data System (ADS)

Tejado, E.; Dias, M.; Correia, J. B.; Palacios, T.; Carvalho, P. A.; Alves, E.; Pastor, J. Y.

2018-01-01

The combination of tungsten carbide and copper as a thermal barrier could effectively reduce the thermal mismatch between tungsten and copper alloy, which are proposed as base armour and heat sink, respectively, in the divertor of future fusion reactors. Furthermore, since the optimum operating temperature windows for these divertor materials do not overlap, a compatible thermal barrier interlayer between them is required to guarantee a smooth thermal transition, which in addition may mitigate radiation damage. The aim of this work is to study the thermo-mechanical properties of WC-Cu cermets fabricated by hot pressing. Focus is placed on the temperature effect and composition dependence, as the volume fraction of copper varies from 25 to 50 and 75 vol%. To explore this behaviour, fracture experiments are performed within a temperature range from room temperature to 800 °C under vacuum. In addition, elastic modulus and thermal expansion coefficient are estimated from these tests. Results reveal a strong dependence of the performance on temperature and on the volume fraction of copper and, surprisingly, a slight percent of Cu (25 vol%) can effectively reduce the large difference in thermal expansion between tungsten and copper alloy, which is a critical point for in service applications. The thermal performance of these materials, together with their mechanical properties could indeed reduce the heat transfer from the PFM to the underlying element while supporting the high thermal stresses of the joint. Thus, the presence of these cermets could allow the reactor to operate above the ductile to brittle transition temperature of tungsten, without compromising the underlying materials.

Surface Anchoring of Nematic Phase on Carbon Nanotubes: Nanostructure of Ultra-High Temperature Materials

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

Ogale, Amod A

2012-04-27

Nuclear energy is a dependable and economical source of electricity. Because fuel supply sources are available domestically, nuclear energy can be a strong domestic industry that can reduce dependence on foreign energy sources. Commercial nuclear power plants have extensive security measures to protect the facility from intruders [1]. However, additional research efforts are needed to increase the inherent process safety of nuclear energy plants to protect the public in the event of a reactor malfunction. The next generation nuclear plant (NGNP) is envisioned to utilize a very high temperature reactor (VHTR) design with an operating temperature of 650-1000°C [2]. Onemore » of the most important safety design requirements for this reactor is that it must be inherently safe, i.e., the reactor must shut down safely in the event that the coolant flow is interrupted [2]. This next-generation Gen IV reactor must operate in an inherently safe mode where the off-normal temperatures may reach 1500°C due to coolant-flow interruption. Metallic alloys used currently in reactor internals will melt at such temperatures. Structural materials that will not melt at such ultra-high temperatures are carbon/graphtic fibers and carbon-matrix composites. Graphite does not have a measurable melting point; it is known to sublime starting about 3300°C. However, neutron radiation-damage effects on carbon fibers are poorly understood. Therefore, the goal of this project is to obtain a fundamental understanding of the role of nanotexture on the properties of resulting carbon fibers and their neutron-damage characteristics. Although polygranular graphite has been used in nuclear environment for almost fifty years, it is not suitable for structural applications because it do not possess adequate strength, stiffness, or toughness that is required of structural components such as reaction control-rods, upper plenum shroud, and lower core-support plate [2,3]. For structural purposes, composites consisting of strong carbon fibers embedded in a carbon matrix are needed. Such carbon/carbon (C/C) composites have been used in aerospace industry to produce missile nose cones, space shuttle leading edge, and aircraft brake-pads. However, radiation-tolerance of such materials is not adequately known because only limited radiation studies have been performed on C/C composites, which suggest that pitch-based carbon fibers have better dimensional stability than that of polyacrylonitrile (PAN) based fibers [4]. The thermodynamically-stable state of graphitic crystalline packing of carbon atoms derived from mesophase pitch leads to a greater stability during neutron irradiation [5]. The specific objectives of this project were: (i) to generating novel carbonaceous nanostructures, (ii) measure extent of graphitic crystallinity and the extent of anisotropy, and (iii) collaborate with the Carbon Materials group at Oak Ridge National Lab to have neutron irradiation studies and post-irradiation examinations conducted on the carbon fibers produced in this research project.« less

Multiplexed chemostat system for quantification of biodiversity and ecosystem functioning in anaerobic digestion

PubMed Central

Plouchart, Diane; Guizard, Guillaume; Latrille, Eric

2018-01-01

Continuous cultures in chemostats have proven their value in microbiology, microbial ecology, systems biology and bioprocess engineering, among others. In these systems, microbial growth and ecosystem performance can be quantified under stable and defined environmental conditions. This is essential when linking microbial diversity to ecosystem function. Here, a new system to test this link in anaerobic, methanogenic microbial communities is introduced. Rigorously replicated experiments or a suitable experimental design typically require operating several chemostats in parallel. However, this is labor intensive, especially when measuring biogas production. Commercial solutions for multiplying reactors performing continuous anaerobic digestion exist but are expensive and use comparably large reactor volumes, requiring the preparation of substantial amounts of media. Here, a flexible system of Lab-scale Automated and Multiplexed Anaerobic Chemostat system (LAMACs) with a working volume of 200 mL is introduced. Sterile feeding, biomass wasting and pressure monitoring are automated. One module containing six reactors fits the typical dimensions of a lab bench. Thanks to automation, time required for reactor operation and maintenance are reduced compared to traditional lab-scale systems. Several modules can be used together, and so far the parallel operation of 30 reactors was demonstrated. The chemostats are autoclavable. Parameters like reactor volume, flow rates and operating temperature can be freely set. The robustness of the system was tested in a two-month long experiment in which three inocula in four replicates, i.e., twelve continuous digesters were monitored. Statistically significant differences in the biogas production between inocula were observed. In anaerobic digestion, biogas production and consequently pressure development in a closed environment is a proxy for ecosystem performance. The precision of the pressure measurement is thus crucial. The measured maximum and minimum rates of gas production could be determined at the same precision. The LAMACs is a tool that enables us to put in practice the often-demanded need for replication and rigorous testing in microbial ecology as well as bioprocess engineering. PMID:29518106

Impact investigation of reactor fuel operating parameters on reactivity for use in burnup credit applications

NASA Astrophysics Data System (ADS)

Sloma, Tanya Noel

When representing the behavior of commercial spent nuclear fuel (SNF), credit is sought for the reduced reactivity associated with the net depletion of fissile isotopes and the creation of neutron-absorbing isotopes, a process that begins when a commercial nuclear reactor is first operated at power. Burnup credit accounts for the reduced reactivity potential of a fuel assembly and varies with the fuel burnup, cooling time, and the initial enrichment of fissile material in the fuel. With regard to long-term SNF disposal and transportation, tremendous benefits, such as increased capacity, flexibility of design and system operations, and reduced overall costs, provide an incentive to seek burnup credit for criticality safety evaluations. The Nuclear Regulatory Commission issued Interim Staff Guidance 8, Revision 2 in 2002, endorsing burnup credit of actinide composition changes only; credit due to actinides encompasses approximately 30% of exiting pressurized water reactor SNF inventory and could potentially be increased to 90% if fission product credit were accepted. However, one significant issue for utilizing full burnup credit, compensating for actinide and fission product composition changes, is establishing a set of depletion parameters that produce an adequately conservative representation of the fuel's isotopic inventory. Depletion parameters can have a significant effect on the isotopic inventory of the fuel, and thus the residual reactivity. This research seeks to quantify the reactivity impact on a system from dominant depletion parameters (i.e., fuel temperature, moderator density, burnable poison rod, burnable poison rod history, and soluble boron concentration). Bounding depletion parameters were developed by statistical evaluation of a database containing reactor operating histories. The database was generated from summary reports of commercial reactor criticality data. Through depletion calculations, utilizing the SCALE 6 code package, several light water reactor assembly designs and in-core locations are analyzed in establishing a combination of depletion parameters that conservatively represent the fuel's isotopic inventory as an initiative to take credit for fuel burnup in criticality safety evaluations for transportation and storage of SNF.