Low cost solar array project silicon materials task. Development of a process for high capacity arc heater production of silicon for solar arrays

NASA Technical Reports Server (NTRS)

Fey, M. G.

1981-01-01

The experimental verification system for the production of silicon via the arc heater-sodium reduction of SiCl4 was designed, fabricated, installed, and operated. Each of the attendant subsystems was checked out and operated to insure performance requirements. These subsystems included: the arc heaters/reactor, cooling water system, gas system, power system, Control & Instrumentation system, Na injection system, SiCl4 injection system, effluent disposal system and gas burnoff system. Prior to introducing the reactants (Na and SiCl4) to the arc heater/reactor, a series of gas only-power tests was conducted to establish the operating parameters of the three arc heaters of the system. Following the successful completion of the gas only-power tests and the readiness tests of the sodium and SiCl4 injection systems, a shakedown test of the complete experimental verification system was conducted.

Reference Reactor Module for the Affordable Fission Surface Power System

NASA Astrophysics Data System (ADS)

Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Amiri, Benjamin W.; Marcille, Thomas F.

2008-01-01

Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The requirements of many surface power applications allow the consideration of systems with much less development risk than most other space reactor applications, because of modest power (10s of kWe) and no driving need for minimal mass (allowing temperatures <1000 K). The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. This paper describes the reference AFSPS reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based, UO2-fueled, liquid metal-cooled fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. One of the important ``affordability'' attributes is that the concept has been designed to minimize both the technical and programmatic safety risk.

Accurate evaluation for the biofilm-activated sludge reactor using graphical techniques

NASA Astrophysics Data System (ADS)

Fouad, Moharram; Bhargava, Renu

2018-05-01

A complete graphical solution is obtained for the completely mixed biofilm-activated sludge reactor (hybrid reactor). The solution consists of a series of curves deduced from the principal equations of the hybrid system after converting them in dimensionless form. The curves estimate the basic parameters of the hybrid system such as suspended biomass concentration, sludge residence time, wasted mass of sludge, and food to biomass ratio. All of these parameters can be expressed as functions of hydraulic retention time, influent substrate concentration, substrate concentration in the bulk, stagnant liquid layer thickness, and the minimum substrate concentration which can maintain the biofilm growth in addition to the basic kinetics of the activated sludge process in which all these variables are expressed in a dimensionless form. Compared to other solutions of such system these curves are simple, easy to use, and provide an accurate tool for analyzing such system based on fundamental principles. Further, these curves may be used as a quick tool to get the effect of variables change on the other parameters and the whole system.

Research and Development Roadmaps for Liquid Metal Cooled Fast Reactors

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

Kim, T. K.; Grandy, C.; Natesan, K.

The United States Department of Energy (DOE) commissioned the development of technology roadmaps for advanced (non-light water reactor) reactor concepts to help focus research and development funding over the next five years. The roadmaps show the research and development needed to support demonstration of an advanced (non-LWR) concept by the early 2030s, consistent with DOE’s Vision and Strategy for the Development and Deployment of Advanced Reactors. The intent is only to convey the technical steps that would be required to achieve such a goal; the means by which DOE will determine whether to invest in specific tasks will be treatedmore » separately. The starting point for the roadmaps is the Technical Readiness Assessment performed as part of an Advanced Test and Demonstration Reactor study released in 2016. The roadmaps were developed based upon a review of technical reports and vendor literature summarizing the technical maturity of each concept and the outstanding research and development needs. Critical path tasks for specific systems were highlighted on the basis of time and resources needed to complete the tasks and the importance of the system to the performance of the reactor concept. The roadmaps are generic, i.e. not specific to a particular vendor’s design but vendor design information may have been used as representative of the concept family. In the event that both near-term and more advanced versions of a concept are being developed, either a single roadmap with multiple branches or separate roadmaps for each version were developed. In each case, roadmaps point to a demonstration reactor (engineering or commercial) and show the activities that must be completed in parallel to support that demonstration in the 2030-2035 window. This report provides the roadmaps for two fast reactor concepts, the Sodium-cooled Fast Reactor (SFR) and the Lead-cooled Fast Reactor (LFR). The SFR technology is mature enough for commercial demonstration by the early 2030s, and the remaining critical paths and R&D needs are generally related to the completion of qualification of fuel and structural materials, validation of reactor design codes and methods, and support of the licensing frameworks. The LFR’s technology is instead less-mature compared to the SFR’s, and will be at the engineering demonstration stage by the early 2030s. Key LFR technology development activities will focus on resolving remaining design challenges and demonstrating the viability of systems and components in the integral system, which will be done in parallel with addressing the gaps shared with SFR technology. The approach and timeline presented here assume that, for the first module demonstration, vendors would pursue a two-step licensing process based on 10CFR Part 50.« less

Series-Bosch Technology for Oxygen Recovery During Lunar or Martian Surface Missions

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, J. Matthew; Rabenberg, Ellen; Stanley, Christine M.; Edmunson, Jennifer; Alleman, James E.; Chen, Kevin; Dumez, Sam

2014-01-01

Long-duration surface missions to the Moon or Mars will require life support systems that maximize resource recovery to minimize resupply from Earth. To address this need, NASA previously proposed a Series-Bosch (S-Bosch) oxygen recovery system, based on the Bosch process, which can theoretically recover 100% of the oxygen from metabolic carbon dioxide. Bosch processes have the added benefits of the potential to recover oxygen from atmospheric carbon dioxide and the use of regolith materials as catalysts, thereby eliminating the need for catalyst resupply from Earth. In 2012, NASA completed an initial design for an S-Bosch development test stand that incorporates two catalytic reactors in series including a Reverse Water-Gas Shift (RWGS) Reactor and a Carbon Formation Reactor (CFR). In 2013, fabrication of system components, with the exception of a CFR, and assembly of the test stand was initiated. Stand-alone testing of the RWGS reactor was completed to compare performance with design models. Continued testing of Lunar and Martian regolith simulants provided sufficient data to design a CFR intended to utilize these materials as catalysts. Finally, a study was conducted to explore the possibility of producing bricks from spent regolith catalysts. The results of initial demonstration testing of the RWGS reactor, results of continued catalyst performance testing of regolith simulants, and results of brick material properties testing are reported. Additionally, design considerations for a regolith-based CFR are discussed.

Series-Bosch Technology for Oxygen Recovery During Lunar or Martian Surface Missions

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, James M.; Stanley, Christine; Edmunson, Jennifer; Dumez, Samuel; Chen, Kevin; Alleman, James E.

2014-01-01

Long-duration surface missions to the Moon or Mars will require life support systems that maximize resource recovery to minimize resupply from Earth. To address this need, NASA previously proposed a Series-Bosch (S-Bosch) oxygen recovery system, based on the Bosch process, which can theoretically recover 100% of the oxygen from metabolic carbon dioxide. Bosch processes have the added benefits of the potential to recover oxygen from atmospheric carbon dioxide and the use of regolith materials as catalysts, thereby eliminating the need for catalyst resupply from Earth. In 2012, NASA completed an initial design for an S-Bosch development test stand that incorporates two catalytic reactors in series including a Reverse Water-Gas Shift (RWGS) Reactor and a Carbon Formation Reactor (CFR). In 2013, fabrication of system components, with the exception of a CFR, and assembly of the test stand was initiated. Stand-alone testing of the RWGS reactor was completed to compare performance with design models. Continued testing of Lunar and Martian regolith simulants provided sufficient data to design a CFR intended to utilize these materials as catalysts. Finally, a study was conducted to explore the possibility of producing bricks from spend regolith catalysts. The results of initial demonstration testing of the RWGS reactor, results of continued catalyst performance testing of regolith simulants, and results of brick material properties testing are reported. Additionally, design considerations for a regolith-based CFR are discussed.

Recycled and virgin plastic carriers in hybrid reactors for wastewater treatment.

PubMed

Paul, Etienne; Wolff, Delmira Beatriz; Ochoa, Juan Carlos; da Costa, Rejane Helena Ribeiro

2007-07-01

The reduction of organic and nitrogen pollution of wastewater was investigated in two hybrid reactors and compared with the reduction obtained by using a conventional activated sludge reactor (ASR) run as a control. Both HR-1 and HR-2 were activated sludge systems where a low-density carrier, P1 (polyethylene) for HR-1 and P2 (recycled plastics) for HR-2, was added. Firstly, the three reactors were operated at 10 days Suspended Solid Retention Time (SRT(SS)), leading to a complete nitrification. Secondly, the SRT(SS) for each reactor was lowered to 3 days. Nitrification was lost for the ASR but remained complete for HR's. Respirometric techniques were used to measure fixed or suspended biomass activities for heterotrophic and autotrophic biomass. More than 90% of the autotrophic activity was found on the supports whatever the SRT(SS) used. The results may underline the role of the carrier geometry or surface characteristics on the autotrophic/heterotrophic microorganism distribution.

Advanced Reactor PSA Methodologies for System Reliability Analysis and Source Term Assessment

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

Grabaskas, D.; Brunett, A.; Passerini, S.

Beginning in 2015, a project was initiated to update and modernize the probabilistic safety assessment (PSA) of the GE-Hitachi PRISM sodium fast reactor. This project is a collaboration between GE-Hitachi and Argonne National Laboratory (Argonne), and funded in part by the U.S. Department of Energy. Specifically, the role of Argonne is to assess the reliability of passive safety systems, complete a mechanistic source term calculation, and provide component reliability estimates. The assessment of passive system reliability focused on the performance of the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedback mechanisms of the metal fuel core. Themore » mechanistic source term assessment attempted to provide a sequence specific source term evaluation to quantify offsite consequences. Lastly, the reliability assessment focused on components specific to the sodium fast reactor, including electromagnetic pumps, intermediate heat exchangers, the steam generator, and sodium valves and piping.« less

Application of biocatalysts to Space Station ECLSS and PMMS water reclamation

NASA Technical Reports Server (NTRS)

Jolly, Clifford D.; Bagdigian, Robert M.

1989-01-01

Immobilized enzyme reactors have been developed and tested for potential water reclamation applications in the Space Station Freedom Environmental Control and Life Support System (ECLSS) and Process Materials Management System (PMMS). The reactors convert low molecular weight organic contaminants found in ECLSS and PMMS wastewaters to compounds that are more efficiently removed by existing technologies. Demonstration of the technology was successfully achieved with two model reactors. A packed bed reactor containing immobilized urease was found to catalyze the complete decomposition of urea to by-products that were subsequently removed using conventional ion exchange results. A second reactor containing immobilized alcohol oxidase showed promising results relative to its ability to convert methanol and ethanol to the corresponding aldehydes for subsequent removal. Preliminary assessments of the application of biocatalysts to ECLSS and PMMS water reclamation sytems are presented.

Dynamic Modeling and Control of Nuclear Reactors Coupled to Closed-Loop Brayton Cycle Systems using SIMULINK{sup TM}

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

Wright, Steven A.; Sanchez, Travis

2005-02-06

The operation of space reactors for both in-space and planetary operations will require unprecedented levels of autonomy and control. Development of these autonomous control systems will require dynamic system models, effective control methodologies, and autonomous control logic. This paper briefly describes the results of reactor, power-conversion, and control models that are implemented in SIMULINK{sup TM} (Simulink, 2004). SIMULINK{sup TM} is a development environment packaged with MatLab{sup TM} (MatLab, 2004) that allows the creation of dynamic state flow models. Simulation modules for liquid metal, gas cooled reactors, and electrically heated systems have been developed, as have modules for dynamic power-conversion componentsmore » such as, ducting, heat exchangers, turbines, compressors, permanent magnet alternators, and load resistors. Various control modules for the reactor and the power-conversion shaft speed have also been developed and simulated. The modules are compiled into libraries and can be easily connected in different ways to explore the operational space of a number of potential reactor, power-conversion system configurations, and control approaches. The modularity and variability of these SIMULINK{sup TM} models provides a way to simulate a variety of complete power generation systems. To date, both Liquid Metal Reactors (LMR), Gas Cooled Reactors (GCR), and electric heaters that are coupled to gas-dynamics systems and thermoelectric systems have been simulated and are used to understand the behavior of these systems. Current efforts are focused on improving the fidelity of the existing SIMULINK{sup TM} modules, extending them to include isotopic heaters, heat pipes, Stirling engines, and on developing state flow logic to provide intelligent autonomy. The simulation code is called RPC-SIM (Reactor Power and Control-Simulator)« less

Final report of the decontamination and decommissioning of the BORAX-V facility turbine building

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

Arave, A.E.; Rodman, G.R.

1992-12-01

The Boiling Water Reactor Experiment (BORAX)-V Facility Turbine Building Decontamination and Decommissioning (D&D) Project is described in this report. The BORAX series of five National Reactor Testing Station (NRTS) reactors pioneered intensive work on boiling water reactor (BWR) experiments conducted between 1953 and 1964. Facility characterization, decision analyses, and D&D plans for the turbine building were prepared from 1979 through 1990. D&D activities of the turbine building systems were initiated in November of 1988 and completed with the demolition and backfill of the concrete foundation in March 1992. Due to the low levels of radioactivity and the absence of loosemore » contamination, the D&D activities were completed with no radiation exposure to the workers. The D&D activities were performed in a manner that no radiological health or safety hazard to the public or to personnel at the Idaho National Engineering Laboratory (INEL) remain.« less

Final report of the decontamination and decommissioning of the BORAX-V facility turbine building

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

Arave, A.E.; Rodman, G.R.

1992-12-01

The Boiling Water Reactor Experiment (BORAX)-V Facility Turbine Building Decontamination and Decommissioning (D D) Project is described in this report. The BORAX series of five National Reactor Testing Station (NRTS) reactors pioneered intensive work on boiling water reactor (BWR) experiments conducted between 1953 and 1964. Facility characterization, decision analyses, and D D plans for the turbine building were prepared from 1979 through 1990. D D activities of the turbine building systems were initiated in November of 1988 and completed with the demolition and backfill of the concrete foundation in March 1992. Due to the low levels of radioactivity and themore » absence of loose contamination, the D D activities were completed with no radiation exposure to the workers. The D D activities were performed in a manner that no radiological health or safety hazard to the public or to personnel at the Idaho National Engineering Laboratory (INEL) remain.« less

Experience with chemical system decontamination by the CORD process and electrochemical decontamination of pipe ends

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

Wille, H.; Bertholdt, H.O.; Operschall, H.

Efforts to reduce occupational radiation exposure during inspection and repair work in nuclear power plants turns steadily increasing attention to the decontamination of systems and components. Due to the advanced age of nuclear power plants resulting in increasing dose rates, the decontamination of components, or rather of complete systems, or loops to protect operating and inspection personnel becomes demanding. Besides, decontaminating complete primary loops is in many cases less difficult than cleaning large components. Based on experience gained in nuclear power plants, an outline of two different decontamination methods performed recently are given. For the decontamination of complete systems ormore » loops, Kraftwerk Union AG has developed CORD, a low-concentration process. For the decontamination performance of a subsystem, such as the steam generator (SG) channel heads of a pressurized water reactor or the recirculation loops of a boiling water reactor the automated mobile decontamination appliance is used. The electrochemical decontamination process is primarily applicable for the treatment of specially limited surface areas.« less

Dynamic characteristics of a VK-50 reactor operating under conditions of the loss of a normal feedwater flow

NASA Astrophysics Data System (ADS)

Semidotskiy, I. I.; Kurskiy, A. S.

2013-12-01

The paper describes the conditions of the ATWS type with virtually complete cessation of the feed-water flow at the operating power level of a reactor of the VK-50 type. Under these conditions, the role of spatial kinetics in the system of feedback between thermohydraulic and nuclear processes with bulk boiling of the coolant in the reactor core is clearly seen. This feature determines the specific character of experimental data obtained and the suitability of their use for verification of the associated codes used for calculating water-water reactors.

ENGINEERING APPLICATIONS OF ANALOG COMPUTERS

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

Bryant, L.T.; Janicke, M.J.; Just, L.C.

1963-10-31

Six experiments from the fields of reactor engineering, heat transfer, and dynamics are presented to illustrate the engineering applications of analog computers. The steps required for producing the analog solution are shown, as well as complete information for duplicating the solution. Graphical results are provided. The experiments include: deceleration of a reactor control rod, pressure variations through a packed bed, reactor kinetics over many decades with thermal feedback, a vibrating system with two degrees of freedom, temperature distribution in a radiating fin, temperature distribution in an infinite slab considering variable thermal properties, and iodine -xenon buildup in a reactor. (M.C.G.)

Advances in boron neutron capture therapy (BNCT) at kyoto university - From reactor-based BNCT to accelerator-based BNCT

NASA Astrophysics Data System (ADS)

Sakurai, Yoshinori; Tanaka, Hiroki; Takata, Takushi; Fujimoto, Nozomi; Suzuki, Minoru; Masunaga, Shinichiro; Kinashi, Yuko; Kondo, Natsuko; Narabayashi, Masaru; Nakagawa, Yosuke; Watanabe, Tsubasa; Ono, Koji; Maruhashi, Akira

2015-07-01

At the Kyoto University Research Reactor Institute (KURRI), a clinical study of boron neutron capture therapy (BNCT) using a neutron irradiation facility installed at the research nuclear reactor has been regularly performed since February 1990. As of November 2014, 510 clinical irradiations were carried out using the reactor-based system. The world's first accelerator-based neutron irradiation system for BNCT clinical irradiation was completed at this institute in early 2009, and the clinical trial using this system was started in 2012. A shift of BCNT from special particle therapy to a general one is now in progress. To promote and support this shift, improvements to the irradiation system, as well as its preparation, and improvements in the physical engineering and the medical physics processes, such as dosimetry systems and quality assurance programs, must be considered. The recent advances in BNCT at KURRI are reported here with a focus on physical engineering and medical physics topics.

Gasification in pulverized coal flames. Second annual progress report, July 1976--August 1977. [Pulverized coal

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

George, P. E.; Lenzer, R. C.; Thomas, J. F.

1977-08-01

This project concerns the production of power and synthesis gases from pulverized coal via suspension gasification. Swirling flow in both concentric jet and cyclone gasifiers will separate oxidation and reduction zones. Gasifier performance will be correlated with internally measured temperature and concentration profiles. The test cell flow system and electrical system, which includes a safety interlock design, has been installed. Calibration of the UTI-30C mass spectrometer and construction of the gas sampling system are complete. Both the coal feeder, which has been calibrated, and the boiler are ready for integration into the test cell flow system. Construction and testing ofmore » the cyclone reactor, including methane combustion experiments, is complete. The confined jet reactor has been designed and construction is underway. Investigation of combustion and gasification modeling techniques has begun.« less

Reference reactor module for NASA's lunar surface fission power system

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

Poston, David I; Kapernick, Richard J; Dixon, David D

Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on themore » lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.« less

Removal of slowly biodegradable COD in combined thermophilic UASB and MBBR systems.

PubMed

Ji, M; Yu, J; Chen, H; Yue, P L

2001-09-01

Starch, cellulose and polyvinyl alcohol (PVA) are common substrates of the slowly biodegradable COD (SBCOD) in industrial wastewaters. Removal of the individual and mixed SbCOD substrates was investigated in a combined system of thermophilic upflow anaerobic sludge blanket (TUASB) reactor (55 degrees C) and aerobic moving bed biofilm reactor (MBBR). The removal mechanisms of the three SBCOD substrates were quite different. Starch-COD was almost equally utilized and removed in the two reactors. Cellulose-COD was completely (97-98%) removed from water in the TUASB reactor by microbial entrapment and sedimentation of the cellulose fibers. PVA alone was hardly biodegraded and removed by the combined reactors. However, PVA-COD could be removed to some extent in a binary solution of starch (77%) plus PVA (23%). The PVA macromolecules in the binary solution actually affected the microbial activity in the TUASB reactor resulting accumulation of volatile fatty acids, which shifted the overall COD removal from the TUASB to the MBBR reactor where SBCOD including PVA-COD was removed. Since the three SBCOD substrates were removed by different mechanisms, the combined reactors showed a better and more stable performance than individual reactors.

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

NONE

During this time period, at WVU, the authors have obtained models for the kinetics of the HAS (higher alcohol synthesis) reaction over the Co-K-MoS{sub 2}/C catalyst. The Rotoberty reactor was then replaced in the reactor system by a plug-flow tubular reactor. Accordingly, the authors re-started the investigations on sulfide catalysts. The authors encountered and solved the leak problem from the sampling valve for the non-sulfided reactor system. They also modified the system to eliminate the condensation problem. Accordingly, they are continuing their kinetic studies on the reduced Mo-Ni-K/C catalysts. They have set up an apparatus for temperature-programmed reduction (TPR) studies,more » and have obtained some interesting results on TPR characterizations. At UCC, the complete characterization of selected catalysts has been started. The authors sent nine selected types of ZnO, Zn/CrO and Zn/Cr/MnO catalysts and supports for BET surface area, SEM, XRD and ICP. They also sent fresh and spent samples of the Engelhard Zn/CrO catalyst impregnated with 3 wt% potassium for ISS and XPS testing. In Task 2, work on the design and optimization portion of this task, as well as on the fuel testing, is completed. All funds have been expended and there are no personnel working on this project.« less

Comparison of the performance of MBBR and SBR systems for the treatment of anaerobic reactor biowaste effluent.

PubMed

Comett-Ambriz, I; Gonzalez-Martinez, S; Wilderer, P

2003-01-01

Anaerobic reactor biowaste effluent was treated with biofilm and activated sludge sequencing batch reactors to compare the performance of both systems. The treatment targets were organic carbon removal and nitrification. The pilot plant was operated in two phases. During the first phase, it was operated like a Moving Bed Biofilm Reactor (MBBR) with the Natrix media, with a specific surface area of 210 m2/m3. The MBBR was operated under Sequencing Batch Reactor (SBR) modality with three 8-hour cycles per day over 70 days. During the second phase of the experiment, the pilot plant was operated over 79 days as a SBR. In both phases the influent was fed to the reactor at a flow rate corresponding to a Hydraulic Retention Time (HRT) of 4 days. Both systems presented a good carbon removal for this specific wastewater. The Chemical Oxygen Demand (COD) total removal was 53% for MBBR and 55% for SBR. MBBR offered a higher dissolved COD removal (40%) than SBR (30%). The limited COD removal achieved is in agreement with the high COD to BOD5 ratio (1/3) of the influent wastewater. In both systems a complete nitrification was obtained. The different efficiencies in both systems are related to the different biomass concentrations.

High throughput semiconductor deposition system

DOEpatents

Young, David L.; Ptak, Aaron Joseph; Kuech, Thomas F.; Schulte, Kevin; Simon, John D.

2017-11-21

A reactor for growing or depositing semiconductor films or devices. The reactor may be designed for inline production of III-V materials grown by hydride vapor phase epitaxy (HVPE). The operating principles of the HVPE reactor can be used to provide a completely or partially inline reactor for many different materials. An exemplary design of the reactor is shown in the attached drawings. In some instances, all or many of the pieces of the reactor formed of quartz, such as welded quartz tubing, while other reactors are made from metal with appropriate corrosion resistant coatings such as quartz or other materials, e.g., corrosion resistant material, or stainless steel tubing or pipes may be used with a corrosion resistant material useful with HVPE-type reactants and gases. Using HVPE in the reactor allows use of lower-cost precursors at higher deposition rates such as in the range of 1 to 5 .mu.m/minute.

Continuous immobilized yeast reactor system for complete beer fermentation using spent grains and corncobs as carrier materials.

PubMed

Brányik, Tomás; Silva, Daniel P; Vicente, António A; Lehnert, Radek; e Silva, João B Almeida; Dostálek, Pavel; Teixeira, José A

2006-12-01

Despite extensive research carried out in the last few decades, continuous beer fermentation has not yet managed to outperform the traditional batch technology. An industrial breakthrough in favour of continuous brewing using immobilized yeast could be expected only on achievement of the following process characteristics: simple design, low investment costs, flexible operation, effective process control and good product quality. The application of cheap carrier materials of by-product origin could significantly lower the investment costs of continuous fermentation systems. This work deals with a complete continuous beer fermentation system consisting of a main fermentation reactor (gas-lift) and a maturation reactor (packed-bed) containing yeast immobilized on spent grains and corncobs, respectively. The suitability of cheap carrier materials for long-term continuous brewing was proved. It was found that by fine tuning of process parameters (residence time, aeration) it was possible to adjust the flavour profile of the final product. Consumers considered the continuously fermented beer to be of a regular quality. Analytical and sensorial profiles of both continuously and batch fermented beers were compared.

Low cost solar array project: Experimental process system development unit for producing semiconductor-grade silicon using the silane-to-silicon process

NASA Technical Reports Server (NTRS)

1981-01-01

The results of the free space reactor experimental work are summarized. Overall, the objectives were achieved and the unit can be confidently scaled to the EPSDU size based on the experimental work and supporting theoretical analyses. The piping and instrumentation of the fluidized bed reactor was completed.

Central waste processing system

NASA Technical Reports Server (NTRS)

Kester, F. L.

1973-01-01

A new concept for processing spacecraft type wastes has been evaluated. The feasibility of reacting various waste materials with steam at temperatures of 538 - 760 C in both a continuous and batch reactor with residence times from 3 to 60 seconds has been established. Essentially complete gasification is achieved. Product gases are primarily hydrogen, carbon dioxide, methane, and carbon monoxide. Water soluble synthetic wastes are readily processed in a continuous tubular reactor at concentrations up to 20 weight percent. The batch reactor is able to process wet and dry wastes at steam to waste weight ratios from 2 to 20. Feces, urine, and synthetic wastes have been successfully processed in the batch reactor.

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

NASA Technical Reports Server (NTRS)

Woerner, L.

1977-01-01

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

A bioreactor system for the nitrogen loop in a Controlled Ecological Life Support System

NASA Technical Reports Server (NTRS)

Saulmon, M. M.; Reardon, K. F.; Sadeh, W. Z.

1996-01-01

As space missions become longer in duration, the need to recycle waste into useful compounds rises dramatically. This problem can be addressed by the development of Controlled Ecological Life Support Systems (CELSS) (i.e., Engineered Closed/Controlled Eco-Systems (ECCES)), consisting of human and plant modules. One of the waste streams leaving the human module is urine. In addition to the reclamation of water from urine, recovery of the nitrogen is important because it is an essential nutrient for the plant module. A 3-step biological process for the recycling of nitrogenous waste (urea) is proposed. A packed-bed bioreactor system for this purpose was modeled, and the issues of reaction step segregation, reactor type and volume, support particle size, and pressure drop were addressed. Based on minimization of volume, a bioreactor system consisting of a plug flow immobilized urease reactor, a completely mixed flow immobilized cell reactor to convert ammonia to nitrite, and a plug flow immobilized cell reactor to produce nitrate from nitrite is recommended. It is apparent that this 3-step bioprocess meets the requirements for space applications.

The IRIS Spool-Type Reactor Coolant Pump

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

Kujawski, J.M.; Kitch, D.M.; Conway, L.E.

2002-07-01

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safetymore » to address the requirements defined by the US DOE for Generation IV reactors. One of the innovative features of the IRIS design is the adoption of a reactor coolant pump (called 'spool' pump) which is completely contained inside the reactor vessel. Background, status and future developments of the IRIS spool pump are presented in this paper. (authors)« less

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

M. J. Appel

This cleanup verification package documents completion of remedial action for the 118-F-3, Minor Construction Burial Ground waste site. This site was an open field covered with cobbles, with no vegetation growing on the surface. The site received irradiated reactor parts that were removed during conversion of the 105-F Reactor from the Liquid 3X to the Ball 3X Project safety systems and received mostly vertical safety rod thimbles and step plugs.

Engine System Model Development for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Nelson, Karl W.; Simpson, Steven P.

2006-01-01

In order to design, analyze, and evaluate conceptual Nuclear Thermal Propulsion (NTP) engine systems, an improved NTP design and analysis tool has been developed. The NTP tool utilizes the Rocket Engine Transient Simulation (ROCETS) system tool and many of the routines from the Enabler reactor model found in Nuclear Engine System Simulation (NESS). Improved non-nuclear component models and an external shield model were added to the tool. With the addition of a nearly complete system reliability model, the tool will provide performance, sizing, and reliability data for NERVA-Derived NTP engine systems. A new detailed reactor model is also being developed and will replace Enabler. The new model will allow more flexibility in reactor geometry and include detailed thermal hydraulics and neutronics models. A description of the reactor, component, and reliability models is provided. Another key feature of the modeling process is the use of comprehensive spreadsheets for each engine case. The spreadsheets include individual worksheets for each subsystem with data, plots, and scaled figures, making the output very useful to each engineering discipline. Sample performance and sizing results with the Enabler reactor model are provided including sensitivities. Before selecting an engine design, all figures of merit must be considered including the overall impacts on the vehicle and mission. Evaluations based on key figures of merit of these results and results with the new reactor model will be performed. The impacts of clustering and external shielding will also be addressed. Over time, the reactor model will be upgraded to design and analyze other NTP concepts with CERMET and carbide fuel cores.

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

Apple juice clarification by immobilized pectolytic enzymes in packed or fluidized bed reactors.

PubMed

Diano, Nadia; Grimaldi, Tiziana; Bianco, Mariangela; Rossi, Sergio; Gabrovska, Katya; Yordanova, Galya; Godjevargova, Tzonka; Grano, Valentina; Nicolucci, Carla; Mita, Luigi; Bencivenga, Umberto; Canciglia, Paolo; Mita, Damiano G

2008-12-10

The catalytic behavior of a mixture of pectic enzymes, covalently immobilized on different supports (glass microspheres, nylon 6/6 pellets, and PAN beads), was analyzed with a pectin aqueous solution that simulates apple juice. The following parameters were investigated: the rate constant at which pectin hydrolysis is conducted, the time (tau(50)) in which the reduction of 50% of the initial viscosity is reached, and the time (tau(comp,dep)) required to obtain complete depectinization. The best catalytic system was proven to be PAN beads, and their pH and temperature behavior were determined. The yields of two bed reactors, packed or fluidized, using the catalytic PAN beads, were compared to the circulation flow rate of real apple juice. The experimental conditions were as follows: pH 4.0, T = 50 degrees C, and beads volume = 20 cm(3). The initial pectin concentration was the one that was present in our apple juice sample. No differences were observed at low circulation rates, while at higher recirculation rates, the time required to obtain complete pectin hydrolysis into the fluidized reactor was found to be 0.25 times smaller than in the packed bed reactor: 131 min for the packed reactors and 41 min for the fluidized reactors.

Modifications to the NRAD Reactor, 1977 to present

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

Weeks, A.A.; Pruett, D.P.; Heidel, C.C.

1986-01-01

Argonne National Laboratory-West, operated by the University of Chicago, is located near Idaho Falls, ID, on the Idaho National Engineering laboratory Site. ANL-West performs work in support of the Liquid Metal Fast Breeder Reactor Program (LMFBR) sponsored by the United States Department of Energy. The NRAD reactor is located at the Argonne Site within the Hot Fuel Examination Facility/North, a large hot cell facility where both non-destructive and destructive examinations are performed on highly irradiated reactor fuels and materials in support of the LMFBR program. The NRAD facility utilizes a 250-kW TRIGA reactor and is completely dedicated to neutron radiographymore » and the development of radiography techniques. Criticality was first achieved at the NRAD reactor in October of 1977. Since that time, a number of modifications have been implemented to improve operational efficiency and radiography production. This paper describes the modifications and changes that significantly improved operational efficiency and reliability of the reactor and the essential auxiliary reactor systems.« less

Research reactor decommissioning experience - concrete removal and disposal -

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

Manning, Mark R.; Gardner, Frederick W.

1990-07-01

Removal and disposal of neutron activated concrete from biological shields is the most significant operational task associated with research reactor decommissioning. During the period of 1985 thru 1989 Chem-Nuclear Systems, Inc. was the prime contractor for complete dismantlement and decommissioning of the Northrop TRIGA Mark F, the Virginia Tech Argonaut, and the Michigan State University TRIGA Mark I Reactor Facilities. This paper discusses operational requirements, methods employed, and results of the concrete removal, packaging, transport and disposal operations for these (3) research reactor decommissioning projects. Methods employed for each are compared. Disposal of concrete above and below regulatory release limitsmore » for unrestricted use are discussed. This study concludes that activated reactor biological shield concrete can be safely removed and buried under current regulations.« less

Microbial community analysis in a combined anaerobic and aerobic digestion system for treatment of cellulosic ethanol production wastewater.

PubMed

Shan, Lili; Yu, Yanling; Zhu, Zebing; Zhao, Wei; Wang, Haiman; Ambuchi, John J; Feng, Yujie

2015-11-01

This study investigated the microbial diversity established in a combined system composed of a continuous stirred tank reactor (CSTR), expanded granular sludge bed (EGSB) reactor, and sequencing batch reactor (SBR) for treatment of cellulosic ethanol production wastewater. Excellent wastewater treatment performance was obtained in the combined system, which showed a high chemical oxygen demand removal efficiency of 95.8% and completely eliminated most complex organics revealed by gas chromatography-mass spectrometry (GC-MS). Denaturing gradient gel electrophoresis (DGGE) analysis revealed differences in the microbial community structures of the three reactors. Further identification of the microbial populations suggested that the presence of Lactobacillus and Prevotella in CSTR played an active role in the production of volatile fatty acids (VFAs). The most diverse microorganisms with analogous distribution patterns of different layers were observed in the EGSB reactor, and bacteria affiliated with Firmicutes, Synergistetes, and Thermotogae were associated with production of acetate and carbon dioxide/hydrogen, while all acetoclastic methanogens identified belonged to Methanosaetaceae. Overall, microorganisms associated with the ability to degrade cellulose, hemicellulose, and other biomass-derived organic carbons were observed in the combined system. The results presented herein will facilitate the development of an improved cellulosic ethanol production wastewater treatment system.

Development work for a borax internal core-catcher for a gas-cooled fast reactor

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

Donne, M.D.; Dorner, S.; Schumacher, G.

1978-07-01

Preliminary thermal calculations show that a corecatcher, which is able to cope with the complete meltdown of the core and blankets of a 1000-MW(electric) gas-cooled fast reactor, appears to be feasible. This core-catcher is based on borax (Na/sub 2/B/sub 4/O/sub 7/) dissolving the oxide fuel and the fission products occurring in oxide form. The borax is contained in steel boxes forming a 2.2-m-thick slab on the base of the reactor cavity inside the prestressed concrete reactor vessel (PCRV), just underneath the reactor core. After a complete meltdown accident, the fission products, in oxide form, are dispersed in the pool formedmore » by the liquid borax. The metallic fission products are contained in the steel lying below the borax pool and in contact with the water-cooled PCRV liner. The volumetric power density of the molten core is conveniently reduced as it is dissolved in the borax, and the resulting heat fluxes at the borders of the pool can be safely carried away through the PCRV liner and its water cooling system.« less

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

Faidy, C.

Practical applications of the leak-before break concept are presently limited in French Pressurized Water Reactors (PWR) compared to Fast Breeder Reactors. Neithertheless, different fracture mechanic demonstrations have been done on different primary, auxiliary and secondary PWR piping systems based on similar requirements that the American NUREG 1061 specifications. The consequences of the success in different demonstrations are still in discussion to be included in the global safety assessment of the plants, such as the consequences on in-service inspections, leak detection systems, support optimization,.... A large research and development program, realized in different co-operative agreements, completes the general approach.

Method of shielding a liquid-metal-cooled reactor

DOEpatents

Sayre, Robert K.

1978-01-01

The primary heat transport system of a nuclear reactor -- particularly for a liquid-metal-cooled fast-breeder reactor -- is shielded and protected from leakage by establishing and maintaining a bed of a powdered oxide closely and completely surrounding all components thereof by passing a gas upwardly therethrough at such a rate as to slightly expand the bed to the extent that the components of the system are able to expand without damage and yet the particles of the bed remain close enough so that the bed acts as a guard vessel for the system. Preferably the gas contains 1 to 10% oxygen and the gas is passed upwardly through the bed at such a rate that the lower portion of the bed is a fixed bed while the upper portion is a fluidized bed, the line of demarcation therebetween being high enough that the fixed bed portion of the bed serves as guard vessel for the system.

Optimization of arsenic removal water treatment system through characterization of terminal electron accepting processes.

PubMed

Upadhyaya, Giridhar; Clancy, Tara M; Brown, Jess; Hayes, Kim F; Raskin, Lutgarde

2012-11-06

Terminal electron accepting process (TEAP) zones developed when a simulated groundwater containing dissolved oxygen (DO), nitrate, arsenate, and sulfate was treated in a fixed-bed bioreactor system consisting of two reactors (reactors A and B) in series. When the reactors were operated with an empty bed contact time (EBCT) of 20 min each, DO-, nitrate-, sulfate-, and arsenate-reducing TEAP zones were located within reactor A. As a consequence, sulfate reduction and subsequent arsenic removal through arsenic sulfide precipitation and/or arsenic adsorption on or coprecipitation with iron sulfides occurred in reactor A. This resulted in the removal of arsenic-laden solids during backwashing of reactor A. To minimize this by shifting the sulfate-reducing zone to reactor B, the EBCT of reactor A was sequentially lowered from 20 min to 15, 10, and 7 min. While 50 mg/L (0.81 mM) nitrate was completely removed at all EBCTs, more than 90% of 300 μg/L (4 μM) arsenic was removed with the total EBCT as low as 27 min. Sulfate- and arsenate-reducing bacteria were identified throughout the system through clone libraries and quantitative PCR targeting the 16S rRNA, dissimilatory (bi)sulfite reductase (dsrAB), and dissimilatory arsenate reductase (arrA) genes. Results of reverse transcriptase (RT) qPCR of partial dsrAB (i.e., dsrA) and arrA transcripts corresponded with system performance. The RT qPCR results indicated colocation of sulfate- and arsenate-reducing activities, in the presence of iron(II), suggesting their importance in arsenic removal.

Development and performance of an alternative biofilter system.

PubMed

Lee, D H; Lau, A K; Pinder, K L

2001-01-01

Step tracer tests were carried out on lab-scale biofilters to determine the residence time distributions (RTDs) of gases passing through two types of biofilters: a standard biofilter with vertical gas flow and a modified biofilter with horizontal gas flow. Results were used to define the flow patterns in the reactors. "Non-ideal flow" indicates that the flow reactors did not behave like either type of ideal reactor: the perfectly stirred reactor [often called a "continuously stirred tank reactor" (CSTR)] or the plug-flow reactor. The horizontal biofilter with back-mixing was able to accommodate a shorter residence time without the usual requirement of greater biofilter surface area for increased biofiltration efficiency. Experimental results indicated that the first bed of the modified biofilter behaved like two CSTRs in series, while the second bed may be represented by two or three CSTRs in series. Because of the flow baffles used in the horizontal biofilter system, its performance was more similar to completely mixed systems, and hence, it could not be modeled as a plug-flow reactor. For the standard biofilter, the number of CSTRs was found to be between 2 and 9 depending on the airflow rate. In terms of NH3 removal efficiency and elimination capacity, the standard biofilter was not as good as the modified system; moreover, the second bed of the modified biofilter exhibited greater removal efficiency than the first bed. The elimination rate increased as biofilter load increased. An opposite trend was exhibited with respect to removal efficiency.

Hanging core support system for a nuclear reactor. [LMFBR

DOEpatents

Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

1984-04-26

For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform.

DETECTION OF COATING FAILURES IN A NEUTRONIC REACTOR

DOEpatents

Snell, A.H.; Allison, S.K.

1958-02-11

This patent relates to water-cooled reactor systems and discloses a means to detect leaks in the jackets of jacketed fuel elements comprising a neutron detector located in the cooling water discharge pipe,the pipe being provided with an enlarged portion for housing the detector so that the latter is completely surrounded by the water in its passage through the pipe, said enlarged portion and detector being shielded from the reactor for the purpose of detecting only those delayed neutrons emitted in the cooling water and due to the latter picking up fission fragments from the defective fuel elements.

Status of the irradiation test vehicle for testing fusion materials in the Advanced Test Reactor

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

Tsai, H.; Gomes, I.C.; Smith, D.L.

1998-09-01

The design of the irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) has been completed. The main application for the ITV is irradiation testing of candidate fusion structural materials, including vanadium-base alloys, silicon carbide composites, and low-activation steels. Construction of the vehicle is underway at the Lockheed Martin Idaho Technology Company (LMITCO). Dummy test trains are being built for system checkout and fine-tuning. Reactor insertion of the ITV with the dummy test trains is scheduled for fall 1998. Barring unexpected difficulties, the ITV will be available for experiments in early 1999.

A complete dosimetry experimental program in support to the core characterization and to the power calibration of the CABRI reactor. A complete dosimetry experimental program in support of the core characterization and of the power calibration of the CABRI reactor

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

Rodiac, F.; Hudelot, JP.; Lecerf, J.

CABRI is an experimental pulse reactor operated by CEA at the Cadarache research center. Since 1978 the experimental programs have aimed at studying the fuel behavior under Reactivity Initiated Accident (RIA) conditions. Since 2003, it has been refurbished in order to be able to provide RIA and LOCA (Loss Of Coolant Accident) experiments in prototypical PWR conditions (155 bar, 300 deg. C). This project is part of a broader scope including an overall facility refurbishment and a safety review. The global modification is conducted by the CEA project team. It is funded by IRSN, which is conducting the CIP experimentalmore » program, in the framework of the OECD/NEA project CIP. It is financed in the framework of an international collaboration. During the reactor restart, commissioning tests are realized for all equipment, systems and circuits of the reactor. In particular neutronics and power commissioning tests will be performed respectively in 2015 and 2016. This paper focuses on the design of a complete and original dosimetry program that was built in support to the CABRI core characterization and to the power calibration. Each one of the above experimental goals will be fully described, as well as the target uncertainties and the forecasted experimental techniques and data treatment. (authors)« less

A new method of two-phase anaerobic digestion for fruit and vegetable waste treatment.

PubMed

Wu, Yuanyuan; Wang, Cuiping; Liu, Xiaoji; Ma, Hailing; Wu, Jing; Zuo, Jiane; Wang, Kaijun

2016-07-01

A novel method of two-phase anaerobic digestion where the acid reactor is operated at low pH 4.0 was proposed and investigated. A completely stirred tank acid reactor and an up-flow anaerobic sludge bed methane reactor were operated to examine the possibility of efficient degradation of lactate and to identify their optimal operating conditions. Lactate with an average concentration of 14.8g/L was the dominant fermentative product and Lactobacillus was the predominant microorganism in the acid reactor. The effluent from the acid reactor was efficiently degraded in the methane reactor and the average methane yield was 261.4ml/gCOD removed. Organisms of Methanosaeta were the predominant methanogen in granular sludge of methane reactor, however, after acclimation hydrogenotrophic methanogens enriched, which benefited for the conversion of lactate to acetate. The two-phase AD system exhibited a low hydraulic retention time of 3.56days and high methane yield of 348.5ml/g VS removed. Copyright © 2016 Elsevier Ltd. All rights reserved.

Low-Dissolved-Oxygen Nitrifying Systems Exploit Ammonia-Oxidizing Bacteria with Unusually High Yields▿

PubMed Central

Bellucci, Micol; Ofiţeru, Irina D.; Graham, David W.; Head, Ian M.; Curtis, Thomas P.

2011-01-01

In wastewater treatment plants, nitrifying systems are usually operated with elevated levels of aeration to avoid nitrification failures. This approach contributes significantly to operational costs and the carbon footprint of nitrifying wastewater treatment processes. In this study, we tested the effect of aeration rate on nitrification by correlating ammonia oxidation rates with the structure of the ammonia-oxidizing bacterial (AOB) community and AOB abundance in four parallel continuous-flow reactors operated for 43 days. Two of the reactors were supplied with a constant airflow rate of 0.1 liter/min, while in the other two units the airflow rate was fixed at 4 liters/min. Complete nitrification was achieved in all configurations, though the dissolved oxygen (DO) concentration was only 0.5 ± 0.3 mg/liter in the low-aeration units. The data suggest that efficient performance in the low-DO units resulted from elevated AOB levels in the reactors and/or putative development of a mixotrophic AOB community. Denaturing gel electrophoresis and cloning of AOB 16S rRNA gene fragments followed by sequencing revealed that the AOB community in the low-DO systems was a subset of the community in the high-DO systems. However, in both configurations the dominant species belonged to the Nitrosomonas oligotropha lineage. Overall, the results demonstrated that complete nitrification can be achieved at low aeration in lab-scale reactors. If these findings could be extended to full-scale plants, it would be possible to minimize the operational costs and greenhouse gas emissions without risk of nitrification failure. PMID:21926211

Removal of oxytetracycline (OTC) in a synthetic pharmaceutical wastewater by a sequential anaerobic multichamber bed reactor (AMCBR)/completely stirred tank reactor (CSTR) system: biodegradation and inhibition kinetics.

PubMed

Sponza, Delia Teresa; Çelebi, Hakan

2012-01-01

An anaerobic multichamber bed reactor (AMCBR) was effective in removing both molasses-chemical oxygen demand (COD), and the antibiotic oxytetracycline (OTC). The maximum COD and OTC removals were 99% in sequential AMCBR/completely stirred tank reactor (CSTR) at an OTC concentration of 300 mg L(-1). 51%, 29% and 9% of the total volatile fatty acid (TVFA) was composed of acetic, propionic acid and butyric acids, respectively. The OTC loading rates at between 22.22 and 133.33 g OTC m(-3) d(-1) improved the hydrolysis of molasses-COD (k), the maximum specific utilization of molasses-COD (k(mh)) and the maximum specific utilization rate of TVFA (k(TVFA)). The direct effect of high OTC loadings (155.56 and -177.78 g OTC m(-3) d(-1)) on acidogens and methanogens were evaluated with Haldane inhibition kinetic. A significant decrease of the Haldane inhibition constant was indicative of increases in toxicity at increasing loading rates. Copyright © 2011 Elsevier Ltd. All rights reserved.

Continuous flow synthesis of ZSM-5 zeolite on the order of seconds

PubMed Central

Liu, Zhendong; Okabe, Kotatsu; Anand, Chokkalingam; Yonezawa, Yasuo; Zhu, Jie; Yamada, Hiroki; Endo, Akira; Yanaba, Yutaka; Yoshikawa, Takeshi; Ohara, Koji; Okubo, Tatsuya; Wakihara, Toru

2016-01-01

The hydrothermal synthesis of zeolites carried out in batch reactors takes a time so long (typically, on the order of days) that the crystallization of zeolites has long been believed to be very slow in nature. We herein present a synthetic process for ZSM-5, an industrially important zeolite, on the order of seconds in a continuous flow reactor using pressurized hot water as a heating medium. Direct mixing of a well-tuned precursor (90 °C) with the pressurized water preheated to extremely high temperature (370 °C) in the millimeter-sized continuous flow reactor resulted in immediate heating to high temperatures (240–300 °C); consequently, the crystallization of ZSM-5 in a seed-free system proceeded to completion within tens of or even several seconds. These results indicate that the crystallization of zeolites can complete in a period on the order of seconds. The subtle design combining a continuous flow reactor with pressurized hot water can greatly facilitate the mass production of zeolites in the future. PMID:27911823

Discussion-preliminary review of the safety aspects of the crossunder line, Project CG-884. Volume 1

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

Jones, S.S.

1960-12-19

In order to reduce both charge-discharge shutdown time and the number of manhours of radiation exposure, Project CGI-884 is being completed at the B, D, DR, F and R Reactors. This consists essentially of installing a large drain line at the bottom of one rear reactor riser. This drain line passes to a control valve and then to the effluent line beyond the downcomer. This system by-passes the crossover downcomer part of the effluent system and eliminates the need for intermittent rear crossheader valving during reactor charge-discharge procedures. Two aspects of this system have been considered, its basic design requirements,more » and operating restrictions to ensure adequate process tube cooling. Because of the complexity of the reactor flow system approximate solutions were used to compare different methods or degrees of operation and establish limits. Despite these approximations, there was sufficient difference in the case results to justify the specific conclusions presented in this report. This report should serve the dual purpose of providing design requirements for the crossunder and also providing the technical criteria necessary for the operating standards for the use of this new system.« less

Helium heater design for the helium direct cycle component test facility. [for gas-cooled nuclear reactor power plant

NASA Technical Reports Server (NTRS)

Larson, V. R.; Gunn, S. V.; Lee, J. C.

1975-01-01

The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

Study of the Open Loop and Closed Loop Oscillator Techniques

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

Imel, George R.; Baker, Benjamin; Riley, Tony

This report presents the progress and completion of a five-year study undertaken at Idaho State University of the measurement of very small worth reactivity samples comparing open and closed loop oscillator techniques.The study conclusively demonstrated the equivalency of the two techniques with regard to uncertainties in reactivity values, i.e., limited by reactor noise. As those results are thoroughly documented in recent publications, in this report we will concentrate on the support work that was necessary. For example, we describe in some detail the construction and calibration of a pilot rod for the closed loop system. We discuss the campaign tomore » measure the required reactor parameters necessary for inverse-kinetics. Finally, we briefly discuss the transfer of the open loop technique to other reactor systems.« less

From biofilm ecology to reactors: a focused review.

PubMed

Boltz, Joshua P; Smets, Barth F; Rittmann, Bruce E; van Loosdrecht, Mark C M; Morgenroth, Eberhard; Daigger, Glen T

2017-04-01

Biofilms are complex biostructures that appear on all surfaces that are regularly in contact with water. They are structurally complex, dynamic systems with attributes of primordial multicellular organisms and multifaceted ecosystems. The presence of biofilms may have a negative impact on the performance of various systems, but they can also be used beneficially for the treatment of water (defined herein as potable water, municipal and industrial wastewater, fresh/brackish/salt water bodies, groundwater) as well as in water stream-based biological resource recovery systems. This review addresses the following three topics: (1) biofilm ecology, (2) biofilm reactor technology and design, and (3) biofilm modeling. In so doing, it addresses the processes occurring in the biofilm, and how these affect and are affected by the broader biofilm system. The symphonic application of a suite of biological methods has led to significant advances in the understanding of biofilm ecology. New metabolic pathways, such as anaerobic ammonium oxidation (anammox) or complete ammonium oxidation (comammox) were first observed in biofilm reactors. The functions, properties, and constituents of the biofilm extracellular polymeric substance matrix are somewhat known, but their exact composition and role in the microbial conversion kinetics and biochemical transformations are still to be resolved. Biofilm grown microorganisms may contribute to increased metabolism of micro-pollutants. Several types of biofilm reactors have been used for water treatment, with current focus on moving bed biofilm reactors, integrated fixed-film activated sludge, membrane-supported biofilm reactors, and granular sludge processes. The control and/or beneficial use of biofilms in membrane processes is advancing. Biofilm models have become essential tools for fundamental biofilm research and biofilm reactor engineering and design. At the same time, the divergence between biofilm modeling and biofilm reactor modeling approaches is recognized.

NASA's Kilopower Reactor Development and the Path to Higher Power Missions

NASA Technical Reports Server (NTRS)

Gibson, Marc A.; Oleson, Steven R.; Poston, David I.; McClure, Patrick

2017-01-01

The development of NASAs Kilopower fission reactor is taking large strides toward flight development with several successful tests completed during its technology demonstration trials. The Kilopower reactors are designed to provide 1-10 kW of electrical power to a spacecraft which could be used for additional science instruments as well as the ability to power electric propulsion systems. Power rich nuclear missions have been excluded from NASA proposals because of the lack of radioisotope fuel and the absence of a flight qualified fission system. NASA has partnered with the Department of Energy's National Nuclear Security Administration to develop the Kilopower reactor using existing facilities and infrastructure to determine if the design is ready for flight development. The 3-year Kilopower project started in 2015 with a challenging goal of building and testing a full-scale flight prototypic nuclear reactor by the end of 2017. As the date approaches, the engineering team shares information on the progress of the technology as well as the enabling capabilities it provides for science and human exploration.

NASA's Kilopower Reactor Development and the Path to Higher Power Missions

NASA Technical Reports Server (NTRS)

Gibson, Marc A.; Oleson, Steven R.; Poston, Dave I.; McClure, Patrick

2017-01-01

The development of NASA's Kilopower fission reactor is taking large strides toward flight development with several successful tests completed during its technology demonstration trials. The Kilopower reactors are designed to provide 1-10 kW of electrical power to a spacecraft which could be used for additional science instruments as well as the ability to power electric propulsion systems. Power rich nuclear missions have been excluded from NASA proposals because of the lack of radioisotope fuel and the absence of a flight qualified fission system. NASA has partnered with the Department of Energy's National Nuclear Security Administration to develop the Kilopower reactor using existing facilities and infrastructure to determine if the design is ready for flight development. The 3-year Kilopower project started in 2015 with a challenging goal of building and testing a full-scale flight prototypic nuclear reactor by the end of 2017. As the date approaches, the engineering team shares information on the progress of the technology as well as the enabling capabilities it provides for science and human exploration.

Comparison of bio-dissolution of spent Ni-Cd batteries by sewage sludge using ferrous ions and elemental sulfur as substrate.

PubMed

Zhao, Ling; Zhu, Nan-Wen; Wang, Xiao-Hui

2008-01-01

Bioleaching of spent Ni-Cd batteries using acidified sewage sludge was carried out in a continuous flow two-step leaching system including an acidifying reactor and a leaching reactor. Two systems operated about 30d to achieve almost complete dissolution of heavy metals Ni, Cd and Co in four Ni-Cd batteries. Ferrous sulphate and elemental sulfur were used as two different substrates to culture indigenous thiobacilli in sewage sludge. pH and ORP of the acidifying reactor was stabilized around 2.3 and 334mV for the iron-oxidizing system and 1.2 and 390mV for the sulfur-oxidizing system. It was opposite to the acidifying reactor, the pH/ORP in the leaching reactor of the iron-oxidizing system was relatively lower/higher than that of the sulphur-oxidizing system in the first 17d. The metal dissolution, in the first 12-16d, was faster in the iron-oxidizing system than in the sulphur-oxidizing system due to the lower pH. In the iron-oxidizing system, the maximum solubilization of cadmium (2500mg l(-1)) and cobalt (260mg l(-1)) can be reached at day 6-8 and the most of metal nickel was leached in the first 16d. But in the sulphur-oxidizing system there was a lag period of 4-8d to reach the maximum solubilization of cadmium and cobalt. The maximum dissolution of nickel hydroxide (1400mg l(-1)) and metallic nickel (2300mg l(-1)) occurred at about day 12 and day 20, respectively.

Design and testing of the reactor-internal hydraulic control rod drive for the nuclear heating plant

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

Batheja, P.; Meier, W.J.; Rau, P.J.

A hydraulically driven control rod is being developed at Kraftwerk Union for integration in the primary system of a small nuclear district heating reactor. An elaborate test program, under way for --3 yr, was initiated with a plexiglass rig to understand the basic principles. A design specification list was prepared, taking reactor boundary conditions and relevant German rules and regulations into account. Subsequently, an atmospheric loop for testing of components at 20 to 90/sup 0/C was erected. The objectives involved optimization of individual components such as a piston/cylinder drive unit, electromagnetic valves, and an ultrasonic position indication system as wellmore » as verification of computer codes. Based on the results obtained, full-scale components were designed and fabricated for a prototype test rig, which is currently in operation. Thus far, all atmospheric tests in this rig have been completed. Investigations under reactor temperature and pressure, followed by endurance tests, are under way. All tests to date have shown a reliable functioning of the hydraulic drive, including a novel ultrasonic position indication system.« less

MYRRHA: A multipurpose nuclear research facility

NASA Astrophysics Data System (ADS)

Baeten, P.; Schyns, M.; Fernandez, Rafaël; De Bruyn, Didier; Van den Eynde, Gert

2014-12-01

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a multipurpose research facility currently being developed at SCK•CEN. MYRRHA is based on the ADS (Accelerator Driven System) concept where a proton accelerator, a spallation target and a subcritical reactor are coupled. MYRRHA will demonstrate the ADS full concept by coupling these three components at a reasonable power level to allow operation feedback. As a flexible irradiation facility, the MYRRHA research facility will be able to work in both critical as subcritical modes. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for GEN IV and fusion reactors, and radioisotope production for medical and industrial applications. MYRRHA will be cooled by lead-bismuth eutectic and will play an important role in the development of the Pb-alloys technology needed for the LFR (Lead Fast Reactor) GEN IV concept. MYRRHA will also contribute to the study of partitioning and transmutation of high-level waste. Transmutation of minor actinides (MA) can be completed in an efficient way in fast neutron spectrum facilities, so both critical reactors and subcritical ADS are potential candidates as dedicated transmutation systems. However critical reactors heavily loaded with fuel containing large amounts of MA pose reactivity control problems, and thus safety problems. A subcritical ADS operates in a flexible and safe manner, even with a core loading containing a high amount of MA leading to a high transmutation rate. In this paper, the most recent developments in the design of the MYRRHA facility are presented.

Superphenix: Restarting, with emphasis on research

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

NONE

1994-04-01

French Prime Minister Edouard Baladour announced on February 22 that the Superphenix fast reactor at Creys-Malville will be allowed to restart. Along with his industry minister, Gerard Longuet, and the environment minister, Michel Bamier, Baladour has accepted the recommendation for reissuance of an operating license submitted to the ministers in January by the nuclear installations safety directorate (DSIN). In announcing their approval for restart, the ministers emphasized that the reactor was to be used as a research and demonstration facility rather than a power station. In particular, they said it should be used to investigate the possibility of using fastmore » reactors as plutonium burners rather than breeders and for the incineration of other long-lived actinide wastes. This point was made also by DSIN Director Andre-Claude Lacoste in his recommendations, but appears to have been given more weight by the government ministers. The ministers also accepted the DSIN recommendation that restart be conditional on completion of improvements to fire protection systems for the secondary sodium coolant ducts. Baladour said that it would take several months to complete this work, but other sources have suggested that the reactor could be ready to operate this month. DSIN has also stipulated that the power level be limited for several months to around 50 percent of the plant`s 1200-MWe potential in order to check out all the systems carefully.« less

Treatment of fruit-juice industry wastewater in a two-stage anaerobic hybrid (AH) reactor system followed by a sequencing batch reactor (SBR).

PubMed

Tawfik, A; El-Kamah, H

2012-01-01

This study has been carried out to assess the performance of a combined system consisting of an anaerobic hybrid (AH) reactor followed by a sequencing batch reactor (SBR) for treatment of fruit-juice industry wastewater at a temperature of 26 degrees C. Three experimental runs were conducted in this investigation. In the first experiment, a single-stage AH reactor was operated at a hydraulic retention time (HRT) of 10.2 h and organic loading rate (OLR) of 11.8 kg COD m(-3) d(-1). The reactor achieved a removal efficiency of 42% for chemical oxygen demand (COD), 50.8% for biochemical oxygen demand (BOD5), 50.3% for volatile fatty acids (VFA) and 56.4% for total suspended solids (TSS). In the second experiment, two AH reactors connected in series achieved a higher removal efficiency for COD (67.4%), BOD5 (77%), and TSS (71.5%) at a total HRT of 20 h and an OLR of 5.9 kg COD m(-3) d(-1). For removal of the remaining portions of COD, BOD5 and TSS from the effluent of the two-stage AH system, a sequencing batch reactor (SBR) was investigated as a post-treatment unit. The reactor achieved a substantial reduction in total COD, resulting in an average effluent concentration of 50 mg L(-1) at an HRT of 11 h and OLR of 5.3 kg COD m(-3) d(-1). Almost complete removal of total BOD5 and oil and grease was achieved, i.e. 10 mg L(-1) and 1.2 mg L(-1), respectively, remained in the final effluent of the SBR.

Serial completely stirred tank reactors for improving biogas production and substance degradation during anaerobic digestion of corn stover.

PubMed

Li, YuQian; Liu, ChunMei; Wachemo, Akiber Chufo; Yuan, HaiRong; Zou, DeXun; Liu, YanPing; Li, XiuJin

2017-07-01

Several completely stirred tank reactors (CSTR) connected in series for anaerobic digestion of corn stover were investigated in laboratory scale. Serial anaerobic digestion systems operated at a total HRT of 40days, and distribution of HRT are 10+30days (HRT10+30d), 20+20days (HRT20+20d), and 30+10days (HRT30+10d) were compared to a conventional one-step CSTR at the same HRT of 40d. The results showed that in HRT10+30d serial system, the process became very unstable at organic load of 50gTS·L -1 . The HRT20+20d and HRT30+10d serial systems improved methane production by 8.3-14.6% compared to the one-step system in all loads of 50, 70, 90gTS·L -1 . The conversion rates of total solid, cellulose, and hemicellulose were increased in serial anaerobic digestion systems compared to single system. The serial systems showed more stable process performance in high organic load. HRT30+10d system showed the best biogas production and conversions among all systems. Copyright © 2017. Published by Elsevier Ltd.

F Reactor Inspection

ScienceCinema

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2018-01-16

Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

F Reactor Inspection

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

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-10-29

Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosuremore » and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."« less

PLUTONIUM-CERIUM ALLOY

DOEpatents

Coffinberry, A.S.

1959-01-01

An alloy is presented for use as a reactor fuel. The binary alloy consists essentially of from about 5 to 90 atomic per cent cerium and the balance being plutonium. A complete phase diagram for the cerium--plutonium system is given.

Hanging core support system for a nuclear reactor

DOEpatents

Burelbach, James P.; Kann, William J.; Pan, Yen-Cheng; Saiveau, James G.; Seidensticker, Ralph W.

1987-01-01

For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform. Motion or radiation sensing detectors can be provide at the lower ends of the tension rods for obtaining pertinent readings proximate the core.

REACTOR VIEWING APPARATUS

DOEpatents

Monk, G.S.

1959-01-13

An optical system is presented that is suitable for viewing objects in a region of relatively high radioactivity, or high neutron activity, such as a neutronic reactor. This optical system will absorb neutrons and gamma rays thereby protecting personnel fronm the harmful biological effects of such penetrating radiations. The optical system is comprised of a viewing tube having a lens at one end, a transparent solid member at the other end and a transparent aqueous liquid completely filling the tube between the ends. The lens is made of a polymerized organic material and the transparent solid member is made of a radiation absorbent material. A shield surrounds the tube betwcen the flanges and is made of a gamma ray absorbing material.

Temperature Swing Adsorption Compressor Development

NASA Technical Reports Server (NTRS)

Finn, John E.; Mulloth, Lila M.; Affleck, Dave L.

2001-01-01

Closing the oxygen loop in an air revitalization system based on four-bed molecular sieve and Sabatier reactor technology requires a vacuum pump-compressor that can take the low-pressure CO, from the 4BMS and compress and store for use by a Sabatier reactor. NASA Ames Research Center proposed a solid-state temperature-swing adsorption (TSA) compressor that appears to meet performance requirements, be quiet and reliable, and consume less power than a comparable mechanical compressor/accumulator combination. Under this task, TSA compressor technology is being advanced through development of a complete prototype system. A liquid-cooled TSA compressor has been partially tested, and the rest of the system is being fabricated. An air-cooled TSA compressor is also being designed.

ENHANCED PRACTICAL PHOTOSYNTHETIC CO2 MITIGATION

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

Dr. Gregory Kremer; Dr. David J. Bayless; Dr. Morgan Vis

2001-07-25

This quarterly report documents significant achievements in the Enhanced Practical Photosynthetic CO{sub 2} Mitigation project during the period from 4/03/2001 through 7/02/2001. Most of the achievements are milestones in our efforts to complete the tasks and subtasks that constitute the project objectives. Note that this version of the quarterly technical report is a revision to add the reports from subcontractors Montana State and Oak Ridge National Laboratories The significant accomplishments for this quarter include: Development of an experimental plan and initiation of experiments to create a calibration curve that correlates algal chlorophyll levels with carbon levels (to simplify future experimentalmore » procedures); Completion of debugging of the slug flow reactor system, and development of a plan for testing the pressure drop of the slug flow reactor; Design and development of a new bioreactor screen design which integrates the nutrient delivery drip system and the harvesting system; Development of an experimental setup for testing the new integrated drip system/harvesting system; Completion of model-scale bioreactor tests examining the effects of CO{sub 2} concentration levels and lighting levels on Nostoc 86-3 growth rates; Completion of the construction of a larger model-scale bioreactor to improve and expand testing capabilities and initiation of tests; Substantial progress on construction of a pilot-scale bioreactor; and Preliminary economic analysis of photobioreactor deployment. Plans for next quarter's work are included in the conclusions. A preliminary economic analysis is included as an appendix.« less

Advanced anaerobic bioconversion of lignocellulosic waste for the melissa life support system

NASA Astrophysics Data System (ADS)

Lissens, G.; Verstraete, W.; Albrecht, T.; Brunner, G.; Creuly, C.; Dussap, G.; Kube, J.; Maerkl, H.; Lasseur, C.

The feasibility of nearly-complete conversion of lignocellulosic waste (70% food crops, 20% faecal matter and 10% green algae) into biogas was investigated in the context of the MELiSSA loop (Micro-Ecological Life Support System Alternative). The treatment comprised a series of processes, i.e. a mesophilic laboratory scale CSTR (continuously stirred tank reactor), an upflow biofilm reactor, a fiber liquefaction reactor employing the rumen bacterium Fibrobacter succinogenes and a hydrothermolysis system in near-critical water. By the one-stage CSTR, a biogas yield of 75% with a specific biogas production of 0.37 l biogas g-1 VSS (volatile suspended solids) added at a RT (hydraulic retention time) of 20-25 d was obtained. Biogas yields could not be increased considerably at higher RT, indicating the depletion of readily available substrate after 25 d. The solids present in the CSTR-effluent were subsequently treated in two ways. Hydrothermal treatment (T ˜ 310-350C, p ˜ 240 bar) resulted in effective carbon liquefaction (50-60% without and 83% with carbon dioxide saturation) and complete sanitation of the residue. Application of the cellulolytic Fibrobacter succinogenes converted remaining cellulose contained in the CSTR-effluent into acetate and propionate mainly. Subsequent anaerobic digestion of the hydrothermolysis and the Fibrobacter hydrolysates allowed conversion of 48-60% and 30%, respectively. Thus, the total process yielded biogas corresponding with conversions up to 90% of the original organic matter. It appears that particularly mesophilic digestion in conjunction with hydrothermolysis offers interesting features for (nearly) the MELiSSA system. The described additional technologies show that complete and hygienic carbon and energy recovery from human waste within MELiSSA is technically feasible, provided that the extra energy needed for the thermal treatment is guaranteed.

Post-treatment of reclaimed waste water based on an electrochemical advanced oxidation process

NASA Technical Reports Server (NTRS)

Verostko, Charles E.; Murphy, Oliver J.; Hitchens, G. D.; Salinas, Carlos E.; Rogers, Tom D.

1992-01-01

The purification of reclaimed water is essential to water reclamation technology life-support systems in lunar/Mars habitats. An electrochemical UV reactor is being developed which generates oxidants, operates at low temperatures, and requires no chemical expendables. The reactor is the basis for an advanced oxidation process in which electrochemically generated ozone and hydrogen peroxide are used in combination with ultraviolet light irradiation to produce hydroxyl radicals. Results from this process are presented which demonstrate concept feasibility for removal of organic impurities and disinfection of water for potable and hygiene reuse. Power, size requirements, Faradaic efficiency, and process reaction kinetics are discussed. At the completion of this development effort the reactor system will be installed in JSC's regenerative water recovery test facility for evaluation to compare this technique with other candidate processes.

Wastewater treatment with submerged fixed bed biofilm reactor systems--design rules, operating experiences and ongoing developments.

PubMed

Schlegel, S; Koeser, H

2007-01-01

Wastewater treatment systems using bio-films that grow attached to a support media are an alternative to the widely used suspended growth activated sludge process. Different fixed growth biofilm reactors are commercially used for the treatment of municipal as well as industrial wastewater. In this paper a fairly new fixed growth biofilm system, the submerged fixed bed biofilm reactor (SFBBR), is discussed. SFBBRs are based on aerated submerged fixed open structured plastic media for the support of the biofilm. They are generally operated without sludge recirculation in order to avoid clogging of the support media and problems with the control of the biofilm. Reactor and process design considerations for these reactors are reviewed. Measures to ensure the development and maintenance of an active biofilm are examined. SFBBRs have been applied successfully to small wastewater treatment plants where complete nitrification but no high degree of denitrification is necessary. For the pre-treatment of industrial wastewater the use of SFBBRs is advantageous, especially in cases of wastewater with high organic loading or high content of compounds with low biodegradability. Performance data from exemplary commercial plants are given. Ongoing research and development efforts aim at achieving a high simultaneous total nitrogen (TN) removal of aerated SFBBRs and at improving the efficiency of TN removal in anoxic SFBBRs.

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.

Characterization of fast neutron spectrum in the TRIGA for hardness testing of electronic components

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

Nelson, George W.

1986-07-01

Argonne National Laboratory-West, operated by the University of Chicago, is located near Idaho Falls, ID, on the Idaho National Engineering Laboratory Site. ANL-West performs work in support of the Liquid Metal Fast Breeder Reactor Program (LMFBR) sponsored by the United States Department of Energy. The NRAD reactor is located at the Argonne Site within the Hot Fuel Examination Facility/North, a large hot cell facility where both non-destructive and destructive examinations are performed on highly irradiated reactor fuels and materials in support of the LMFBR program. The NRAD facility utilizes a 250-kW TRIGA reactor and is completely dedicated to neutron radiographymore » and the development of radiography techniques. Criticality was first achieved at the NRAD reactor in October of 1977. Since that time, a number of modifications have been implemented to improve operational efficiency and radiography production. This paper describes the modifications and changes that significantly improved operational efficiency and reliability of the reactor and the essential auxiliary reactor systems. (author)« less

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

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

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

2016-11-01

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

Analysis of closed cycle megawatt class space power systems with nuclear reactor heat sources

NASA Technical Reports Server (NTRS)

Juhasz, A. J.; Jones, B. I.

1987-01-01

The analysis and integration studies of multimegawatt nuclear power conversion systems for potential SDI applications is presented. A study is summarized which considered 3 separate types of power conversion systems for steady state power generation with a duty requirement of 1 yr at full power. The systems considered are based on the following conversion cycles: direct and indirect Brayton gas turbine, direct and indirect liquid metal Rankine, and in core thermionic. A complete mass analysis was performed for each system at power levels ranging from 1 to 25 MWe for both heat pipe and liquid droplet radiator options. In the modeling of common subsystems, reactor and shield calculations were based on multiparameter correlation and an in-house analysis for the heat rejection and other subsystems.

Evaluation of a Decentralized Wastewater Treatment Technology. INTERNATIONAL WASTEWATER SYSTEMS, INC. MODEL 6000 SEQUENCING BATCH REACTOR SYSTEM

EPA Science Inventory

Evaluation of the IWS Model 6000 SBR began in April 2004 when one SBR was taken off line and cleaned. The verification testing started July 1, 2004 and proceeded without interruption through June 30, 2005. All sixteen four-day sampling events were completed as scheduled, yielding...

Effect of air-assisted backwashing on the performance of an anaerobic fixed-bed bioreactor that simultaneously removes nitrate and arsenic from drinking water sources.

PubMed

Upadhyaya, Giridhar; Clancy, Tara M; Snyder, Kathryn V; Brown, Jess; Hayes, Kim F; Raskin, Lutgarde

2012-03-15

Contaminant removal from drinking water sources under reducing conditions conducive for the growth of denitrifying, arsenate reducing, and sulfate reducing microbes using a fixed-bed bioreactor may require oxygen-free gas (e.g., N2 gas) during backwashing. However, the use of air-assisted backwashing has practical advantages, including simpler operation, improved safety, and lower cost. A study was conducted to evaluate whether replacing N2 gas with air during backwashing would impact performance in a nitrate and arsenic removing anaerobic bioreactor system that consisted of two biologically active carbon reactors in series. Gas-assisted backwashing, comprised of 2 min of gas injection to fluidize the bed and dislodge biomass and solid phase products, was performed in the first reactor (reactor A) every two days. The second reactor (reactor B) was subjected to N2 gas-assisted backwashing every 3-4 months. Complete removal of 50 mg/L NO3- was achieved in reactor A before and after the switch from N2-assisted backwashing (NAB) to air-assisted backwashing (AAB). Substantial sulfate removal was achieved with both backwashing strategies. Prolonged practice of AAB (more than two months), however, diminished sulfate reduction in reactor B somewhat. Arsenic removal in reactor A was impacted slightly by long-term use of AAB, but arsenic removals achieved by the entire system during NAB and AAB periods were not significantly different (p>0.05) and arsenic concentrations were reduced from approximately 200 μg/L to below 20 μg/L. These results indicate that AAB can be implemented in anaerobic nitrate and arsenic removal systems. Copyright © 2011 Elsevier Ltd. All rights reserved.

Fuel processing in integrated micro-structured heat-exchanger reactors

NASA Astrophysics Data System (ADS)

Kolb, G.; Schürer, J.; Tiemann, D.; Wichert, M.; Zapf, R.; Hessel, V.; Löwe, H.

Micro-structured fuel processors are under development at IMM for different fuels such as methanol, ethanol, propane/butane (LPG), gasoline and diesel. The target application are mobile, portable and small scale stationary auxiliary power units (APU) based upon fuel cell technology. The key feature of the systems is an integrated plate heat-exchanger technology which allows for the thermal integration of several functions in a single device. Steam reforming may be coupled with catalytic combustion in separate flow paths of a heat-exchanger. Reactors and complete fuel processors are tested up to the size range of 5 kW power output of a corresponding fuel cell. On top of reactor and system prototyping and testing, catalyst coatings are under development at IMM for numerous reactions such as steam reforming of LPG, ethanol and methanol, catalytic combustion of LPG and methanol, and for CO clean-up reactions, namely water-gas shift, methanation and the preferential oxidation of carbon monoxide. These catalysts are investigated in specially developed testing reactors. In selected cases 1000 h stability testing is performed on catalyst coatings at weight hourly space velocities, which are sufficiently high to meet the demands of future fuel processing reactors.

Effectiveness of solar disinfection using batch reactors with non-imaging aluminium reflectors under real conditions: Natural well-water and solar light.

PubMed

Navntoft, C; Ubomba-Jaswa, E; McGuigan, K G; Fernández-Ibáñez, P

2008-12-11

Inactivation kinetics are reported for suspensions of Escherichia coli in well-water using compound parabolic collector (CPC) mirrors to enhance the efficiency of solar disinfection (SODIS) for batch reactors under real, solar radiation (cloudy and cloudless) conditions. On clear days, the system with CPC reflectors achieved complete inactivation (more than 5-log unit reduction in bacterial population to below the detection limit of 4CFU/mL) one hour sooner than the system fitted with no CPC. On cloudy days, only systems fitted with CPCs achieved complete inactivation. Degradation of the mirrors under field conditions was also evaluated. The reflectivity of CPC systems that had been in use outdoors for at least 3 years deteriorated in a non-homogeneous fashion. Reflectivity values for these older systems were found to vary between 27% and 72% compared to uniform values of 87% for new CPC systems. The use of CPC has been proven to be a good technological enhancement to inactivate bacteria under real conditions in clear and cloudy days. A comparison between enhancing optics and thermal effect is also discussed.

Core design of a direct-cycle, supercritical-water-cooled fast breeder reactor

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

Jevremovic, T.; Oka, Yoshiaki; Koshizuka, Seiichi

1994-10-01

The conceptual design of a direct-cycle fast breeder reactor (FBR) core cooled by supercritical water is carried out as a step toward a low-cost FBR plant. The supercritical water does not exhibit change of phase. The turbines are directly driven by the core outlet coolant. In comparison with a boiling water reactor (BWR), the recirculation systems, steam separators, and dryers are eliminated. The reactor system is much simpler than the conventional steam-cooled FBRs, which adopted Loeffler boilers and complicated coolant loops for generating steam and separating it from water. Negative complete and partial coolant void reactivity are provided without muchmore » deterioration in the breeding performances by inserting thin zirconium-hydride layers between the seeds and blankets in a radially heterogeneous core. The net electric power is 1245 MW (electric). The estimated compound system doubling time is 25 yr. The discharge burnup is 77.7 GWd/t, and the refueling period is 15 months with a 73% load factor. The thermal efficiency is high (41.5%), an improvement of 24% relative to a BWR's. The pressure vessel is not thick at 30.3 cm.« less

Operator Support System Design forthe Operation of RSG-GAS Research Reactor

NASA Astrophysics Data System (ADS)

Santoso, S.; Situmorang, J.; Bakhri, S.; Subekti, M.; Sunaryo, G. R.

2018-02-01

The components of RSG-GAS main control room are facing the problem of material ageing and technology obsolescence as well, and therefore the need for modernization and refurbishment are essential. The modernization in control room can be applied on the operator support system which bears the function in providing information for assisting the operator in conducting diagnosis and actions. The research purpose is to design an operator support system for RSG-GAS control room. The design was developed based on the operator requirement in conducting task operation scenarios and the reactor operation characteristics. These scenarios include power operation, low power operation and shutdown/scram reactor. The operator support system design is presented in a single computer display which contains structure and support system elements e.g. operation procedure, status of safety related components and operational requirements, operation limit condition of parameters, alarm information, and prognosis function. The prototype was developed using LabView software and consisted of components structure and features of the operator support system. Information of each component in the operator support system need to be completed before it can be applied and integrated in the RSG-GAS main control room.

Elimination of water pathogens with solar radiation using an automated sequential batch CPC reactor.

PubMed

Polo-López, M I; Fernández-Ibáñez, P; Ubomba-Jaswa, E; Navntoft, C; García-Fernández, I; Dunlop, P S M; Schmid, M; Byrne, J A; McGuigan, K G

2011-11-30

Solar disinfection (SODIS) of water is a well-known, effective treatment process which is practiced at household level in many developing countries. However, this process is limited by the small volume treated and there is no indication of treatment efficacy for the user. Low cost glass tube reactors, together with compound parabolic collector (CPC) technology, have been shown to significantly increase the efficiency of solar disinfection. However, these reactors still require user input to control each batch SODIS process and there is no feedback that the process is complete. Automatic operation of the batch SODIS process, controlled by UVA-radiation sensors, can provide information on the status of the process, can ensure the required UVA dose to achieve complete disinfection is received and reduces user work-load through automatic sequential batch processing. In this work, an enhanced CPC photo-reactor with a concentration factor of 1.89 was developed. The apparatus was automated to achieve exposure to a pre-determined UVA dose. Treated water was automatically dispensed into a reservoir tank. The reactor was tested using Escherichia coli as a model pathogen in natural well water. A 6-log inactivation of E. coli was achieved following exposure to the minimum uninterrupted lethal UVA dose. The enhanced reactor decreased the exposure time required to achieve the lethal UVA dose, in comparison to a CPC system with a concentration factor of 1.0. Doubling the lethal UVA dose prevented the need for a period of post-exposure dark inactivation and reduced the overall treatment time. Using this reactor, SODIS can be automatically carried out at an affordable cost, with reduced exposure time and minimal user input. Copyright © 2011 Elsevier B.V. All rights reserved.

BIOWINOL TECHNOLOGIES: A HYBRID GREEN PROCESS FOR BIOFUEL PRODUCTION – PHASE 2

EPA Science Inventory

The development of hollow fiber membrane (HFM) reactor will result in improved gas utilization that will positively impact overall process efficiencies. Successful completion of this project could result in the development of many decentralized biofuel production systems near ...

Continuous AE crack monitoring of a dissimilar metal weldment at Limerick Unit 1

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

Hutton, P.H.; Friesel, M.A.; Dawson, J.F.

1993-12-01

Acoustic emission (AE) technology for continuous surveillance of a reactor component(s) to detect crack initiation and/or crack growth has been developed at Pacific Northwest Laboratory (PNL). The technology was validated off-reactor in several major tests, but it had not been validated by monitoring crack growth on an operating reactor system. A flaw indication was identified during normal inservice inspection of piping at Philadelphia Electric Company (PECO) Limerick Unit 1 reactor during the 1989 refueling outage. Evaluation of the flaw indication showed that it could remain in place during the subsequent fuel cycle without compromising safety. The existence of this flawmore » indication offered a long sought opportunity to validate AE surveillance to detect and evaluate crack growth during reactor operation. AE instrumentation was installed by PNL and PECO to monitor the flaw indication during two complete fuel cycles. This report discusses the results obtained from the AE monitoring over the period May 1989 to March 1992 (two fuel cycles).« less

Alternative nuclear technologies

NASA Astrophysics Data System (ADS)

Schubert, E.

1981-10-01

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

Optimal design of an activated sludge plant: theoretical analysis

NASA Astrophysics Data System (ADS)

Islam, M. A.; Amin, M. S. A.; Hoinkis, J.

2013-06-01

The design procedure of an activated sludge plant consisting of an activated sludge reactor and settling tank has been theoretically analyzed assuming that (1) the Monod equation completely describes the growth kinetics of microorganisms causing the degradation of biodegradable pollutants and (2) the settling characteristics are fully described by a power law. For a given reactor height, the design parameter of the reactor (reactor volume) is reduced to the reactor area. Then the sum total area of the reactor and the settling tank is expressed as a function of activated sludge concentration X and the recycled ratio α. A procedure has been developed to calculate X opt, for which the total required area of the plant is minimum for given microbiological system and recycled ratio. Mathematical relations have been derived to calculate the α-range in which X opt meets the requirements of F/ M ratio. Results of the analysis have been illustrated for varying X and α. Mathematical formulae have been proposed to recalculate the recycled ratio in the events, when the influent parameters differ from those assumed in the design.

Biocatalytic methanation of hydrogen and carbon dioxide in an anaerobic three-phase system.

PubMed

Burkhardt, M; Koschack, T; Busch, G

2015-02-01

A new type of anaerobic trickle-bed reactor was used for biocatalytic methanation of hydrogen and carbon dioxide under mesophilic temperatures and ambient pressure in a continuous process. The conversion of gaseous substrates through immobilized hydrogenotrophic methanogenic archaea in a biofilm is a unique feature of this type of reactor. Due to the formation of a three-phase system on the carrier surface and operation as a plug flow reactor without gas recirculation, a complete reaction could be observed. With a methane concentration higher than c(CH4) = 98%, the product gas exhibits a very high quality. A specific methane production of P(CH4) = 1.49 Nm(3)/(m(3)(SV) d) was achieved at a hydraulic loading rate of LR(H2) = 6.0 Nm(3)/(m(3)(SV) d). The relation between trickle flow through the reactor and productivity could be shown. An application for methane enrichment in combination with biogas facilities as a source of carbon dioxide has also been positively proven. Copyright © 2014 Elsevier Ltd. All rights reserved.

Multi-unit Operations in Non-Nuclear Systems: Lessons Learned for Small Modular Reactors

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

OHara J. M.; Higgins, J.; DAgostino, A.

2012-01-17

The nuclear-power community has reached the stage of proposing advanced reactor designs to support power generation for decades to come. Small modular reactors (SMRs) are one approach to meet these energy needs. While the power output of individual reactor modules is relatively small, they can be grouped to produce reactor sites with different outputs. Also, they can be designed to generate hydrogen, or to process heat. Many characteristics of SMRs are quite different from those of current plants and may be operated quite differently. One difference is that multiple units may be operated by a single crew (or a singlemore » operator) from one control room. The U.S. Nuclear Regulatory Commission (NRC) is examining the human factors engineering (HFE) aspects of SMRs to support licensing reviews. While we reviewed information on SMR designs to obtain information, the designs are not completed and all of the design and operational information is not yet available. Nor is there information on multi-unit operations as envisioned for SMRs available in operating experience. Thus, to gain a better understanding of multi-unit operations we sought the lesson learned from non-nuclear systems that have experience in multi-unit operations, specifically refineries, unmanned aerial vehicles and tele-intensive care units. In this paper we report the lessons learned from these systems and the implications for SMRs.« less

Design principles of a simple and safe 200-MW(thermal) nuclear district heating plant

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

Goetzmann, C.; Bittermann, D.; Gobel, A.

Kraftwerk Union AG has almost completed the development of a dedicated 200-MW(thermal) nuclear district heating plant to provide environmentally clean energy at a predictably low cost. The concept can easily be adapted to meet power requirements within the 100- to 500-MW(thermal) range. This technology is the product of the experience gained with large pressurized water reactor and boiling water reactor power plants, with respect to both plant and fuel performance. The major development task is that of achieving sufficiently low capital cost by tailoring components and systems designed for large plants to the specific requirements of district heating. These requirementsmore » are small absolute power, low temperatures and pressures, and modest load following, all of which result in the characteristics that are summarized. A fully integrated primary system with natural circulation permits a very compact reactor building containing all safety-related systems and components. Plant safety is essentially guaranteed by inherent features. The reactor containment is tightly fitted around the reactor pressure vessel in such a way that, in the event of any postulated coolant leak, the core cannot become uncovered, even temporarily. Shutdown is assured by gravity drop of the control rods mounted above the core. Decay heat is removed from the core by means of natural circulation via dedicated intermediate circuits of external aircoolers.« less

Plum Brook Reactor Facility Control Room during Facility Startup

NASA Image and Video Library

1961-02-21

Operators test the National Aeronautics and Space Administration’s (NASA) Plum Brook Reactor Facility systems in the months leading up to its actual operation. The “Reactor On” signs are illuminated but the reactor core was not yet ready for chain reactions. Just a couple weeks after this photograph, Plum Brook Station held a media open house to unveil the 60-megawatt test reactor near Sandusky, Ohio. More than 60 members of the print media and radio and television news services met at the site to talk with community leaders and representatives from NASA and Atomic Energy Commission. The Plum Brook reactor went critical for the first time on the evening of June 14, 1961. It was not until April 1963 that the reactor reached its full potential of 60 megawatts. The reactor control room, located on the second floor of the facility, was run by licensed operators. The operators manually operated the shim rods which adjusted the chain reaction in the reactor core. The regulating rods could partially or completely shut down the reactor. The control room also housed remote area monitoring panels and other monitoring equipment that allowed operators to monitor radiation sensors located throughout the facility and to scram the reactor instantly if necessary. The color of the indicator lights corresponded with the elevation of the detectors in the various buildings. The reactor could also shut itself down automatically if the monitors detected any sudden irregularities.

VICTORIA: A mechanistic model for radionuclide behavior in the reactor coolant system

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

Schaperow, J.H.; Bixler, N.E.

1996-12-31

VICTORIA is the U.S. Nuclear Regulatory Commission`s (NRC`s) mechanistic, best-estimate code for analysis of fission product release from the core and subsequent transport in the reactor vessel and reactor coolant system. VICTORIA requires thermal-hydraulic data (i.e., temperatures, pressures, and velocities) as input. In the past, these data have been taken from the results of calculations from thermal-hydraulic codes such as SCDAP/RELAP5, MELCOR, and MAAP. Validation and assessment of VICTORIA 1.0 have been completed. An independent peer review of VICTORIA, directed by Brookhaven National Laboratory and supported by experts in the areas of fuel release, fission product chemistry, and aerosol physics,more » has been undertaken. This peer review, which will independently assess the code`s capabilities, is nearing completion with the peer review committee`s final report expected in Dec 1996. A limited amount of additional development is expected as a result of the peer review. Following this additional development, the NRC plans to release VICTORIA 1.1 and an updated and improved code manual. Future plans mainly involve use of the code for plant calculations to investigate specific safety issues as they arise. Also, the code will continue to be used in support of the Phebus experiments.« less

Aircraft and Bases Powered by Compact Nuclear Reactors: Solutions to Projecting Power in Highly Contested Environments and Fossil Fuel Dependence

DTIC Science & Technology

2015-05-01

pushed the depletion date past 2100.21 David Archibald, author of books and papers on climate science and a fellow at the Institute of World...Politics, does not predict explicitly the date of complete exhaustion, but he does note that humans have consumed about half of the world’s supply.22...deuterium, and lithium are plentiful on the earth and in the solar system. As far as fuel for existing and future fission reactors, uranium and

Bioaugmentation of activated sludge towards 3-chloroaniline removal with a mixed bacterial population carrying a degradative plasmid.

PubMed

Bathe, Stephan; Schwarzenbeck, Norbert; Hausner, Martina

2009-06-01

A bioaugmentation approach combining several strategies was applied to achieve degradation of 3-chloroaniline (3CA) in semicontinuous activated sludge reactors. In a first step, a 3CA-degrading Comamonas testosteroni strain carrying the degradative plasmid pNB2 was added to a biofilm reactor, and complete 3CA degradation together with spread of the plasmid within the indigenous biofilm population was achieved. A second set of reactors was then bioaugmented with either a suspension of biofilm cells removed from the carrier material or with biofilm-containing carrier material. 3CA degradation was established rapidly in all bioaugmented reactors, followed by a slow adaptation of the non-bioaugmented control reactors. In response to variations in 3CA concentration, all reactors exhibited temporary performance breakdowns. Whereas duplicates of the control reactors deviated in their behaviour, the bioaugmented reactors appeared more reproducible in their performance and population dynamics. Finally, the carrier-bioaugmented reactors showed an improved performance in the presence of high 3CA influent concentrations over the suspension-bioaugmented reactors. In contrast, degradation in one control reactor failed completely, but was rapidly established in the remaining control reactor.

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

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

Bucknor, Matthew; Hu, Rui; Lisowski, Darius

2016-04-17

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

Application of a Systems Engineering Approach to Support Space Reactor Development

NASA Astrophysics Data System (ADS)

Wold, Scott

2005-02-01

In 1992, approximately 25 Russian and 12 U.S. engineers and technicians were involved in the transport, assembly, inspection, and testing of over 90 tons of Russian equipment associated with the Thermionic System Evaluation Test (TSET) Facility. The entire Russian Baikal Test Stand, consisting of a 5.79 m tall vacuum chamber and related support equipment, was reassembled and tested at the TSET facility in less than four months. In November 1992, the first non-nuclear operational test of a complete thermionic power reactor system in the U.S. was accomplished three months ahead of schedule and under budget. A major factor in this accomplishment was the application of a disciplined top-down systems engineering approach and application of a spiral development model to achieve the desired objectives of the TOPAZ International Program (TIP). Systems Engineering is a structured discipline that helps programs and projects conceive, develop, integrate, test and deliver products and services that meet customer requirements within cost and schedule. This paper discusses the impact of Systems Engineering and a spiral development model on the success of the TOPAZ International Program and how the application of a similar approach could help ensure the success of future space reactor development projects.

Thermonuclear inverse magnetic pumping power cycle for stellarator reactor

DOEpatents

Ho, Darwin D.; Kulsrud, Russell M.

1991-01-01

The plasma column in a stellarator is compressed and expanded alternatively in minor radius. First a plasma in thermal balance is compressed adiabatically. The volume of the compressed plasma is maintained until the plasma reaches a new thermal equilibrium. The plasma is then expanded to its original volume. As a result of the way a stellarator works, the plasma pressure during compression is less than the corresponding pressure during expansion. Therefore, negative work is done on the plasma over a complete cycle. This work manifests itself as a back-voltage in the toroidal field coils. Direct electrical energy is obtained from this voltage. Alternatively, after the compression step, the plasma can be expanded at constant pressure. The cycle can be made self-sustaining by operating a system of two stellarator reactors in tandem. Part of the energy derived from the expansion phase of a first stellarator reactor is used to compress the plasma in a second stellarator reactor.

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

Boyack, B.E.; Steiner, J.L.; Harmony, S.C.

The PIUS Advanced Reactor is a 640-MW(e) pressurized-water reactor developed by Asea Brown Boveri. A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity normally is controlled by the boron concentration in the coolant and the temperature of the moderator coolant. Analyses of five initiating events have been completed on the basis of calculations performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. The initiating events analyzed are (1) reactor scram, (2) loss of off-site power (3) main steam-line break, (4) small-break loss of coolant, and (5) large-break loss of coolant. Inmore » addition to the baseline calculation for each sequence, sensitivity studies were performed to explore the response of the PIUS reactor to severe off-normal conditions having a very low probability of occurrence. The sensitivity studies provide insights into the robustness of the design.« less

Gasification in pulverized coal flames. First annual progress report, July 1975--June 1976

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

Lenzer, R. C.; George, P. E.; Thomas, J. F.

1976-07-01

This project concerns the production of power and synthesis gas from pulverized coal via suspension gasification. Swirling flow in both concentric jet and cyclone gasifiers will separate oxidation and reduction zones. Gasifier performance will be correlated with internally measured temperature and concentration profiles. A literature review of vortex and cyclone reactors is complete. Preliminary reviews of confined jet reactors and pulverized coal reaction models have also been completed. A simple equilibrium model for power gas production is in agreement with literature correlations. Cold gas efficiency is not a suitable performance parameter for combined cycle operation. The coal handling facility, equippedmore » with crusher, pulverizer and sieve shaker, is in working order. Test cell flow and electrical systems have been designed, and most of the equipment has been received. Construction of the cyclone gasifier has begun. A preliminary design for the gas sampling system, which will utilize a UTI Q-30C mass spectrometer, has been developed.« less

Reducing numerical costs for core wide nuclear reactor CFD simulations by the Coarse-Grid-CFD

NASA Astrophysics Data System (ADS)

Viellieber, Mathias; Class, Andreas G.

2013-11-01

Traditionally complete nuclear reactor core simulations are performed with subchannel analysis codes, that rely on experimental and empirical input. The Coarse-Grid-CFD (CGCFD) intends to replace the experimental or empirical input with CFD data. The reactor core consists of repetitive flow patterns, allowing the general approach of creating a parametrized model for one segment and composing many of those to obtain the entire reactor simulation. The method is based on a detailed and well-resolved CFD simulation of one representative segment. From this simulation we extract so-called parametrized volumetric forces which close, an otherwise strongly under resolved, coarsely-meshed model of a complete reactor setup. While the formulation so far accounts for forces created internally in the fluid others e.g. obstruction and flow deviation through spacers and wire wraps, still need to be accounted for if the geometric details are not represented in the coarse mesh. These are modelled with an Anisotropic Porosity Formulation (APF). This work focuses on the application of the CGCFD to a complete reactor core setup and the accomplishment of the parametrization of the volumetric forces.

Parametric study of natural circulation flow in molten salt fuel in molten salt reactor

NASA Astrophysics Data System (ADS)

Pauzi, Anas Muhamad; Cioncolini, Andrea; Iacovides, Hector

2015-04-01

The Molten Salt Reactor (MSR) is one of the most promising system proposed by Generation IV Forum (GIF) for future nuclear reactor systems. Advantages of the MSR are significantly larger compared to other reactor system, and is mainly achieved from its liquid nature of fuel and coolant. Further improvement to this system, which is a natural circulating molten fuel salt inside its tube in the reactor core is proposed, to achieve advantages of reducing and simplifying the MSR design proposed by GIF. Thermal hydraulic analysis on the proposed system was completed using a commercial computation fluid dynamics (CFD) software called FLUENT by ANSYS Inc. An understanding on theory behind this unique natural circulation flow inside the tube caused by fission heat generated in molten fuel salt and tube cooling was briefly introduced. Currently, no commercial CFD software could perfectly simulate natural circulation flow, hence, modeling this flow problem in FLUENT is introduced and analyzed to obtain best simulation results. Results obtained demonstrate the existence of periodical transient nature of flow problem, hence improvements in tube design is proposed based on the analysis on temperature and velocity profile. Results show that the proposed system could operate at up to 750MW core power, given that turbulence are enhanced throughout flow region, and precise molten fuel salt physical properties could be defined. At the request of the authors and the Proceedings Editor the name of the co-author Andrea Cioncolini was corrected from Andrea Coincolini. The same name correction was made in the Acknowledgement section on page 030004-10 and in reference number 4. The updated article was published on 11 May 2015.

NEUTRONIC REACTOR

DOEpatents

Fermi, E.; Zinn, W.H.; Anderson, H.L.

1958-09-16

Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

Decommissioning of the High Flux Beam Reactor at Brookhaven National Laboratory.

PubMed

Hu, Jih-Perng; Reciniello, Richard N; Holden, Norman E

2012-08-01

The High Flux Beam Reactor (HFBR) at the Brookhaven National Laboratory was a heavy-water cooled and moderated reactor that achieved criticality on 31 October 1965. It operated at a power level of 40 mega-watts. An equipment upgrade in 1982 allowed operations at 60 mega-watts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 mega-watts. The HFBR was shut down in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of ground water from wells located adjacent to the reactor's spent fuel pool. The reactor remained shut down for almost 3 y for safety and environmental reviews. In November 1999, the United States Department of Energy decided to permanently shut down the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome, which still contains the irradiated reactor vessel is presently under 24/7 surveillance for safety. Details of the HFBR's cleanup performed during 1999-2009, to allow the BNL facilities to be re-accessed by the public, will be described in the paper.

76 FR 68514 - Request for a License To Export Reactor Components

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-04

... NUCLEAR REGULATORY COMMISSION Request for a License To Export Reactor Components Pursuant to 10.../docket Number Westinghouse Electric Company Complete reactor 12 Perform seismic China. LLC, August 18... qualification equipment. of AP1000 (design) nuclear reactors. For the Nuclear Regulatory Commission. Dated this...

Development of NSSS Thermal-Hydraulic Model for KNPEC-2 Simulator Using the Best-Estimate Code RETRAN-3D

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

Kim, Kyung-Doo; Jeong, Jae-Jun; Lee, Seung-Wook

The Nuclear Steam Supply System (NSSS) thermal-hydraulic model adopted in the Korea Nuclear Plant Education Center (KNPEC)-2 simulator was provided in the early 1980s. The reference plant for KNPEC-2 is the Yong Gwang Nuclear Unit 1, which is a Westinghouse-type 3-loop, 950 MW(electric) pressurized water reactor. Because of the limited computational capability at that time, it uses overly simplified physical models and assumptions for a real-time simulation of NSSS thermal-hydraulic transients. This may entail inaccurate results and thus, the possibility of so-called ''negative training,'' especially for complicated two-phase flows in the reactor coolant system. To resolve the problem, we developedmore » a realistic NSSS thermal-hydraulic program (named ARTS code) based on the best-estimate code RETRAN-3D. The systematic assessment of ARTS has been conducted by both a stand-alone test and an integrated test in the simulator environment. The non-integrated stand-alone test (NIST) results were reasonable in terms of accuracy, real-time simulation capability, and robustness. After successful completion of the NIST, ARTS was integrated with a 3-D reactor kinetics model and other system models. The site acceptance test (SAT) has been completed successively and confirmed to comply with the ANSI/ANS-3.5-1998 simulator software performance criteria. This paper presents our efforts for the ARTS development and some test results of the NIST and SAT.« less

Fission Surface Power Systems (FSPS) Project Final Report for the Exploration Technology Development Program (ETDP): Fission Surface Power, Transition Face to Face

NASA Technical Reports Server (NTRS)

Palac, Donald T.

2011-01-01

The Fission Surface Power Systems Project became part of the ETDP on October 1, 2008. Its goal was to demonstrate fission power system technology readiness in an operationally relevant environment, while providing data on fission system characteristics pertinent to the use of a fission power system on planetary surfaces. During fiscal years 08 to 10, the FSPS project activities were dominated by hardware demonstrations of component technologies, to verify their readiness for inclusion in the fission surface power system. These Pathfinders demonstrated multi-kWe Stirling power conversion operating with heat delivered via liquid metal NaK, composite Ti/H2O heat pipe radiator panel operations at 400 K input water temperature, no-moving-part electromagnetic liquid metal pump operation with NaK at flight-like temperatures, and subscale performance of an electric resistance reactor simulator capable of reproducing characteristics of a nuclear reactor for the purpose of system-level testing, and a longer list of component technologies included in the attached report. Based on the successful conclusion of Pathfinder testing, work began in 2010 on design and development of the Technology Demonstration Unit (TDU), a full-scale 1/4 power system-level non-nuclear assembly of a reactor simulator, power conversion, heat rejection, instrumentation and controls, and power management and distribution. The TDU will be developed and fabricated during fiscal years 11 and 12, culminating in initial testing with water cooling replacing the heat rejection system in 2012, and complete testing of the full TDU by the end of 2014. Due to its importance for Mars exploration, potential applicability to missions preceding Mars missions, and readiness for an early system-level demonstration, the Enabling Technology Development and Demonstration program is currently planning to continue the project as the Fission Power Systems project, including emphasis on the TDU completion and testing.

Reactor/Brayton power systems for nuclear electric spacecraft

NASA Technical Reports Server (NTRS)

Layton, J. P.

1980-01-01

Studies are currently underway to assess the technological feasibility of a nuclear-reactor-powered spacecraft propelled by electric thrusters. This vehicle would be capable of performing detailed exploration of the outer planets of the solar system during the remainder of this century. The purpose of this study was to provide comparative information on a closed cycle gas turbine power conversion system. The results have shown that the performance is very competitive and that a 400 kWe space power system is dimensionally compatible with a single Space Shuttle launch. Performance parameters of system mass and radiator area were determined for systems from 100 to 1000 kWe. A 400 kWe reference system received primary attention. The components of this system were defined and a conceptual layout was developed with encouraging results. The preliminary mass determination for the complete power system was very close to the desired goal of 20 kg/kWe. Use of more advanced technology (higher turbine inlet temperature) will substantially improve system performance characteristics.

The Bosch Process-Performance of a Developmental Reactor and Experimental Evaluation of Alternative Catalysts

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, J. Matthew

2010-01-01

Bosch-based reactors have been in development at NASA since the 1960's. Traditional operation involves the reduction of carbon dioxide with hydrogen over a steel wool catalyst to produce water and solid carbon. While the system is capable of completely closing the loop on oxygen and hydrogen for Atmosphere Revitalization, steel wool requires a reaction temperature of 650C or higher for optimum performance. The single pass efficiency of the reaction over steel wool has been shown to be less than 10% resulting in a high recycle stream. Finally, the formation of solid carbon on steel wool ultimately fouls the catalyst necessitating catalyst resupply. These factors result in high mass, volume and power demands for a Bosch system. Interplanetary transportation and surface exploration missions of the moon, Mars, and near-earth objects will require higher levels of loop closure than current technology cannot provide. A Bosch system can provide the level of loop closure necessary for these long-term missions if mass, volume, and power can be kept low. The keys to improving the Bosch system lie in reactor and catalyst development. In 2009, the National Aeronautics and Space Administration refurbished a circa 1980's developmental Bosch reactor and built a sub-scale Bosch Catalyst Test Stand for the purpose of reactor and catalyst development. This paper describes the baseline performance of two commercially available steel wool catalysts as compared to performance reported in the 1960's and 80's. Additionally, the results of sub-scale testing of alternative Bosch catalysts, including nickel- and cobalt-based catalysts, are discussed.

10 CFR 2.621 - Acceptance and docketing of application for early review of site suitability issues in a combined...

Code of Federal Regulations, 2012 CFR

2012-01-01

... Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as...) The Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation... of Nuclear Reactor Regulation, as appropriate, that they are complete. (c) If part one of the...

91. ARAIII. GCRE reactor building (ARA608) at 48 percent completion. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

91. ARA-III. GCRE reactor building (ARA-608) at 48 percent completion. Camera faces west end of building; shows roll-up door. High bay section on right view. Petro-chem heater stack extends above roof of low-bay section on left. Excavation for 13, 8 kv electrical conduit in foreground. January 20, 1959. Ineel photo no. 59-322. Photographer: Jack L. Anderson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Performance of intermittent aeration reactor on NH4-N removal from groundwater resources.

PubMed

Khanitchaidecha, W; Nakamura, T; Sumino, T; Kazama, F

2010-01-01

To study the effect of intermittent aeration period on ammonium-nitrogen (NH4-N) removal from groundwater resources, synthetic groundwater was prepared and three reactors were operated under different conditions--"reactor A" under continuous aeration, "reactor B" under 6 h intermittent aeration, and "reactor C" under 2 h intermittent aeration. To facilitate denitrification simultaneously with nitrification, "acetate" was added as an external carbon source with step-wise increase from 0.5 to 1.5 C/N ratio, where C stands for total carbon content in the system, and N for NH4-N concentration in the synthetic groundwater. Results show that complete NH4-N removal was obtained in "reactor B" and "reactor C" at 1.3 and 1.5 C/N ratio respectively; and partial NH4-N removal in "reactor A". These results suggest that intermittent aeration at longer interval could enhance the reactor performance on NH4-N removal in terms of efficiency and low external carbon requirement. Because of consumption of internal carbon by the process, less amount of external carbon is required. Further increase in carbon in a form of acetate (1.5 to 2.5 C/N ratios) increases removal rate (represented by reaction rate coefficient (k) of kinetic equation) as well as occurrence of free cells. It suggests that the operating condition at reactor B with 1.3 C/N ratio is more appropriate for long-term operation at a pilot-scale.

Using single-chamber microbial fuel cells as renewable power sources of electro-Fenton reactors for organic pollutant treatment.

PubMed

Zhu, Xiuping; Logan, Bruce E

2013-05-15

Electro-Fenton reactions can be very effective for organic pollutant degradation, but they typically require non-sustainable electrical power to produce hydrogen peroxide. Two-chamber microbial fuel cells (MFCs) have been proposed for pollutant treatment using Fenton-based reactions, but these types of MFCs have low power densities and require expensive membranes. Here, more efficient dual reactor systems were developed using a single-chamber MFC as a low-voltage power source to simultaneously accomplish H2O2 generation and Fe(2+) release for the Fenton reaction. In tests using phenol, 75 ± 2% of the total organic carbon (TOC) was removed in the electro-Fenton reactor in one cycle (22 h), and phenol was completely degraded to simple and readily biodegradable organic acids. Compared to previously developed systems based on two-chamber MFCs, the degradation efficiency of organic pollutants was substantially improved. These results demonstrate that this system is an energy-efficient and cost-effective approach for industrial wastewater treatment of certain pollutants. Copyright © 2013 Elsevier B.V. All rights reserved.

Investigation of the possibility of using residual heat reactor energy

NASA Astrophysics Data System (ADS)

Aminov, R. Z.; Yurin, V. E.; Bessonov, V. N.

2017-11-01

The largest contribution to the probable frequency of core damage is blackout events. The main component of the heat capacity at each reactor within a few minutes following a blackout is the heat resulting from the braking of beta-particles and the transfer of gamma-ray energy by the fission fragments and their decay products, which is known as the residual heat. The power of the residual heat changes gradually over a long period of time and for a VVER-1000 reactor is about 15-20 MW of thermal power over 72 hours. Current cooldown systems increase the cost of the basic nuclear power plants (NPP) funds without changing the amount of electricity generated. Such systems remain on standby, accelerating the aging of the equipment and accordingly reducing its reliability. The probability of system failure increases with the duration of idle time. Furthermore, the reactor residual heat energy is not used. A proposed system for cooling nuclear power plants involves the use of residual thermal power to supply the station’s own needs in emergency situations accompanied by a complete blackout. The thermal power of residual heat can be converted to electrical energy through an additional low power steam turbine. In normal mode, the additional steam turbine generates electricity, which makes it possible to ensure spare NPP and a return on the investment in the reservation system. In this work, experimental data obtained from a Balakovo NPP was analyzed to determine the admissibility of cooldown of the reactors through the 2nd circuit over a long time period, while maintaining high-level parameters for the steam generated by the steam generators.

Microbial ureolysis in the seawater-catalysed urine phosphorus recovery system: Kinetic study and reactor verification.

PubMed

Tang, Wen-Tao; Dai, Ji; Liu, Rulong; Chen, Guang-Hao

2015-12-15

Our previous study has confirmed the feasibility of using seawater as an economical precipitant for urine phosphorus (P) precipitation. However, we still understand very little about the ureolysis in the Seawater-based Urine Phosphorus Recovery (SUPR) system despite its being a crucial step for urine P recovery. In this study, batch experiments were conducted to investigate the kinetics of microbial ureolysis in the seawater-urine system. Indigenous bacteria from urine and seawater exhibited relatively low ureolytic activity, but they adapted quickly to the urine-seawater mixture during batch cultivation. During cultivation, both the abundance and specific ureolysis rate of the indigenous bacteria were greatly enhanced as confirmed by a biomass-dependent Michaelis-Menten model. The period for fully ureolysis was decreased from 180 h to 2.5 h after four cycles of cultivation. Based on the successful cultivation, a lab-scale SUPR reactor was set up to verify the fast ureolysis and efficient P recovery in the SUPR system. Nearly complete urine P removal was achieved in the reactor in 6 h without adding any chemicals. Terminal Restriction Fragment Length Polymorphism (TRFLP) analysis revealed that the predominant groups of bacteria in the SUPR reactor likely originated from seawater rather than urine. Moreover, batch tests confirmed the high ureolysis rates and high phosphorus removal efficiency induced by cultivated bacteria in the SUPR reactor under seawater-to-urine mixing ratios ranging from 1:1 to 9:1. This study has proved that the enrichment of indigenous bacteria in the SUPR system can lead to sufficient ureolytic activity for phosphate precipitation, thus providing an efficient and economical method for urine P recovery. Copyright © 2015 Elsevier Ltd. All rights reserved.

Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

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

A. Joseph Palmer; Gerry L. McCormick; Shannon J. Corrigan

2010-06-01

2010 International Congress on Advances in Nuclear Power Plants (ICAPP’10) ANS Annual Meeting Imbedded Topical San Diego, CA June 13 – 17, 2010 Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR) Author: A. Joseph Palmer, Mechanical Engineer, Irradiation Test Programs, 208-526-8700, Joe.Palmer@INL.gov Affiliation: Idaho National Laboratory P.O. Box 1625, MS-3840 Idaho Falls, ID 83415 INL/CON-10-17680 ABSTRACT Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has beenmore » restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed a Readiness Assessment in November 2009. The HSIS is a key component of the ATR National Scientific User Facility (NSUF) operated by Battelle Energy Alliance, LLC and is available to a wide variety of university researchers for nuclear fuels and materials experiments as well as medical isotope research and production.« less

Treatment of synthetic refinery wastewater in anoxic-aerobic sequential moving bed reactors and sulphur recovery.

PubMed

Mallick, Subrat Kumar; Chakraborty, Saswati

2017-11-10

Objective of the present study was to simultaneously biodegrade synthetic petroleum refinery wastewater containing phenol (750 mg/L), sulphide (750 mg/L), hydrocarbon (as emulsified diesel of 300 mg/L), ammonia-nitrogen (350 mg/L) at pH >9 in anoxic-aerobic sequential moving bed reactors. The optimum mixing speed of anoxic reactor was observed at 20 rpm and beyond that, removal rate remained constant. In anoxic reactor the minimum hydraulic retention time was observed to be 2 days for complete removal of sulphide, 40-50% removal of phenol and total hydrocarbons and 52% of sulphur recovery. The optimum HRT of aerobic moving bed reactor was observed as 16 h (total HRT of 64 h for anoxic and aerobic reactors) for complete removals of phenol, total hydrocarbons, COD (chemical oxygen demand) and ammonia-nitrogen with nitrification.

Comparison of Alternative Hydrogen Donors for Anaerobic Reductive Dechlorination of Tetrachloroethene

DTIC Science & Technology

1998-01-01

feature two - stage anaerobic reductive dechlorination of highly chlorinated compounds coupled with aerobic (sometimes co- metabolic) treatment of the...activity at some naturally attenuated sites. Fathepure and Vogel [76] used a two - stage anaerobic-aerobic reactor system to treat hexachlorobenzene, PCE...Complete removal of the chloroethenes by the two - stage system was observed using pyruvate, formate, or lactose as electron donor for the dechlorinating

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.

Light Water Breeder Reactor fuel rod design and performance characteristics (LWBR Development Program)

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

Campbell, W.R.; Giovengo, J.F.

1987-10-01

Light Water Breeder Reactor (LWBR) fuel rods were designed to provide a reliable fuel system utilizing thorium/uranium-233 mixed-oxide fuel while simultaneously minimizing structural material to enhance fuel breeding. The fuel system was designed to be capable of operating successfully under both load follow and base load conditions. The breeding objective required thin-walled, low hafnium content Zircaloy cladding, tightly spaced fuel rods with a minimum number of support grid levels, and movable fuel rod bundles to supplant control rods. Specific fuel rod design considerations and their effects on performance capability are described. Successful completion of power operations to over 160 percentmore » of design lifetime including over 200 daily load follow cycles has proven the performance capability of the fuel system. 68 refs., 19 figs., 44 tabs.« less

Technical Bases to Aid in the Decision of Conducting Full Power Ground Nuclear Tests for Space Fission Reactors

NASA Astrophysics Data System (ADS)

Hixson, Laurie L.; Houts, Michael G.; Clement, Steven D.

2004-02-01

The extent to which, if any, full power ground nuclear testing of space reactors should be performed has been a point of discussion within the industry for decades. Do the benefits outweigh the risks? Are there equivalent alternatives? Can a test facility be constructed (or modified) in a reasonable amount of time? Is the test article an accurate representation of the flight system? Are the costs too restrictive? The obvious benefits of full power ground nuclear testing; obtaining systems integrated reliability data on a full-scale, complete end-to-end system; come at some programmatic risk. Safety related information is not obtained from a full-power ground nuclear test. This paper will discuss and assess these and other technical considerations essential in the decision to conduct full power ground nuclear-or alternative-tests.

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

NONE

The objective of Task 1 is to prepare and evaluate catalysts and to develop efficient reactor systems for the selective conversion of hydrogen-lean synthesis gas to alcohol fuel extender and octane enhancers. Task 1 is subdivided into three separate subtasks: laboratory and equipment setup; catalysis research; and reaction engineering and modeling. Research at West Virginia University (WVU) is focused on molybdenum-based catalysts for higher alcohol synthesis. Parallel research carried out at Union Carbide Corporation (UCC) is focused on transition-metal-oxide catalysts. During this time period, at WVU, we tried several methods to eliminate problems related to condensation of heavier products whenmore » reduced Mo-Ni-K/C materials were used as catalysts. We then resumed our kinetic study on the reduced Mo-Ni-K/C materials were used as catalysts. We then resumed our kinetic study on the reduced Mo-Ni-K/C catalysts. We have also obtained same preliminary results in our attempts to analyze quantitatively the temperature-programmed reduction spectra for C- supported Mo-based catalysts. We have completed the kinetic study for the sulfided Co-K-MoS{sub 2}/C catalyst. We have compared the results of methanol synthesis using the membrane reactor with those using a simple plug-flow reactor. At UCC, the complete characterization of selected catalysts has been completed. The results suggest that catalyst pretreatment under different reducing conditions yield different surface compositions and thus different catalytic reactivities.« less

Safety and core design of large liquid-metal cooled fast breeder reactors

NASA Astrophysics Data System (ADS)

Qvist, Staffan Alexander

In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

Expert system for maintenance management of a boiling water reactor power plant

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

Hong Shen; Liou, L.W.; Levine, S.

1992-01-01

An expert system code has been developed for the maintenance of two boiling water reactor units in Berwick, Pennsylvania, that are operated by the Pennsylvania Power and Light Company (PP and L). The objective of this expert system code, where the knowledge of experienced operators and engineers is captured and implemented, is to support the decisions regarding which components can be safely and reliably removed from service for maintenance. It can also serve as a query-answering facility for checking the plant system status and for training purposes. The operating and maintenance information of a large number of support systems, whichmore » must be available for emergencies and/or in the event of an accident, is stored in the data base of the code. It identifies the relevant technical specifications and management rules for shutting down any one of the systems or removing a component from service to support maintenance. Because of the complexity and time needed to incorporate a large number of systems and their components, the first phase of the expert system develops a prototype code, which includes only the reactor core isolation coolant system, the high-pressure core injection system, the instrument air system, the service water system, and the plant electrical system. The next phase is scheduled to expand the code to include all other systems. This paper summarizes the prototype code and the design concept of the complete expert system code for maintenance management of all plant systems and components.« less

Single Channel Testing for Characterization of the Direct Gas Cooled Reactor and the SAFE-100 Heat Exchanger

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

Bragg-Sitton, S.M.; Propulsion Research Center, NASA Marshall Space Flight Center, Huntsville, AL 35812; Kapernick, R.

2004-02-04

Experiments have been designed to characterize the coolant gas flow in two space reactor concepts that are currently under investigation by NASA Marshall Space Flight Center and Los Alamos National Laboratory: the direct-drive gas-cooled reactor (DDG) and the SAFE-100 heatpipe-cooled reactor (HPR). For the DDG concept, initial tests have been completed to measure pressure drop versus flow rate for a prototypic core flow channel, with gas exiting to atmospheric pressure conditions. The experimental results of the completed DDG tests presented in this paper validate the predicted results to within a reasonable margin of error. These tests have resulted in amore » re-design of the flow annulus to reduce the pressure drop. Subsequent tests will be conducted with the re-designed flow channel and with the outlet pressure held at 150 psi (1 MPa). Design of a similar test for a nominal flow channel in the HPR heat exchanger (HPR-HX) has been completed and hardware is currently being assembled for testing this channel at 150 psi. When completed, these test programs will provide the data necessary to validate calculated flow performance for these reactor concepts (pressure drop and film temperature rise)« less

Optimization of post-column reactor radius in capillary high performance liquid chromatography Effect of chromatographic column diameter and particle diameter

PubMed Central

Xu, Hongjuan; Weber, Stephen G.

2006-01-01

A post-column reactor consisting of a simple open tube (Capillary Taylor Reactor) affects the performance of a capillary LC in two ways: stealing pressure from the column and adding band spreading. The former is a problem for very small radius reactors, while the latter shows itself for large reactor diameters. We derived an equation that defines the observed number of theoretical plates (Nobs) taking into account the two effects stated above. Making some assumptions and asserting certain conditions led to a final equation with a limited number of variables, namely chromatographic column radius, reactor radius and chromatographic particle diameter. The assumptions and conditions are that the van Deemter equation applies, the mass transfer limitation is for intraparticle diffusion in spherical particles, the velocity is at the optimum, the analyte’s retention factor, k′, is zero, the post-column reactor is only long enough to allow complete mixing of reagents and analytes and the maximum operating pressure of the pumping system is used. Optimal ranges of the reactor radius (ar) are obtained by comparing the number of observed theoretical plates (and theoretical plates per time) with and without a reactor. Results show that the acceptable reactor radii depend on column diameter, particle diameter, and maximum available pressure. Optimal ranges of ar become narrower as column diameter increases, particle diameter decreases or the maximum pressure is decreased. When the available pressure is 4000 psi, a Capillary Taylor Reactor with 12 μm radius is suitable for all columns smaller than 150 μm (radius) packed with 2–5 μm particles. For 1 μm packing particles, only columns smaller than 42.5 μm (radius) can be used and the reactor radius needs to be 5 μm. PMID:16494886

Hydrogen sulfide oxidation by a microbial consortium in a recirculation reactor system: sulfur formation under oxygen limitation and removal of phenols.

PubMed

Alcantara, Sergio; Velasco, Antonio; Muñoz, Ana; Cid, Juan; Revah, Sergio; Razo-Flores, Elías

2004-02-01

Wastewater from petroleum refining may contain a number of undesirable contaminants including sulfides, phenolic compounds, and ammonia. The concentrations of these compounds must be reduced to acceptable levels before discharge. Sulfur formation and the effect of selected phenolic compounds on the sulfide oxidation were studied in autotrophic aerobic cultures. A recirculation reactor system was implemented to improve the elemental sulfur recovery. The relation between oxygen and sulfide was determined calculating the O2/S2- loading rates (Q(O2)/Q(S)2- = Rmt), which adequately defined the operation conditions to control the sulfide oxidation. Sulfur-producing steady states were achieved at Rmt ranging from 0.5 to 1.5. The maximum sulfur formation occurred at Rmt of 0.5 where 85% of the total sulfur added to the reactor as sulfide was transformed to elemental sulfur and 90% of it was recovered from the bottom of the reactor. Sulfide was completely oxidized to sulfate (Rmt of 2) in a stirred tank reactor, even when a mixture of phenolic compounds was present in the medium. Microcosm experiments showed that carbon dioxide production increased in the presence of the phenols, suggesting that these compounds were oxidized and that they may have been used as carbon and energy source by heterotrophic microorganisms present in the consortium.

Nitrate removal from high strength nitrate-bearing wastes in granular sludge sequencing batch reactors.

PubMed

Krishna Mohan, Tulasi Venkata; Renu, Kadali; Nancharaiah, Yarlagadda Venkata; Satya Sai, Pedapati Murali; Venugopalan, Vayalam Purath

2016-02-01

A 6-L sequencing batch reactor (SBR) was operated for development of granular sludge capable of denitrification of high strength nitrates. Complete and stable denitrification of up to 5420 mg L(-1) nitrate-N (2710 mg L(-1) nitrate-N in reactor) was achieved by feeding simulated nitrate waste at a C/N ratio of 3. Compact and dense denitrifying granular sludge with relatively stable microbial community was developed during reactor operation. Accumulation of large amounts of nitrite due to incomplete denitrification occurred when the SBR was fed with 5420 mg L(-1) NO3-N at a C/N ratio of 2. Complete denitrification could not be achieved at this C/N ratio, even after one week of reactor operation as the nitrite levels continued to accumulate. In order to improve denitrification performance, the reactor was fed with nitrate concentrations of 1354 mg L(-1), while keeping C/N ratio at 2. Subsequently, nitrate concentration in the feed was increased in a step-wise manner to establish complete denitrification of 5420 mg L(-1) NO3-N at a C/N ratio of 2. The results show that substrate concentration plays an important role in denitrification of high strength nitrate by influencing nitrite accumulation. Complete denitrification of high strength nitrates can be achieved at lower substrate concentrations, by an appropriate acclimatization strategy. Copyright © 2015 The Society for Biotechnology, Japan. Published by Elsevier B.V. All rights reserved.

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

Pauzi, Anas Muhamad, E-mail: Anas@uniten.edu.my; Cioncolini, Andrea; Iacovides, Hector

The Molten Salt Reactor (MSR) is one of the most promising system proposed by Generation IV Forum (GIF) for future nuclear reactor systems. Advantages of the MSR are significantly larger compared to other reactor system, and is mainly achieved from its liquid nature of fuel and coolant. Further improvement to this system, which is a natural circulating molten fuel salt inside its tube in the reactor core is proposed, to achieve advantages of reducing and simplifying the MSR design proposed by GIF. Thermal hydraulic analysis on the proposed system was completed using a commercial computation fluid dynamics (CFD) software calledmore » FLUENT by ANSYS Inc. An understanding on theory behind this unique natural circulation flow inside the tube caused by fission heat generated in molten fuel salt and tube cooling was briefly introduced. Currently, no commercial CFD software could perfectly simulate natural circulation flow, hence, modeling this flow problem in FLUENT is introduced and analyzed to obtain best simulation results. Results obtained demonstrate the existence of periodical transient nature of flow problem, hence improvements in tube design is proposed based on the analysis on temperature and velocity profile. Results show that the proposed system could operate at up to 750MW core power, given that turbulence are enhanced throughout flow region, and precise molten fuel salt physical properties could be defined. At the request of the authors and the Proceedings Editor the name of the co-author Andrea Cioncolini was corrected from Andrea Coincolini. The same name correction was made in the Acknowledgement section on page 030004-10 and in reference number 4. The updated article was published on 11 May 2015.« less

Milestone M3FT-15OR0203112. Build redesigned HFIR rabbit capsules and make ready for insertion for irradiation in HFIR

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

Howard, Richard H; McDuffee, Joel Lee; Okuniewski, Maria A.

2015-09-01

This report details the fabrication and delivery of two Fuel Cycle Research and Development irradiation capsules (FCRP20 and FCRP03), with associated quality assurance documentation, to the High Flux Isotope Reactor. The capsules and documentation were delivered by September 30, 2015, thus meeting the deadline for milestone M3FT-15OR0203112. These irradiation experiments irradiate metal parallelepiped specimens that may consist of various compositions including uranium metal, steel, etc. This document contains a copy of the completed capsule fabrication request sheets, which detail all constituent components, pertinent drawings, etc., along with a detailed summary of the capsule assembly process performed by the Thermal Hydraulicsmore » and Irradiation Engineering Group (THIEG) in the Reactor and Nuclear Systems Division. A complete fabrication package record is maintained by THIEG and is available upon request.« less

FROM CONCEPT TO REALITY, IN-SITU DECOMMISSIONING OF THE P AND R REACTORS AT THE SAVANNAH RIVER SITE

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

Musall, J.; Blankenship, J.; Griffin, W.

2012-01-09

SRS recently completed an approximately three year effort to decommission two SRS reactors: P-Reactor (Building 105-P) and R-Reactor (Building 105-R). Completed in December 2011, the concurrent decommissionings marked the completion of two relatively complex and difficult facility disposition projects at the SRS. Buildings 105-P and 105-R began operating as production reactors in the early 1950s with the mission of producing weapons material (e.g., tritium and plutonium-239). The 'P' Reactor and was shutdown in 1991 while the 'R' Reactor and was shutdown in 1964. In the intervening period between shutdown and deactivation & decommissioning (D&D), Buildings 105-P and 105-R saw limitedmore » use (e.g., storage of excess heavy water and depleted uranium oxide). For Building 105-P, deactivation was initiated in April 2007 and was essentially complete by June 2010. For Building 105-R, deactivation was initiated in August 2008 and was essentially complete by September 2010. For both buildings, the primary objective of deactivation was to remove/mitigate hazards associated with the remaining hazardous materials, and thus prepare the buildings for in-situ decommissioning. Deactivation removed the following hazardous materials to the extent practical: combustibles/flammables, residual heavy water, acids, friable asbestos (as needed to protect workers performing deactivation and decommissioning), miscellaneous chemicals, lead/brass components, Freon(reg sign), oils, mercury/PCB containing components, mold and some radiologically-contaminated equipment. In addition to the removal of hazardous materials, deactivation included the removal of hazardous energy, exterior metallic components (representing an immediate fall hazard), and historical artifacts along with the evaporation of water from the two Disassembly Basins. Finally, so as to facilitate occupancy during the subsequent in-situ decommissioning, deactivation implemented repairs to the buildings and provided temporary power.« less

Cometabolic degradation of ethyl mercaptan by phenol-utilizing Ralstonia eutropha in suspended growth and gas-recycling trickle-bed reactor.

PubMed

Sedighi, Mahsa; Zamir, Seyed Morteza; Vahabzadeh, Farzaneh

2016-01-01

The degradability of ethyl mercaptan (EM), by phenol-utilizing cells of Ralstonia eutropha, in both suspended and immobilized culture systems, was investigated in the present study. Free-cells experiments conducted at EM concentrations ranging from 1.25 to 14.42 mg/l, showed almost complete removal of EM at concentrations below 10.08 mg/l, which is much higher than the maximum biodegradable EM concentration obtained in experiments that did not utilize phenol as the primary substrate, i.e. 2.5 mg/l. The first-order kinetic rate constant (kSKS) for EM biodegradation by the phenol-utilizing cells (1.7 l/g biomass/h) was about 10 times higher than by cells without phenol utilization. Immobilized-cells experiments performed in a gas recycling trickle-bed reactor packed with kissiris particles at EM concentrations ranging from 1.6 to 36.9 mg/l, showed complete removal at all tested concentrations in a much shorter time, compared with free cells. The first-order kinetic rate constant (rmaxKs) for EM utilization was 0.04 l/h for the immobilized system compared to 0.06 for the suspended-growth culture, due to external mass transfer diffusion. Diffusion limitation was decreased by increasing the recycling-liquid flow rate from 25 to 65 ml/min. The removed EM was almost completely mineralized according to TOC and sulfate measurements. Shut down and starvation experiments revealed that the reactor could effectively handle the starving conditions and was reliable for full-scale application. Copyright © 2015 Elsevier Ltd. All rights reserved.

Anaerobic Biodegradation Of Methyl tert-Butyl Ether Under Iron-Reducing Conditions In Batch And Continuous-Flow Cultures

EPA Science Inventory

The feasibility of biodegradation of the fuel oxygenate methyl tert-butyl ether (MTBE) under iron-reducing conditions was explored in batch and continuous-flow systems. A porous pot completely-mixed reactor was seeded with diverse cultures and operated under iron-reducing...

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

NONE

The objective of Task I is to prepare and evaluate catalysts and to develop efficient reactor systems for the selective conversion of hydrogen-lean synthesis gas to alcohol fuel extenders and octane enhancers. In Task 1, during this reporting period, we encountered and solved a problem in the analysis of the reaction products containing a small amount of heavy components. Subsequently, we continued with the major thrusts of the program. We analyzed the results from our preliminary studies on the packed-bed membrane reactor using the BASF methanol synthesis catalyst. We developed a quantitative model to describe the performance of the reactor.more » The effect of varying permeances and the effect of catalyst aging are being incorporated into the model. Secondly, we resumed our more- detailed parametric studies on selected non-sulfide Mo-based catalysts. Finally, we continue with the analysis of data from the kinetic study of a sulfided carbon-supported potassium-doped molybdenum-cobalt catalyst in the Rotoberty reactor. We have completed catalyst screening at UCC. The complete characterization of selected catalysts has been started. In Task 2, the fuel blends of alcohol and unleaded test gas 96 (UTG 96) have been made and tests have been completed. The testing includes knock resistance tests and emissions tests. Emissions tests were conducted when the engine was optimized for the particular blend being tested (i.e. where the engine produced the most power when running on the blend in question). The data shows that the presence of alcohol in the fuel increases the fuel`s ability to resist knock. Because of this, when the engine was optimized for use with alcohol blends, the engine produced more power and lower emission rates.« less

Biomass-derived Syngas Utilization for Fuels and Chemicals - Final Report

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

Dayton, David C

2010-03-24

Executive Summary The growing gap between petroleum production and demand, mounting environmental concerns, and increasing fuel prices have stimulated intense interest in research and development (R&D) of alternative fuels, both synthetic and bio-derived. Currently, the most technically defined thermochemical route for producing alternative fuels from lignocellulosic biomass involves gasification/reforming of biomass to produce syngas (carbon monoxide [CO] + hydrogen [H2]), followed by syngas cleaning, Fischer-Tropsch synthesis (FTS) or mixed alcohol synthesis, and some product upgrading via hydroprocessing or separation. A detailed techno-economic analysis of this type of process has recently been published [1] and it highlights the need for technicalmore » breakthroughs and technology demonstration for gas cleanup and fuel synthesis. The latter two technical barrier areas contribute 40% of the total thermochemical ethanol cost and 70% of the production cost, if feedstock costs are factored out. Developing and validating technologies that reduce the capital and operating costs of these unit operations will greatly reduce the risk for commercializing integrated biomass gasification/fuel synthesis processes for biofuel production. The objective of this project is to develop and demonstrate new catalysts and catalytic processes that can efficiently convert biomass-derived syngas into diesel fuel and C2-C4 alcohols. The goal is to improve the economics of the processes by improving the catalytic activity and product selectivity, which could lead to commercialization. The project was divided into 4 tasks: Task 1: Reactor Systems: Construction of three reactor systems was a project milestone. Construction of a fixed-bed microreactor (FBR), a continuous stirred tank reactor (CSTR), and a slurry bubble column reactor (SBCR) were completed to meet this milestone. Task 2: Iron Fischer-Tropsch (FT) Catalyst: An attrition resistant iron FT catalyst will be developed and tested. Task 3: Chemical Synthesis: Promising process routes will be identified for synthesis of selected chemicals from biomass-derived syngas. A project milestone was to select promising mixed alcohol catalysts and screen productivity and performance in a fixed bed micro-reactor using bottled syngas. This milestone was successfully completed in collaboration withour catalyst development partner. Task 4: Modeling, Engineering Evaluation, and Commercial Assessment: Mass and energy balances of conceptual commercial embodiment for FT and chemical synthesis were completed.« less

Fabrication and Testing of a Modular Micro-Pocket Fission Detector Instrumentation System for Test Nuclear Reactors

NASA Astrophysics Data System (ADS)

Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Roberts, Jeremy A.; Unruh, Troy C.; McGregor, Douglas S.

2018-01-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node Micro-Pocket Fission Detector assembly will enhance nuclear research capabilities. A modular, four-node Micro-Pocket Fission Detector array was designed, fabricated, and tested at Kansas State University. The array was constructed from materials that do not significantly perturb the neutron flux in the reactor core. All four sensor nodes were equally spaced axially in the array to span the fuel-region of the reactor core. The array was filled with neon gas, serving as an ionization medium in the small cavities of the Micro-Pocket Fission Detectors. The modular design of the instrument facilitates the testing and deployment of numerous sensor arrays. The unified design drastically improved device ruggedness and simplified construction from previous designs. Five 8-mm penetrations in the upper grid plate of the Kansas State University TRIGA Mk. II research nuclear reactor were utilized to deploy the array between fuel elements in the core. The Micro-Pocket Fission Detector array was coupled to an electronic support system which has been specially developed to support pulse-mode operation. The Micro-Pocket Fission Detector array composed of four sensors was used to monitor local neutron flux at a constant reactor power of 100 kWth at different axial locations simultaneously. The array was positioned at five different radial locations within the core to emulate the deployment of multiple arrays and develop a 2-dimensional measurement of neutron flux in the reactor core.

Application of a fuzzy logic control system for continuous anaerobic digestion of low buffered, acidic energy crops as mono-substrate.

PubMed

Scherer, P; Lehmann, K; Schmidt, O; Demirel, B

2009-02-15

A fuzzy logic control (FLC) system was developed at the Hamburg University of Applied Sciences (HAW Hamburg) for operation of biogas reactors running on energy crops. Three commercially available measuring parameters, namely pH, the methane (CH4) content, and the specific gas production rate (spec. GPR = m(3)/kg VS/day) were included. The objective was to avoid stabilization of pH with use of buffering supplements, like lime or manure. The developed FLC system can cover most of all applications, such as a careful start-up process and a gentle recovery strategy after a severe reactor failure, also enabling a process with a high organic loading rate (OLR) and a low hydraulic retention time (HRT), that is, a high throughput anaerobic digestion process with a stable pH and CH4 content. A precondition for a high load process was the concept of interval feeding, for example, with 8 h of interval. The FLC system was proved to be reliable during the long term fermentation studies over 3 years in one-stage, completely stirred tank reactors (CSTR) with acidic beet silage as mono-input (pH 3.3-3.4). During fermentation of the fodder beet silage (FBS), a stable HRT of 6.0 days with an OLR of up to 15 kg VS/m(3)/day and a volumetric GPR of 9 m(3)/m(3)/day could be reached. The FLC enabled an automatic recovery of the digester after two induced severe reactor failures. In another attempt to prove the feasibility of the FLC, substrate FBS was changed to sugar beet silage (SBS), which had a substantially lower buffering capacity than that of the FBS. With SBS, the FLC accomplished a stable fermentation at a pH level between 6.5 and 6.6, and a volatile fatty acid level (VFA) below 500 mg/L, but the FLC had to interact and to change the substrate dosage permanently. In a further experiment, the reactor temperature was increased from 41 to 50 degrees C. Concomitantly, the specific GPR, pH and CH4 dropped down. Finally, the FLC automatically enabled a complete recovery in 16 days.

Implementing a Nuclear Power Plant Model for Evaluating Load-Following Capability on a Small Grid

NASA Astrophysics Data System (ADS)

Arda, Samet Egemen

A pressurized water reactor (PWR) nuclear power plant (NPP) model is introduced into Positive Sequence Load Flow (PSLF) software by General Electric in order to evaluate the load-following capability of NPPs. The nuclear steam supply system (NSSS) consists of a reactor core, hot and cold legs, plenums, and a U-tube steam generator. The physical systems listed above are represented by mathematical models utilizing a state variable lumped parameter approach. A steady-state control program for the reactor, and simple turbine and governor models are also developed. Adequacy of the isolated reactor core, the isolated steam generator, and the complete PWR models are tested in Matlab/Simulink and dynamic responses are compared with the test results obtained from the H. B. Robinson NPP. Test results illustrate that the developed models represents the dynamic features of real-physical systems and are capable of predicting responses due to small perturbations of external reactivity and steam valve opening. Subsequently, the NSSS representation is incorporated into PSLF and coupled with built-in excitation system and generator models. Different simulation cases are run when sudden loss of generation occurs in a small power system which includes hydroelectric and natural gas power plants besides the developed PWR NPP. The conclusion is that the NPP can respond to a disturbance in the power system without exceeding any design and safety limits if appropriate operational conditions, such as achieving the NPP turbine control by adjusting the speed of the steam valve, are met. In other words, the NPP can participate in the control of system frequency and improve the overall power system performance.

In Situ and ex Situ Catalytic Pyrolysis of Pine in a Bench-Scale Fluidized Bed Reactor System

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

Iisa, Kristiina; French, Richard J.; Orton, Kellene A.

In situ and ex situ catalytic pyrolysis were compared in a system with two 2-in. bubbling fluidized bed reactors. Pine was pyrolyzed in the system with a catalyst, HZSM-5 with a silica-to-alumina ratio of 30, placed either in the first (pyrolysis) reactor or the second (upgrading) reactor. Both the pyrolysis and upgrading temperatures were 500 degrees C, and the weight hourly space velocity was 1.1 h -1. Five catalytic cycles were completed in each experiment. The catalytic cycles were continued until oxygenates in the vapors became dominant. The catalyst was then oxidized, after which a new catalytic cycle was begun.more » The in situ configuration gave slightly higher oil yield but also higher oxygen content than the ex situ configuration, which indicates that the catalyst deactivated faster in the in situ configuration than the ex situ configuration. Analysis of the spent catalysts confirmed higher accumulation of metals in the in situ experiment. In all experiments, the organic oil mass yields varied between 14 and 17% and the carbon efficiencies between 20 and 25%. The organic oxygen concentrations in the oils were 16-18%, which represented a 45% reduction compared to corresponding noncatalytic pyrolysis oils prepared in the same fluidized bed reactor system. GC/MS analysis showed the oils to contain one- to four-ring aromatic hydrocarbons and a variety of oxygenates (phenols, furans, benzofurans, methoxyphenols, naphthalenols, indenols). Lastly, high fractions of oxygen were rejected as water, CO, and CO 2, which indicates the importance of dehydration, decarbonylation, and decarboxylation reactions. Light gases were the major sources of carbon losses, followed by char and coke.« less

In Situ and ex Situ Catalytic Pyrolysis of Pine in a Bench-Scale Fluidized Bed Reactor System

DOE PAGES

Iisa, Kristiina; French, Richard J.; Orton, Kellene A.; ...

2016-02-03

In situ and ex situ catalytic pyrolysis were compared in a system with two 2-in. bubbling fluidized bed reactors. Pine was pyrolyzed in the system with a catalyst, HZSM-5 with a silica-to-alumina ratio of 30, placed either in the first (pyrolysis) reactor or the second (upgrading) reactor. Both the pyrolysis and upgrading temperatures were 500 degrees C, and the weight hourly space velocity was 1.1 h -1. Five catalytic cycles were completed in each experiment. The catalytic cycles were continued until oxygenates in the vapors became dominant. The catalyst was then oxidized, after which a new catalytic cycle was begun.more » The in situ configuration gave slightly higher oil yield but also higher oxygen content than the ex situ configuration, which indicates that the catalyst deactivated faster in the in situ configuration than the ex situ configuration. Analysis of the spent catalysts confirmed higher accumulation of metals in the in situ experiment. In all experiments, the organic oil mass yields varied between 14 and 17% and the carbon efficiencies between 20 and 25%. The organic oxygen concentrations in the oils were 16-18%, which represented a 45% reduction compared to corresponding noncatalytic pyrolysis oils prepared in the same fluidized bed reactor system. GC/MS analysis showed the oils to contain one- to four-ring aromatic hydrocarbons and a variety of oxygenates (phenols, furans, benzofurans, methoxyphenols, naphthalenols, indenols). Lastly, high fractions of oxygen were rejected as water, CO, and CO 2, which indicates the importance of dehydration, decarbonylation, and decarboxylation reactions. Light gases were the major sources of carbon losses, followed by char and coke.« less

10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Physical Protection of Irradiated Reactor Fuel in Transit... Irradiated Reactor Fuel in Transit, Training Program Subject Schedule Pursuant to the provision of § 73.37 of... reactor fuel is required to assure that individuals used as shipment escorts have completed a training...

10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Physical Protection of Irradiated Reactor Fuel in Transit... Irradiated Reactor Fuel in Transit, Training Program Subject Schedule Pursuant to the provision of § 73.37 of... reactor fuel is required to assure that individuals used as shipment escorts have completed a training...

10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems required...

10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems required...

Condensation model for the ESBWR passive condensers

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

Revankar, S. T.; Zhou, W.; Wolf, B.

2012-07-01

In the General Electric's Economic simplified boiling water reactor (GE-ESBWR) the passive containment cooling system (PCCS) plays a major role in containment pressure control in case of an loss of coolant accident. The PCCS condenser must be able to remove sufficient energy from the reactor containment to prevent containment from exceeding its design pressure following a design basis accident. There are three PCCS condensation modes depending on the containment pressurization due to coolant discharge; complete condensation, cyclic venting and flow through mode. The present work reviews the models and presents model predictive capability along with comparison with existing data frommore » separate effects test. The condensation models in thermal hydraulics code RELAP5 are also assessed to examine its application to various flow modes of condensation. The default model in the code predicts complete condensation well, and basically is Nusselt solution. The UCB model predicts through flow well. None of condensation model in RELAP5 predict complete condensation, cyclic venting, and through flow condensation consistently. New condensation correlations are given that accurately predict all three modes of PCCS condensation. (authors)« less

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

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

Stoner, K.J.

1999-11-05

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

A KINETIC MODEL FOR H2O2/UV PROCESS IN A COMPLETELY MIXED BATCH REACTOR. (R825370C076)

EPA Science Inventory

A dynamic kinetic model for the advanced oxidation process (AOP) using hydrogen peroxide and ultraviolet irradiation (H₂O₂/UV) in a completely mixed batch reactor (CMBR) is developed. The model includes the known elementary chemical and photochemical reac...

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

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

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

2004-02-04

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

Design and Status of RERTR Irradiation Tests in the Advanced Test Reactor

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

Daniel M. Wachs; Richard G. Ambrosek; Gray Chang

2006-10-01

Irradiation testing of U-Mo based fuels is the central component of the Reduced Enrichment for Research and Test Reactors (RERTR) program fuel qualification plan. Several RERTR tests have recently been completed or are planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory in Idaho Falls, ID. Four mini-plate experiments in various stages of completion are described in detail, including the irradiation test design, objectives, and irradiation conditions. Observations made during and after the in-reactor RERTR-7A experiment breach are summarized. The irradiation experiment design and planned irradiation conditions for full-size plate test are described. Progressmore » toward element testing will be reviewed.« less

Availability analysis of an HTGR fuel recycle facility. Summary report

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

Sharmahd, J.N.

1979-11-01

An availability analysis of reprocessing systems in a high-temperature gas-cooled reactor (HTGR) fuel recycle facility was completed. This report summarizes work done to date to define and determine reprocessing system availability for a previously planned HTGR recycle reference facility (HRRF). Schedules and procedures for further work during reprocessing development and for HRRF design and construction are proposed in this report. Probable failure rates, transfer times, and repair times are estimated for major system components. Unscheduled down times are summarized.

PATHFINDER ATOMIC POWER PLANT TECHNICAL PROGRESS REPORT FOR JULY 1, 1959- SEPTEMBER 30, 1959

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

None

1960-10-31

ABS>Fuel Element Research and Development. Dynamic and static corrosion tests on 8001 Al were completed. Annealmmmg of 1100 cladding on 5083 and M400 cladding on X2219 were tested at 500 deg C, and investigation continued on producing X8101 Al alloy cladding in tube plates by extrusion. Boiler fuel element capsule irradiation tests and subassembly tests are described Heat transfer loop studies and fuel fabrication for the critical facility are reported. Boiler fuel element mechanical design and testing progress is desc ribed. and the superheater fuel element temperature evaluating routine is discussed. Low- enrichment superheater fuel element development included design studiesmore » and stainless steel powder and UO/sub 2/ powder fabrication studies Reactor Mechanical Studies. Research is reported on vessel and structure design, fabrication, and testing, recirculation system design, steam separator tests, and control rod studies. Nuclear Analysis. Reactor physics studies are reported on nuclear constants, baffle plate analysis, comparison of core representations, delayed neutron fraction. and shielding analysis of the reactor building. Reactor and system dynamics and critical experiments were also studied. Chemistry. Progress is reported on recombiner. radioactive gas removal and storage, ion exchanger and radiochemical processing. (For preceding period see ACNP-5915.) (T.R.H.)« less

Overview of the Westinghouse Small Modular Reactor building layout

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

Cronje, J. M.; Van Wyk, J. J.; Memmott, M. J.

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the third in a series of four papers, which describe the design and functionality of the Westinghouse SMR. It focuses in particular upon the plant building layout and modular design of the Westinghouse SMR. In the development of small modular reactors, the building layout is an area where the safety of themore » plant can be improved by applying new design approaches. This paper will present an overview of the Westinghouse SMR building layout and indicate how the design features improve the safety and robustness of the plant. The Westinghouse SMR is designed with no shared systems between individual reactor units. The main buildings inside the security fence are the nuclear island, the rad-waste building, the annex building, and the turbine building. All safety related equipment is located in the nuclear island, which is a seismic class 1 building. To further enhance the safety and robustness of the design, the reactor, containment, and most of the safety related equipment are located below grade on the nuclear island. This reduces the possibility of severe damage from external threats or natural disasters. Two safety related ultimate heat sink (UHS) water tanks that are used for decay heat removal are located above grade, but are redundant and physically separated as far as possible for improved safety. The reactor and containment vessel are located below grade in the center of the nuclear island. The rad-waste and other radioactive systems are located on the bottom floors to limit the radiation exposure to personnel. The Westinghouse SMR safety trains are completely separated into four unconnected quadrants of the building, with access between quadrants only allowed above grade. This is an improvement to conventional reactor design since it prevents failures of multiple trains during floods or fires and other external events. The main control room is located below grade, with a remote shutdown room in a different quadrant. All defense in depth systems are placed on the nuclear island, primarily above grade, while the safety systems are located on lower floors. The economics of the Westinghouse SMR challenges the established approach of large Light Water Reactors (LWR) that utilized the economies of scale to reach economic competitiveness. To serve the market expectation of smaller capital investment and cost competitive energy, a modular design approach is implemented within the Westinghouse SMR. The Westinghouse SMR building layout integrates the three basic design constraints of modularization; transportation, handling and module-joining technology. (authors)« less

Westinghouse Small Modular Reactor nuclear steam supply system design

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

Memmott, M. J.; Harkness, A. W.; Van Wyk, J.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development andmore » integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam generator, and eight reactor coolant pumps (RCP). The containment vessel is 27.1 m (89 ft) long and 9.8 m (32 ft) in diameter, and is designed to withstand pressures up to 1.7 MPa (250 psi). It is completely submerged in a pool of water serving as a heat sink and radiation shield. Housed within the containment are four combined core makeup tanks (CMT)/passive residual heat removal (PRHR) heat exchangers, two in-containment pools (ICP), two ICP tanks and four valves which function as the automatic depressurization system (ADS). The PRHR heat exchangers are thermally connected to two different ultimate heat sink (UHS) tanks which provide transient cooling capabilities. (authors)« less

10 CFR 2.603 - Acceptance and docketing of application for early review of site suitability issues in a...

Code of Federal Regulations, 2012 CFR

2012-01-01

... processing, the Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor... days. (b)(1) The Director of the Office of New Reactors or the Director of the Office of Nuclear... Nuclear Reactor Regulation, as appropriate, that they are complete. (c) If part one of the application is...

Cross-Section Measurements in the Fast Neutron Energy Range

NASA Astrophysics Data System (ADS)

Plompen, Arjan

2006-04-01

Generation IV focuses research for advanced nuclear reactors on six concepts. Three of these concepts, the lead, gas and sodium fast reactors (LFR, GFR and SFR) have fast neutron spectra, whereas a fourth, the super-critical water reactor (SCWR), can be configured to have a fast spectrum. Such fast neutron spectra are essential to meet the sustainability objective of GenIV. Nuclear data requirements for GenIV concepts will therefore emphasize the energy region from about 1 keV to 10 MeV. Here, the potential is illustrated of the GELINA neutron time-of-flight facility and the Van de Graaff laboratory at IRMM to measure the relevant nuclear data in this energy range: the total, capture, fission and inelastic-scattering cross sections. In particular, measurement results will be shown for lead and bismuth inelastic scattering for which the need was recently expressed in a quantitative way by Aliberti et al. for Accelerator Driven Systems. Even without completion of the quantitative assessment of the data needs for GenIV concepts at ANL it is clear that this particular effort is of relevance to LFR system studies.

Nuclear Engine System Simulation (NESS). Volume 1: Program user's guide

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.

1993-03-01

A Nuclear Thermal Propulsion (NTP) engine system design analysis tool is required to support current and future Space Exploration Initiative (SEI) propulsion and vehicle design studies. Currently available NTP engine design models are those developed during the NERVA program in the 1960's and early 1970's and are highly unique to that design or are modifications of current liquid propulsion system design models. To date, NTP engine-based liquid design models lack integrated design of key NTP engine design features in the areas of reactor, shielding, multi-propellant capability, and multi-redundant pump feed fuel systems. Additionally, since the SEI effort is in the initial development stage, a robust, verified NTP analysis design tool could be of great use to the community. This effort developed an NTP engine system design analysis program (tool), known as the Nuclear Engine System Simulation (NESS) program, to support ongoing and future engine system and stage design study efforts. In this effort, Science Applications International Corporation's (SAIC) NTP version of the Expanded Liquid Engine Simulation (ELES) program was modified extensively to include Westinghouse Electric Corporation's near-term solid-core reactor design model. The ELES program has extensive capability to conduct preliminary system design analysis of liquid rocket systems and vehicles. The program is modular in nature and is versatile in terms of modeling state-of-the-art component and system options as discussed. The Westinghouse reactor design model, which was integrated in the NESS program, is based on the near-term solid-core ENABLER NTP reactor design concept. This program is now capable of accurately modeling (characterizing) a complete near-term solid-core NTP engine system in great detail, for a number of design options, in an efficient manner. The following discussion summarizes the overall analysis methodology, key assumptions, and capabilities associated with the NESS presents an example problem, and compares the results to related NTP engine system designs. Initial installation instructions and program disks are in Volume 2 of the NESS Program User's Guide.

Nuclear Engine System Simulation (NESS). Volume 1: Program user's guide

NASA Technical Reports Server (NTRS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.

1993-01-01

A Nuclear Thermal Propulsion (NTP) engine system design analysis tool is required to support current and future Space Exploration Initiative (SEI) propulsion and vehicle design studies. Currently available NTP engine design models are those developed during the NERVA program in the 1960's and early 1970's and are highly unique to that design or are modifications of current liquid propulsion system design models. To date, NTP engine-based liquid design models lack integrated design of key NTP engine design features in the areas of reactor, shielding, multi-propellant capability, and multi-redundant pump feed fuel systems. Additionally, since the SEI effort is in the initial development stage, a robust, verified NTP analysis design tool could be of great use to the community. This effort developed an NTP engine system design analysis program (tool), known as the Nuclear Engine System Simulation (NESS) program, to support ongoing and future engine system and stage design study efforts. In this effort, Science Applications International Corporation's (SAIC) NTP version of the Expanded Liquid Engine Simulation (ELES) program was modified extensively to include Westinghouse Electric Corporation's near-term solid-core reactor design model. The ELES program has extensive capability to conduct preliminary system design analysis of liquid rocket systems and vehicles. The program is modular in nature and is versatile in terms of modeling state-of-the-art component and system options as discussed. The Westinghouse reactor design model, which was integrated in the NESS program, is based on the near-term solid-core ENABLER NTP reactor design concept. This program is now capable of accurately modeling (characterizing) a complete near-term solid-core NTP engine system in great detail, for a number of design options, in an efficient manner. The following discussion summarizes the overall analysis methodology, key assumptions, and capabilities associated with the NESS presents an example problem, and compares the results to related NTP engine system designs. Initial installation instructions and program disks are in Volume 2 of the NESS Program User's Guide.

A Microwave Thermostatic Reactor for Processing Liquid Materials Based on a Heat-Exchanger.

PubMed

Zhou, Yongqiang; Zhang, Chun; Xie, Tian; Hong, Tao; Zhu, Huacheng; Yang, Yang; Liu, Changjun; Huang, Kama

2017-10-08

Microwaves have been widely used in the treatment of different materials. However, the existing adjustable power thermostatic reactors cannot be used to analyze materials characteristics under microwave effects. In this paper, a microwave thermostatic chemical reactor for processing liquid materials is proposed, by controlling the velocity of coolant based on PLC (programmable logic controller) in different liquid under different constant electric field intensity. A nonpolar coolant (Polydimethylsiloxane), which is completely microwave transparent, is employed to cool the liquid materials. Experiments are performed to measure the liquid temperature using optical fibers, the results show that the precision of temperature control is at the range of ±0.5 °C. Compared with the adjustable power thermostatic control system, the effect of electric field changes on material properties are avoided and it also can be used to detect the properties of liquid materials and special microwave effects.

A Microwave Thermostatic Reactor for Processing Liquid Materials Based on a Heat-Exchanger

PubMed Central

Zhou, Yongqiang; Zhang, Chun; Xie, Tian; Hong, Tao; Yang, Yang; Liu, Changjun; Huang, Kama

2017-01-01

Microwaves have been widely used in the treatment of different materials. However, the existing adjustable power thermostatic reactors cannot be used to analyze materials characteristics under microwave effects. In this paper, a microwave thermostatic chemical reactor for processing liquid materials is proposed, by controlling the velocity of coolant based on PLC (programmable logic controller) in different liquid under different constant electric field intensity. A nonpolar coolant (Polydimethylsiloxane), which is completely microwave transparent, is employed to cool the liquid materials. Experiments are performed to measure the liquid temperature using optical fibers, the results show that the precision of temperature control is at the range of ±0.5 °C. Compared with the adjustable power thermostatic control system, the effect of electric field changes on material properties are avoided and it also can be used to detect the properties of liquid materials and special microwave effects. PMID:28991195

Role of nickel in high rate methanol degradation in anaerobic granular sludge bioreactors

PubMed Central

Fermoso, Fernando G.; Collins, Gavin; Bartacek, Jan; O’Flaherty, Vincent

2008-01-01

The effect of nickel deprivation from the influent of a mesophilic (30°C) methanol fed upflow anaerobic sludge bed (UASB) reactor was investigated by coupling the reactor performance to the evolution of the Methanosarcina population of the bioreactor sludge. The reactor was operated at pH 7.0 and an organic loading rate (OLR) of 5–15 g COD l−1 day−1 for 191 days. A clear limitation of the specific methanogenic activity (SMA) on methanol due to the absence of nickel was observed after 129 days of bioreactor operation: the SMA of the sludge in medium with the complete trace metal solution except nickel amounted to 1.164 (±0.167) g CH4-COD g VSS−1 day−1 compared to 2.027 (±0.111) g CH4-COD g VSS−1 day−1 in a medium with the complete (including nickel) trace metal solution. The methanol removal efficiency during these 129 days was 99%, no volatile fatty acid (VFA) accumulation was observed and the size of the Methanosarcina population increased compared to the seed sludge. Continuation of the UASB reactor operation with the nickel limited sludge lead to incomplete methanol removal, and thus methanol accumulation in the reactor effluent from day 142 onwards. This methanol accumulation subsequently induced an increase of the acetogenic activity in the UASB reactor on day 160. On day 165, 77% of the methanol fed to the system was converted to acetate and the Methanosarcina population size had substantially decreased. Inclusion of 0.5 μM Ni (dosed as NiCl2) to the influent from day 165 onwards lead to the recovery of the methanol removal efficiency to 99% without VFA accumulation within 2 days of bioreactor operation. PMID:18247139

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

James K. Neathery; Gary Jacobs; Amitava Sarkar

In the previous reporting period, modifications were completed for integrating a continuous wax filtration system for a 4 liter slurry bubble column reactor. During the current reporting period, a shakedown of the system was completed. Several problems were encountered with the progressive cavity pump used to circulate the wax/catalyst slurry though the cross-flow filter element and reactor. During the activation of the catalyst with elevated temperature (> 270 C) the elastomer pump stator released sulfur thereby totally deactivating the iron-based catalyst. Difficulties in maintaining an acceptable leak rate from the pump seal and stator housing were also encountered. Consequently, themore » system leak rate exceeded the expected production rate of wax; therefore, no online filtration could be accomplished. Work continued regarding the characterization of ultra-fine catalyst structures. The effect of carbidation on the morphology of iron hydroxide oxide particles was the focus of the study during this reporting period. Oxidation of Fe (II) sulfate results in predominantly {gamma}-FeOOH particles which have a rod-shaped (nano-needles) crystalline structure. Carbidation of the prepared {gamma}-FeOOH with CO at atmospheric pressure produced iron carbides with spherical layered structure. HRTEM and EDS analysis revealed that carbidation of {gamma}-FeOOH particles changes the initial nano-needles morphology and generates ultrafine carbide particles with irregular spherical shape.« less

Trace Assessment for BWR ATWS Analysis

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

Cheng, L.Y.; Diamond, D.; Arantxa Cuadra, Gilad Raitses, Arnold Aronson

2010-04-22

A TRACE/PARCS input model has been developed in order to be able to analyze anticipated transients without scram (ATWS) in a boiling water reactor. The model is based on one developed previously for the Browns Ferry reactor for doing loss-of-coolant accident analysis. This model was updated by adding the control systems needed for ATWS and a core model using PARCS. The control systems were based on models previously developed for the TRAC-B code. The PARCS model is based on information (e.g., exposure and moderator density (void) history distributions) obtained from General Electric Hitachi and cross sections for GE14 fuel obtainedmore » from an independent source. The model is able to calculate an ATWS, initiated by the closure of main steam isolation valves, with recirculation pump trip, water level control, injection of borated water from the standby liquid control system and actuation of the automatic depres-surization system. The model is not considered complete and recommendations are made on how it should be improved.« less

Fast reactor power plant design having heat pipe heat exchanger

DOEpatents

Huebotter, P.R.; McLennan, G.A.

1984-08-30

The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

Fast reactor power plant design having heat pipe heat exchanger

DOEpatents

Huebotter, Paul R.; McLennan, George A.

1985-01-01

The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

Evaluation of Municipal Wastewater Treatment Plant Activated Sludge for Biodegradation of Propylene Glycol as an Aircraft Deicing Fluid

DTIC Science & Technology

2012-03-01

Propylene Glycol Deicer Biodegredation Kinetics: Complete-Mix Stirred Tank Reactors , Filter, and Fluidized Bed . Journal of Environmental...scale sequencing batch reactor containing municipal waste water treatment facility activated sludge (AS) performing simultaneous organic carbon...Sequencing Batch Reactor Operation ..................................................................... 13 PG extraction from AS

POWER-BURST FACILITY (PBF) CONCEPTUAL DESIGN

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

Wasserman, A.A.; Johnson, S.O.; Heffner, R.E.

1963-06-21

A description is presented of the conceptual design of a high- performance, pulsed reactor called the Power Burst Facility (PBF). This reactor is designed to generate power bursts with initial asymptotic periods as short as 1 msec, producing energy releases large enough to destroy entire fuel subassemblies placed in a capsule or flow loop mounted in the reactor, all without damage to the reactor itself. It will be used primarily to evaluate the consequences and hazards of very rapid destructive accidents in reactors representing the entire range of current nuclear technology as applied to power generation, propulsion, and testing. Itmore » will also be used to carry out detailed studies of nondestructive reactivity feedback mechanisms in the shortperiod domain. The facility was designed to be sufficiently flexible to accommodate future cores of even more advanced design. The design for the first reactor core is based upon proven technology; hence, completion of the final design of this core will involve no significant development delays. Construction of the PBF is proposed to begin in September 1984, and is expected to take approximately 20 months to complete. (auth)« less

Dismantling of Loop-Type Channel Equipment of MR Reactor in NRC 'Kurchatov Institute' - 13040

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

Volkov, Victor; Danilovich, Alexey; Zverkov, Yuri

2013-07-01

In 2009 the project of decommissioning of MR and RTF reactors was developed and approved by the Expert Authority of the Russian Federation (Gosexpertiza). The main objective of the decommissioning works identified in this project: - complete dismantling of reactor equipment and systems; - decontamination of reactor premises and site in accordance with the established sanitary and hygienic standards. At the preparatory stage (2008-2010) of the project the following works were executed: loop-type channels' dismantling in the storage pool; experimental fuel assemblies' removal from spent fuel repositories in the central hall; spent fuel assembly removal from the liquid-metal-cooled loop-type channelmore » of the reactor core and its placement into the SNF repository; and reconstruction of engineering support systems to the extent necessary for reactor decommissioning. The project assumes three main phases of dismantling and decontamination: - dismantling of equipment/pipelines of cooling circuits and loop-type channels, and auxiliary reactor equipment (2011-2012); - dismantling of equipment in underground reactor premises and of both MR and RTF in-vessel devices (2013-2014); - decontamination of reactor premises; rehabilitation of the reactor site; final radiation survey of reactor premises, loop-type channels and site; and issuance of the regulatory authorities' de-registration statement (2015). In 2011 the decommissioning license for the two reactors was received and direct MR decommissioning activities started. MR primary pipelines and loop-type facilities situated in the underground reactor hall were dismantled. Works were also launched to dismantle the loop-type channels' equipment in underground reactor premises; reactor buildings were reconstructed to allow removal of dismantled equipment; and the MR/RTF decommissioning sequence was identified. In autumn 2011 - spring 2012 results of dismantling activities performed are: - equipment from underground rooms (No. 66, 66A, 66B, 72, 64, 63) - as well as from water and gas loop corridors - was dismantled, with the total radwaste weight of 53 tons and the total removed activity of 5,0 x 10{sup 10} Bq; - loop-type channel equipment from underground reactor hall premises was dismantled; - 93 loop-type channels were characterized, chopped and removed, with radwaste of 2.6 x 10{sup 13} Bq ({sup 60}Co) and 1.5 x 10{sup 13} Bq ({sup 137}Cs) total activity removed from the reactor pool, fragmented and packaged. Some of this waste was placed into the high-level waste (HLW) repository of the Center. Dismantling works were executed with application of remotely operated mechanisms, which promoted decrease of radiation impact on the personnel. The average individual dose for the personnel was 1.9 mSv/year in 2011, and the collective dose is estimated as 0.0605 man x Sv/year. (authors)« less

DEMINERALIZER BUILDING, TRA608. INSTALLATION OF SAMPLING AND OTHER INSTRUMENTS COMPLETES ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

DEMINERALIZER BUILDING, TRA-608. INSTALLATION OF SAMPLING AND OTHER INSTRUMENTS COMPLETES DEMINERALIZER UNITS ALONG NORTH WALL. CAMERA FACES EAST. CARD IN LOWER RIGHT WAS INSERTED BY INL PHOTOGRAPHER TO COVER AN OBSOLETE SECURITY RESTRICTION PRINTED ON THE ORIGINAL NEGATIVE. INL NEGATIVE NO. 3996A. Unknown Photographer, 12/28/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Impact of thorium based molten salt reactor on the closure of the nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Jaradat, Safwan Qasim Mohammad

Molten salt reactor (MSR) is one of six reactors selected by the Generation IV International Forum (GIF). The liquid fluoride thorium reactor (LFTR) is a MSR concept based on thorium fuel cycle. LFTR uses liquid fluoride salts as a nuclear fuel. It uses 232Th and 233U as the fertile and fissile materials, respectively. Fluoride salt of these nuclides is dissolved in a mixed carrier salt of lithium and beryllium (FLiBe). The objective of this research was to complete feasibility studies of a small commercial thermal LFTR. The focus was on neutronic calculations in order to prescribe core design parameter such as core size, fuel block pitch (p), fuel channel radius, fuel path, reflector thickness, fuel salt composition, and power. In order to achieve this objective, the applicability of Monte Carlo N-Particle Transport Code (MCNP) to MSR modeling was verified. Then, a prescription for conceptual small thermal reactor LFTR and relevant calculations were performed using MCNP to determine the main neutronic parameters of the core reactor. The MCNP code was used to study the reactor physics characteristics for the FUJI-U3 reactor. The results were then compared with the results obtained from the original FUJI-U3 using the reactor physics code SRAC95 and the burnup analysis code ORIPHY2. The results were comparable with each other. Based on the results, MCNP was found to be a reliable code to model a small thermal LFTR and study all the related reactor physics characteristics. The results of this study were promising and successful in demonstrating a prefatory small commercial LFTR design. The outcome of using a small core reactor with a diameter/height of 280/260 cm that would operate for more than five years at a power level of 150 MWth was studied. The fuel system 7LiF - BeF2 - ThF4 - UF4 with a (233U/ 232Th) = 2.01 % was the candidate fuel for this reactor core.

Design Skills and Prototyping for Defense Systems

DTIC Science & Technology

2015-04-30

however, the utility of prototyping has had a demonstrably mixed record in defense acquisition. Some programs, such as the Manhattan Project , were...almost completely undefined. The first production reactors for the Manhattan Project suffered a near- catastrophic engineering design flaw stemming...architecture, as was seen in the F-117 and Manhattan Project development efforts. Architectural Prototyping Simply maintaining design teams or developing

The Behavior of Matter under Nonequilibrium Conditions: Fundamental Aspects and Applications in Energy-oriented Problems: Progress Report for Period September 1984--November 1987

DOE R&D Accomplishments Database

Prigogine, I.

1987-10-07

This report briefly discusses progress on the following topics: state selection dynamics; polymerization under nonequilibrium conditions; inhomogeneous fluctuations in hydrodynamics and in completely mixed reactors; homoclinic bifurcations and mixed-mode oscillations; intrinsic randomness and spontaneous symmetry breaking in explosive systems; and microscopic means of irreversibility.

Genome-based microbial ecology of anammox granules in a full-scale wastewater treatment system.

PubMed

Speth, Daan R; In 't Zandt, Michiel H; Guerrero-Cruz, Simon; Dutilh, Bas E; Jetten, Mike S M

2016-03-31

Partial-nitritation anammox (PNA) is a novel wastewater treatment procedure for energy-efficient ammonium removal. Here we use genome-resolved metagenomics to build a genome-based ecological model of the microbial community in a full-scale PNA reactor. Sludge from the bioreactor examined here is used to seed reactors in wastewater treatment plants around the world; however, the role of most of its microbial community in ammonium removal remains unknown. Our analysis yielded 23 near-complete draft genomes that together represent the majority of the microbial community. We assign these genomes to distinct anaerobic and aerobic microbial communities. In the aerobic community, nitrifying organisms and heterotrophs predominate. In the anaerobic community, widespread potential for partial denitrification suggests a nitrite loop increases treatment efficiency. Of our genomes, 19 have no previously cultivated or sequenced close relatives and six belong to bacterial phyla without any cultivated members, including the most complete Omnitrophica (formerly OP3) genome to date.

Genome-based microbial ecology of anammox granules in a full-scale wastewater treatment system

PubMed Central

Speth, Daan R.; in 't Zandt, Michiel H.; Guerrero-Cruz, Simon; Dutilh, Bas E.; Jetten, Mike S. M.

2016-01-01

Partial-nitritation anammox (PNA) is a novel wastewater treatment procedure for energy-efficient ammonium removal. Here we use genome-resolved metagenomics to build a genome-based ecological model of the microbial community in a full-scale PNA reactor. Sludge from the bioreactor examined here is used to seed reactors in wastewater treatment plants around the world; however, the role of most of its microbial community in ammonium removal remains unknown. Our analysis yielded 23 near-complete draft genomes that together represent the majority of the microbial community. We assign these genomes to distinct anaerobic and aerobic microbial communities. In the aerobic community, nitrifying organisms and heterotrophs predominate. In the anaerobic community, widespread potential for partial denitrification suggests a nitrite loop increases treatment efficiency. Of our genomes, 19 have no previously cultivated or sequenced close relatives and six belong to bacterial phyla without any cultivated members, including the most complete Omnitrophica (formerly OP3) genome to date. PMID:27029554

Conceptual design study of the moderate size superconducting spherical tokamak power plant

NASA Astrophysics Data System (ADS)

Gi, Keii; Ono, Yasushi; Nakamura, Makoto; Someya, Youji; Utoh, Hiroyasu; Tobita, Kenji; Ono, Masayuki

2015-06-01

A new conceptual design of the superconducting spherical tokamak (ST) power plant was proposed as an attractive choice for tokamak fusion reactors. We reassessed a possibility of the ST as a power plant using the conservative reactor engineering constraints often used for the conventional tokamak reactor design. An extensive parameters scan which covers all ranges of feasible superconducting ST reactors was completed, and five constraints which include already achieved plasma magnetohydrodynamic (MHD) and confinement parameters in ST experiments were established for the purpose of choosing the optimum operation point. Based on comparison with the estimated future energy costs of electricity (COEs) in Japan, cost-effective ST reactors can be designed if their COEs are smaller than 120 mills kW-1 h-1 (2013). We selected the optimized design point: A = 2.0 and Rp = 5.4 m after considering the maintenance scheme and TF ripple. A self-consistent free-boundary MHD equilibrium and poloidal field coil configuration of the ST reactor were designed by modifying the neutral beam injection system and plasma profiles. The MHD stability of the equilibrium was analysed and a ramp-up scenario was considered for ensuring the new ST design. The optimized moderate-size ST power plant conceptual design realizes realistic plasma and fusion engineering parameters keeping its economic competitiveness against existing energy sources in Japan.

High-rate two-phase process for the anaerobic degradation of cellulose, employing rumen microorganisms for an efficient acidogenesis

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

Gijzen, H.J.; Zwart, K.B.; Verhagen, F.J.M.

1988-04-05

A novel two-stage anaerobic process for the microbial conversion of cellulose into biogas has been developed. In the first phase, a mixed population of rumen bacteria and ciliates was used in the hydrolysis and fermentation of cellulose. The volatile fatty acids (VFA) produced in this acidogenic reactor were subsequently converted into biogas in a UASB-type methanogenic reactor. A stepwise increase of the loading rate from 11.9 to 25.8 g volatile solids/L reactor volume/day (g VS/L/day) did not affect the degradation efficiency in the acidogenic reactor, whereas the methanogenic reactor appeared to be overloaded at the highest loading rate. Cellulose digestionmore » was almost complete at all loading rates applied. The two-stage anaerobic process was also tested with a closed fluid circuit. In this instance total methane production was 0.438 L CH/sub 4//g VS added, which is equivalent to 98% of the theoretical value. The application of rumen microorganisms in combination with a high-rate methane reactor is proposed as a means of efficient anaerobic degradation of cellulosic residues to methane. Because this newly developed two-phase system is based on processes and microorganisms from the ruminant, it will be referred to as Rumen Derived Anaerobic Digestion (RUDAD)-process.« less

Space Nuclear Power Plant Pre-Conceptual Design Report, For Information

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

B. Levine

2006-01-27

This letter transmits, for information, the Project Prometheus Space Nuclear Power Plant (SNPP) Pre-Conceptual Design Report completed by the Naval Reactors Prime Contractor Team (NRPCT). This report documents the work pertaining to the Reactor Module, which includes integration of the space nuclear reactor with the reactor radiation shield, energy conversion, and instrumentation and control segments. This document also describes integration of the Reactor Module with the Heat Rejection segment, the Power Conditioning and Distribution subsystem (which comprise the SNPP), and the remainder of the Prometheus spaceship.

ETR, TRA642. ETR COMPLEX NEARLY COMPLETE. CAMERA FACES NORTHWEST, PROBABLY ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR, TRA-642. ETR COMPLEX NEARLY COMPLETE. CAMERA FACES NORTHWEST, PROBABLY FROM TOP DECK OF COOLING TOWER. SHADOW IS CAST BY COOLING TOWER UNITS OFF LEFT OF VIEW. HIGH-BAY REACTOR BUILDING IS SURROUNDED BY ITS ATTACHED SERVICES: ELECTRICAL (TRA-648), HEAT EXCHANGER (TRA-644 WITH U-SHAPED YARD), AND COMPRESSOR (TRA-643). THE CONTROL BUILDING (TRA-647) ON THE NORTH SIDE IS HIDDEN FROM VIEW. AT UPPER RIGHT IS MTR BUILDING, TRA-603. INL NEGATIVE NO. 56-3798. Jack L. Anderson, Photographer, 11/26/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

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

Baratta, A.J.

1997-07-01

To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts andmore » engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.« less

Radioactive waste from decommissioning of fast reactors (through the example of BN-800)

NASA Astrophysics Data System (ADS)

Rybin, A. A.; Momot, O. A.

2017-01-01

Estimation of volume of radioactive waste from operating and decommissioning of fast reactors is introduced. Preliminary estimation has shown that the volume of RW from decommissioning of BN-800 is amounted to 63,000 cu. m. Comparison of the amount of liquid radioactive waste derived from operation of different reactor types is performed. Approximate costs of all wastes disposal for complete decommissioning of BN-800 reactor are estimated amounting up to approx. 145 million.

Advanced oxidation process-biological system for wastewater containing a recalcitrant pollutant.

PubMed

Oller, I; Malato, S; Sánchez-Pérez, J A; Maldonado, M I; Gernjak, W; Pérez-Estrada, L A

2007-01-01

Two advanced oxidation processes (AOPs), ozonation and photo-Fenton, combined with a pilot aerobic biological reactor at field scale were employed for the treatment of industrial non-biodegradable saline wastewater (TOC around 200 mgL(-1)) containing a biorecalcitrant compound, alpha-methylphenylglycine (MPG), at a concentration of 500 mgL(-1). Ozonation experiments were performed in a 50-L reactor with constant inlet ozone of 21.9 g m(-3). Solar photo-Fenton tests were carried out in a 75-L pilot plant made up of four compound parabolic collector (CPC) units. The catalyst concentration employed in this system was 20 mgL(-1) of Fe2+ and the H2O2 concentration was kept in the range of 200-500mgL(-1). Complete degradation of MPG was attained after 1,020 min of ozone treatment, while only 195 min were required for photo-Fenton. Samples from different stages of both AOPs were taken for Zahn-Wellens biocompatibility tests. Biodegradability enhancement of the industrial saline wastewater was confirmed (>70% biodegradability). Biodegradable compounds generated during the preliminary oxidative processes were biologically mineralised in a 170-L aerobic immobilised biomass reactor (IBR). The global efficiency of both AOP/biological combined systems was 90% removal of an initial TOC of over 500 mgL(-1).

Analysis of Loss-of-Coolant Accidents in the NIST Research Reactor - Early Phase

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

Baek, Joo S.; Diamond, David

A study of the fuel temperature during the early phase of a loss-of-coolant accident (LOCA) in the NIST research reactor (NBSR) was completed. Previous studies had been reported in the preliminary safety analysis report for the conversion of the NBSR from high-enriched uranium (HEU) fuel to low-enriched (LEU) fuel. Those studies had focused on the most vulnerable LOCA situation, namely, a double-ended guillotine break in the time period after reactor trip when water is drained from either the coolant channels inside the fuel elements or the region outside the fuel elements. The current study fills in a gap in themore » analysis which is the early phase of the event when there may still be water present but the reactor is at power or immediately after reactor trip and pumps have tripped. The calculations were done, for both the current HEU-fueled core and the proposed LEU core, with the TRACE thermal-hydraulic systems code. Several break locations and different break sizes were considered. In all cases the increase in the clad (or fuel meat) temperature was relatively small so that a large margin to the temperature threshold for blistering (the Safety Limit for the NBSR) remained.« less

Nitrification in a completely stirred tank reactor treating the liquid phase of digestate: The way towards rational use of nitrogen.

PubMed

Svehla, Pavel; Radechovska, Helena; Pacek, Lukas; Michal, Pavel; Hanc, Ales; Tlustos, Pavel

2017-06-01

The nitrification of the liquid phase of digestate (LPD) was conducted using a 5L completely stirred tank reactor (CSTR) in two independent periods (P1 - without pH control; P2 - with pH control). The possibility of minimizing nitrogen losses during the application of LPD to the soil as well as during long-term storage or thermal thickening of LPD using nitrification was discussed. Moreover, the feasibility of applying the nitrification of LPD to the production of electron acceptors for biological desulfurization of biogas was assessed. Despite an extremely high average concentration of ammonia and COD in LPD reaching 2470 and 9080mg/L, respectively, nitrification was confirmed immediately after the start-up of the CSTR. N-NO 3 - concentration reached 250mg/L only two days after the start of P1. On the other hand, P1 demonstrated that working without pH control is a risk because of the free nitrous acid (FNA) inhibition towards nitrite oxidizing bacteria (NOB) resulting in massive nitrite accumulation. Up to 30.9mg/L of FNA was present in the reactor during P1, where the NOB started to be inhibited even at 0.15mg/L of FNA. During P2, the control of pH at 7.0 resulted in nitrogen oxidation efficiency reaching 98.3±1.5% and the presence of N-NO 3 - among oxidized nitrogen 99.6±0.4%. The representation of volatile free ammonia within total nitrogen was reduced more than 1000 times comparing with raw LPD under these conditions. Thus, optimum characteristics of the tested system from the point of view of minimizing the nitrogen losses as well as production of electron acceptors for the desulfurization of biogas were gained in this phase of reactor operation. Based on the results of the experiments, potential improvements and modifications of the tested system were suggested. Copyright © 2017 Elsevier Ltd. All rights reserved.

Small space reactor power systems for unmanned solar system exploration missions

NASA Technical Reports Server (NTRS)

Bloomfield, Harvey S.

1987-01-01

A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model.

The current status of fluoride salt cooled high temperature reactor (FHR) technology and its overlap with HIF target chamber concepts

NASA Astrophysics Data System (ADS)

Scarlat, Raluca O.; Peterson, Per F.

2014-01-01

The fluoride salt cooled high temperature reactor (FHR) is a class of fission reactor designs that use liquid fluoride salt coolant, TRISO coated particle fuel, and graphite moderator. Heavy ion fusion (HIF) can likewise make use of liquid fluoride salts, to create thick or thin liquid layers to protect structures in the target chamber from ablation by target X-rays and damage from fusion neutron irradiation. This presentation summarizes ongoing work in support of design development and safety analysis of FHR systems. Development work for fluoride salt systems with application to both FHR and HIF includes thermal-hydraulic modeling and experimentation, salt chemistry control, tritium management, salt corrosion of metallic alloys, and development of major components (e.g., pumps, heat exchangers) and gas-Brayton cycle power conversion systems. In support of FHR development, a thermal-hydraulic experimental test bay for separate effects (SETs) and integral effect tests (IETs) was built at UC Berkeley, and a second IET facility is under design. The experiments investigate heat transfer and fluid dynamics and they make use of oils as simulant fluids at reduced scale, temperature, and power of the prototypical salt-cooled system. With direct application to HIF, vortex tube flow was investigated in scaled experiments with mineral oil. Liquid jets response to impulse loading was likewise studied using water as a simulant fluid. A set of four workshops engaging industry and national laboratory experts were completed in 2012, with the goal of developing a technology pathway to the design and licensing of a commercial FHR. The pathway will include experimental and modeling efforts at universities and national laboratories, requirements for a component test facility for reliability testing of fluoride salt equipment at prototypical conditions, requirements for an FHR test reactor, and development of a pre-conceptual design for a commercial reactor.

Simulation of the MELiSSA closed loop system as a tool to define its integration strategy

NASA Astrophysics Data System (ADS)

Poughon, Laurent; Farges, Berangere; Dussap, Claude-Gilles; Godia, Francesc; Lasseur, Christophe

Inspired from a terrestrial ecosystem, MELiSSA (Micro Ecological Life Support System Alternative) is a project of closed life support system future long-term manned missions (Moon and Mars bases). Started on ESA in 1989, this 5 compartments concept has evolved following a mechanistic engineering approach for acquiring both theoretical and technical knowledge. In its current state of development the project can now start to demonstrate the MELiSSA loop concept at a pilot scale. Thus an integration strategy for a MELiSSA Pilot Plant (MPP) was defined, describing the different phases for tests and connections between compartments. The integration steps should be started in 2008 and be completed with a complete operational loop in 2015, which final objective is to achieve a closed liquid and gas loop with 100 Although the integration logic could start with the most advanced processes in terms of knowledge and hardware development, this logic needs to be completed by high politic of simulation. Thanks to this simulation exercise, the effective demonstrations of each independent process and its progressive coupling with others will be performed in operational conditions as close as possible to the final configuration. The theoretical approach described in this paper is based on mass balance models of each of the MELiSSA biological compartments which are used to simulate each integration step and the complete MPP loop itself. These simulations will help to identify criticalities of each integration steps and to check the consistencies between objectives, flows, recycling efficiencies and sizing of the pilot reactors. A MPP scenario compatible with the current knowledge of the operation of the pilot reactors was investigated and the theoretical performances of the system compared to the objectives of the MPP. From this scenario the most important milestone steps in the integration are highlighted and their behaviour can be simulated.

Toxicity of nonylphenol diethoxylate in lab-scale anaerobic digesters.

PubMed

Bozkurt, Hande; Sanin, F Dilek

2014-06-01

Nonylphenol compounds have high commercial, industrial and domestic uses owing to their surface active properties. In addition to their toxic, carcinogenic and persistent characteristics; they have drawn the attention of scientists lately due to their endocrine disrupting properties. Their widespread use and disposal cause them to enter wastewater treatment systems at high concentrations. Since they are highly persistent and hydrophobic, they accumulate mostly on sludge. In this study using Anaerobic Toxicity Assay (ATA) tests, the toxicity of a model nonylphenol compound, nonylphenol diethoxylate (NP2EO), for anaerobic digestion of sludge was determined. The test bottles were dosed with NP2EO in acetone, with concentrations ranging from 1 mg L(-1) to 30 mg L(-1). During the tests, gas productions and compositions in terms of methane and carbon dioxide were monitored. To be able to judge about the fate, the target compounds were extracted from water and sludge and analyzed using GC/MS. The sludge samples used for assembling the reactors were found to contain NP and NP1EO but no NP2EO. After the assay was completed, all the NP2EO spiked into the live reactors was found to disappear. The increase seen in NP1EO and NP and further accumulation of NP in the system, indicated the conversion of NP2EO to these metabolites. On the other hand, no conversion was observed in abiotic reactors. Inhibition of NP2EO for anaerobic microorganisms was not observed throughout the tests considering the biogas production of the test reactors in comparison to the control reactors. Copyright © 2013 Elsevier Ltd. All rights reserved.

Special Purpose Nuclear Reactor (5 MW) for Reliable Power at Remote Sites Assessment Report

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

Sterbentz, James William; Werner, James Elmer; McKellar, Michael George

The Phenomena Identification and Ranking Table (PIRT) technique was conducted on the Special Purpose Reactor nuclear plant design. The PIRT is a structured process to identify safety-relevant/safety-significant phenomena and assess the importance and knowledge base by ranking the phenomena. The Special Purpose Reactor is currently in the conceptual design stage. The candidate reactor has a solid monolithic stainless steel core with an array of heat pipes and fuel pellets embedded in the monolith. The heat pipes are used to remove heat from the core using simple, reliable, and well-characterized physics (capillarity, boiling, and condensation). In the initial design, one heatmore » exchanger is used for the working fluid that produces energy, and a second heat exchanger is used to remove decay heat in emergency or shutdown conditions. In addition, a power conversion cycle such as an open-air Brayton system is available as an option for power conversion and process heat. This report summarizes and documents the process and scope of the four PIRT reviews, noting the major activities and conclusions. The identified phenomena, analyses, rationales, and associated ratings are presented along with a summary of the findings from the four individual PIRTs, namely (1) Reactor Accident and Normal Operations, (2) Heat Pipes, (3) Materials, and (4) Power Conversion. The PIRT reports for these four major system areas evaluated are attached as appendixes to this report and provide considerably more detail about each assessment as well as a more complete listing of the phenomena that were evaluated.« less

Evaluation of Heat Recuperation in a Concentric Hydrogen Reduction Reactor

NASA Technical Reports Server (NTRS)

Linne, Diane; Kleinhenz, Julie; Hegde, Uday

2012-01-01

Heat recuperation in an ISRU reactor system involves the recovery of heat from a reacted regolith batch by transferring this energy into a batch of fresh regolith. One concept for a hydrogen reduction reactor is a concentric chamber design where heat is transferred from the inner, reaction chamber into fresh regolith in the outer, recuperation chamber. This concept was tested and analyzed to define the overall benefit compared to a more traditional single chamber batch reactor. Data was gathered for heat-up and recuperation in the inner chamber alone, simulating a single chamber design, as well as recuperation into the outer chamber, simulating a dual chamber design. Experimental data was also used to improve two analytical models, with good agreement for temperature behavior during recuperation, calculated mass of the reactor concepts, and energy required during heat-up. The five tests, performed using JSC-1A regolith simulant, also explored the effectiveness of helium gas fluidization, hydrogen gas fluidization, and vibrational fluidization. Results indicate that higher hydrogen volumetric flow rates are required compared to helium for complete fluidization and mixing, and that vibrational fluidization may provide equivalent mixing while eliminating the need to flow large amounts of excess hydrogen. Analysis of the total energy required for heat-up and steady-state operations for a variety of conditions and assumptions shows that the dual-chamber concept requires the same or more energy than the single chamber concept. With no clear energy savings, the added mass and complexity of the dual-chamber makes it unlikely that this design concept will provide any added benefit to the overall ISRU oxygen production system.

Steel slag carbonation in a flow-through reactor system: the role of fluid-flux.

PubMed

Berryman, Eleanor J; Williams-Jones, Anthony E; Migdisov, Artashes A

2015-01-01

Steel production is currently the largest industrial source of atmospheric CO2. As annual steel production continues to grow, the need for effective methods of reducing its carbon footprint increases correspondingly. The carbonation of the calcium-bearing phases in steel slag generated during basic oxygen furnace (BOF) steel production, in particular its major constituent, larnite {Ca2SiO4}, which is a structural analogue of olivine {(MgFe)2SiO4}, the main mineral subjected to natural carbonation in peridotites, offers the potential to offset some of these emissions. However, the controls on the nature and efficiency of steel slag carbonation are yet to be completely understood. Experiments were conducted exposing steel slag grains to a CO2-H2O mixture in both batch and flow-through reactors to investigate the impact of temperature, fluid flux, and reaction gradient on the dissolution and carbonation of steel slag. The results of these experiments show that dissolution and carbonation of BOF steel slag are more efficient in a flow-through reactor than in the batch reactors used in most previous studies. Moreover, they show that fluid flux needs to be optimized in addition to grain size, pressure, and temperature, in order to maximize the efficiency of carbonation. Based on these results, a two-stage reactor consisting of a high and a low fluid-flux chamber is proposed for CO2 sequestration by steel slag carbonation, allowing dissolution of the slag and precipitation of calcium carbonate to occur within a single flow-through system. Copyright © 2014. Published by Elsevier B.V.

Systems definition space based power conversion systems: Executive summary

NASA Technical Reports Server (NTRS)

1977-01-01

Potential space-located systems for the generation of electrical power for use on earth were investigated. These systems were of three basic types: (1) systems producing electrical power from solar energy; (2) systems producing electrical power from nuclear reactors; (3) systems for augmenting ground-based solar power plants by orbital sunlight reflectors. Configurations implementing these concepts were developed through an optimization process intended to yield the lowest cost for each. A complete program was developed for each concept, identifying required production rates, quantities of launches, required facilities, etc. Each program was costed in order to provide the electric power cost appropriate to each concept.

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.

Kilopower: Small and Affordable Fission Power Systems for Space

NASA Technical Reports Server (NTRS)

Mason, Lee; Palac, Don; Gibson, Marc

2017-01-01

The Nuclear Systems Kilopower Project was initiated by NASA's Space Technology Mission Directorate Game Changing Development Program in fiscal year 2015 to demonstrate subsystem-level technology readiness of small space fission power in a relevant environment (Technology Readiness Level 5) for space science and human exploration power needs. The Nuclear Systems Kilopower Project centerpiece is the Kilopower Reactor Using Stirling Technology (KRUSTY) test, which consists of the development and testing of a fission ground technology demonstrator of a 1 kWe-class fission power system. The technologies to be developed and validated by KRUSTY are extensible to space fission power systems from 1 to 10 kWe, which can enable higher power future potential deep space science missions, as well as modular surface fission power systems for exploration. The Kilopower Project is cofounded by NASA and the Department of Energy National Nuclear Security Administration (NNSA).KRUSTY include the reactor core, heat pipes to transfer the heat from the core to the power conversion system, and the power conversion system. Los Alamos National Laboratory leads the design of the reactor, and the Y-12 National Security Complex is fabricating it. NASA Glenn Research Center (GRC) has designed, built, and demonstrated the balance of plant heat transfer and power conversion portions of the KRUSTY experiment. NASA MSFC developed an electrical reactor simulator for non-nuclear testing, and the design of the reflector and shielding for nuclear testing. In 2016, an electrically heated non-fissionable Depleted Uranium (DU) core was tested at GRC in a configuration identical to the planned nuclear test. Once the reactor core has been fabricated and shipped to the Device Assembly Facility at the NNSAs Nevada National Security Site, the KRUSTY nuclear experiment will be assembled and tested. Completion of the KRUSTY experiment will validate the readiness of 1 to 10 kWe space fission technology for NASAs future requirements for sunlight-independent space power. An early opportunity for demonstration of In-Situ Resource Utilization (ISRU) capability on the surface of Mars is currently being considered for 2026 launch. Since a space fission system is the leading option for power generation for the first Mars human outpost, a smaller version of a planetary surface fission power system could be built to power the ISRU demonstration and ensure its end-to-end validity. Planning is underway to start the hardware development of this subscale flight demonstrator in 2018.

A cost/benefit analysis of commercial fusion-fission hybrid reactor development

NASA Astrophysics Data System (ADS)

Kostoff, Ronald N.

1983-04-01

A simple algorithm was developed that allows rapid computation of the ratio, R, of present worth of benefits to present worth of hybrid R&D program costs as a function of potential hybrid unit electricity cost savings, discount rate, electricity demand growth rate, total hybrid R&D program cost, and time to complete a demonstration reactor. In the sensitivity study, these variables were assigned nominal values (unit electricity cost savings of 4 mills/kW-hr, discount rate of 4%/year, growth rate of 2.25%/year, total R&D program cost of 20 billion, and time to complete a demonstration reactor of 30 years), and the variable of interest was varied about its nominal value. Results show that R increases with decreasing discount rate and increasing unit electricity savings and ranges from 4 to 94 as discount rate ranges from 5 to 3%/year and unit electricity savings range from 2 to 6 mills/kW-hr. R increases with increasing growth rate and ranges from 3 to 187 as growth rate ranges from 1 to 3.5%/year and unit electricity cost savings range from 2 to 6 mills/kW-hr. R attains a maximum value when plotted against time to complete a demonstration reactor. The location of this maximum value occurs at shorter completion times as discount rate increases, and this optimal completion time ranges from 20 years for a discount rate of 4%/year to 45 years for a discount rate of 3%/year.

METHOD FOR SENSING DEGREE OF FLUIDIZATION IN FLUIDIZED BED

DOEpatents

Levey, R.P. Jr.; Fowler, A.H.

1961-12-12

A method is given for detecting, indicating, and controlling the degree of fluidization in a fluid-bed reactor into which powdered material is fed. The method comprises admitting of gas into the reactor, inserting a springsupported rod into the powder bed of the reactor, exciting the rod to vibrate at its resonant frequency, deriving a signal responsive to the amplitude of vibi-ation of the rod and spring, the signal being directiy proportional to the rate of flow of the gas through the reactor, displaying the signal to provide an indication of the degree of fluidization within the reactor, and controlling the rate of gas flow into the reactor until said signal stabilizes at a constant value to provide substantially complete fluidization within the reactor. (AEC)

Transient Approximation of SAFE-100 Heat Pipe Operation

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Reid, Robert S.

2005-01-01

Engineers at Los Alamos National Laboratory (LANL) have designed several heat pipe cooled reactor concepts, ranging in power from 15 kWt to 800 kWt, for both surface power systems and nuclear electric propulsion systems. The Safe, Affordable Fission Engine (SAFE) is now being developed in a collaborative effort between LANL and NASA Marshall Space Flight Center (NASA/MSFC). NASA is responsible for fabrication and testing of non-nuclear, electrically heated modules in the Early Flight Fission Test Facility (EFF-TF) at MSFC. In-core heat pipes must be properly thawed as the reactor power starts. Computational models have been developed to assess the expected operation of a specific heat pipe design during start-up, steady state operation, and shutdown. While computationally intensive codes provide complete, detailed analyses of heat pipe thaw, a relatively simple. concise routine can also be applied to approximate the response of a heat pipe to changes in the evaporator heat transfer rate during start-up and power transients (e.g., modification of reactor power level) with reasonably accurate results. This paper describes a simplified model of heat pipe start-up that extends previous work and compares the results to experimental measurements for a SAFE-100 type heat pipe design.

Sodium Heat Pipe Module Processing For the SAFE-100 Reactor Concept

NASA Technical Reports Server (NTRS)

Martin, James; Salvail, Pat

2003-01-01

To support development and hardware-based testing of various space reactor concepts, the Early Flight Fission-Test Facility (EFF-TF) team established a specialized glove box unit with ancillary systems to handle/process alkali metals. Recently, these systems have been commissioned with sodium supporting the fill of stainless steel heat pipe modules for use with a 100 kW thermal heat pipe reactor design. As part of this effort, procedures were developed and refined to govern each segment of the process covering: fill, leak check, vacuum processing, weld closeout, and final "wet in". A series of 316 stainless steel modules, used as precursors to the actual 321 stainless steel modules, were filled with 35 +/- 1 grams of sodium using a known volume canister to control the dispensed mass. Each module was leak checked to less than10(exp -10) std cc/sec helium and vacuum conditioned at 250 C to assist in the removal of trapped gases. A welding procedure was developed to close out the fill stem preventing external gases from entering the evacuated module. Finally the completed modules were vacuum fired at 750 C allowing the sodium to fully wet the internal surface and wick structure of the heat pipe module.

Sodium Heat Pipe Module Processing For the SAFE-100 Reactor Concept

NASA Astrophysics Data System (ADS)

Martin, James; Salvail, Pat

2004-02-01

To support development and hardware-based testing of various space reactor concepts, the Early Flight Fission-Test Facility (EFF-TF) team established a specialized glove box unit with ancillary systems to handle/process alkali metals. Recently, these systems have been commissioned with sodium supporting the fill of stainless steel heat pipe modules for use with a 100 kW thermal heat pipe reactor design. As part of this effort, procedures were developed and refined to govern each segment of the process covering: fill, leak check, vacuum processing, weld closeout, and final ``wet in''. A series of 316 stainless steel modules, used as precursors to the actual 321 stainless steel modules, were filled with 35 +/-1 grams of sodium using a known volume canister to control the dispensed mass. Each module was leak checked to <10-10 std cc/sec helium and vacuum conditioned at 250 °C to assist in the removal of trapped gases. A welding procedure was developed to close out the fill stem preventing external gases from entering the evacuated module. Finally the completed modules were vacuum fired at 750 °C allowing the sodium to fully wet the internal surface and wick structure of the heat pipe module.

Advanced phosphorus recovery using a novel SBR system with granular sludge in simultaneous nitrification, denitrification and phosphorus removal process.

PubMed

Lu, Yong-Ze; Wang, Hou-Feng; Kotsopoulos, Thomas A; Zeng, Raymond J

2016-05-01

In this study, a novel process for phosphorus (P) recovery without excess sludge production from granular sludge in simultaneous nitrification-denitrification and P removal (SNDPR) system is presented. Aerobic microbial granules were successfully cultivated in an alternating aerobic-anaerobic sequencing batch reactor (SBR) for removing P and nitrogen (N). Dense and stable granular sludge was created, and the SBR system showed good performance in terms of P and N removal. The removal efficiency was approximately 65.22 % for N, and P was completely removed under stable operating conditions. Afterward, new operating conditions were applied in order to enhance P recovering without excess sludge production. The initial SBR system was equipped with a batch reactor and a non-woven cloth filter, and 1.37 g of CH3COONa·3H2O was added to the batch reactor after mixing it with 1 L of sludge derived from the SBR reactor to enhance P release in the liquid fraction, this comprises the new system configuration. Under the new operating conditions, 93.19 % of the P contained in wastewater was released in the liquid fraction as concentrated orthophosphate from part of granular sludge. This amount of P could be efficiently recovered in the form of struvite. Meanwhile, a deterioration of the denitrification efficiency was observed and the granules were disintegrated into smaller particles. The biomass concentration in the system increased firstly and then maintained at 4.0 ± 0.15 gVSS/L afterward. These results indicate that this P recovery operating (PRO) mode is a promising method to recover P in a SNDPR system with granular sludge. In addition, new insights into the granule transformation when confronted with high chemical oxygen demand (COD) load were provided.

Study Gives Good Odds on Nuclear Reactor Safety

ERIC Educational Resources Information Center

Russell, Cristine

1974-01-01

Summarized is data from a recent study on nuclear reactor safety completed by Norman C. Rasmussen and others. Non-nuclear events are about 10,000 times more likely to produce large accidents than nuclear plants. (RH)

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.

Treatment of real wastewater using co-culture of immobilized Chlorella vulgaris and suspended activated sludge.

PubMed

Mujtaba, Ghulam; Lee, Kisay

2017-09-01

The use of algal-bacterial symbiotic association establishes a sustainable and cost-effective strategy in wastewater treatment. Using municipal wastewater, the removal performances of inorganic nutrients (nitrogen and phosphorus) and organic pollutants were investigated by the co-culture system having different inoculum ratios (R) of suspended activated sludge to alginate-immobilized microalgae Chlorella vulgaris. The co-culture reactors with lower R ratios obtained more removal of nitrogen than in pure culture of C. vulgaris. The reactor with R = 0.5 (sludge/microalgae) showed the highest performance representing 66% removal after 24 h and 95% removal after 84 h. Phosphorus was completely eliminated (100%) in the co-culture system with inoculum ratios of 0.5 and 1.0 after 24 h and in the pure C. vulgaris culture after 36 h. The COD level was greatly reduced in the activated sludge reactor, while, it was increasing in pure C. vulgaris culture after 24 h of incubation. However, COD was almost stabilized after 24 h in the reactors with high R ratios such as 2.0, 5.0, and 10 due to the higher concentration of activated sludge. The growth of C. vulgaris was promoted from 0.03 g/L/d to 0.05 g/L/d in the co-culture of low inoculum ratios such as R = 0.5, implying that there exist an optimum inoculum ratio in the co-culture system in order to achieve efficient removal of nutrients. Copyright © 2017 Elsevier Ltd. All rights reserved.

A novel plant protection strategy for transient reactors

NASA Astrophysics Data System (ADS)

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

A novel plant protection system designed for use in the TREAT Upgrade (TU) reactor is described. The TU reactor is designed for controlled transient operation in the testing of reactor fuel behavior under simulated reactor accident conditions. Safe operation of the reactor is of paramount importance and the Plant Protection System (PPS) had to be designed to exacting requirements. Researchers believe that the strategy developed for the TU has potential application to the multimegawatt space reactors and represents the state of the art in terrestrial transient reactor protection systems.

Application of a 2-step process for the biological treatment of sulfidic spent caustics.

PubMed

de Graaff, Marco; Klok, Johannes B M; Bijmans, Martijn F M; Muyzer, Gerard; Janssen, Albert J H

2012-03-01

This research demonstrates the feasibility and advantages of a 2-step process for the biological treatment of sulfidic spent caustics under halo-alkaline conditions (i.e. pH 9.5; Na(+) = 0.8 M). Experiments with synthetically prepared solutions were performed in a continuously fed system consisting of two gas-lift reactors in series operated at aerobic conditions at 35 °C. The detoxification of sulfide to thiosulfate in the first step allowed the successful biological treatment of total-S loading rates up to 33 mmol L(-1) day(-1). In the second, biological step, the remaining sulfide and thiosulfate was completely converted to sulfate by haloalkaliphilic sulfide oxidizing bacteria. Mathematical modeling of the 2-step process shows that under the prevailing conditions an optimal reactor configuration consists of 40% 'abiotic' and 60% 'biological' volume, whilst the total reactor volume is 22% smaller than for the 1-step process. Copyright © 2011 Elsevier Ltd. All rights reserved.

Treatment of natural rubber processing wastewater using a combination system of a two-stage up-flow anaerobic sludge blanket and down-flow hanging sponge system.

PubMed

Tanikawa, D; Syutsubo, K; Hatamoto, M; Fukuda, M; Takahashi, M; Choeisai, P K; Yamaguchi, T

2016-01-01

A pilot-scale experiment of natural rubber processing wastewater treatment was conducted using a combination system consisting of a two-stage up-flow anaerobic sludge blanket (UASB) and a down-flow hanging sponge (DHS) reactor for more than 10 months. The system achieved a chemical oxygen demand (COD) removal efficiency of 95.7% ± 1.3% at an organic loading rate of 0.8 kg COD/(m(3).d). Bacterial activity measurement of retained sludge from the UASB showed that sulfate-reducing bacteria (SRB), especially hydrogen-utilizing SRB, possessed high activity compared with methane-producing bacteria (MPB). Conversely, the acetate-utilizing activity of MPB was superior to SRB in the second stage of the reactor. The two-stage UASB-DHS system can reduce power consumption by 95% and excess sludge by 98%. In addition, it is possible to prevent emissions of greenhouse gases (GHG), such as methane, using this system. Furthermore, recovered methane from the two-stage UASB can completely cover the electricity needs for the operation of the two-stage UASB-DHS system, accounting for approximately 15% of the electricity used in the natural rubber manufacturing process.

Coupled reactor kinetics and heat transfer model for heat pipe cooled reactors

NASA Astrophysics Data System (ADS)

Wright, Steven A.; Houts, Michael

2001-02-01

Heat pipes are often proposed as cooling system components for small fission reactors. SAFE-300 and STAR-C are two reactor concepts that use heat pipes as an integral part of the cooling system. Heat pipes have been used in reactors to cool components within radiation tests (Deverall, 1973); however, no reactor has been built or tested that uses heat pipes solely as the primary cooling system. Heat pipe cooled reactors will likely require the development of a test reactor to determine the main differences in operational behavior from forced cooled reactors. The purpose of this paper is to describe the results of a systems code capable of modeling the coupling between the reactor kinetics and heat pipe controlled heat transport. Heat transport in heat pipe reactors is complex and highly system dependent. Nevertheless, in general terms it relies on heat flowing from the fuel pins through the heat pipe, to the heat exchanger, and then ultimately into the power conversion system and heat sink. A system model is described that is capable of modeling coupled reactor kinetics phenomena, heat transfer dynamics within the fuel pins, and the transient behavior of heat pipes (including the melting of the working fluid). This paper focuses primarily on the coupling effects caused by reactor feedback and compares the observations with forced cooled reactors. A number of reactor startup transients have been modeled, and issues such as power peaking, and power-to-flow mismatches, and loading transients were examined, including the possibility of heat flow from the heat exchanger back into the reactor. This system model is envisioned as a tool to be used for screening various heat pipe cooled reactor concepts, for designing and developing test facility requirements, for use in safety evaluations, and for developing test criteria for in-pile and out-of-pile test facilities. .

Nitrate removal from groundwater by cooperating heterotrophic with autotrophic denitrification in a biofilm-electrode reactor.

PubMed

Zhao, Yingxin; Feng, Chuanping; Wang, Qinghong; Yang, Yingnan; Zhang, Zhenya; Sugiura, Norio

2011-09-15

An intensified biofilm-electrode reactor (IBER) combining heterotrophic and autotrophic denitrification was developed for treatment of nitrate contaminated groundwater. The reactor was evaluated with synthetic groundwater (NO(3)(-)-N50 mg L(-1)) under different hydraulic retention times (HRTs), carbon to nitrogen ratios (C/N) and electric currents (I). The experimental results demonstrate that high nitrate and nitrite removal efficiency (100%) were achieved at C/N = 1, HRT = 8h, and I = 10 mA. C/N ratios were reduced from 1 to 0.5 and the applied electric current was changed from 10 to 100 mA, showing that the optimum running condition was C/N = 0.75 and I = 40 mA, under which over 97% of NO(3)(-)-N was removed and organic carbon (methanol) was completely consumed in treated water. Simultaneously, the denitrification mechanism in this system was analyzed through pH variation in effluent. The CO(2) produced from the anode acted as a good pH buffer, automatically controlling pH in the reaction zone. The intensified biofilm-electrode reactor developed in the study was effective for the treatment of groundwater polluted by nitrate. Copyright © 2011 Elsevier B.V. All rights reserved.

Electrically Heated Testing of the Kilowatt Reactor Using Stirling Technology (KRUSTY) Experiment Using a Depleted Uranium Core

NASA Technical Reports Server (NTRS)

Briggs, Maxwell H.; Gibson, Marc A.; Sanzi, James

2017-01-01

The Kilopower project aims to develop and demonstrate scalable fission-based power technology for systems capable of delivering 110 kW of electric power with a specific power ranging from 2.5 - 6.5 Wkg. This technology could enable high power science missions or could be used to provide surface power for manned missions to the Moon or Mars. NASA has partnered with the Department of Energys National Nuclear Security Administration, Los Alamos National Labs, and Y-12 National Security Complex to develop and test a prototypic reactor and power system using existing facilities and infrastructure. This technology demonstration, referred to as the Kilowatt Reactor Using Stirling TechnologY (KRUSTY), will undergo nuclear ground testing in the summer of 2017 at the Nevada Test Site. The 1 kWe variation of the Kilopower system was chosen for the KRUSTY demonstration. The concept for the 1 kWe flight system consist of a 4 kWt highly enriched Uranium-Molybdenum reactor operating at 800 degrees Celsius coupled to sodium heat pipes. The heat pipes deliver heat to the hot ends of eight 125 W Stirling convertors producing a net electrical output of 1 kW. Waste heat is rejected using titanium-water heat pipes coupled to carbon composite radiator panels. The KRUSTY test, based on this design, uses a prototypic highly enriched uranium-molybdenum core coupled to prototypic sodium heat pipes. The heat pipes transfer heat to two Advanced Stirling Convertors (ASC-E2s) and six thermal simulators, which simulate the thermal draw of full scale power conversion units. Thermal simulators and Stirling engines are gas cooled. The most recent project milestone was the completion of non-nuclear system level testing using an electrically heated depleted uranium (non-fissioning) reactor core simulator. System level testing at the Glenn Research Center (GRC) has validated performance predictions and has demonstrated system level operation and control in a test configuration that replicates the one to be used at the Device Assembly Facility (DAF) at the Nevada National Security Site. Fabrication, assembly, and testing of the depleted uranium core has allowed for higher fidelity system level testing at GRC, and has validated the fabrication methods to be used on the highly enriched uranium core that will supply heat for the DAF KRUSTY demonstration.

DEFE0023863 Final Report, Technology for GHG Emission Reduction and CostCompetitive MilSpec Jet Fuel Production using CTL

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

Hartvigsen, Joseph J; Dimick, Paul; Laumb, Jason D

Ceramatec Inc, in collaboration with IntraMicron (IM), the Energy & Environmental Research Center (EERC) and Sustainable Energy Solutions, LLC (SES), have completed a three-year research project integrating their respective proprietary technologies in key areas to demonstrate production of a jet fuel from coal and biomass sources. The project goals and objectives were to demonstrate technology capable of producing a “commercially-viable quantity” of jet fuel and make significant progress toward compliance with Section 526 of the Energy Independence and Security Act of 2007 (EISA 2007 §526) lifecycle greenhouse gas (GHG) emissions requirements. The Ceramatec led team completed the demonstration of nominalmore » 2 bbl/day Fischer-Tropsch (FT) synthesis pilot plant design, capable of producing a nominal 1 bbl/day in the Jet-A/JP-8 fraction. This production rate would have a capacity of 1,000 gallons of jet fuel per month and provide the design basis of a 100 bbl/day module producing over 2,000 gallons of jet fuel per day. Co-gasification of coal-biomass blends enables a reduction of lifecycle greenhouse gas emissions from equivalent conventional petroleum derived fuel basis. Due to limits of biomass availability within an economic transportation range, implementation of a significant biomass feed fraction will require smaller plants than current world scale CTL and GTL FT plants. Hence a down-scaleable design is essential. The pilot plant design leverages Intramicron’s MicroFiber Entrapped Catalyst (MFEC) support which increases the catalyst bed thermal conductivity two orders of magnitude, allowing thermal management of the FT reaction exotherm in much larger reactor tubes. In this project, single tube reactors having 4.5 inch outer diameter and multi-tube reactors having 4 inch outer diameters were operated, with productivities as high as 1.5 gallons per day per linear foot of reactor tube. A significant reduction in tube count results from the use of large diameter reactor tubes, with an associated reduction in reactor cost. The pilot plant was designed with provisions for product collection capable of operating with conventional wax producing FT catalysts but was operated with a Chevron hybrid wax-free FT catalyst. Process simplification enabled by elimination of the wax hydrocracking process unit provides economic advantages in scaling to biomass capable plant sizes. Intramicron also provided a sulfur capture system based on their Oxidative Sulfur Removal (OSR) catalyst process. The integrated sulfur removal and FT systems were operated with syngas produced by the Transport Reactor Development Unit (TRDU) gasifier at the University of North Dakota EERC. SES performed modeling of their cryogenic carbon capture process on the energy, cost and CO2 emissions impact of the Coal-biomass synthetic fuel process.« less

Comparison of Reductive Dechlorination of Chlorinated Ethylene in Batch and Continuous-Flow Reactor

NASA Astrophysics Data System (ADS)

Park, S.; Jonghwan, L.; Hong, U.; Kim, N.; Ahn, H.; Lee, S.; Kim, Y.

2010-12-01

A 1.28 L-Batch reactor and continuous-flow stirred tank reactor (CFSTR) fed with formate and trichloriethene (TCE) were operated for 120 days and 72 days, respectively, to study the effect of formate as electron donor on reductive dechlorination of TCE to cis-1,2-dichloroethylene (c-DCE), vinyl chloride (VC), and ethylene (ETH). In batch reactor, injected 60 μmol TCE was completely degraded in presence of 20% hydrogen gas (H2) in less than 8 days by Evanite culture (300 mg-soluble protein) with ability to completely degrade tetrachloroethene (PCE) and TCE to ETH under anaerobic conditions. To determine the effect of formate as electron donor instead of H2, about 3 or 11 mmol of formate injected into batch-reactor every 15 days was enough to support H2 for dechlorination of c-DCE to VC and ETH. Soluble protein concentration of Evanite culture during the batch test increased from 300 mg to 688 mg for 120 days. In CFSTR test, TCE was fed continuously at 9.9 ppm (75.38 μmol/L) and the influent formate feed concentration increased stepwise from 1.3 mmol/L to 14.3 mmol/L. Injected TCE was accumulated at HRT 18 days for 13 days, but TCE was completed degraded at HRT 36 days without accumulation during left of experiment period, getting H2 from fermentative hydrogen production of injected formate. Although c-DCE was also accumulated for 23 days after CFSTR operation, it reached steady-state without accumulation in presence of excessive formate. However, since c-DCE in CFSTR was not completely dechlorinated, we will determine the transcriptional level of enzyme involved in reductive dechlorination of TCE, c-DCE, and VC in our future work.

Heat dissipating nuclear reactor with metal liner

DOEpatents

Gluekler, E.L.; Hunsbedt, A.; Lazarus, J.D.

1985-11-21

A nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel is described in this disclosure. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

Heat dissipating nuclear reactor with metal liner

DOEpatents

Gluekler, Emil L.; Hunsbedt, Anstein; Lazarus, Jonathan D.

1987-01-01

Disclosed is a nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

Computer study of emergency shutdowns of a 60-kilowatt reactor Brayton space power system

NASA Technical Reports Server (NTRS)

Tew, R. C.; Jefferies, K. S.

1974-01-01

A digital computer study of emergency shutdowns of a 60-kWe reactor Brayton power system was conducted. Malfunctions considered were (1) loss of reactor coolant flow, (2) loss of Brayton system gas flow, (3)turbine overspeed, and (4) a reactivity insertion error. Loss of reactor coolant flow was the most serious malfunction for the reactor. Methods for moderating the reactor transients due to this malfunction are considered.

Reduction of produced elementary sulfur in denitrifying sulfide removal process.

PubMed

Zhou, Xu; Liu, Lihong; Chen, Chuan; Ren, Nanqi; Wang, Aijie; Lee, Duu-Jong

2011-05-01

Denitrifying sulfide removal (DSR) processes simultaneously convert sulfide, nitrate, and chemical oxygen demand from industrial wastewater into elemental sulfur, dinitrogen gas, and carbon dioxide, respectively. The failure of a DSR process is signaled by high concentrations of sulfide in reactor effluent. Conventionally, DSR reactor failure is blamed for overcompetition for heterotroph to autotroph communities. This study indicates that the elementary sulfur produced by oxidizing sulfide that is a recoverable resource from sulfide-laden wastewaters can be reduced back to sulfide by sulfur-reducing Methanobacterium sp. The Methanobacterium sp. was stimulated with excess organic carbon (acetate) when nitrite was completely consumed by heterotrophic denitrifiers. Adjusting hydraulic retention time of a DSR reactor when nitrite is completely consumed provides an additional control variable for maximizing DSR performance.

Design Report for the ½ Scale Air-Cooled RCCS Tests in the Natural convection Shutdown heat removal Test Facility (NSTF)

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

Lisowski, D. D.; Farmer, M. T.; Lomperski, S.

The Natural convection Shutdown heat removal Test Facility (NSTF) is a large scale thermal hydraulics test facility that has been built at Argonne National Laboratory (ANL). The facility was constructed in order to carry out highly instrumented experiments that can be used to validate the performance of passive safety systems for advanced reactor designs. The facility has principally been designed for testing of Reactor Cavity Cooling System (RCCS) concepts that rely on natural convection cooling for either air or water-based systems. Standing 25-m in height, the facility is able to supply up to 220 kW at 21 kW/m 2 tomore » accurately simulate the heat fluxes at the walls of a reactor pressure vessel. A suite of nearly 400 data acquisition channels, including a sophisticated fiber optic system for high density temperature measurements, guides test operations and provides data to support scaling analysis and modeling efforts. Measurements of system mass flow rate, air and surface temperatures, heat flux, humidity, and pressure differentials, among others; are part of this total generated data set. The following report provides an introduction to the top level-objectives of the program related to passively safe decay heat removal, a detailed description of the engineering specifications, design features, and dimensions of the test facility at Argonne. Specifications of the sensors and their placement on the test facility will be provided, along with a complete channel listing of the data acquisition system.« less

Passive cooling safety system for liquid metal cooled nuclear reactors

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.; Hui, Marvin M.; Berglund, Robert C.

1991-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

Indirect passive cooling system for liquid metal cooled nuclear reactors

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1990-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

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

MH Lane

2006-02-15

This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

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

Nano-metal oxides: Exposure and engineering control assessment.

PubMed

Garcia, Alberto; Eastlake, Adrienne; Topmiller, Jennifer L; Sparks, Christopher; Martinez, Kenneth; Geraci, Charles L

2017-09-01

In January 2007, the National Institute for Occupational Safety and Health (NIOSH) conducted a field study to evaluate process specific emissions during the production of ENMs. This study was performed using the nanoparticle emission assessment technique (NEAT). During this study, it was determined that ENMs were released during production and cleaning of the process reactor. Airborne concentrations of silver, nickel, and iron were found both in the employee's personal breathing zone and area samples during reactor cleaning. At the completion of this initial survey, it was suggested that a flanged attachment be added to the local exhaust ventilation system.  NIOSH re-evaluated the facility in December 2011 to assess worker exposures following an increase in production rates. This study included a fully comprehensive emissions, exposure, and engineering control evaluation of the entire process. This study made use of the nanoparticle exposure assessment technique (NEAT 2.0). Data obtained from filter-based samples and direct reading instruments indicate that reactor cleanout increased the overall particle concentration in the immediate area. However, it does not appear that these concentrations affect areas outside of the production floor. As the distance between the reactor and the sample location increased, the observed particle number concentration decreased, creating a concentration gradient with respect to the reactor. The results of this study confirm that the flanged attachment on the local exhaust ventilation system served to decrease exposure potential.  Given the available toxicological data of the metals evaluated, caution is warranted. One should always keep in mind that occupational exposure levels were not developed specifically for nanoscale particles. With data suggesting that certain nanoparticles may be more toxic than the larger counterparts of the same material; employers should attempt to control emissions of these particles at the source, to limit the potential for exposure.

Validation of the WIMSD4M cross-section generation code with benchmark results

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

Deen, J.R.; Woodruff, W.L.; Leal, L.E.

1995-01-01

The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment Research and Test Reactor (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the WIMSD4M cross-section librariesmore » for reactor modeling of fresh water moderated cores. The results of calculations performed with multigroup cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory (ORNL) unreflected HEU critical spheres, the TRX LEU critical experiments, and calculations of a modified Los Alamos HEU D{sub 2}O moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented.« less

Visible-Light-Responsive Photocatalysis: Ag-Doped TiO2 Catalyst Development and Reactor Design Testing

NASA Technical Reports Server (NTRS)

Coutts, Janelle L.; Hintze, Paul E.; Meier, Anne; Shah, Malay G.; Devor, Robert W.; Surma, Jan M.; Maloney, Phillip R.; Bauer, Brint M.; Mazyck, David W.

2016-01-01

In recent years, the alteration of titanium dioxide to become visible-light-responsive (VLR) has been a major focus in the field of photocatalysis. Currently, bare titanium dioxide requires ultraviolet light for activation due to its band gap energy of 3.2 eV. Hg-vapor fluorescent light sources are used in photocatalytic oxidation (PCO) reactors to provide adequate levels of ultraviolet light for catalyst activation; these mercury-containing lamps, however, hinder the use of this PCO technology in a spaceflight environment due to concerns over crew Hg exposure. VLR-TiO2 would allow for use of ambient visible solar radiation or highly efficient visible wavelength LEDs, both of which would make PCO approaches more efficient, flexible, economical, and safe. Over the past three years, Kennedy Space Center has developed a VLR Ag-doped TiO2 catalyst with a band gap of 2.72 eV and promising photocatalytic activity. Catalyst immobilization techniques, including incorporation of the catalyst into a sorbent material, were examined. Extensive modeling of a reactor test bed mimicking air duct work with throughput similar to that seen on the International Space Station was completed to determine optimal reactor design. A bench-scale reactor with the novel catalyst and high-efficiency blue LEDs was challenged with several common volatile organic compounds (VOCs) found in ISS cabin air to evaluate the system's ability to perform high-throughput trace contaminant removal. The ultimate goal for this testing was to determine if the unit would be useful in pre-heat exchanger operations to lessen condensed VOCs in recovered water thus lowering the burden of VOC removal for water purification systems.

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

The detector system of the Daya Bay reactor neutrino experiment

DOE PAGES

An, F. P.

2015-12-15

The Daya Bay experiment was the first to report simultaneous measurements of reactor antineutrinos at multiple baselines leading to the discovery of ν¯e oscillations over km-baselines. Subsequent data has provided the world's most precise measurement of sin 22θ 13 and the effective mass splitting Δm 2 ee. The experiment is located in Daya Bay, China where the cluster of six nuclear reactors is among the world's most prolific sources of electron antineutrinos. Multiple antineutrino detectors are deployed in three underground water pools at different distances from the reactor cores to search for deviations in the antineutrino rate and energy spectrummore » due to neutrino mixing. Instrumented with photomultiplier tubes, the water pools serve as shielding against natural radioactivity from the surrounding rock and provide efficient muon tagging. Arrays of resistive plate chambers over the top of each pool provide additional muon detection. The antineutrino detectors were specifically designed for measurements of the antineutrino flux with minimal systematic uncertainty. Relative detector efficiencies between the near and far detectors are known to better than 0.2%. With the unblinding of the final two detectors’ baselines and target masses, a complete description and comparison of the eight antineutrino detectors can now be presented. This study describes the Daya Bay detector systems, consisting of eight antineutrino detectors in three instrumented water pools in three underground halls, and their operation through the first year of eight detector data-taking.« less

SPES-2, an experimental program to support the AP600 development

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

Tarantini, M.; Medich, C.

1995-09-01

In support of the development of the AP600 reactor, ENEA, ENEL, ANSALDO and Westinghouse have signed a research agreement. In the framework of this agreement a complex Full Height Full Pressure (FHFP) integral system testing program has been planned on SPES-2 facility. The main purpose of this paper is to point out the status of the test program; describe the hot per-operational test performed and the complete test matrix, giving all the necessary references on the work already published. Two identical Small Break LOCA transients, performed with Pressurizer to Core Make-up Tank (PRZ-CMT) balance line (Test S00203) and without PRZ-CMTmore » balance line (Test S00303) are then compared, to show how the SPES-2 facility can contribute in confirming the new AP600 reactor design choices and can give useful indications to designers. Although the detailed analysis of test data has not been completed, some consideration on the analytical tools utilized and on the SPES-2 capability to simulate the reference plant is then drawn.« less

Advanced anaerobic bioconversion of lignocellulosic waste for bioregenerative life support following thermal water treatment and biodegradation by Fibrobacter succinogenes.

PubMed

Lissens, Geert; Verstraete, Willy; Albrecht, Tobias; Brunner, Gerd; Creuly, Catherine; Seon, Jerome; Dussap, Gilles; Lasseur, Christophe

2004-06-01

The feasibility of nearly-complete conversion of lignocellulosic waste (70% food crops, 20% faecal matter and 10% green algae) into biogas was investigated in the context of a life support project. The treatment comprised a series of processes, i.e., a mesophilic laboratory scale CSTR (continuously stirred tank reactor), an upflow biofilm reactor, a fiber liquefaction reactor employing the rumen bacterium Fibrobacter succinogenes and a hydrothermolysis system in near-critical water. By the one-stage CSTR, a biogas yield of 75% with a specific biogas production of 0.37 l biogas g(-1) VSS (volatile suspended solids) added at a RT (hydraulic retention time) of 20-25 d was obtained. Biogas yields could not be increased considerably at higher RT, indicating the depletion of readily available substrate after 25 d. The solids present in the CSTR-effluent were subsequently treated in two ways. Hydrothermal treatment (T approximately 310-350 degrees C, p approximately 240 bar) resulted in effective carbon liquefaction (50-60% without and 83% with carbon dioxide saturation) and complete sanitation of the residue. Application of the cellulolytic Fibrobacter succinogenes converted remaining cellulose contained in the CSTR-effluent into acetate and propionate mainly. Subsequent anaerobic digestion of the hydrothermolysis and the Fibrobacter hydrolysates allowed conversion of 48-60% and 30%, respectively. Thus, the total process yielded biogas corresponding with conversions up to 90% of the original organic matter. It appears that particularly mesophilic digestion in conjunction with hydrothermolysis at near-critical conditions offers interesting features for (nearly) complete and hygienic carbon and energy recovery from human waste in a bioregenerative life support context.

Complete degradation of Orange G by electrolysis in sub-critical water.

PubMed

Yuksel, Asli; Sasaki, Mitsuru; Goto, Motonobu

2011-06-15

Complete degradation of azo dye Orange G was studied using a 500 mL continuous flow reactor made of SUS 316 stainless steel. In this system, a titanium reactor wall acted as a cathode and a titanium plate-type electrode was used as an anode in a subcritical reaction medium. This hydrothermal electrolysis process provides an environmentally friendly route that does not use any organic solvents or catalysts to remove organic pollutants from wastewater. Reactions were carried out from 30 to 90 min residence times at a pressure of 7 MPa, and at different temperatures of 180-250°C by applying various direct currents ranging from 0.5 to 1A. Removal of dye from the product solution and conversion of TOC increased with increasing current value. Moreover, the effect of salt addition on degradation of Orange G and TOC conversion was investigated, because in real textile wastewater, many salts are also included together with dye. Addition of Na(2)CO(3) resulted in a massive degradation of the dye itself and complete mineralization of TOC, while NaCl and Na(2)SO(4) obstructed the removal of Orange G. Greater than 99% of Orange G was successfully removed from the product solution with a 98% TOC conversion. Copyright © 2011 Elsevier B.V. All rights reserved.

Adsorption and biodegradation of 2-chlorophenol by mixed culture using activated carbon as a supporting medium-reactor performance and model verification

NASA Astrophysics Data System (ADS)

Lin, Yen-Hui

2017-11-01

A non-steady-state mathematical model system for the kinetics of adsorption and biodegradation of 2-chlorophenol (2-CP) by attached and suspended biomass on activated carbon process was derived. The mechanisms in the model system included 2-CP adsorption by activated carbon, 2-CP mass transport diffusion in biofilm, and biodegradation by attached and suspended biomass. Batch kinetic tests were performed to determine surface diffusivity of 2-CP, adsorption parameters for 2-CP, and biokinetic parameters of biomass. Experiments were conducted using a biological activated carbon (BAC) reactor system with high recycled rate to approximate a completely mixed flow reactor for model verification. Concentration profiles of 2-CP by model predictions indicated that biofilm bioregenerated the activated carbon by lowering the 2-CP concentration at the biofilm-activated carbon interface as the biofilm grew thicker. The removal efficiency of 2-CP by biomass was approximately 98.5% when 2-CP concentration in the influent was around 190.5 mg L-1 at a steady-state condition. The concentration of suspended biomass reached up to about 25.3 mg L-1 while the thickness of attached biomass was estimated to be 636 μm at a steady-state condition by model prediction. The experimental results agree closely with the results of the model predictions.

Solvent refined coal reactor quench system

DOEpatents

Thorogood, Robert M.

1983-01-01

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

Solvent refined coal reactor quench system

DOEpatents

Thorogood, R.M.

1983-11-08

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

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.

Method for passive cooling liquid metal cooled nuclear reactors, and system thereof

DOEpatents

Hunsbedt, Anstein; Busboom, Herbert J.

1991-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel.

Structural properties of lead-lithium alloys

NASA Astrophysics Data System (ADS)

Khambholja, S. G.; Satikunvar, D. D.; Abhishek, Agraj; Thakore, B. Y.

2018-05-01

Lead-Lihtium alloys have found large number of applications as liquid metal coolants in nuclear reactors. Large number of experimental work is reported for this system. However, complete theoretical description is still rare. In this scenario, we in the present work report the study of ground state properties of Lead-Lithium system. The present study is performed using plane wave pseudopotential density functional theory as implemented in Quantum ESPRESSO package. The theoretical findings are in agreement with previously reported experimental data. Some conclusions are drawn based on present study, which will be helpful for a comprehensive study.

Systems Based Approaches for Thermochemical Conversion of Biomass to Bioenergy and Bioproducts

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

Taylor, Steven

2016-07-11

Auburn’s Center for Bioenergy and Bioproducts conducts research on production of synthesis gas for use in power generation and the production of liquid fuels. The overall goal of our gasification research is to identify optimal processes for producing clean syngas to use in production of fuels and chemicals from underutilized agricultural and forest biomass feedstocks. This project focused on construction and commissioning of a bubbling-bed fluidized-bed gasifier and subsequent shakedown of the gasification and gas cleanup system. The result of this project is a fully commissioned gasification laboratory that is conducting testing on agricultural and forest biomass. Initial tests onmore » forest biomass have served as the foundation for follow-up studies on gasification under a more extensive range of temperatures, pressures, and oxidant conditions. The laboratory gasification system consists of a biomass storage tank capable of holding up to 6 tons of biomass; a biomass feeding system, with loss-in-weight metering system, capable of feeding biomass at pressures up to 650 psig; a bubbling-bed fluidized-bed gasification reactor capable of operating at pressures up to 650 psig and temperatures of 1500oF with biomass flowrates of 80 lb/hr and syngas production rates of 37 scfm; a warm-gas filtration system; fixed bed reactors for gas conditioning; and a final quench cooling system and activated carbon filtration system for gas conditioning prior to routing to Fischer-Tropsch reactors, or storage, or venting. This completed laboratory enables research to help develop economically feasible technologies for production of biomass-derived synthesis gases that will be used for clean, renewable power generation and for production of liquid transportation fuels. Moreover, this research program provides the infrastructure to educate the next generation of engineers and scientists needed to implement these technologies.« 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.

A miniature fuel reformer system for portable power sources

NASA Astrophysics Data System (ADS)

Dolanc, Gregor; Belavič, Darko; Hrovat, Marko; Hočevar, Stanko; Pohar, Andrej; Petrovčič, Janko; Musizza, Bojan

2014-12-01

A miniature methanol reformer system has been designed and built to technology readiness level exceeding a laboratory prototype. It is intended to feed fuel cells with electric power up to 100 W and contains a complete setup of the technological elements: catalytic reforming and PROX reactors, a combustor, evaporators, actuation and sensing elements, and a control unit. The system is engineered not only for performance and quality of the reformate, but also for its lightweight and compact design, seamless integration of elements, low internal electric consumption, and safety. In the paper, the design of the system is presented by focussing on its miniaturisation, integration, and process control.

High-intensity power-resolved radiation imaging of an operational nuclear reactor.

PubMed

Beaumont, Jonathan S; Mellor, Matthew P; Villa, Mario; Joyce, Malcolm J

2015-10-09

Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors.

High-intensity power-resolved radiation imaging of an operational nuclear reactor

PubMed Central

Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

2015-01-01

Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

Final Technical Report for GO15056 Millennium Cell: Development of an Advanced Chemical Hydrogen Storage and Generation System

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

Moreno, Oscar

The objectives of this project are to increase system storage capacity by improving hydrogen generation from concentrated sodium borohydride, with emphasis on reactor and system engineering; to complete a conceptual system design based on sodium borohydride that will include key technology improvements to enable a hydrogen fuel system that will meet the systembased storage capacity of 1.2 kWh/L (36 g H2/L) and 1.5 kWh/kg (45 g H2/kg), by the end of FY 2007; and to utilize engineering expertise to guide Center research in both off-board chemical hydride regeneration and on-board hydrogen generation systems.

Hydroxyapatite crystallization from a highly concentrated phosphate solution using powdered converter slag as a seed material.

PubMed

Kim, Eung-Ho; Yim, Soo-Bin; Jung, Ho-Chan; Lee, Eok-Jae

2006-08-25

A system for recovering phosphorus from membrane-filtrate from a sludge reduction process containing high phosphorus concentrations was developed. In this system, referred to as the completely mixed phosphorus crystallization reactor, powdered converter slag was used as a seed material. In a preliminary experiment, the optimal pH range for metastable crystallization of phosphorus from membrane-filtrate containing about 100mg/L PO(4)-P was found to be 6.6-7.0. The laboratory scale completely mixed phosphorus crystallization reactor, actually operated in pH range of 6.8-7.6 for influent 72.9 mg/L PO(4)-P, achieved an average efficiency of phosphorus removal from the membrane-filtrate of 52.4% during a 30-day experiment. Mixed-liquor suspended solids (MLSS) measurements revealed that, out of 0.24 kg PO(4)-P in the original membrane-filtrate fed into the reactor, 0.12 kg PO(4)-P was recovered on the seed particles after 30 days. X-ray diffraction (XRD) pattern and Fourier transform infrared (FT-IR) spectra of the crystalline material deposited on the seed particles showed peaks consistent with hydroxyapatite. Scanning electron micrograph (SEM) images exhibited that finely distributed crystalline material was formed on the surfaces of seed particles. Energy dispersive X-ray spectroscopy (EDS) mapping analysis revealed that the molar composition ratio of Ca/P of the crystalline material was 1.84. The Ca/P molar ratio>1.67 for crystalline substance might result from the presence of CaCO(3) on the crystalline surfaces. A particle size distribution analysis showed that the average particle size increased from 22 microm for the original converter slag seed particles, to 94 microm after 30 days of phosphorus crystallization. Collectively, the present results suggest that the proposed phosphorus crystallization recovery system is an effective tool for recycling phosphorus from phosphate solution.

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

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

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

1985-01-01

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

Characterization and treatment of Denizli landfill leachate using anaerobic hybrid/aerobic CSTR systems.

PubMed

Ağdağ, Osman Nuri

2011-01-01

Leachate generated in municipal solid waste landfill contains large amounts of organic and inorganic contaminants. In the scope of the study, characterization and anaerobic/aerobic treatability of leachate from Denizli (Turkey) Sanitary Landfill were investigated. Time-based fluctuations in characteristics of leachate were monitored during a one-year period. In characterization study; chemical oxygen demand (COD), biochemical oxygen demand (BOD) dissolved oxygen, temperature, pH, alkalinity, volatile fatty acids, total nitrogen, NH4-N, BOD5/COD ratio, suspended solid, inert COD, anaerobic toxicity assay and heavy metals concentrations in leachate were monitored. Average COD, BOD and NH4-N concentration in leachate were measured as 18034 mg/l, 11504 mg/l and 454 mg/l, respectively. Generally, pollution parameters in leachate were higher in summer and relatively lower in winter due to dilution by precipitation. For treatment of leachate, two different reactors, namely anaerobic hybrid and aerobic completely stirred tank reactor (CSTR) having effective volumes of 17.7 and 10.5 litres, respectively, were used. After 41 days of start-up period, leachate was loaded to hybrid reactor at 10 different organic loading rates (OLRs). OLR was increased by increasing COD concentrations. COD removal efficiency of hybrid reactor was carried out at a maximum of 91%. A percentage of 96% of residual COD was removed in the aerobic reactor. NH4-N removal rate in CSTR was quite high. In addition, high methane content was obtained as 64% in the hybrid reactor. At the end of the study, after 170 operation days, it can be said that the hybrid reactor and CSTR were very effective for leachate treatment.

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.

Identification of limiting case between DBA and SBDBA (CL break area sensitivity): A new model for the boron injection system

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

Gonzalez Gonzalez, R.; Petruzzi, A.; D'Auria, F.

2012-07-01

Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and (e.g., oblique Control Rods, Positive Void coefficient) required a developed and validated complex three dimensional (3D) neutron kinetics (NK) coupled thermal hydraulic (TH) model. Reactor shut-down is obtained by oblique CRs and, during accidental conditions, by an emergency shut-down system (JDJ) injecting a highly concentrated boron solution (boron clouds) in the moderator tank, the boron clouds reconstruction is obtained using a CFD (CFX) code calculation. A complete LBLOCA calculation implies the application of the RELAP5-3D{sup C} system code. Within the framework of themore » third Agreement 'NA-SA - Univ. of Pisa' a new RELAP5-3D control system for the boron injection system was developed and implemented in the validated coupled RELAP5-3D/NESTLE model of the Atucha 2 NPP. The aim of this activity is to find out the limiting case (maximum break area size) for the Peak Cladding Temperature for LOCAs under fixed boundary conditions. (authors)« less

ETRMTR MECHANICAL SERVICES BUILDING, TRA653. CAMERA FACING NORTHWEST AS BUILDING ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR-MTR MECHANICAL SERVICES BUILDING, TRA-653. CAMERA FACING NORTHWEST AS BUILDING WAS NEARLY COMPLETE. INL NEGATIVE NO. 57-3653. K. Mansfield, Photographer, 7/22/1957 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

10 CFR 2.102 - Administrative review of application.

Code of Federal Regulations, 2011 CFR

2011-01-01

... completion of its review. (b) The Director, Office of Nuclear Reactor Regulation, Director, Office of New... as requested by the Commission. (c) The Director, Office of Nuclear Reactor Regulation, Director... Energy NUCLEAR REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF...

10 CFR 2.102 - Administrative review of application.

Code of Federal Regulations, 2012 CFR

2012-01-01

... completion of its review. (b) The Director, Office of Nuclear Reactor Regulation, Director, Office of New... as requested by the Commission. (c) The Director, Office of Nuclear Reactor Regulation, Director... Energy NUCLEAR REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF...

Neutron radiography experiments for verification of soluble boron mixing and transport modeling under natural circulation conditions

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

Feltus, M.A.; Morlang, G.M.

1996-06-01

The use of neutron radiography for visualization of fluid flow through flow visualization modules has been very successful. Current experiments at the Penn State Breazeale Reactor serve to verify the mixing and transport of soluble boron under natural flow conditions as would be experienced in a pressurized water reactor. Different flow geometries have been modeled including holes, slots, and baffles. Flow modules are constructed of aluminum box material 1 1/2 inches by 4 inches in varying lengths. An experimental flow system was built which pumps fluid to a head tank and natural circulation flow occurs from the head tank throughmore » the flow visualization module to be radiographed. The entire flow system is mounted on a portable assembly to allow placement of the flow visualization module in front of the neutron beam port. A neutron-transparent fluorinert fluid is used to simulate water at different densities. Boron is modeled by gadolinium oxide powder as a tracer element, which is placed in a mixing assembly and injected into the system by remote operated electric valve, once the reactor is at power. The entire sequence is recorded on real-time video. Still photographs are made frame-by-frame from the video tape. Computers are used to digitally enhance the video and still photographs. The data obtained from the enhancement will be used for verification of simple geometry predictions using the TRAC and RELAP thermal-hydraulic codes. A detailed model of a reactor vessel inlet plenum, downcomer region, flow distribution area and core inlet is being constructed to model the AP600 plenum. Successive radiography experiments of each section of the model under identical conditions will provide a complete vessel/core model for comparison with the thermal-hydraulic codes.« less

Integral reactor system and method for fuel cells

DOEpatents

Fernandes, Neil Edward; Brown, Michael S.; Cheekatamaria, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F.

2017-03-07

A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert higher hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

Integral reactor system and method for fuel cells

DOEpatents

Fernandes, Neil Edward; Brown, Michael S; Cheekatamarla, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F

2013-11-19

A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

Comparing the performance and operation stability of an SBR and MBBR for single-stage nitritation-anammox treating wastewater with high organic load.

PubMed

Lackner, Susanne; Horn, Harald

2013-01-01

Single stage nitritation-anammox reactors have gained increasing attention for their application in municipal and industrial wastewater treatment. The most commonly used system in municipal reject water treatment is at present the sequencing batch reactor (SBR), the moving-bed biofilm reactor (MBBR) is the second most common. However, little is known about their applicability to industrial wastewaters with high C/N ratios. This study presents a comparative approach to evaluate the performance of these two systems by changing the influent from reject water (C:N ratio 1:1) stepwise to an industrial wastewater (C:N ratio 3:1). An intentionally induced temperature drop that led to nitrite accumulation was also tested. The results showed that the MBBR (1.9 kg-N m(-3) d(-1)) was superior to the SBR (0.5 kg-N m(-3) d(-1)) with at maximum up to four times higher volumetric nitrogen removal rates. Both systems accumulated nitrite (> 100 mg-N L(-1)) during the temperature drop from 30 degrees C to as low as 18 degrees C (MBBR) and 20 degrees C (SBR), which subsequently resulted in almost complete loss in the removal capacities. However, the previous removal rates could be re-established in both systems within approximately 40 days. In comparison, the MBBR showed the more stable and higher performance even though higher nitrite concentrations (up to 500 mg-N L(-1)) were encountered. Overall, MBBR operation and handling was also easier and the system was more robust to disturbances compared to the SBR.

Rapid total volatile organic carbon quantification from microbial fermentation using a platinum catalyst and proton transfer reaction-mass spectrometry.

PubMed

Schoen, Heidi R; Peyton, Brent M; Knighton, W Berk

2016-12-01

A novel analytical system was developed to rapidly and accurately quantify total volatile organic compound (VOC) production from microbial reactor systems using a platinum catalyst and a sensitive CO 2 detector. This system allows nearly instantaneous determination of total VOC production by utilizing a platinum catalyst to completely and quantitatively oxidize headspace VOCs to CO 2 in coordination with a CO 2 detector. Measurement of respiratory CO 2 by bypassing the catalyst allowed the total VOC content to be determined from the difference in the two signals. To the best of our knowledge, this is the first instance of a platinum catalyst and CO 2 detector being used to quantify the total VOCs produced by a complex bioreactor system. Continuous recording of these CO 2 data provided a record of respiration and total VOC production throughout the experiments. Proton transfer reaction-mass spectrometry (PTR-MS) was used to identify and quantify major VOCs. The sum of the individual compounds measured by PTR-MS can be compared to the total VOCs quantified by the platinum catalyst to identify potential differences in detection, identification and calibration. PTR-MS measurements accounted on average for 94 % of the total VOC carbon detected by the platinum catalyst and CO 2 detector. In a model system, a VOC producing endophytic fungus Nodulisporium isolate TI-13 was grown in a solid state reactor utilizing the agricultural byproduct beet pulp as a substrate. Temporal changes in production of major volatile compounds (ethanol, methanol, acetaldehyde, terpenes, and terpenoids) were quantified by PTR-MS and compared to the total VOC measurements taken with the platinum catalyst and CO 2 detector. This analytical system provided fast, consistent data for evaluating VOC production in the nonhomogeneous solid state reactor system.

THE COOLING REQUIREMENTS AND PROCESS SYSTEMS OF THE SOUTH AFRICAN RESEARCH REACTOR, SAFARI 1

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

Colley, J.R.

1962-12-01

The SAFARI 1 research reactor is cooled and moderated by light water. There are three process systems, a primary water system which cools the reactor core and surroundings, a pool water system, and a secondary water system which removes the heat from the primary and pool systems. The cooling requirements for the reactor core and experimental facilities are outlined, and the cooling and purification functions of the three process systems are described. (auth)

Thermochemical reactor systems and methods

DOEpatents

Lipinski, Wojciech; Davidson, Jane Holloway; Chase, Thomas Richard

2016-11-29

Thermochemical reactor systems that may be used to produce a fuel, and methods of using the thermochemical reactor systems, utilizing a reactive cylindrical element, an optional energy transfer cylindrical element, an inlet gas management system, and an outlet gas management system.

Covariance generation and uncertainty propagation for thermal and fast neutron induced fission yields

NASA Astrophysics Data System (ADS)

Terranova, Nicholas; Serot, Olivier; Archier, Pascal; De Saint Jean, Cyrille; Sumini, Marco

2017-09-01

Fission product yields (FY) are fundamental nuclear data for several applications, including decay heat, shielding, dosimetry, burn-up calculations. To be safe and sustainable, modern and future nuclear systems require accurate knowledge on reactor parameters, with reduced margins of uncertainty. Present nuclear data libraries for FY do not provide consistent and complete uncertainty information which are limited, in many cases, to only variances. In the present work we propose a methodology to evaluate covariance matrices for thermal and fast neutron induced fission yields. The semi-empirical models adopted to evaluate the JEFF-3.1.1 FY library have been used in the Generalized Least Square Method available in CONRAD (COde for Nuclear Reaction Analysis and Data assimilation) to generate covariance matrices for several fissioning systems such as the thermal fission of U235, Pu239 and Pu241 and the fast fission of U238, Pu239 and Pu240. The impact of such covariances on nuclear applications has been estimated using deterministic and Monte Carlo uncertainty propagation techniques. We studied the effects on decay heat and reactivity loss uncertainty estimation for simplified test case geometries, such as PWR and SFR pin-cells. The impact on existing nuclear reactors, such as the Jules Horowitz Reactor under construction at CEA-Cadarache, has also been considered.

HORIZONTAL BOILING REACTOR SYSTEM

DOEpatents

Treshow, M.

1958-11-18

Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

Identification of Some of the Major Groups of Bacteria in Efficient and Nonefficient Biological Phosphorus Removal Activated Sludge Systems

PubMed Central

Bond, Philip L.; Erhart, Robert; Wagner, Michael; Keller, Jürg; Blackall, Linda L.

1999-01-01

To investigate the bacteria that are important to phosphorus (P) removal in activated sludge, microbial populations were analyzed during the operation of a laboratory-scale reactor with various P removal performances. The bacterial population structure, analyzed by fluorescence in situ hybridization (FISH) with oligonucleotides probes complementary to regions of the 16S and 23S rRNAs, was associated with the P removal performance of the reactor. At one stage of the reactor operation, chemical characterization revealed that extremely poor P removal was occurring. However, like in typical P-removing sludges, complete anaerobic uptake of the carbon substrate occurred. Bacteria inhibiting P removal overwhelmed the reactor, and according to FISH, bacteria of the β subclass of the class Proteobacteria other than β-1 or β-2 were dominant in the sludge (58% of the population). Changes made to the operation of the reactor led to the development of a biomass population with an extremely good P removal capacity. The biochemical transformations observed in this sludge were characteristic of typical P-removing activated sludge. The microbial population analysis of the P-removing sludge indicated that bacteria of the β-2 subclass of the class Proteobacteria and actinobacteria were dominant (55 and 35%, respectively), therefore implicating bacteria from these groups in high-performance P removal. The changes in operation that led to the improved performance of the reactor included allowing the pH to rise during the anaerobic period, which promoted anaerobic phosphate release and possibly caused selection against non-phosphate-removing bacteria. PMID:10473419

Autonomous Control of Space Nuclear Reactors

NASA Technical Reports Server (NTRS)

Merk, John

2013-01-01

Nuclear reactors to support future robotic and manned missions impose new and innovative technological requirements for their control and protection instrumentation. Long-duration surface missions necessitate reliable autonomous operation, and manned missions impose added requirements for failsafe reactor protection. There is a need for an advanced instrumentation and control system for space-nuclear reactors that addresses both aspects of autonomous operation and safety. The Reactor Instrumentation and Control System (RICS) consists of two functionally independent systems: the Reactor Protection System (RPS) and the Supervision and Control System (SCS). Through these two systems, the RICS both supervises and controls a nuclear reactor during normal operational states, as well as monitors the operation of the reactor and, upon sensing a system anomaly, automatically takes the appropriate actions to prevent an unsafe or potentially unsafe condition from occurring. The RPS encompasses all electrical and mechanical devices and circuitry, from sensors to actuation device output terminals. The SCS contains a comprehensive data acquisition system to measure continuously different groups of variables consisting of primary measurement elements, transmitters, or conditioning modules. These reactor control variables can be categorized into two groups: those directly related to the behavior of the core (known as nuclear variables) and those related to secondary systems (known as process variables). Reliable closed-loop reactor control is achieved by processing the acquired variables and actuating the appropriate device drivers to maintain the reactor in a safe operating state. The SCS must prevent a deviation from the reactor nominal conditions by managing limitation functions in order to avoid RPS actions. The RICS has four identical redundancies that comply with physical separation, electrical isolation, and functional independence. This architecture complies with the safety requirements of a nuclear reactor and provides high availability to the host system. The RICS is intended to interface with a host computer (the computer of the spacecraft where the reactor is mounted). The RICS leverages the safety features inherent in Earth-based reactors and also integrates the wide range neutron detector (WRND). A neutron detector provides the input that allows the RICS to do its job. The RICS is based on proven technology currently in use at a nuclear research facility. In its most basic form, the RICS is a ruggedized, compact data-acquisition and control system that could be adapted to support a wide variety of harsh environments. As such, the RICS could be a useful instrument outside the scope of a nuclear reactor, including military applications where failsafe data acquisition and control is required with stringent size, weight, and power constraints.

Passive cooling system for liquid metal cooled nuclear reactors with backup coolant flow path

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1993-01-01

A liquid metal cooled nuclear fission reactor plant having a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during reactor shutdown, or heat produced during a mishap. This reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary system when rendered inoperable.

Systems definition space-based power conversion systems. [for satellite power transmission to earth

NASA Technical Reports Server (NTRS)

1976-01-01

Potential space-located systems for the generation of electrical power for use on Earth are discussed and include: (1) systems producing electrical power from solar energy; (2) systems producing electrical power from nuclear reactors; and (3) systems for augmenting ground-based solar power plants by orbital sunlight reflectors. Systems (1) and (2) would utilize a microwave beam system to transmit their output to Earth. Configurations implementing these concepts were developed through an optimization process intended to yield the lowest cost for each. A complete program was developed for each concept, identifying required production rates, quantities of launches, required facilities, etc. Each program was costed in order to provide the electric power cost appropriate to each concept.

Anaerobic co-digestion of food waste and landfill leachate in single-phase batch reactors

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

Liao, Xiaofeng; Zhu, Shuangyan; Zhong, Delai

Highlights: • Anaerobic co-digestion strategy for food waste treatment at OLR 41.8 g VS/L. • A certain amount of raw leachate effectively relieved acidic inhibition. • The study showed that food waste was completely degraded. - Abstract: In order to investigate the effect of raw leachate on anaerobic digestion of food waste, co-digestions of food waste with raw leachate were carried out. A series of single-phase batch mesophilic (35 ± 1 °C) anaerobic digestions were performed at a food waste concentration of 41.8 g VS/L. The results showed that inhibition of biogas production by volatile fatty acids (VFA) occurred withoutmore » raw leachate addition. A certain amount of raw leachate in the reactors effectively relieved acidic inhibition caused by VFA accumulation, and the system maintained stable with methane yield of 369–466 mL/g VS. Total ammonia nitrogen introduced into the digestion systems with initial 2000–3000 mgNH{sub 4}–N/L not only replenished nitrogen for bacterial growth, but also formed a buffer system with VFA to maintain a delicate biochemical balance between the acidogenic and methanogenic microorganisms. UV spectroscopy and fluorescence excitation–emission matrix spectroscopy data showed that food waste was completely degraded. We concluded that using raw leachate for supplement water addition and pH modifier on anaerobic digestion of food waste was effective. An appropriate fraction of leachate could stimulate methanogenic activity and enhance biogas production.« less

Analyses of the reflector tank, cold source, and beam tube cooling for ANS reactor. [Advanced Neutron Source (ANS)

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

Marland, S.

1992-07-01

This report describes my work as an intern with Martin Marietta Energy Systems, Inc., in the summer of 1991. I was assigned to the Reactor Technology Engineering Department, working on the Advanced Neutron Source (ANS). My first project was to select and analyze sealing systems for the top of the diverter/reflector tank. This involved investigating various metal seals and calculating the forces necessary to maintain an adequate seal. The force calculations led to an analysis of several bolt patterns and lockring concepts that could be used to maintain a seal on the vessel. Another project involved some pressure vessel stressmore » calculations and the calculation of the center of gravity for the cold source assembly. I also completed some sketches of possible cooling channel patterns for the inner vessel of the cold source. In addition, I worked on some thermal design analyses for the reflector tank and beam tubes, including heat transfer calculations and assisting in Patran and Pthermal analyses. To supplement the ANS work, I worked on other projects. I completed some stress/deflection analyses on several different beams. These analyses were done with the aid of CAASE, a beam-analysis software package. An additional project involved bending analysis on a carbon removal system. This study was done to find the deflection of a complex-shaped beam when loaded with a full waste can.« less

Low cost solar aray project: Experimental process system development unit for producing semiconductor-grade silicon using the silane-to-silicon process

NASA Technical Reports Server (NTRS)

1981-01-01

This phase consists of the engineering design, fabrication, assembly, operation, economic analysis, and process support R&D for an Experimental Process System Development Unit (EPSDU). The mechanical bid package was issued and the bid responses are under evaluation. Similarly, the electrical bid package was issued, however, responses are not yet due. The majority of all equipment is on order or has been received at the EPSDU site. The pyrolysis/consolidation process design package was issued. Preparation of process and instrumentation diagram for the free-space reactor was started. In the area of melting/consolidation, Kayex successfully melted chunk silicon and have produced silicon shot. The free-space reactor powder was successfully transported pneumatically from a storage bin to the auger feeder twenty-five feet up and was melted. The fluid-bed PDU has successfully operated at silane feed concentrations up to 21%. The writing of the operating manual has started. Overall, the design phase is nearing completion.

Cultivation of aerobic granular sludge for rubber wastewater treatment.

PubMed

Rosman, Noor Hasyimah; Nor Anuar, Aznah; Othman, Inawati; Harun, Hasnida; Sulong Abdul Razak, Muhammad Zuhdi; Elias, Siti Hanna; Mat Hassan, Mohd Arif Hakimi; Chelliapan, Shreesivadass; Ujang, Zaini

2013-02-01

Aerobic granular sludge (AGS) was successfully cultivated at 27±1 °C and pH 7.0±1 during the treatment of rubber wastewater using a sequential batch reactor system mode with complete cycle time of 3 h. Results showed aerobic granular sludge had an excellent settling ability and exhibited exceptional performance in the organics and nutrients removal from rubber wastewater. Regular, dense and fast settling granule (average diameter, 1.5 mm; settling velocity, 33 m h(-1); and sludge volume index, 22.3 mL g(-1)) were developed in a single reactor. In addition, 96.5% COD removal efficiency was observed in the system at the end of the granulation period, while its ammonia and total nitrogen removal efficiencies were up to 94.7% and 89.4%, respectively. The study demonstrated the capabilities of AGS development in a single, high and slender column type-bioreactor for the treatment of rubber wastewater. Copyright © 2012 Elsevier Ltd. All rights reserved.

Spatiotemporal patterns in reaction-diffusion system and in a vibrated granular bed

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

Swinney, H.L.; Lee, K.J.; McCormick, W.D.

Experiments on a quasi-two-dimensional reaction-diffusion system reveal transitions from a uniform state to stationary hexagonal, striped, and rhombic spatial patterns. For other reactor conditions lamellae and self-replicating spot patterns are observed. These patterns form in continuously fed thin gel reactors that can be maintained indefinitely in well-defined nonequilibrium states. Reaction-diffusion models with two chemical species yield patterns similar to those observed in the experiments. Pattern formation is also being examined in vertically oscillated thin granular layers (typically 3-30 particle diameters deep). For small acceleration amplitudes, a granular layer is flat, but above a well-defined critical acceleration amplitude, spatial patterns spontaneouslymore » form. Disordered time-dependent granular patterns are observed as well as regular patterns of squares, stripes, and hexagons. A one-dimensional model consisting of a completely inelastic ball colliding with a sinusoidally oscillating platform provides a semi-quantitative description of most of the observed bifurcations between the different spatiotemporal regimes.« less

Influence of denitrification reactor retention time distribution (RTD) on dissolved oxygen control and nitrogen removal efficiency.

PubMed

Raboni, Massimo; Gavasci, Renato; Viotti, Paolo

2015-01-01

Low concentrations of dissolved oxygen (DO) are usually found in biological anoxic pre-denitrification reactors, causing a reduction in nitrogen removal efficiency. Therefore, the reduction of DO in such reactors is fundamental for achieving good nutrient removal. The article shows the results of an experimental study carried out to evaluate the effect of the anoxic reactor hydrodynamic model on both residual DO concentration and nitrogen removal efficiency. In particular, two hydrodynamic models were considered: the single completely mixed reactor and a series of four reactors that resemble plug-flow behaviour. The latter prove to be more effective in oxygen consumption, allowing a lower residual DO concentration than the former. The series of reactors also achieves better specific denitrification rates and higher denitrification efficiency. Moreover, the denitrification food to microrganism (F:M) ratio (F:MDEN) demonstrates a relevant synergic action in both controlling residual DO and improving the denitrification performance.

Shutdown system for a nuclear reactor

DOEpatents

Groh, E.F.; Olson, A.P.; Wade, D.C.; Robinson, B.W.

1984-06-05

An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion. 8 figs.

Shutdown system for a nuclear reactor

DOEpatents

Groh, Edward F.; Olson, Arne P.; Wade, David C.; Robinson, Bryan W.

1984-01-01

An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion.

Digital computer study of nuclear reactor thermal transients during startup of 60-kWe Brayton power conversion system

NASA Technical Reports Server (NTRS)

Jefferies, K. S.; Tew, R. C.

1974-01-01

A digital computer study was made of reactor thermal transients during startup of the Brayton power conversion loop of a 60-kWe reactor Brayton power system. A startup procedure requiring the least Brayton system complication was tried first; this procedure caused violations of design limits on key reactor variables. Several modifications of this procedure were then found which caused no design limit violations. These modifications involved: (1) using a slower rate of increase in gas flow; (2) increasing the initial reactor power level to make the reactor respond faster; and (3) appropriate reactor control drum manipulation during the startup transient.

Surveillance application using patten recognition software at the EBR-II Reactor Facility

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

Olson, D.L.

1992-01-01

The System State Analyzer (SSA) is a software based pattern recognition system. For the past several year this system has been used at Argonne National Laboratory's Experimental Breeder Reactor 2 (EBR-2) reactor for detection of degradation and other abnormalities in plant systems. Currently there are two versions of the SSA being used at EBR-2. One version of SSA is used for daily surveillance and trending of the reactor delta-T and startups of the reactor. Another version of the SSA is the QSSA which is used to monitor individual systems of the reactor such as the Secondary Sodium System, Secondary Sodiummore » Pumps, and Steam Generator. This system has been able to detect problems such as signals being affected by temperature variations due to a failing temperature controller.« less

Surveillance application using patten recognition software at the EBR-II Reactor Facility

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

Olson, D.L.

1992-05-01

The System State Analyzer (SSA) is a software based pattern recognition system. For the past several year this system has been used at Argonne National Laboratory`s Experimental Breeder Reactor 2 (EBR-2) reactor for detection of degradation and other abnormalities in plant systems. Currently there are two versions of the SSA being used at EBR-2. One version of SSA is used for daily surveillance and trending of the reactor delta-T and startups of the reactor. Another version of the SSA is the QSSA which is used to monitor individual systems of the reactor such as the Secondary Sodium System, Secondary Sodiummore » Pumps, and Steam Generator. This system has been able to detect problems such as signals being affected by temperature variations due to a failing temperature controller.« less

Space shuttle aps propellant thermal conditioner study

NASA Technical Reports Server (NTRS)

Fulton, D. L.

1973-01-01

An analytical and experimental effort was completed to evaluate a baffle type thermal conditioner for superheating O2 and H2 at supercritical pressures. The thermal conditioner consisted of a heat exchanger and an integral reactor (gas generator) operating on O2/H2 propellants. Primary emphasis was placed on the hydrogen conditioner with some effort on the oxygen conditioner and a study completed of alternate concepts for use in conditioning oxygen. A hydrogen conditioner was hot fire tested under a range of conditions to establish ignition, heat exchange and response parameters. A parallel technology task was completed to further evaluate the integral reactor and heat exchanger with the side mounted electrical spark igniter.

Comparing the new generation accelerator driven subcritical reactor system (ADS) to traditional critical reactors

NASA Astrophysics Data System (ADS)

Kemah, Elif; Akkaya, Recep; Tokgöz, Seyit Rıza

2017-02-01

In recent years, the accelerator driven subcritical reactors have taken great interest worldwide. The Accelerator Driven System (ADS) has been used to produce neutron in subcritical state by the external proton beam source. These reactors, which are hybrid systems, are important in production of clean and safe energy and conversion of radioactive waste. The ADS with the selection of reliability and robust target materials have been the new generation of fission reactors. In addition, in the ADS Reactors the problems of long-lived radioactive fission products and waste actinides encountered in the fission process of the reactor during incineration can be solved, and ADS has come to the forefront of thorium as fuel for the reactors.

Liquid metal cooled nuclear reactor plant system

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1993-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

Morphological evolution of copper nanoparticles: Microemulsion reactor system versus batch reactor system

NASA Astrophysics Data System (ADS)

Xia, Ming; Tang, Zengmin; Kim, Woo-Sik; Yu, Taekyung; Park, Bum Jun

2017-07-01

In the synthesis of nanoparticles, the reaction rate is important to determine the morphology of nanoparticles. We investigated morphology evolution of Cu nanoparticles in this two different reactors, microemulsion reactor and batch reactor. In comparison with the batch reactor system, the enhanced mass and heat transfers in the emulsion system likely led to the relatively short nucleation time and the highly homogeneous environment in the reaction mixture, resulting in suppressing one or two dimensional growth of the nanoparticles. We believe that this work can offer a good model system to quantitatively understand the crystal growth mechanism that depends strongly on the local monomer concentration, the efficiency of heat transfer, and the relative contribution of the counter ions (Br- and Cl-) as capping agents.

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

NONE

Digital instrumentation and controls system technique is being introduced in new constructed research reactor or life extension of older research reactor. Digital systems are easy to change and optimize but the validated process for them is required. Also, to reduce project risk or cost, we have to make it sure that configuration and control functions is right before the commissioning phase on research reactor. For this purpose, simulators have been widely used in developing control systems in automotive and aerospace industries. In these literatures, however, very few of these can be found regarding test on the control system of researchmore » reactor with simulator. Therefore, this paper proposes a simulation platform to verify the performance of RRS (Reactor Regulating System) for research reactor. This simulation platform consists of the reactor simulation model and the interface module. This simulation platform is applied to I and C upgrade project of TRIGA reactor, and many problems of RRS configuration were found and solved. And it proved that the dynamic performance testing based on simulator enables significant time saving and improves economics and quality for RRS in the system test phase. (authors)« 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.

Reactor antineutrino fluxes – Status and challenges

DOE PAGES

Huber, Patrick

2016-04-22

Here, we describe the current understanding of reactor antineutrino fluxes and point out some recent developments. This is not intended to be a complete review of this vast topic but merely a selection of observations and remarks, which despite their incompleteness, will highlight the status and the challenges of this field.

10 CFR 50.82 - Termination of license.

Code of Federal Regulations, 2011 CFR

2011-01-01

... to the NRC, consistent with the requirements of § 50.4(b)(8); (ii) Once fuel has been permanently... fuel from the reactor vessel, or when a final legally effective order to permanently cease operations... emplacement or retention of fuel into the reactor vessel. (3) Decommissioning will be completed within 60...

Reactor performance and microbial community dynamics during anaerobic co-digestion of municipal wastewater sludge with restaurant grease waste at steady state and overloading stages.

PubMed

Razaviarani, Vahid; Buchanan, Ian D

2014-11-01

Linkage between reactor performance and microbial community dynamics was investigated during mesophilic anaerobic co-digestion of restaurant grease waste (GTW) with municipal wastewater sludge (MWS) using 10L completely mixed reactors and a 20day SRT. Test reactors received a mixture of GTW and MWS while control reactors received only MWS. Addition of GTW to the test reactors enhanced the biogas production and methane yield by up to 65% and 120%, respectively. Pyrosequencing revealed that Methanosaeta and Methanomicrobium were the dominant acetoclastic and hydrogenotrophic methanogen genera, respectively, during stable reactor operation. The number of Methanosarcina and Methanomicrobium sequences increased and that of Methanosaeta declined when the proportion of GTW in the feed was increased to cause an overload condition. Under this overload condition, the pH, alkalinity and methane production decreased and VFA concentrations increased dramatically. Candidatus cloacamonas, affiliated within phylum Spirochaetes, were the dominant bacterial genus at all reactor loadings. Copyright © 2014 Elsevier Ltd. All rights reserved.

Electrochemical processing of solid waste

NASA Technical Reports Server (NTRS)

Bockris, J. OM.; Hitchens, G. D.; Kaba, L.

1988-01-01

The investigation into electrolysis as a means of waste treatment and recycling on manned space missions is described. The electrochemical reactions of an artificial fecal waste mixture was examined. Waste electrolysis experiments were performed in a single compartment reactor, on platinum electrodes, to determine conditions likely to maximize the efficiency of oxidation of fecal waste material to CO2. The maximum current efficiencies for artificial fecal waste electrolysis to CO2 was found to be around 50 percent in the test apparatus. Experiments involving fecal waste oxidation on platinum indicates that electrodes with a higher overvoltage for oxygen evolution such as lead dioxide will give a larger effective potential range for organic oxidation reactions. An electrochemical packed column reactor was constructed with lead dioxide as electrode material. Preliminary experiments were performed using a packed-bed reactor and continuous flow techniques showing this system may be effective in complete oxidation of fecal material. The addition of redox mediator Ce(3+)/Ce(4+) enhances the oxidation process of biomass components. Scientific literature relevant to biomass and fecal waste electrolysis were reviewed.

Methyl chloride via oxyhydrochlorination of methane: A building block for chemicals and fuels from natural gas

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

Benson, R.L.; Brown, S.S.D.; Ferguson, S.P.

1995-12-31

The objectives of this program are to (a) develop a process for converting natural gas to methyl chloride via an oxyhydrochlorination route using highly selective, stable catalysts in a fixed-bed, (b) design a reactor capable of removing the large amount of heat generated in the process so as to control the reaction, (c) develop a recovery system capable of removing the methyl chloride from the product stream and (d) determine the economics and commercial viability of the process. The general approach has been as follows: (a) design and build a laboratory scale reactor, (b) define and synthesize suitable OHC catalystsmore » for evaluation, (c) select first generation OHC catalyst for Process Development Unit (PDU) trials, (d) design, construct and startup PDU, (e) evaluate packed bed reactor design, (f) optimize process, in particular, product recovery operations, (g) determine economics of process, (h) complete preliminary engineering design for Phase II and (i) make scale-up decision and formulate business plan for Phase II. Conclusions regarding process development and catalyst development are presented.« less

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

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

E.T. Robinson; John Sirman; Prasad Apte

2005-05-01

This final report summarizes work accomplished in the Program from January 1, 2001 through December 31, 2004. Most of the key technical objectives for this program were achieved. A breakthrough material system has lead to the development of an OTM (oxygen transport membrane) compact planar reactor design capable of producing either syngas or hydrogen. The planar reactor shows significant advantages in thermal efficiency and a step change reduction in costs compared to either autothermal reforming or steam methane reforming with CO{sub 2} recovery. Syngas derived ultra-clean transportation fuels were tested in the Nuvera fuel cell modular pressurized reactor and inmore » International Truck and Engine single cylinder test engines. The studies compared emission and engine performance of conventional base fuels to various formulations of ultra-clean gasoline or diesel fuels. A proprietary BP oxygenate showed significant advantage in both applications for reducing emissions with minimal impact on performance. In addition, a study to evaluate new fuel formulations for an HCCI engine was completed.« less

Characterization of biofilm in 200W fluidized bed reactors

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

Lee, Michelle H.; Saurey, Sabrina D.; Lee, Brady D.

2014-09-29

Contaminated groundwater beneath the 200 West Area at the Hanford Site in Southeast Washington is currently being treated using a pump and treat system to remove organics, inorganics, radionuclides, and metals. A granular activated carbon-based fluidized bed reactor (FBR) has been added to remove nitrate, hexavalent chromium and carbon tetrachloride. Initial analytical results indicated the microorganisms effectively reduced many of the contaminants to less than cleanup levels. However shortly thereafter operational upsets of the FBR include carbon carry over, over production of microbial extracellular polymeric substance (biofilm) materials, and over production of hydrogen sulfide. As a result detailed investigations weremore » undertaken to understand the functional diversity and activity of the microbial community present in the FBR over time. Molecular analyses including terminal restriction fragment length polymorphism analysis, quantitative polymerase chain reaction and fluorescent in situ hybridization analyses were performed on the microbial community extracted from the biofilm within the bed and from the inoculum, to determine functional dynamics of the FBR bed over time and following operational changes. Findings from these analyses indicated: 1) the microbial community within the bed was completely different than community used for inoculation, and was likely from the groundwater; 2) analyses early in the testing showed an FBR community dominated by a few Curvibacter and Flavobacterium species; 3) the final sample taken indicated that the microbial community in the FBR bed had become more diverse; and 4) qPCR analyses indicated that bacteria involved in nitrogen cycling, including denitrifiers and anaerobic ammonia oxidizing bacteria, were dominant in the bed. These results indicate that molecular tools can be powerful for determining functional diversity within FBR type reactors. Coupled with micronutrient, influent and effluent chemistry evaluations, a more complete understanding of the balance between system additions (nutrients, groundwater) and biology can be achieved, thus increasing long-term predictions of performance. These analyses uniquely provide information that can be used in optimizing the overall performance, efficiency, and stability of the system both in real time as well as over the long-term, as the system design is altered or improved and/or new streams are added.« less

A liquid-metal filling system for pumped primary loop space reactors

NASA Astrophysics Data System (ADS)

Crandall, D. L.; Reed, W. C.

Some concepts for the SP-100 space nuclear power reactor use liquid metal as the primary coolant in a pumped loop. Prior to filling ground engineering test articles or reactor systems, the liquid metal must be purified and circulated through the reactor primary system to remove contaminants. If not removed, these contaminants enhance corrosion and reduce reliability. A facility was designed and built to support Department of Energy Liquid Metal Fast Breeder Reactor tests conducted at the Idaho National Engineering Laboratory. This test program used liquid sodium to cool nuclear fuel in in-pile experiments; thus, a system was needed to store and purify sodium inventories and fill the experiment assemblies. This same system, with modifications and potential changeover to lithium or sodium-potassium (NaK), can be used in the Space Nuclear Power Reactor Program. This paper addresses the requirements, description, modifications, operation, and appropriateness of using this liquid-metal system to support the SP-100 space reactor program.

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

Brunett, A. J.; Fei, T.; Strons, P. S.

The Transient Reactor Test Facility (TREAT), located at Idaho National Laboratory (INL), is a test facility designed to evaluate the performance of reactor fuels and materials under transient accident conditions. The facility, an air-cooled, graphite-moderated reactor designed to utilize fuel containing high-enriched uranium (HEU), has been in non-operational standby status since 1994. Currently, in support of the missions of the Department of Energy (DOE) National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Reactor Conversion Program, a new core design is being developed for TREAT that will utilize low-enriched uranium (LEU). The primary objective of this conversion effort ismore » to design an LEU core that is capable of meeting the performance characteristics of the existing HEU core. Minimal, if any, changes are anticipated for the supporting systems (e.g. reactor trip system, filtration/cooling system, etc.); therefore, the LEU core must also be able to function with the existing supporting systems, and must also satisfy acceptable safety limits. In support of the LEU conversion effort, a range of ancillary safety analyses are required to evaluate the LEU core operation relative to that of the existing facility. These analyses cover neutronics, shielding, and thermal hydraulic topics that have been identified as having the potential to have reduced safety margins due to conversion to LEU fuel, or are required to support the required safety analyses documentation. The majority of these ancillary tasks have been identified in [1] and [2]. The purpose of this report is to document the ancillary safety analyses that have been performed at Argonne National Laboratory during the early stages of the LEU design effort, and to describe ongoing and anticipated analyses. For all analyses presented in this report, methodologies are utilized that are consistent with, or improved from, those used in analyses for the HEU Final Safety Analysis Report (FSAR) [3]. Depending on the availability of historical data derived from HEU TREAT operation, results calculated for the LEU core are compared to measurements obtained from HEU TREAT operation. While all analyses in this report are largely considered complete and have been reviewed for technical content, it is important to note that all topics will be revisited once the LEU design approaches its final stages of maturity. For most safety significant issues, it is expected that the analyses presented here will be bounding, but additional calculations will be performed as necessary to support safety analyses and safety documentation. It should also be noted that these analyses were completed as the LEU design evolved, and therefore utilized different LEU reference designs. Preliminary shielding, neutronic, and thermal hydraulic analyses have been completed and have generally demonstrated that the various LEU core designs will satisfy existing safety limits and standards also satisfied by the existing HEU core. These analyses include the assessment of the dose rate in the hodoscope room, near a loaded fuel transfer cask, above the fuel storage area, and near the HEPA filters. The potential change in the concentration of tramp uranium and change in neutron flux reaching instrumentation has also been assessed. Safety-significant thermal hydraulic items addressed in this report include thermally-induced mechanical distortion of the grid plate, and heating in the radial reflector.« less

Thermodynamics of soluble fission products cesium and iodine in the Molten Salt Reactor

NASA Astrophysics Data System (ADS)

Capelli, E.; Beneš, O.; Konings, R. J. M.

2018-04-01

The present study describes the full thermodynamic assessment of the Li,Cs,Th//F,I system. The existing database for the relevant fluoride salts considered as fuel for the Molten Salt Reactor (MSR) has been extended with two key fission products, cesium and iodine. A complete evaluation of all the common-ion binary and ternary sub-systems of the LiF-ThF4-CsF-LiI-ThI4-CsI system has been performed and the optimized parameters are presented in this work. New equilibrium data have been measured using Differential Scanning Calorimetry and were used to assess the reciprocal ternary systems and confirm the extrapolated phase diagrams. The developed database significantly contributes to the understanding of the behaviour of cesium and iodine in the MSR, which strongly depends on their concentration and chemical form. Cesium bonded with fluorine is well retained in the fuel mixture while in the form of CsI the solubility of these elements is very limited. Finally, the influence of CsI and CsF on the physico-chemical properties of the fuel mixture was calculated as function of composition.

Cost-Effective Systems for Atomic Layer Deposition

ERIC Educational Resources Information Center

Lubitz, Michael; Medina, Phillip A., IV; Antic, Aleks; Rosin, Joseph T.; Fahlman, Bradley D.

2014-01-01

Herein, we describe the design and testing of two different home-built atomic layer deposition (ALD) systems for the growth of thin films with sub-monolayer control over film thickness. The first reactor is a horizontally aligned hot-walled reactor with a vacuum purging system. The second reactor is a vertically aligned cold-walled reactor with a…

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... behavior of the reactor system during a loss-of-coolant accident. Comparisons to applicable experimental...

Is there a functional neural correlate of individual differences in cardiovascular reactivity?

PubMed

Gianaros, Peter J; May, J Christopher; Siegle, Greg J; Jennings, J Richard

2005-01-01

The present study tested whether individuals who differ in the magnitude of their blood pressure reactions to a behavioral stressor also differ in their stressor-induced patterns of functional neural activation. Sixteen participants (7 men, 9 women aged 47 to 72 years) were classified as high (n = 8) or low (n = 8) blood pressure reactors by the magnitude and temporal consistency of their systolic blood pressure (SBP) reaction to a Stroop color-word interference stressor. Both high and low SBP reactors completed this Stroop stressor while their task-related changes in blood pressure and functional neural activity were assessed in a blocked functional magnetic resonance imaging design. In both high and low SBP reactors, the Stroop-stressor engaged the anterior cingulate, orbitofrontal, insular, posterior parietal, and the dorsolateral prefrontal regions of the cortex, the thalamus, and the cerebellum. Compared with low reactors, however, high reactors not only showed a larger magnitude increase in SBP to the Stroop stressor, but also an increased activation of the posterior cingulate cortex. A behavioral stressor that is used widely in cardiovascular reactivity research, the Stroop stressor, engages brain systems that are thought to support both stressor processing and cardiovascular reactivity. Increased activation of the posterior cingulate, a brain region implicated in vigilance to the environment and evaluative emotional processes, may be a functional neural correlate of an individual's tendency to show large-magnitude (exaggerated) blood pressure reactions to behavioral stressors.

Performance Evaluation of Staged Bosch Process for CO2 Reduction to Produce Life Support Consumables

NASA Technical Reports Server (NTRS)

Vilekar, Saurabh A.; Hawley, Kyle; Junaedi, Christian; Walsh, Dennis; Roychoudhury, Subir; Abney. Morgan B.; Mansell, James M.

2012-01-01

Utilizing carbon dioxide to produce water and hence oxygen is critical for sustained manned missions in space, and to support both NASA's cabin Atmosphere Revitalization System (ARS) and In-Situ Resource Utilization (ISRU) concepts. For long term missions beyond low Earth orbit, where resupply is significantly more difficult and costly, open loop ARS, like Sabatier, consume inputs such as hydrogen. The Bosch process, on the other hand, has the potential to achieve complete loop closure and is hence a preferred choice. However, current single stage Bosch reactor designs suffer from a large recycle penalty due to slow reaction rates and the inherent limitation in approaching thermodynamic equilibrium. Developmental efforts are seeking to improve upon the efficiency (hence reducing the recycle penalty) of current single stage Bosch reactors which employ traditional steel wool catalysts. Precision Combustion, Inc. (PCI), with support from NASA, has investigated the potential for utilizing catalysts supported over short-contact time Microlith substrates for the Bosch reaction to achieve faster reaction rates, higher conversions, and a reduced recycle flows. Proof-of-concept testing was accomplished for a staged Bosch process by splitting the chemistry in two separate reactors, first being the reverse water-gas-shift (RWGS) and the second being the carbon formation reactor (CFR) via hydrogenation and/or Boudouard. This paper presents the results from this feasibility study at various operating conditions. Additionally, results from two 70 hour durability tests for the RWGS reactor are discussed.

Carbofuran removal in continuous-photocatalytic reactor: Reactor optimization, rate-constant determination and carbofuran degradation pathway analysis.

PubMed

Vishnuganth, M A; Remya, Neelancherry; Kumar, Mathava; Selvaraju, N

2017-05-04

Carbofuran (CBF) removal in a continuous-flow photocatalytic reactor with granular activated carbon supported titanium dioxide (GAC-TiO 2 ) catalyst was investigated. The effects of feed flow rate, TiO 2 concentration and addition of supplementary oxidants on CBF removal were investigated. The central composite design (CCD) was used to design the experiments and to estimate the effects of feed flow rate and TiO 2 concentration on CBF removal. The outcome of CCD experiments demonstrated that reactor performance was influenced mainly by feed flow rate compared to TiO 2 concentration. A second-order polynomial model developed based on CCD experiments fitted the experimental data with good correlation (R 2 ∼ 0.964). The addition of 1 mL min -1 hydrogen peroxide has shown complete CBF degradation and 76% chemical oxygen demand removal under the following operating conditions of CBF ∼50 mg L -1 , TiO 2 ∼5 mg L -1 and feed flow rate ∼82.5 mL min -1 . Rate constant of the photodegradation process was also calculated by applying the kinetic data in pseudo-first-order kinetics. Four major degradation intermediates of CBF were identified using GC-MS analysis. As a whole, the reactor system and GAC-TiO 2 catalyst used could be constructive in cost-effective CBF removal with no impact to receiving environment through getaway of photocatalyst.

Trench fast reactor design using the microcomputer

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

Rohach, A.F.; Sankoorikal, J.T.; Schmidt, R.R.

1987-01-01

This project is a study of alternative liquid-metal-cooled fast power reactor system concepts. Specifically, an unconventional primary system is being conceptually designed and evaluated. The project design is based primarily on microcomputer analysis through the use of computational modules. The reactor system concept is a long, narrow pool with a long, narrow reactor called a trench-type pool reactor in it. The reactor consists of five core-blanket modules in a line. Specific power is to be modest, permitting long fuel residence time. Two fuel cycles are currently being considered. The reactor design philosophy is that of the inherently safe concept. Thismore » requires transient analysis dependent on reactivity coefficients: prompt fuel, including Doppler and expansion, fuel expansion, sodium temperature and void, and core expansion. Conceptual reactor design is done on a microcomputer. A part of the trench reactor project is to develop a microcomputer-based system that can be used by the user for scoping studies and design. Current development includes the neutronics and fuel management aspects of the design. Thermal-hydraulic analysis and economics are currently being incorporated into the microcomputer system. The system is menu-driven including preparation of program input data and of output data for displays in graphics form.« less

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

Microwave Plasma Hydrogen Recovery System

NASA Technical Reports Server (NTRS)

Atwater, James; Wheeler, Richard, Jr.; Dahl, Roger; Hadley, Neal

2010-01-01

A microwave plasma reactor was developed for the recovery of hydrogen contained within waste methane produced by Carbon Dioxide Reduction Assembly (CRA), which reclaims oxygen from CO2. Since half of the H2 reductant used by the CRA is lost as CH4, the ability to reclaim this valuable resource will simplify supply logistics for longterm manned missions. Microwave plasmas provide an extreme thermal environment within a very small and precisely controlled region of space, resulting in very high energy densities at low overall power, and thus can drive high-temperature reactions using equipment that is smaller, lighter, and less power-consuming than traditional fixed-bed and fluidized-bed catalytic reactors. The high energy density provides an economical means to conduct endothermic reactions that become thermodynamically favorable only at very high temperatures. Microwave plasma methods were developed for the effective recovery of H2 using two primary reaction schemes: (1) methane pyrolysis to H2 and solid-phase carbon, and (2) methane oligomerization to H2 and acetylene. While the carbon problem is substantially reduced using plasma methods, it is not completely eliminated. For this reason, advanced methods were developed to promote CH4 oligomerization, which recovers a maximum of 75 percent of the H2 content of methane in a single reactor pass, and virtually eliminates the carbon problem. These methods were embodied in a prototype H2 recovery system capable of sustained high-efficiency operation. NASA can incorporate the innovation into flight hardware systems for deployment in support of future long-duration exploration objectives such as a Space Station retrofit, Lunar outpost, Mars transit, or Mars base. The primary application will be for the recovery of hydrogen lost in the Sabatier process for CO2 reduction to produce water in Exploration Life Support systems. Secondarily, this process may also be used in conjunction with a Sabatier reactor employed to stockpile life-support oxygen as well as propellant and fuel production from Martian atmospheric CO2

Design and Test Plans for a Non-Nuclear Fission Power System Technology Demonstration Unit

NASA Technical Reports Server (NTRS)

Mason, Lee; Palac, Donald; Gibson, Marc; Houts, Michael; Warren, John; Werner, James; Poston, David; Qualls, Arthur Lou; Radel, Ross; Harlow, Scott

2012-01-01

A joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) team is developing concepts and technologies for affordable nuclear Fission Power Systems (FPSs) to support future exploration missions. A key deliverable is the Technology Demonstration Unit (TDU). The TDU will assemble the major elements of a notional FPS with a non-nuclear reactor simulator (Rx Sim) and demonstrate system-level performance in thermal vacuum. The Rx Sim includes an electrical resistance heat source and a liquid metal heat transport loop that simulates the reactor thermal interface and expected dynamic response. A power conversion unit (PCU) generates electric power utilizing the liquid metal heat source and rejects waste heat to a heat rejection system (HRS). The HRS includes a pumped water heat removal loop coupled to radiator panels suspended in the thermal-vacuum facility. The basic test plan is to subject the system to realistic operating conditions and gather data to evaluate performance sensitivity, control stability, and response characteristics. Upon completion of the testing, the technology is expected to satisfy the requirements for Technology Readiness Level 6 (System Demonstration in an Operational and Relevant Environment) based on the use of high-fidelity hardware and prototypic software tested under realistic conditions and correlated with analytical predictions.

Design and Test Plans for a Non-Nuclear Fission Power System Technology Demonstration Unit

NASA Astrophysics Data System (ADS)

Mason, L.; Palac, D.; Gibson, M.; Houts, M.; Warren, J.; Werner, J.; Poston, D.; Qualls, L.; Radel, R.; Harlow, S.

A joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) team is developing concepts and technologies for affordable nuclear Fission Power Systems (FPSs) to support future exploration missions. A key deliverable is the Technology Demonstration Unit (TDU). The TDU will assemble the major elements of a notional FPS with a non-nuclear reactor simulator (Rx Sim) and demonstrate system-level performance in thermal vacuum. The Rx Sim includes an electrical resistance heat source and a liquid metal heat transport loop that simulates the reactor thermal interface and expected dynamic response. A power conversion unit (PCU) generates electric power utilizing the liquid metal heat source and rejects waste heat to a heat rejection system (HRS). The HRS includes a pumped water heat removal loop coupled to radiator panels suspended in the thermal-vacuum facility. The basic test plan is to subject the system to realistic operating conditions and gather data to evaluate performance sensitivity, control stability, and response characteristics. Upon completion of the testing, the technology is expected to satisfy the requirements for Technology Readiness Level 6 (System Demonstration in an Operational and Relevant Environment) based on the use of high-fidelity hardware and prototypic software tested under realistic conditions and correlated with analytical predictions.

The Rockwell SR-100G reactor turboelectric space power system

NASA Technical Reports Server (NTRS)

Anderson, R. V.

1985-01-01

During FY 1982 and 1983, Rockwell International performed system and subsystem studies for space reactor power systems. These studies drew on the expertise gained from the design and flight of the SNAP-10A space nuclear reactor system. These studies, performed for the SP-100 Program, culminated in the selection of a reactor-turboelectric (gas Brayton) system for the SP-100 application; this system is called the SR-100G. This paper describes the features of the system and provides references where more detailed information can be obtained.

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.

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.

Biodegradation of pulp and paper mill effluent using anaerobic followed by aerobic digestion.

PubMed

Bishnoi, Narsi R; Khumukcham, R K; Kumar, Rajender

2006-05-01

An experimental study was carried to find out the degradability of black liquor of pulp and paper mill wastewater for biomethanogenesis in continuous stirred tank reactor (CSTR) and followed by activated sludge process (ASP). Continuous stirred tank reactor was used in present study for anaerobic digestion of black liquor, while completely mixed activated sludge system was used for aerobic digestion. A maximum methane production was found up to 430 ml/day, chemical oxygen demand was reduced up to 64% and total volatile fatty acid increased up to 1500 mg/l from 975 mg/l at 7.3 pH, 37 degrees C temperature and 8 days hydraulic retention time during anaerobic digestion. In activated sludge process (aerobic digestion) chemical oxygen demand and biological oxygen demand reduction were 81% and 86% respectively at 72 hr hydraulic retention time.

Revised FINAL–REPORT NO. 2: INDEPENDENT CONFIRMATORY SURVEY SUMMARY AND RESULTS FOR THE ENRICO FERMI ATOMIC POWER PLANT, UNIT 1, NEWPORT, MICHIGAN (DOCKET NO. 50 16; RFTA 10-004) 2018-SR-02-1

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

Erika Bailey

2011-10-27

The Enrico Fermi Atomic Power Plant, Unit 1 (Fermi 1) was a fast breeder reactor design that was cooled by sodium and operated at essentially atmospheric pressure. On May 10, 1963, the Atomic Energy Commission (AEC) granted an operating license, DPR-9, to the Power Reactor Development Company (PRDC), a consortium specifically formed to own and operate a nuclear reactor at the Fermi 1 site. The reactor was designed for a maximum capability of 430 megawatts (MW); however, the maximum reactor power with the first core loading (Core A) was 200 MW. The primary system was filled with sodium in Decembermore » 1960 and criticality was achieved in August 1963. The reactor was tested at low power during the first couple years of operation. Power ascension testing above 1 MW commenced in December 1965 immediately following the receipt of a high-power operating license. In October 1966 during power ascension, zirconium plates at the bottom of the reactor vessel became loose and blocked sodium coolant flow to some fuel subassemblies. Two subassemblies started to melt and the reactor was manually shut down. No abnormal releases to the environment occurred. Forty-two months later after the cause had been determined, cleanup completed, and the fuel replaced, Fermi 1 was restarted. However, in November 1972, PRDC made the decision to decommission Fermi 1 as the core was approaching its burn-up limit. The fuel and blanket subassemblies were shipped off-site in 1973. Following that, the secondary sodium system was drained and sent off-site. The radioactive primary sodium was stored on-site in storage tanks and 55 gallon (gal) drums until it was shipped off-site in 1984. The initial decommissioning of Fermi 1 was completed in 1975. Effective January 23, 1976, DPR-9 was transferred to the Detroit Edison Company (DTE) as a 'possession only' license (DTE 2010a). This report details the confirmatory activities performed during the second Oak Ridge Institute for Science and Education (ORISE) site visit to Fermi 1 in November 2010. The survey was strategically planned during a Unit 2 (Fermi 2) outage to take advantage of decreased radiation levels that were observed and attributed to Fermi 2 from the operating unit during the first site visit. However, during the second visit there were elevated radiation levels observed and attributed to the partially dismantled Fermi 1 reactor vessel and a waste storage box located on the 3rd floor of the Fermi 1 Turbine Building. Confirmatory surveys (unshielded) performed directly in the line of sight of these areas were affected. The objective of the confirmatory survey was to verify that the final radiological conditions were accurately and adequately described in Final Status Survey (FSS) documentation, relative to the established release criteria. This objective was achieved by performing document reviews, as well as independent measurements and sampling. Specifically, documentation of the planning, implementation, and results of the FSS were evaluated; side-by-side FSS measurement and source comparisons were performed; site areas were evaluated relative to appropriate FSS classification; and areas were assessed for residual, undocumented contamination.« less

An optically accessible pyrolysis microreactor

NASA Astrophysics Data System (ADS)

Baraban, J. H.; David, D. E.; Ellison, G. Barney; Daily, J. W.

2016-01-01

We report an optically accessible pyrolysis micro-reactor suitable for in situ laser spectroscopic measurements. A radiative heating design allows for completely unobstructed views of the micro-reactor along two axes. The maximum temperature demonstrated here is only 1300 K (as opposed to 1700 K for the usual SiC micro-reactor) because of the melting point of fused silica, but alternative transparent materials will allow for higher temperatures. Laser induced fluorescence measurements on nitric oxide are presented as a proof of principle for spectroscopic characterization of pyrolysis conditions.

An optically accessible pyrolysis microreactor

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

Baraban, J. H.; Ellison, G. Barney; David, D. E.

2016-01-15

We report an optically accessible pyrolysis micro-reactor suitable for in situ laser spectroscopic measurements. A radiative heating design allows for completely unobstructed views of the micro-reactor along two axes. The maximum temperature demonstrated here is only 1300 K (as opposed to 1700 K for the usual SiC micro-reactor) because of the melting point of fused silica, but alternative transparent materials will allow for higher temperatures. Laser induced fluorescence measurements on nitric oxide are presented as a proof of principle for spectroscopic characterization of pyrolysis conditions.

Megawatt Class Nuclear Space Power Systems (MCNSPS) conceptual design and evaluation report. Volume 1: Objectives, summary results and introduction

NASA Technical Reports Server (NTRS)

Wetch, J. R.

1988-01-01

The objective was to determine which reactor, conversion, and radiator technologies would best fulfill future Megawatt Class Nuclear Space Power System Requirements. Specifically, the requirement was 10 megawatts for 5 years of full power operation and 10 years systems life on orbit. A variety of liquid metal and gas cooled reactors, static and dynamic conversion systems, and passive and dynamic radiators were considered. Four concepts were selected for more detailed study. The concepts are: a gas cooled reactor with closed cycle Brayton turbine-alternator conversion with heat pipe and pumped tube-fin heat rejection; a lithium cooled reactor with a free piston Stirling engine-linear alternator and a pumped tube-fin radiator; a lithium cooled reactor with potassium Rankine turbine-alternator and heat pipe radiator; and a lithium cooled incore thermionic static conversion reactor with a heat pipe radiator. The systems recommended for further development to meet a 10 megawatt long life requirement are the lithium cooled reactor with the K-Rankine conversion and heat pipe radiator, and the lithium cooled incore thermionic reactor with heat pipe radiator.

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

SNAP 10A FS-3 reactor performance

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

Hawley, J.P.; Johnson, R.A.

1966-08-15

SNAP 10FS-3 was the first flight-qualified SNAP reactor system to be operated in a simulated space environment. Prestart-up qualification testing, automatic start-up, endurance period performance, extended operation test and reactor shutdown are described as they affected, or were affected by, overall reactor performance. Performance of the reactor control system and the diagnostic instrumentation is critically evaluted.

Turbulence coefficients and stability studies for the coaxial flow or dissimiliar fluids. [gaseous core nuclear reactors

NASA Technical Reports Server (NTRS)

Weinstein, H.; Lavan, Z.

1975-01-01

Analytical investigations of fluid dynamics problems of relevance to the gaseous core nuclear reactor program are presented. The vortex type flow which appears in the nuclear light bulb concept is analyzed along with the fluid flow in the fuel inlet region for the coaxial flow gaseous core nuclear reactor concept. The development of numerical methods for the solution of the Navier-Stokes equations for appropriate geometries is extended to the case of rotating flows and almost completes the gas core program requirements in this area. The investigations demonstrate that the conceptual design of the coaxial flow reactor needs further development.

Comparative study between single core model and detail core model of CFD modelling on reactor core cooling behaviour

NASA Astrophysics Data System (ADS)

Darmawan, R.

2018-01-01

Nuclear power industry is facing uncertainties since the occurrence of the unfortunate accident at Fukushima Daiichi Nuclear Power Plant. The issue of nuclear power plant safety becomes the major hindrance in the planning of nuclear power program for new build countries. Thus, the understanding of the behaviour of reactor system is very important to ensure the continuous development and improvement on reactor safety. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last four decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. Recently, 2D and 3D system level codes such as CFD are being explored. This paper discusses a comparative study on two different approaches of CFD modelling on reactor core cooling behaviour.

Hybrid Plasma Reactor/Filter for Transportable Collective Protection Systems

DTIC Science & Technology

2011-03-01

protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to provide a broader envelope of...conventional methods for collective protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to...protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to provide a broader

Training courses on neutron detection systems on the ISIS research reactor: on-site and through internet training

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

Lescop, B.; Badeau, G.; Ivanovic, S.

Today, ISIS research reactor is an essential tool for Education and Training programs organized by the National Institute for Nuclear Science and Technology (INSTN) from CEA. In the field of nuclear instrumentation, the INSTN offers both, theoretical courses and training courses on the use of neutron detection systems taking advantage of the ISIS research reactor for the supply of a wide range of neutron fluxes. This paper describes the content of the training carried out on the use of neutron detectors and detection systems, on-site or remote. The ISIS reactor is a 700 kW open core pool type reactor. Themore » facility is very flexible since neutron detectors can be inserted into the core or its vicinity, and be used at different levels of power according to the needs of the course. Neutron fluxes, typically ranging from 1 to 10{sup 12} n/cm{sup 2}.s, can be obtained for the characterisation of the neutron detectors and detection systems. For the monitoring of the neutron density at low level of power, the Instrumentation and Control (I and C) system of the reactor is equipped with two detection systems, named BN1 and BN2. Each way contains a fission chamber, type CFUL01, connected to an electronic system type SIREX.The system works in pulse mode and exhibits two outputs: the counting rate and the doubling time. For the high level of power, the I and C is equipped with two detection systems HN1 and HN2.Each way contain a boron ionization chamber (type CC52) connected to an electronics system type SIREX. The system works in current mode and has two outputs: the current and the doubling time. For each mode, the trainees can observe and measure the signal at the different stages of the electronic system, with an oscilloscope. They can understand the role of each component of the detection system: detector, cable and each electronic block. The limitation of the detection modes and their operating range can be established from the measured signal. The trainees can also modify the settings of the electronic system, such as the high voltage and the discrimination level in order to obtain all the characteristic curves of the detectors. These curves are used to define the right setting of the electronic system and to discuss the expected degradation of the detector signal resulting from the detector damage under the integrated neutron and gamma fluxes. Moreover, in addition to the study of the neutron detection systems itself, the integration of the measurements made by these detection systems in the logic of the safety system of the nuclear reactor is also addressed. Providing the trainees with an extensive overview of each part of the neutron monitoring instrumentation apply to a nuclear reactor, hands-on measurements on the ISIS reactor play a major role in ensuring a practical and comprehensive understanding of the neutron detection system and their integration in the safety system of nuclear reactors. It also gives a solid background for the follow up and the development of the neutron detection systems. In addition to on-reactor training, Internet Reactor Laboratory capability has been implemented on the ISIS reactor in 2014. For the Internet Reactor Laboratory an extensive video conference system has been implemented on ISIS reactor. The system includes 4 cameras and the transmission of the video signal given by the supervision system of the reactor which records and processes the data of the reactor. According to the pedagogic needs during the training courses, the lecturer on the ISIS reactor chooses to broadcast the relevant information at each stage of the course. For example, graph showing the histogram of the counting and current as a function of the time, or the electrical signal observed on the oscilloscope, can be broadcasted trough internet. By interacting through the video conference, the remote classroom is able to ask for changes in the reactor power or settings of the detection systems. They can also ask for the broadcast of some particular information. At the guest institution, the information is displayed in two parts or screens, as shown in the Figure 3. Concerning the interaction with - and the feedback from - the remote classroom, the camera of the video system in the remote classroom is used to ensure the contact between the trainees and the lecturer and reactor operators. Thus, the Internet Reactor Laboratory is complementary to the on reactor training courses. It allows distant learning, reducing the overall cost of the course when this is necessary. It can efficiently be used for the development of the human resources needed by the nuclear industry and the nuclear programs in countries without research reactors.« 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.

Modeling and simulation of CANDU reactor and its regulating system

NASA Astrophysics Data System (ADS)

Javidnia, Hooman

Analytical computer codes are indispensable tools in design, optimization, and control of nuclear power plants. Numerous codes have been developed to perform different types of analyses related to the nuclear power plants. A large number of these codes are designed to perform safety analyses. In the context of safety analyses, the control system is often neglected. Although there are good reasons for such a decision, that does not mean that the study of control systems in the nuclear power plants should be neglected altogether. In this thesis, a proof of concept code is developed as a tool that can be used in the design. optimization. and operation stages of the control system. The main objective in the design of this computer code is providing a tool that is easy to use by its target audience and is capable of producing high fidelity results that can be trusted to design the control system and optimize its performance. Since the overall plant control system covers a very wide range of processes, in this thesis the focus has been on one particular module of the the overall plant control system, namely, the reactor regulating system. The center of the reactor regulating system is the CANDU reactor. A nodal model for the reactor is used to represent the spatial neutronic kinetics of the core. The nodal model produces better results compared to the point kinetics model which is often used in the design and analysis of control system for nuclear reactors. The model can capture the spatial effects to some extent. although it is not as detailed as the finite difference methods. The criteria for choosing a nodal model of the core are: (1) the model should provide more detail than point kinetics and capture spatial effects, (2) it should not be too complex or overly detailed to slow down the simulation and provide details that are extraneous or unnecessary for a control engineer. Other than the reactor itself, there are auxiliary models that describe dynamics of different phenomena related to the transfer of the energy from the core. The main function of the reactor regulating system is to control the power of the reactor. This is achieved by using a set of detectors. reactivity devices. and digital control algorithms. Three main reactivity devices that are activated during short-term or intermediate-term transients are modeled in this thesis. The main elements of the digital control system are implemented in accordance to the program specifications for the actual control system in CANDU reactors. The simulation results are validated against requirements of the reactor regulating system. actual plant data. and pre-validated data from other computer codes. The validation process shows that the simulation results can be trusted in making engineering decisions regarding the reactor regulating system and prediction of the system performance in response to upset conditions or disturbances. KEYWORDS: CANDU reactors. reactor regulating system. nodal model. spatial kinetics. reactivity devices. simulation.

Potential of Electric Power Production from Microbial Fuel Cell (MFC) in Evapotranspiration Reactor for Leachate Treatment Using Alocasia macrorrhiza Plant and Eleusine indica Grass

NASA Astrophysics Data System (ADS)

Zaman, Badrus; Wardhana, Irawan Wisnu

2018-02-01

Microbial fuel cell is one of attractive electric power generator from nature bacterial activity. While, Evapotranspiration is one of the waste water treatment system which developed to eliminate biological weakness that utilize the natural evaporation process and bacterial activity on plant roots and plant media. This study aims to determine the potential of electrical energy from leachate treatment using evapotranspiration reactor. The study was conducted using local plant, namely Alocasia macrorrhiza and local grass, namely Eleusine Indica. The system was using horizontal MFC by placing the cathodes and anodes at different chamber (i.e. in the leachate reactor and reactor with plant media). Carbon plates was used for chatode-anodes material with size of 40 cm x 10 cm x1 cm. Electrical power production was measure by a digital multimeter for 30 days reactor operation. The result shows electric power production was fluctuated during reactor operation from all reactors. The electric power generated from each reactor was fluctuated, but from the reactor using Alocasia macrorrhiza plant reach to 70 μwatt average. From the reactor using Eleusine Indica grass was reached 60 μwatt average. Electric power production fluctuation is related to the bacterial growth pattern in the soil media and on the plant roots which undergo the adaptation process until the middle of the operational period and then in stable growth condition until the end of the reactor operation. The results indicate that the evapotranspiration reactor using Alocasia macrorrhiza plant was 60-95% higher electric power potential than using Eleusine Indica grass in short-term (30-day) operation. Although, MFC system in evapotranspiration reactor system was one of potential system for renewable electric power generation.

Co-digestion of manure and whey for in situ biogas upgrading by the addition of H(2): process performance and microbial insights.

PubMed

Luo, Gang; Angelidaki, Irini

2013-02-01

In situ biogas upgrading was conducted by introducing H(2) directly to the anaerobic reactor. As H(2) addition is associated with consumption of the CO(2) in the biogas reactor, pH increased to higher than 8.0 when manure alone was used as substrate. By co-digestion of manure with acidic whey, the pH in the anaerobic reactor with the addition of hydrogen could be maintained below 8.0, which did not have inhibition to the anaerobic process. The H(2) distribution systems (diffusers with different pore sizes) and liquid mixing intensities were demonstrated to affect the gas-liquid mass transfer of H(2) and the biogas composition. The best biogas composition (75:6.6:18.4) was obtained at stirring speed 150 rpm and using ceramic diffuser, while the biogas in the control reactor consisted of CH(4) and CO(2) at a ratio of 55:45. The consumed hydrogen was almost completely converted to CH(4), and there was no significant accumulation of VFA in the effluent. The study showed that addition of hydrogen had positive effect on the methanogenesis, but had no obvious effect on the acetogenesis. Both hydrogenotrophic methanogenic activity and the concentration of coenzyme F(420) involved in methanogenesis were increased. The archaeal community was also altered with the addition of hydrogen, and a Methanothermobacter thermautotrophicus related band appeared in a denaturing gradient gel electrophoresis gel from the sample of the reactor with hydrogen addition. Though the addition of hydrogen increased the dissolved hydrogen concentration, the degradation of propionate was still thermodynamically feasible at the reactor conditions.

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.

Demonstration of Robustness and Integrated Operation of a Series-Bosch System

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, Matthew J.; Stanley, Christine; Barnett, Bill; Junaedi, Christian; Vilekar, Saurabh A.; Ryan, Kent

2016-01-01

Manned missions beyond low Earth orbit will require highly robust, reliable, and maintainable life support systems that maximize recycling of water and oxygen. Bosch technology is one option to maximize oxygen recovery, in the form of water, from metabolically-produced carbon dioxide (CO2). A two stage approach to Bosch, called Series-Bosch, reduces metabolic CO2 with hydrogen (H2) to produce water and solid carbon using two reactors: a Reverse Water-Gas Shift (RWGS) reactor and a carbon formation (CF) reactor. Previous development efforts demonstrated the stand-alone performance of a NASA-designed RWGS reactor designed for robustness against carbon formation, two membrane separators intended to maximize single pass conversion of reactants, and a batch CF reactor with both transit and surface catalysts. In the past year, Precision Combustion, Inc. (PCI) developed and delivered a RWGS reactor for testing at NASA. The reactor design was based on their patented Microlith® technology and was first evaluated under a Phase I Small Business Innovative Research (SBIR) effort in 2010. The RWGS reactor was recently evaluated at NASA to compare its performance and operating conditions with NASA's RWGS reactor. The test results will be provided in this paper. Separately, in 2015, a semi-continuous CF reactor was designed and fabricated at NASA based on the results from batch CF reactor testing. The batch CF reactor and the semi-continuous CF reactor were individually integrated with an upstream RWGS reactor to demonstrate the system operation and to evaluate performance. Here, we compare the performance and robustness to carbon formation of both RWGS reactors. We report the results of the integrated operation of a Series-Bosch system and we discuss the technology readiness level.

77 FR 55877 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-09-11

...-492- 3668; email: [email protected] . NRC's Agencywide Documents Access and Management System... Systems for Light-Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Regulatory guide; issuance... Systems for Boiling Water Reactor Power Plants.'' This regulatory guide is being revised to: (1) Expand...

Reforming results of a novel radial reactor for a solid oxide fuel cell system with anode off-gas recirculation

NASA Astrophysics Data System (ADS)

Bosch, Timo; Carré, Maxime; Heinzel, Angelika; Steffen, Michael; Lapicque, François

2017-12-01

A novel reactor of a natural gas (NG) fueled, 1 kW net power solid oxide fuel cell (SOFC) system with anode off-gas recirculation (AOGR) is experimentally investigated. The reactor operates as pre-reformer, is of the type radial reactor with centrifugal z-flow, has the shape of a hollow cylinder with a volume of approximately 1 L and is equipped with two different precious metal wire-mesh catalyst packages as well as with an internal electric heater. Reforming investigations of the reactor are done stand-alone but as if the reactor would operate within the total SOFC system with AOGR. For the tests presented here it is assumed that the SOFC system runs on pure CH4 instead of NG. The manuscript focuses on the various phases of reactor operation during the startup process of the SOFC system. Startup process reforming experiments cover reactor operation points at which it runs on an oxygen to carbon ratio at the reactor inlet (ϕRI) of 1.2 with air supplied, up to a ϕRI of 2.4 without air supplied. As confirmed by a Monte Carlo simulation, most of the measured outlet gas concentrations are in or close to equilibrium.

NUCLEAR REACTOR CONTROL SYSTEM

DOEpatents

Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

1959-11-01

A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

Integrated Response Time Evaluation Methodology for the Nuclear Safety Instrumentation System

NASA Astrophysics Data System (ADS)

Lee, Chang Jae; Yun, Jae Hee

2017-06-01

Safety analysis for a nuclear power plant establishes not only an analytical limit (AL) in terms of a measured or calculated variable but also an analytical response time (ART) required to complete protective action after the AL is reached. If the two constraints are met, the safety limit selected to maintain the integrity of physical barriers used for preventing uncontrolled radioactivity release will not be exceeded during anticipated operational occurrences and postulated accidents. Setpoint determination methodologies have actively been developed to ensure that the protective action is initiated before the process conditions reach the AL. However, regarding the ART for a nuclear safety instrumentation system, an integrated evaluation methodology considering the whole design process has not been systematically studied. In order to assure the safety of nuclear power plants, this paper proposes a systematic and integrated response time evaluation methodology that covers safety analyses, system designs, response time analyses, and response time tests. This methodology is applied to safety instrumentation systems for the advanced power reactor 1400 and the optimized power reactor 1000 nuclear power plants in South Korea. The quantitative evaluation results are provided herein. The evaluation results using the proposed methodology demonstrate that the nuclear safety instrumentation systems fully satisfy corresponding requirements of the ART.

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

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

Edwin A. Harvego; Michael G. McKellar

2011-11-01

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

Utilization of Stop-flow Micro-tubing Reactors for the Development of Organic Transformations.

PubMed

Toh, Ren Wei; Li, Jie Sheng; Wu, Jie

2018-01-04

A new reaction screening technology for organic synthesis was recently demonstrated by combining elements from both continuous micro-flow and conventional batch reactors, coined stop-flow micro-tubing (SFMT) reactors. In SFMT, chemical reactions that require high pressure can be screened in parallel through a safer and convenient way. Cross-contamination, which is a common problem in reaction screening for continuous flow reactors, is avoided in SFMT. Moreover, the commercially available light-permeable micro-tubing can be incorporated into SFMT, serving as an excellent choice for light-mediated reactions due to a more effective uniform light exposure, compared to batch reactors. Overall, the SFMT reactor system is similar to continuous flow reactors and more superior than batch reactors for reactions that incorporate gas reagents and/or require light-illumination, which enables a simple but highly efficient reaction screening system. Furthermore, any successfully developed reaction in the SFMT reactor system can be conveniently translated to continuous-flow synthesis for large scale production.

Axi-symmetrical flow reactor for .sup.196 Hg photochemical enrichment

DOEpatents

Grossman, Mark W.

1991-01-01

The present invention is directed to an improved photochemical reactor useful for the isotopic enrichment of a predetermined isotope of mercury, especially, .sup.196 Hg. Specifically, two axi-symmetrical flow reactors were constructed according to the teachings of the present invention. These reactors improve the mixing of the reactants during the photochemical enrichment process, affording higher yields of the desired .sup.196 Hg product. Measurements of the variation of yield (Y) and enrichment factor (E) along the flow axis of these reactors indicates very substantial improvement in process uniformity compared to previously used photochemical reactor systems. In one preferred embodiment of the present invention, the photoreactor system was built such that the reactor chamber was removable from the system without disturbing the location of either the photochemical lamp or the filter employed therewith.

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.-

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

A comparison of the physical, chemical, and biological properties of sludges from a complete-mix activated sludge reactor and a submerged membrane bioreactor.

PubMed

Merlo, Rion P; Trussell, R Shane; Hermanowicz, Slawomir W; Jenkins, David

2007-03-01

The properties of sludges from a pilot-scale submerged membrane bioreactor (SMBR) and two bench-scale complete-mix, activated sludge (CMAS) reactors treating municipal primary effluent were determined. Compared with the CMAS sludges, the SMBR sludge contained a higher amount of soluble microbial products (SMP) and colloidal material attributed to the use of a membrane for solid-liquid separation; a higher amount nocardioform bacteria, resulting from efficient foam trapping; and a lower amount of extracellular polymeric substances (EPS), possibly because there was no selective pressure for the sludge to settle. High aeration rates in both the CMAS and SMBR reactors produced sludges with higher numbers of smaller particles. Normalized capillary suction time values for the SMBR sludge were lower than for the CMAS sludges, possibly because of its lower EPS content.

Qualification and characterization of electronics of the fast neutron Hodoscope detectors using neutrons from CABRI core

NASA Astrophysics Data System (ADS)

Mirotta, S.; Guillot, J.; Chevalier, V.; Biard, B.

2018-01-01

The study of Reactivity Initiated Accidents (RIA) is important to determine up to which limits nuclear fuels can withstand such accidents without clad failure. The CABRI International Program (CIP), conducted by IRSN under an OECD/NEA agreement, has been launched to perform representative RIA Integral Effect Tests (IET) on real irradiated fuel rods in prototypical Pressurized Water Reactors (PWR) conditions. For this purpose, the CABRI experimental pulse reactor, operated by CEA in Cadarache, France, has been strongly renovated, and equipped with a pressurized water loop. The behavior of the test rod, located in that loop in the center of the driver core, is followed in real time during the power transients thanks to the hodoscope, a unique online fuel motion monitoring system, and one of the major distinctive features of CABRI. The hodoscope measures the fast neutrons emitted by the tested rod during the power pulse with a complete set of 153 Fission Chambers and 153 Proton Recoil Counters. During the CABRI facility renovation, the electronic chain of these detectors has been upgraded. In this paper, the performance of the new system is presented describing gain calibration methodology in order to get maximal Signal/Noise ratio for amplification modules, threshold tuning methodology for the discrimination modules (old and new ones), and linear detectors response limit versus different reactor powers for the whole electronic chain.

Bioleaching of spent Ni-Cd batteries by continuous flow system: effect of hydraulic retention time and process load.

PubMed

Zhao, Ling; Yang, Dong; Zhu, Nan-Wen

2008-12-30

Spent Ni-Cd batteries bring a severe environmental problem that needs to be solved urgently. A novel continuous flow two-step leaching system based on bioleaching was introduced to dissolve heavy metals in batteries. It consists of an acidifying reactor which was used to culture indigenous thiobacilli and a leaching reactor which was used to leach metals from spent batteries. The indigenous acidophilic thiobacilli in sewage sludge was used as the microorganisms and the sludge itself as culture medium. Bioleaching tests at different hydraulic retention time (HRT) and process load in the leaching reactor were performed. The results showed that the longer the HRT (1, 3, 6, 9 and 15 days) was, the more time required to achieve the complete leaching of Ni, Cd and Co. The maximum dissolution of cadmium and cobalt was achieved at higher pH values (3.0-4.5) while the leaching of nickel hydroxide and nickel in metallic form (Ni0) were obtained separately in different acidity (pH 2.5-3.5). It cost about 25, 30 and more than 40 days to remove all of the three heavy metals with the process load of two, four and eight Ni-Cd batteries under the conditions that the ingoing bio-sulphuric acid was 1Ld(-1) and HRT was 3 days.

Megawatt Class Nuclear Space Power Systems (MCNSPS) conceptual design and evaluation report. Volume 4: Concepts selection, conceptual designs, recommendations

NASA Technical Reports Server (NTRS)

Wetch, J. R.

1988-01-01

A study was conducted by NASA Lewis Research Center for the Triagency SP-100 program office. The objective was to determine which reactor, conversion and radiator technologies would best fulfill future Megawatt Class Nuclear Space Power System Requirements. The requirement was 10 megawatts for 5 years of full power operation and 10 years system life on orbit. A variety of liquid metal and gas cooled reactors, static and dynamic conversion systems, and passive and dynamic radiators were considered. Four concepts were selected for more detailed study: (1) a gas cooled reactor with closed cycle Brayton turbine-alternator conversion with heatpipe and pumped tube fin rejection, (2) a Lithium cooled reactor with a free piston Stirling engine-linear alternator and a pumped tube-fin radiator,(3) a Lithium cooled reactor with a Potassium Rankine turbine-alternator and heat pipe radiator, and (4) a Lithium cooled incore thermionic static conversion reactor with a heat pipe radiator. The systems recommended for further development to meet a 10 megawatt long life requirement are the Lithium cooled reactor with the K-Rankine conversion and heat pipe radiator, and the Lithium cooled incore thermionic reactor with heat pipe radiator.

Advantages of Production of New Fissionable Nuclides for the Nuclear Power Industry in Hybrid Fusion-Fission Reactors

NASA Astrophysics Data System (ADS)

Tsibulskiy, V. F.; Andrianova, E. A.; Davidenko, V. D.; Rodionova, E. V.; Tsibulskiy, S. V.

2017-12-01

A concept of a large-scale nuclear power engineering system equipped with fusion and fission reactors is presented. The reactors have a joint fuel cycle, which imposes the lowest risk of the radiation impact on the environment. The formation of such a system is considered within the framework of the evolution of the current nuclear power industry with the dominance of thermal reactors, gradual transition to the thorium fuel cycle, and integration into the system of the hybrid fusion-fission reactors for breeding nuclear fuel for fission reactors. Such evolution of the nuclear power engineering system will allow preservation of the existing structure with the dominance of thermal reactors, enable the reprocessing of the spent nuclear fuel (SNF) with low burnup, and prevent the dangerous accumulation of minor actinides. The proposed structure of the nuclear power engineering system minimizes the risk of radioactive contamination of the environment and the SNF reprocessing facilities, decreasing it by more than one order of magnitude in comparison with the proposed scheme of closing the uranium-plutonium fuel cycle based on the reprocessing of SNF with high burnup from fast reactors.

Applicability of 100kWe-class of space reactor power systems to NASA manned space station missions

NASA Technical Reports Server (NTRS)

Silverman, S. W.; Willenberg, H. J.; Robertson, C.

1985-01-01

An assessment is made of a manned space station operating with sufficiently high power demands to require a multihundred kilowatt range electrical power system. The nuclear reactor is a competitor for supplying this power level. Load levels were selected at 150kWe and 300kWe. Interactions among the reactor electrical power system, the manned space station, the space transportation system, and the mission were evaluated. The reactor shield and the conversion equipment were assumed to be in different positions with respect to the station; on board, tethered, and on a free flyer platform. Mission analyses showed that the free flyer concept resulted in unacceptable costs and technical problems. The tethered reactor providing power to an electrolyzer for regenerative fuel cells on the space station, results in a minimum weight shield and can be designed to release the reactor power section so that it moves to a high altitude orbit where the decay period is at least 300 years. Placing the reactor on the station, on a structural boom is an attractive design, but heavier than the long tethered reactor design because of the shield weight for manned activity near the reactor.

Strategy to identify the causes and to solve a sludge granulation problem in methanogenic reactors: application to a full-scale plant treating cheese wastewater.

PubMed

Macarie, Hervé; Esquivel, Maricela; Laguna, Acela; Baron, Olivier; El Mamouni, Rachid; Guiot, Serge R; Monroy, Oscar

2017-08-26

Granulation of biomass is at the basis of the operation of the most successful anaerobic systems (UASB, EGSB and IC reactors) applied worldwide for wastewater treatment. Despite of decades of studies of the biomass granulation process, it is still not fully understood and controlled. "Degranulation/lack of granulation" is a problem that occurs sometimes in anaerobic systems resulting often in heavy loss of biomass and poor treatment efficiencies or even complete reactor failure. Such a problem occurred in Mexico in two full-scale UASB reactors treating cheese wastewater. A close follow-up of the plant was performed to try to identify the factors responsible for the phenomenon. Basically, the list of possible causes to a granulation problem that were investigated can be classified amongst nutritional, i.e. related to wastewater composition (e.g. deficiency or excess of macronutrients or micronutrients, too high COD proportion due to proteins or volatile fatty acids, high ammonium, sulphate or fat concentrations), operational (excessive loading rate, sub- or over-optimal water upflow velocity) and structural (poor hydraulic design of the plant). Despite of an intensive search, the causes of the granulation problems could not be identified. The present case remains however an example of the strategy that must be followed to identify these causes and could be used as a guide for plant operators or consultants who are confronted with a similar situation independently of the type of wastewater. According to a large literature based on successful experiments at lab scale, an attempt to artificially granulate the industrial reactor biomass through the dosage of a cationic polymer was also tested but equally failed. Instead of promoting granulation, the dosage caused a heavy sludge flotation. This shows that the scaling of such a procedure from lab to real scale cannot be advised right away unless its operability at such a scale can be demonstrated.

Reactor water cleanup system

DOEpatents

Gluntz, Douglas M.; Taft, William E.

1994-01-01

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2011 CFR

2011-01-01

.... Containment inspection. B. Repordkeeping of test results. I. Introduction One of the conditions of all... following: A. Type A test—1. Pretest requirements. (a) Containment inspection in accordance with V. A. shall.... During the period between the completion of one Type A test and the initiation of the containment...

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

Logan, B.G.

A recently completed two-year study of a commercial tandem mirror reactor design (Mirror Advanced Reactor Study (MARS)) is briefly reviewed. The end plugs are designed for trapped particle stability, MHD ballooning, balanced geodesic curvature, and small radial electric fields in the central cell. New technologies such as lithium-lead blankets, 24T hybrid coils, gridless direct converters and plasma halo vacuum pumps are highlighted.

PBF Reactor Building (PER620). Detail of fuel test assembly in ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620). Detail of fuel test assembly in preparation for test. When complete, it will fit into in-pile tube. The maximum outside diameter of which must be about 8.25 inches. Date: 1982. INEEL negative no. 82-4908 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Cleanup Verification Package for the 118-F-7, 100-F Miscellaneous Hardware Storage Vault

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

M. J. Appel

2006-11-02

This cleanup verification package documents completion of remedial action for the 118-F-7, 100-F Miscellaneous Hardware Storage Vault. The site consisted of an inactive solid waste storage vault used for temporary storage of slightly contaminated reactor parts that could be recovered and reused for the 100-F Area reactor operations.

10 CFR 2.629 - Finality of partial decision on site suitability issues in a combined license proceeding.

Code of Federal Regulations, 2012 CFR

2012-01-01

... complete and acceptable for docketing under § 2.101(a)(3), the Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, shall determine whether the... a combined license proceeding. 2.629 Section 2.629 Energy NUCLEAR REGULATORY COMMISSION RULES OF...

10 CFR 2.629 - Finality of partial decision on site suitability issues in a combined license proceeding.

Code of Federal Regulations, 2011 CFR

2011-01-01

... complete and acceptable for docketing under § 2.101(a)(3), the Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, shall determine whether the... a combined license proceeding. 2.629 Section 2.629 Energy NUCLEAR REGULATORY COMMISSION RULES OF...

Summary of NR Program Prometheus Efforts

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

J Ashcroft; C Eshelman

2006-02-08

The Naval Reactors Program led work on the development of a reactor plant system for the Prometheus space reactor program. The work centered on a 200 kWe electric reactor plant with a 15-20 year mission applicable to nuclear electric propulsion (NEP). After a review of all reactor and energy conversion alternatives, a direct gas Brayton reactor plant was selected for further development. The work performed subsequent to this selection included preliminary nuclear reactor and reactor plant design, development of instrumentation and control techniques, modeling reactor plant operational features, development and testing of core and plant material options, and development ofmore » an overall project plan. Prior to restructuring of the program, substantial progress had been made on defining reference plant operating conditions, defining reactor mechanical, thermal and nuclear performance, understanding the capabilities and uncertainties provided by material alternatives, and planning non-nuclear and nuclear system testing. The mission requirements for the envisioned NEP missions cannot be accommodated with existing reactor technologies. Therefore concurrent design, development and testing would be needed to deliver a functional reactor system. Fuel and material performance beyond the current state of the art is needed. There is very little national infrastructure available for fast reactor nuclear testing and associated materials development and testing. Surface mission requirements may be different enough to warrant different reactor design approaches and development of a generic multi-purpose reactor requires substantial sacrifice in performance capability for each mission.« less

Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

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

Murata, K.K.; Williams, D.C.; Griffith, R.O.

1997-12-01

The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of themore » input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.« less

Planning, Preparation, and Transport of the High-Enriched Uranium Spent Nuclear Fuel from the Czech Republic to the Russian Federation

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

M. J. Tyacke; I. Bolshinsky; Frantisek Svitak

The United States, Russian Federation, and the International Atomic Energy Agency have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program, which is part of the Global Threat Reduction Initiative. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. In February 2003, the RRRFR Program began discussions with the Nuclear Research Institute (NRI) in Rež, Czech Republic, about returning their HEU spent nuclear fuel to the Russian Federation for reprocessing. In March 2005, the U.S. Department ofmore » Energy signed a contract with NRI to perform all activities needed for transporting their HEU spent nuclear fuel to Russia. After 2 years of intense planning, preparations, and coordination at NRI and with three other countries, numerous organizations and agencies, and a Russian facility, this shipment is scheduled for completion before the end of 2007. This paper will provide a summary of activities completed for making this international shipment. This paper contains an introduction and background of the RRRFR Program and the NRI shipment project. It summarizes activities completed in preparation for the shipment, including facility preparations at NRI in Rež and FSUE “Mayak” in Ozyorsk, Russia; a new transportation cask system; regulatory approvals; transportation planning and preparation in the Czech Republic, Slovakia, Ukraine, and the Russian Federation though completion of the Unified Project and Special Ecological Programs. The paper also describes fuel loading and cask preparations at NRI and final preparations/approvals for transporting the shipment across the Czech Republic, Slovakia, Ukraine, and the Russian Federation to FSUE Mayak where the HEU spent nuclear fuel will be processed, the uranium will be downblended and made into low-enriched uranium fuel for commercial reactor use, and the high-level waste from the processing will be stabilized and stored for less than 20 years before being sent back to the Czech Republic for final disposition. Finally, the paper contains a section for the summary and conclusions.« less

PM-1 NUCLEAR POWER PLANT PROGRAM. Quarterly Progress Report No. 2 for June 1 to August 31, 1959

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

Sieg, J.S.; Smith, E.H.

1959-10-01

The objective of the contract is the design, development, fabrication, installation, and initial testing and operation of a prepackaged air- transportable pressurized water reactor nuclear power plant, the PM-1. The specified output is 1 Mwe and 7 million Btu/hr of heat. The plant is to be operational by March 1962. The principal efforts were completion of the plant parametric study and preparation of the preliminary design. A summary of design parameters is given. Systems development work included study and selection of packages for full-scale testing, a survey of in-core instrumentation techniques, control and instrumentation development, and development of components formore » the steam generator, condenser, and turbine generator, which are not commercially available. Reactor development work included completion of the parametric zeropower experiments and preparrtions for a flexible zeropower test program, a revision of plans for irradiation testing PM-1 fuel elements, initiation of a reactor flow test program, outliring of a heat tnansfer test program, completion of the seven-tube test section (SETCH-1) tests, and evaluation of control rod actuators leading to specification of a magnetic jack-type control rod drive similar to that reported in ANL-5768. Completion of the prelimirary design led to initiation of the final design effort, which will be the principal activity during the next two project quarters. Preparations for core fabrication included procurement of core cladding material for the zero-power teat core, arrangement with a subcontractor to convent UF/sub 6/ to UO/sub 2/ and to commence delivery of the oxide during the next quarter, development of fuel element fabrication and ultrasonic testing techniques, study of control rod materials, UO/sub 2/ recovery techniques, and boron analysis methods. Preliminary work on site preparation was pursued with receipt of USAEC approval for a location on the eastern slope of Warren Peak at Sundance, Wyoming. A survey of this site is underway. A preliminary Hazards Summary Report is in preparation. (For preceding period see MND-M-1812.) (auth)« 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

Laser-Ultrasonic Testing and its Applications to Nuclear Reactor Internals

NASA Astrophysics Data System (ADS)

Ochiai, M.; Miura, T.; Yamamoto, S.

2008-02-01

A new nondestructive testing technique for surface-breaking microcracks in nuclear reactor components based on laser-ultrasonics is developed. Surface acoustic wave generated by Q-switched Nd:YAG laser and detected by frequency-stabilized long pulse laser coupled with confocal Fabry-Perot interferometer is used to detect and size the cracks. A frequency-domain signal processing is developed to realize accurate sizing capability. The laser-ultrasonic testing allows the detection of surface-breaking microcrack having a depth of less than 0.1 mm, and the measurement of their depth with an accuracy of 0.2 mm when the depth exceeds 0.5 mm including stress corrosion cracking. The laser-ultrasonic testing system combined with laser peening system, which is another laser-based maintenance technology to improve surface stress, for inner surface of small diameter tube is developed. The generation laser in the laser-ultrasonic testing system can be identical to the laser source of the laser peening. As an example operation of the system, the system firstly works as the laser-ultrasonic testing mode and tests the inner surface of the tube. If no cracks are detected, the system then changes its work mode to the laser peening and improves surface stress to prevent crack initiation. The first nuclear industrial application of the laser-ultrasonic testing system combined with the laser peening was completed in Japanese nuclear power plant in December 2004.

An Optically Accessible Pyrolysis Microreactor

NASA Astrophysics Data System (ADS)

Baraban, Joshua H.; David, Donald E.; Ellison, Barney; Daily, John W.

2016-06-01

We report an optically accessible pyrolysis micro-reactor suitable for in situ laser spectroscopic measurements. A radiative heating design allows for completely unobstructed views of the micro-reactor along two axes. The maximum temperature demonstrated here is only 1300 K (as opposed to 1700 K for the usual SiC micro-reactor) because of the melting point of fused silica, but alternative transparent materials will allow for higher temperatures. Laser induced fluorescence measurements on nitric oxide are presented as a proof of principle for spectroscopic characterization of pyrolysis conditions. (This work has been published in J. H. Baraban, D. E. David, G. B. Ellison, and J. W. Daily. An Optically Accessible Pyrolysis Micro-Reactor. Review of Scientific Instruments, 87(1):014101, 2016.)

Performances and microbial features of an aerobic packed-bed biofilm reactor developed to post-treat an olive mill effluent from an anaerobic GAC reactor

PubMed Central

Bertin, Lorenzo; Colao, Maria Chiara; Ruzzi, Maurizio; Marchetti, Leonardo; Fava, Fabio

2006-01-01

Background Olive mill wastewater (OMW) is the aqueous effluent of olive oil producing processes. Given its high COD and content of phenols, it has to be decontaminated before being discharged. Anaerobic digestion is one of the most promising treatment process for such an effluent, as it combines high decontamination efficiency with methane production. The large scale anaerobic digestion of OMWs is normally conducted in dispersed-growth reactors, where however are generally achieved unsatisfactory COD removal and methane production yields. The possibility of intensifying the performance of the process using a packed bed biofilm reactor, as anaerobic treatment alternative, was demonstrated. Even in this case, however, a post-treatment step is required to further reduce the COD. In this work, a biological post-treatment, consisting of an aerobic biological "Manville" silica bead-packed bed aerobic reactor, was developed, tested for its ability to complete COD removal from the anaerobic digestion effluents, and characterized biologically through molecular tools. Results The aerobic post-treatment was assessed through a 2 month-continuous feeding with the digested effluent at 50.42 and 2.04 gl-1day-1 of COD and phenol loading rates, respectively. It was found to be a stable process, able to remove 24 and 39% of such organic loads, respectively, and to account for 1/4 of the overall decontamination efficiency displayed by the anaerobic-aerobic integrated system when fed with an amended OMW at 31.74 and 1.70 gl-1day-1 of COD and phenol loading rates, respectively. Analysis of 16S rRNA gene sequences of biomass samples from the aerobic reactor biofilm revealed that it was colonized by Rhodobacterales, Bacteroidales, Pseudomonadales, Enterobacteriales, Rhodocyclales and genera incertae sedis TM7. Some taxons occurring in the influent were not detected in the biofilm, whereas others, such as Paracoccus, Pseudomonas, Acinetobacter and Enterobacter, enriched significantly in the biofilter throughout the treatment. Conclusion The silica-bead packed bed biofilm reactor developed and characterized in this study was able to significantly decontaminate anaerobically digested OMWs. Therefore, the application of an integrated anaerobic-aerobic process resulted in an improved system for valorization and decontamination of OMWs. PMID:16595023

Reactor vessel support system. [LMFBR

DOEpatents

Golden, M.P.; Holley, J.C.

1980-05-09

A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

Nuclear Thermal Propulsion: A Joint NASA/DOE/DOD Workshop

NASA Technical Reports Server (NTRS)

Clark, John S. (Editor)

1991-01-01

Papers presented at the joint NASA/DOE/DOD workshop on nuclear thermal propulsion are compiled. The following subject areas are covered: nuclear thermal propulsion programs; Rover/NERVA and NERVA systems; Low Pressure Nuclear Thermal Rocket (LPNTR); particle bed reactor nuclear rocket; hybrid propulsion systems; wire core reactor; pellet bed reactor; foil reactor; Droplet Core Nuclear Rocket (DCNR); open cycle gas core nuclear rockets; vapor core propulsion reactors; nuclear light bulb; Nuclear rocket using Indigenous Martian Fuel (NIMF); mission analysis; propulsion and reactor technology; development plans; and safety issues.

Feasibility of an on-line fission-gas-leak detection system

NASA Technical Reports Server (NTRS)

Lustig, P. H.

1973-01-01

Calculations were made to determine if a cladding failure could be detected in a 100-kW zirconium hydride reactor primary system by monitoring the highly radioactive NaK coolant for the presence of I-131. The system is to be completely sealed. A leak of 0.01 percent from a single fuel pin was postulated. The 0.364-MeV gamma of I-131 could be monitored on an almost continuous basis, while its presence could be varified by using a longer counting time for the 0.638-MeV gamma. A lithium-drifted germanium detector would eliminate radioactive corrosion product interference that could occur with a sodium iodide scintillation detector.

Cleanup Verification Package for the 118-F-1 Burial Ground

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

E. J. Farris and H. M. Sulloway

2008-01-10

This cleanup verification package documents completion of remedial action for the 118-F-1 Burial Ground on the Hanford Site. This burial ground is a combination of two locations formerly called Minor Construction Burial Ground No. 2 and Solid Waste Burial Ground No. 2. This waste site received radioactive equipment and other miscellaneous waste from 105-F Reactor operations, including dummy elements and irradiated process tubing; gun barrel tips, steel sleeves, and metal chips removed from the reactor; filter boxes containing reactor graphite chips; and miscellaneous construction solid waste.

TREAT Reactor Control and Protection System

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

Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.

1985-01-01

The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS).more » The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab.« less

A Novel Fuel/Reactor Cycle to Implement the 300 Years Nuclear Waste Policy Approach - 12377

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

Carelli, M.D.; Franceschini, F.; Lahoda, E.J.

2012-07-01

A thorium-based fuel cycle system can effectively burn the currently accumulated commercial used nuclear fuel and move to a sustainable equilibrium where the actinide levels in the high level waste are low enough to yield a radiotoxicity after 300 years lower than that of the equivalent uranium ore. The second step of the Westinghouse approach to solving the waste 'problem' has been completed. The thorium fuel cycle has indeed the potential of burning the legacy TRU and achieve the waste objective proposed. Initial evaluations have been started for the third step, development and selection of appropriate reactors. Indications are thatmore » the probability of show-stoppers is rather remote. It is, therefore, believed that development of the thorium cycle and associated technologies will provide a permanent solution to the waste management. Westinghouse is open to the widest collaboration to make this a reality. (authors)« less

A Pilot Study Investigating the Effects of Advanced Nuclear Power Plant Control Room Technologies: Methods and Qualitative Results

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

BLanc, Katya Le; Powers, David; Joe, Jeffrey

2015-08-01

Control room modernization is an important part of life extension for the existing light water reactor fleet. None of the 99 currently operating commercial nuclear power plants in the U.S. has completed a full-scale control room modernization to date. Nuclear power plant main control rooms for the existing commercial reactor fleet remain significantly analog, with only limited digital modernizations. Upgrades in the U.S. do not achieve the full potential of newer technologies that might otherwise enhance plant and operator performance. The goal of the control room upgrade benefits research is to identify previously overlooked benefits of modernization, identify candidate technologiesmore » that may facilitate such benefits, and demonstrate these technologies through human factors research. This report describes a pilot study to test upgrades to the Human Systems Simulation Laboratory at INL.« less

Safety and licensing of a small modular gas-cooled reactor system

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

Brown, N.W.; Kelley, A.P. Jr.

A modular side-by-side high-temperature gas-cooled reactor (SBS-HTGR) is being developed by Interatom/Kraftwerk Union (KWU). The General Electric Company and Interatom/KWU entered into a proprietary working agreement to continue develop jointly of the SBS-HTGR. A study on adapting the SBS-HTGR for application in the US has been completed. The study investigated the safety characteristics and the use of this type of design in an innovative approach to licensing. The safety objective guiding the design of the modular SBS-HTGR is to control radionuclide release by the retention of fission products within the fuel particles with minimal reliance on active design features. Themore » philosophy on which this objective is predicated is that by providing a simple safety case, the safety criteria can be demonstrated as being met with high confidence through conduct of a full-scale module safety test.« less

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

None

During this time period, at WVU, we tried several methods to eliminate problems related to condensation of heavier products when reduced Mo-Ni-K/C materials were used as catalysts. We then resumed our kinetic study on the reduced Mo-Ni-K/C catalysts. We have also obtained same preliminary results in our attempts to analyze quantitatively the temperature-programmed reduction (TPR) spectra for C-supported Mo-based catalysts. We have completed the kinetic study for the sulfided Co-K-MoS /C catalyst. We have compared the results of methanol synthesis 2 using the membrane reactor with those using a simple plug-flow reactor. At UCC, the complete characterization of selected catalystsmore » has been completed. The results suggest that catalyst pretreatment under different reducing conditions yield different surface compositions and thus different catalytic reactivities.« less

Axi-symmetrical flow reactor for [sup 196]Hg photochemical enrichment

DOEpatents

Grossman, M.W.

1991-04-30

The present invention is directed to an improved photochemical reactor useful for the isotopic enrichment of a predetermined isotope of mercury, especially, [sup 196]Hg. Specifically, two axi-symmetrical flow reactors were constructed according to the teachings of the present invention. These reactors improve the mixing of the reactants during the photochemical enrichment process, affording higher yields of the desired [sup 196]Hg product. Measurements of the variation of yield (Y) and enrichment factor (E) along the flow axis of these reactors indicates very substantial improvement in process uniformity compared to previously used photochemical reactor systems. In one preferred embodiment of the present invention, the photoreactor system was built such that the reactor chamber was removable from the system without disturbing the location of either the photochemical lamp or the filter employed therewith. 10 figures.

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

Wichman, K.; Tsao, J.; Mayfield, M.

The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRCmore » and the International Piping Integrity Research Group is also briefly summarized.« less

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

Fitzpatrick, F.C.; Gray, D.D.; Hyndman, J.R.

The thermal, ecological, and social impacts of a 40-reactor NEC are compared to impacts from four 10-reactor NECs and ten 4-reactor power plants. The comparison was made for surrogate sites in western Tennessee. The surrogate site for the 40-reactor NEC is located on Kentucky Lake. A layout is postulated for ten clusters of four reactors each with 2.5-mile spacing between clusters. The plants use natural-draft cooling towers. A transmission system is proposed for delivering the power (48,000 MW) to five load centers. Comparable transmission systems are proposed for the 10-reactor NECs and the 4-reactor dispersed sites delivering power to themore » same load centers. (auth)« less

Test Results from a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

NASA Technical Reports Server (NTRS)

Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.; Godfroy, Thomas J.

2010-01-01

Component level testing of power conversion units proposed for use in fission surface power systems has typically been done using relatively simple electric heaters for thermal input. These heaters do not adequately represent the geometry or response of proposed reactors. As testing of fission surface power systems transitions from the component level to the system level it becomes necessary to more accurately replicate these reactors using reactor simulators. The Direct Drive Gas-Brayton Power Conversion Unit test activity at the NASA Glenn Research Center integrates a reactor simulator with an existing Brayton test rig. The response of the reactor simulator to a change in Brayton shaft speed is shown as well as the response of the Brayton to an insertion of reactivity, corresponding to a drum reconfiguration. The lessons learned from these tests can be used to improve the design of future reactor simulators which can be used in system level fission surface power tests.

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.

Loss-of-flow-without-scram tests in Experimental Breeder Reactor-II and comparison with pretest predictions

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

Chang, L.K.; Mohr, D.; Planchon, H.P.

This article discusses a series of successful loss-of-flow-without-scram tests conducted in Experimental Breeder Reactor-II (EBR-II), a metal-fueled, sodium-cooled fast reactor. These May 1985 tests demonstrated the capability of the EBR to reduce reactor power passively during a loss of flow and to maintain reactor temperatures within bounds without any reliance on an active safety system. The tests were run from reduced power to ensure that temperatures could be maintained well below the fuel-clad eutectic temperature. Good agreement was found between selected test data and pretest predictions made with the EBR-II system analysis code NATDEMO and the hot channel analysis codemore » HOTCHAN. The article also discusses safety assessments of the tests as well as modifications required on the EBR-II reactor safety system for conducting required on the EBR-II reactor safety system for the conducting the tests.« less

Validation of the WIMSD4M cross-section generation code with benchmark results

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

Leal, L.C.; Deen, J.R.; Woodruff, W.L.

1995-02-01

The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment for Research and Test (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the procedure to generatemore » cross-section libraries for reactor analyses and calculations utilizing the WIMSD4M code. To do so, the results of calculations performed with group cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory(ORNL) unreflected critical spheres, the TRX critical experiments, and calculations of a modified Los Alamos highly-enriched heavy-water moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented.« less

A quantified dosing ALD reactor with in-situ diagnostics for surface chemistry studies

NASA Astrophysics Data System (ADS)

Larrabee, Thomas J.

A specialized atomic layer deposition (ALD) reactor has been constructed to serve as an instrument to simultaneously study the surface chemistry of the ALD process, and perform ALD as is conventionally done in continuum flow of inert gas. This reactor is uniquely useful to gain insight into the ALD process because of the combination of its precise, controllable, and quantified dosing/microdosing capability; its in-situ quadrupole mass spectrometer for gas composition analysis; its pair of highly-sensitive in-situ quartz crystal microbalances (QCMs); and its complete spectrum of pressures and operating conditions --- from viscous to molecular flow regimes. Control of the dose is achieved independently of the conditions by allowing a reactant gas to fill a fixed volume and measured pressure, which is held at a controlled temperature, and subsequently dosed into the system by computer controlled pneumatic valves. Absolute reactant exposure to the substrate and QCMs is unambiguously calculated from the molecular impingement flux, and its relationship to dose size is established, allowing means for easily intentionally reproducing specific exposures. Methods for understanding atomic layer growth and adsorption phenomena, including the precursor sticking probability, dynamics of molecular impingement, size of dose, and other operating variables are for the first time quantitatively related to surface reaction rates by mass balance. Extensive characterization of the QCM as a measurement tool for adsorption under realistic ALD conditions has been examined, emphasizing the state-of-the-art and importance of QCM system features required. Finally, the importance of dose-quantification and microdosing has been contextualized in view of the ALD literature, underscoring the significance of more precise condition specification in establishing a better basis for reactor and reactant comparison.

Sustainable Mars Sample Return

NASA Technical Reports Server (NTRS)

Alston, Christie; Hancock, Sean; Laub, Joshua; Perry, Christopher; Ash, Robert

2011-01-01

The proposed Mars sample return mission will be completed using natural Martian resources for the majority of its operations. The system uses the following technologies: In-Situ Propellant Production (ISPP), a methane-oxygen propelled Mars Ascent Vehicle (MAV), a carbon dioxide powered hopper, and a hydrogen fueled balloon system (large balloons and small weather balloons). The ISPP system will produce the hydrogen, methane, and oxygen using a Sabatier reactor. a water electrolysis cell, water extracted from the Martian surface, and carbon dioxide extracted from the Martian atmosphere. Indigenous hydrogen will fuel the balloon systems and locally-derived methane and oxygen will fuel the MAV for the return of a 50 kg sample to Earth. The ISPP system will have a production cycle of 800 days and the estimated overall mission length is 1355 days from Earth departure to return to low Earth orbit. Combining these advanced technologies will enable the proposed sample return mission to be executed with reduced initial launch mass and thus be more cost efficient. The successful completion of this mission will serve as the next step in the advancement of Mars exploration technology.

Enhanced biodegradation of hexachlorocyclohexane in upflow anaerobic sludge blanket reactor using methanol as an electron donor.

PubMed

Bhatt, Praveena; Kumar, M Suresh; Mudliar, Sandeep; Chakrabarti, Tapan

2008-05-01

Anaerobic dechlorination of technical grade hexachlorocyclohexane (THCH) was studied in a continuous upflow anaerobic sludge blanket (UASB) reactor with methanol as a supplementary substrate and electron donor. A reactor without methanol served as the experimental control. The inlet feed concentration of THCH in both the experimental and the control UASB reactor was 100 mg l(-1). After 60 days of continuous operation, the removal of THCH was >99% in the methanol-supplemented reactor as compared to 20-35% in the control reactor. THCH was completely dechlorinated in the methanol fed reactor at 48 h HRT after 2 months of continuous operation. This period was also accompanied by increase in biomass in the reactor, which was not observed in the experimental control. Batch studies using other supplementary substrates as well as electron donors namely acetate, butyrate, formate and ethanol showed lower % dechlorination (<85%) and dechlorination rates (<3 mg g(-1)d(-1)) as compared to methanol (98%, 5 mg g(-1)d(-1)). The optimum concentration of methanol required, for stable dechlorination of THCH (100 mg l(-1)) in the UASB reactor, was found to be 500 mg l(-1). Results indicate that addition of methanol as electron donor enhances dechlorination of THCH at high inlet concentration, and is also required for stable UASB reactor performance.

Comparative assessment of out-of-core nuclear thermionic power systems

NASA Technical Reports Server (NTRS)

Estabrook, W. C.; Koenig, D. R.; Prickett, W. Z.

1975-01-01

The hardware selections available for fabrication of a nuclear electric propulsion stage for planetary exploration were explored. The investigation was centered around a heat-pipe-cooled, fast-spectrum nuclear reactor for an out-of-core power conversion system with sufficient detail for comparison with the in-core system studies completed previously. A survey of competing power conversion systems still indicated that the modular reliability of thermionic converters makes them the desirable choice to provide the 240-kWe end-of-life power for at least 20,000 full power hours. The electrical energy will be used to operate a number of mercury ion bombardment thrusters with a specific impulse in the range of about 4,000-5,000 seconds.

Cooling Performance Analysis of ThePrimary Cooling System ReactorTRIGA-2000Bandung

NASA Astrophysics Data System (ADS)

Irianto, I. D.; Dibyo, S.; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The conversion of reactor fuel type will affect the heat transfer process resulting from the reactor core to the cooling system. This conversion resulted in changes to the cooling system performance and parameters of operation and design of key components of the reactor coolant system, especially the primary cooling system. The calculation of the operating parameters of the primary cooling system of the reactor TRIGA 2000 Bandung is done using ChemCad Package 6.1.4. The calculation of the operating parameters of the cooling system is based on mass and energy balance in each coolant flow path and unit components. Output calculation is the temperature, pressure and flow rate of the coolant used in the cooling process. The results of a simulation of the performance of the primary cooling system indicate that if the primary cooling system operates with a single pump or coolant mass flow rate of 60 kg/s, it will obtain the reactor inlet and outlet temperature respectively 32.2 °C and 40.2 °C. But if it operates with two pumps with a capacity of 75% or coolant mass flow rate of 90 kg/s, the obtained reactor inlet, and outlet temperature respectively 32.9 °C and 38.2 °C. Both models are qualified as a primary coolant for the primary coolant temperature is still below the permitted limit is 49.0 °C.

Reactor water cleanup system

DOEpatents

Gluntz, D.M.; Taft, W.E.

1994-12-20

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

Improved vortex reactor system

DOEpatents

Diebold, James P.; Scahill, John W.

1995-01-01

An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

Simultaneous removal of aniline, nitrogen and phosphorus in aniline-containing wastewater treatment by using sequencing batch reactor.

PubMed

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

2016-05-01

The high removal efficiencies of traditional biological aniline-degrading systems always lead to accumulation of ammonium. In this study, simultaneous removal of aniline, nitrogen and phosphorus in a single sequencing batch reactor was achieved by using anaerobic/aerobic/anoxic (A/O/A) operational process. The removal efficiencies of COD, NH4(+)-N, TN, TP were over 95.80%, 83.03%, 87.13%, 90.95%, respectively in most cases with 250mgL(-1) of initial aniline at 6h cycle when DO was 5.5±0.5mgL(-1). Aniline was able to be completely degraded when initial concentrations were less than 750mgL(-1). When DO increased, the removal rate of NH4(+)-N and TP slightly increased along with the moderate decrease of removal efficiencies of TN. The variation of HRT had obvious influence on removal performance of pollutants. The system showed high removal efficiencies of aniline, COD and nutrients during the variation of operating conditions, which might contribute to disposal of aniline-rich industrial wastewater. Copyright © 2016 Elsevier Ltd. All rights reserved.

Automated one-step DNA sequencing based on nanoliter reaction volumes and capillary electrophoresis.

PubMed

Pang, H M; Yeung, E S

2000-08-01

An integrated system with a nano-reactor for cycle-sequencing reaction coupled to on-line purification and capillary gel electrophoresis has been demonstrated. Fifty nanoliters of reagent solution, which includes dye-labeled terminators, polymerase, BSA and template, was aspirated and mixed with the template inside the nano-reactor followed by cycle-sequencing reaction. The reaction products were then purified by a size-exclusion chromatographic column operated at 50 degrees C followed by room temperature on-line injection of the DNA fragments into a capillary for gel electrophoresis. Over 450 bases of DNA can be separated and identified. As little as 25 nl reagent solution can be used for the cycle-sequencing reaction with a slightly shorter read length. Significant savings on reagent cost is achieved because the remaining stock solution can be reused without contamination. The steps of cycle sequencing, on-line purification, injection, DNA separation, capillary regeneration, gel-filling and fluidic manipulation were performed with complete automation. This system can be readily multiplexed for high-throughput DNA sequencing or PCR analysis directly from templates or even biological materials.

Seismic attenuation system for a nuclear reactor

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

Liszkai, Tamas; Cadell, Seth

A system for attenuating seismic forces includes a reactor pressure vessel containing nuclear fuel and a containment vessel that houses the reactor pressure vessel. Both the reactor pressure vessel and the containment vessel include a bottom head. Additionally, the system includes a base support to contact a support surface on which the containment vessel is positioned in a substantially vertical orientation. An attenuation device is located between the bottom head of the reactor pressure vessel and the bottom head of the containment vessel. Seismic forces that travel from the base support to the reactor pressure vessel via the containment vesselmore » are attenuated by the attenuation device in a direction that is substantially lateral to the vertical orientation of the containment vessel.« less

Cooling system for a nuclear reactor

DOEpatents

Amtmann, Hans H.

1982-01-01

A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

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.

Nuclear reactor cavity floor passive heat removal system

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

Edwards, Tyler A.; Neeley, Gary W.; Inman, James B.

A nuclear reactor includes a reactor core disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor and includes a concrete floor located underneath the nuclear reactor. An ex vessel corium retention system includes flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels. In some embodiments the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluidmore » communication with the interior of the radiological containment at a second elevation higher than the first elevation. The radiological containment may include a reactor cavity containing a lower portion of the pressure vessel, wherein the concrete floor located underneath the nuclear reactor is the reactor cavity floor.« less

Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

DOEpatents

McDermott, D.J.; Schrader, K.J.; Schulz, T.L.

1994-05-03

The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

DOEpatents

McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

1994-01-01

The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

Systems and methods for managing shared-path instrumentation and irradiation targets in a nuclear reactor

DOEpatents

Heinold, Mark R.; Berger, John F.; Loper, Milton H.; Runkle, Gary A.

2015-12-29

Systems and methods permit discriminate access to nuclear reactors. Systems provide penetration pathways to irradiation target loading and offloading systems, instrumentation systems, and other external systems at desired times, while limiting such access during undesired times. Systems use selection mechanisms that can be strategically positioned for space sharing to connect only desired systems to a reactor. Selection mechanisms include distinct paths, forks, diverters, turntables, and other types of selectors. Management methods with such systems permits use of the nuclear reactor and penetration pathways between different systems and functions, simultaneously and at only distinct desired times. Existing TIP drives and other known instrumentation and plant systems are useable with access management systems and methods, which can be used in any nuclear plant with access restrictions.

Performance analysis of a continuous serpentine flow reactor for electrochemical oxidation of synthetic and real textile wastewater: Energy consumption, mass transfer coefficient and economic analysis.

PubMed

Pillai, Indu M Sasidharan; Gupta, Ashok K

2017-05-15

A continuous flow electrochemical reactor was developed, and its application was tested for the treatment of textile wastewater. A parallel plate configuration with serpentine flow was chosen for the continuous flow reactor. Uniparameter optimization was carried out for electrochemical oxidation of synthetic and real textile wastewater (collected from the inlet of the effluent treatment plant). Chemical Oxygen Demand (COD) removal efficiency of 90% was achieved for synthetic textile wastewater (initial COD - 780 mg L -1 ) at a flow rate of 500 mL h -1 (retention time of 6 h) and a current density of 1.15 mA cm -2 and the energy consumption for the degradation was 9.2 kWh (kg COD) -1 . The complete degradation of real textile wastewater (initial COD of 368 mg L -1 ) was obtained at a current density of 1.15 mA cm -2 , NaCl concentration of 1 g L -1 and retention time of 6 h. Energy consumption and mass transfer coefficient of the reactions were calculated. The continuous flow reactor performed better than batch reactor with reference to energy consumption and economy. The overall treatment cost for complete COD removal of real textile wastewater was 5.83 USD m -3 . Copyright © 2017 Elsevier Ltd. All rights reserved.

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.

Pepsin immobilized in dextran-modified fused-silica capillaries for on-line protein digestion and peptide mapping.

PubMed

Stigter, E C A; de Jong, G J; van Bennekom, W P

2008-07-07

On-line digestion of proteins under acidic conditions was studied using micro-reactors consisting of dextran-modified fused-silica capillaries with covalently immobilized pepsin. The proteins used in this study differed in molecular weight, isoelectric point and sample composition. The injected protein samples were completely digested in 3 min and the digest was analyzed with micro-high performance liquid chromatography (HPLC) and tandem mass spectrometry (MS/MS). The different proteins present in the samples could be identified with a Mascot database search on the basis of auto-MS/MS data. It proved also to be possible to digest and analyze protein mixtures with a sequence coverage of 55% and 97% for the haemoglobin beta- and alpha-chain, respectively, and 35-55% for the various casein variants. Protease auto-digestion, sample carry-over and loss of signal due to adsorption of the injected proteins were not observed. The backpressure of the reactor is low which makes coupling to systems such as Surface Plasmon Resonance biosensors, which do not tolerate too high pressure, possible. The reactor was stable for at least 40 days when used continuously.

New systematic methodology for incorporating dynamic heat transfer modelling in multi-phase biochemical reactors.

PubMed

Fernández-Arévalo, T; Lizarralde, I; Grau, P; Ayesa, E

2014-09-01

This paper presents a new modelling methodology for dynamically predicting the heat produced or consumed in the transformations of any biological reactor using Hess's law. Starting from a complete description of model components stoichiometry and formation enthalpies, the proposed modelling methodology has integrated successfully the simultaneous calculation of both the conventional mass balances and the enthalpy change of reaction in an expandable multi-phase matrix structure, which facilitates a detailed prediction of the main heat fluxes in the biochemical reactors. The methodology has been implemented in a plant-wide modelling methodology in order to facilitate the dynamic description of mass and heat throughout the plant. After validation with literature data, as illustrative examples of the capability of the methodology, two case studies have been described. In the first one, a predenitrification-nitrification dynamic process has been analysed, with the aim of demonstrating the easy integration of the methodology in any system. In the second case study, the simulation of a thermal model for an ATAD has shown the potential of the proposed methodology for analysing the effect of ventilation and influent characterization. Copyright © 2014 Elsevier Ltd. All rights reserved.

In-Pile Qualification of the Fast-Neutron-Detection-System

NASA Astrophysics Data System (ADS)

Fourmentel, D.; Villard, J.-F.; Destouches, C.; Geslot, B.; Vermeeren, L.; Schyns, M.

2018-01-01

In order to improve measurement techniques for neutron flux assessment, a unique system for online measurement of fast neutron flux has been developed and recently qualified in-pile by the French Alternative Energies and Atomic Energy Commission (CEA) in cooperation with the Belgian Nuclear Research Centre (SCK•ECEN). The Fast-Neutron-Detection-System (FNDS) has been designed to monitor accurately high-energy neutrons flux (E > 1 MeV) in typical Material Testing Reactor conditions, where overall neutron flux level can be as high as 1015 n.cm-2.s-1 and is generally dominated by thermal neutrons. Moreover, the neutron flux is coupled with a high gamma flux of typically a few 1015 γ.cm-2.s-1, which can be highly disturbing for the online measurement of neutron fluxes. The patented FNDS system is based on two detectors, including a miniature fission chamber with a special fissile material presenting an energy threshold near 1 MeV, which can be 242Pu for MTR conditions. Fission chambers are operated in Campbelling mode for an efficient gamma rejection. FNDS also includes a specific software that processes measurements to compensate online the fissile material depletion and to adjust the sensitivity of the detectors, in order to produce a precise evaluation of both thermal and fast neutron flux even after long term irradiation. FNDS has been validated through a two-step experimental program. A first set of tests was performed at BR2 reactor operated by SCK•CEN in Belgium. Then a second test was recently completed at ISIS reactor operated by CEA in France. FNDS proved its ability to measure online the fast neutron flux with an overall accuracy better than 5%.

Low Thrust, Deep Throttling, US/CIS Integrated NTRE

NASA Astrophysics Data System (ADS)

Culver, Donald W.; Kolganov, Vyacheslav; Rochow, Richard F.

1994-07-01

In 1993 our international team performed a follow-on ``Nuclear Thermal Rocket Engine (NTRE) Extended Life Feasibility Assessment'' study for the Nuclear Propulsion Office (NPO) at NASAs Lewis Research Center. The main purpose of this study was to complete the 1992 study matrix to assess NTRE designs at thrust levels of 22.5, 11.3, and 6.8 tonnes, using Commonwealth of Independent States (CIS) reactor technology. An additional Aerojet goal was to continue improving the NTRE concept we had generated. Deep throttling, mission performance optimized engine design parametrics, and reliability/cost enhancing engine system simplifications were studied, because they seem to be the last three basic design improvements sorely needed by post-NERVA NTRE. Deep throttling improves engine life by eliminating damaging thermal and mechanical shocks caused by after-cooling with pulsed coolant flow. Alternately, it improves mission performance with steady flow after-cooling by minimizing reactor over-cooling. Deep throttling also provides a practical transition from high pressures and powers of the high thrust power cycle to the low pressures and powers of our electric power generating mode. Two deep throttling designs are discussed; a workable system that was studied and a simplified system that is recommended for future study. Mission-optimized engine thrust/weight (T/W) and Isp predictions are included along with system flow schemes and concept sketches.

NE-213-scintillator-based neutron detection system for diagnostic measurements of energy spectra for neutrons having energies greater than or equal to 0.8 MeV created during plasma operations at the Princeton Tokamak Fusion Test Reactor

NASA Astrophysics Data System (ADS)

Dickens, J. K.; Hill, N. W.; Hou, F. S.; McConnell, J. W.; Spencer, R. R.; Tsang, F. Y.

1985-08-01

A system for making diagnostic measurements of the energy spectra of greater than or equal to 0.8-MeV neutrons produced during plasma operations of the Princeton Tokamak Fusion Test Reactor (TFTR) has been fabricated and tested and is presently in operation in the TFTR Test Cell Basement. The system consists of two separate detectors, each made up of cells containing liquid NE-213 scintillator attached permanently to RCA-8850 photomultiplier tubes. Pulses obtained from each photomultiplier system are amplified and electronically analyzed to identify and separate those pulses due to neutron-induced events in the detector from those due to photon-induced events in the detector. Signals from each detector are routed to two separate Analog-to-Digital Converters, and the resulting digitized information, representing: (1) the raw neutron-spectrum data; and (2) the raw photon-spectrum data, are transmited to the CICADA data-acquisition computer system of the TFTR. Software programs have been installed on the CICADA system to analyze the raw data to provide moderate-resolution recreations of the energy spectrum of the neutron and photon fluences incident on the detector during the operation of the TFTR. A complete description of, as well as the operation of, the hardware and software is given in this report.

Advanced direct coal liquefaction concepts. Quarterly report, April 1, 1993--June 30, 1993

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

Berger, D.J.; Parker, R.J.; Simpson, P.L.

Construction and commissioning of the bench unit for operation of the first stage of the process was completed. Solubilization of Black Thunder coal using carbon monoxide and steam was successfully demonstrated in the counterflow reactor system. The results were comparable with those obtained in the autoclave with the exception that coal solubilization at the same nominal residence times was slightly lower. The bench unit has now been modified for two stage operation. The Wilsonville process derived solvent for Black Thunder coal (V-1074) was found to be essentially as stable as the previous solvent used in the autoclave runs (V-178 +more » 320) at reactor conditions. This solvent (V-1074) is, therefore, being used in the bench unit tests. Carbon monoxide may be replaced by synthesis gas for the coal solubilization step in the process. However, in autoclave tests, coal conversion was found to be dependent on the amount of carbon monoxide present in the synthesis gas. Coal conversions ranged from 88% for pure carbon monoxide to 67% for a 25:75 carbon monoxide/hydrogen mixture at equivalent conditions. Two stage liquefaction tests were completed in the autoclave using a disposable catalyst (FeS) and hydrogen in the second stage. Increased coal conversion, higher gas and oil and lower asphaltene and preasphaltene yields were observed as expected. However, no hydrogen consumption was observed in the second stage. Other conditions, in particular, alternate catalyst systems will be explored.« less

Degradation of pharmaceuticals from membrane biological reactor sludge with Trametes versicolor.

PubMed

Llorens-Blanch, Guillem; Badia-Fabregat, Marina; Lucas, Daniel; Rodriguez-Mozaz, Sara; Barceló, Damià; Pennanen, Taina; Caminal, Gloria; Blánquez, Paqui

2015-02-01

Emerging contaminants are a wide group of chemical products that are found at low concentrations in the environment. These contaminants can be either natural, e.g., estrogens, or synthetics, such as pesticides and pharmaceuticals, which can enter the environment through the water and sludge from wastewater treatment plants (WWTP). The growth of Trametes versicolor on membrane biological reactor (MBR) sludge in bioslurry systems at the Erlenmeyer scale was assessed and its capacity for removing pharmaceutical and personal care products (PPCPs) was evaluated. The ability of the fungus to remove hydrochlorothiazide (HZT) from liquid media cultures was initially assessed. Consequently, different bioslurry media (complete nutrient, glucose and no-nutrient addition) and conditions (sterile and non-sterile) were tested, and the removal of spiked HZT was monitored under each condition. The highest spiked HZT removal was assessed under non-sterile conditions without nutrient addition (93.2%). Finally, the removal assessment of a broad set of pharmaceuticals was performed in non-spiked bioslurry. Under non-sterile conditions, the fungus was able to completely degrade 12 out of the 28 drugs initially detected in the MBR sludge, achieving an overall degradation of 66.9%. Subsequent microbial analysis showed that the microbial diversity increased after 15 days of treatment, but there was still some T. versicolor in the bioslurry. Results showed that T. versicolor can be used to remove PPCPs in bioslurry systems under non-sterile conditions, without extra nutrients in the media, and in matrices as complex as an MBR sludge.

Reactor application of an improved bundle divertor

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

Yang, T.F.; Ruck, G.W.; Lee, A.Y.

1978-11-01

A Bundle Divertor was chosen as the impurity control and plasma exhaust system for the beam driven Demonstration Tokamak Hybrid Reactor - DTHR. In the context of a preconceptual design study of the reactor and associated facility a bundle divertor concept was developed and integrated into the reactor system. The overall system was found feasible and scalable for reactors with intermediate torodial field strengths on axis. The important design characteristics are: the overall average current density of the divertor coils is 0.73 kA for each tesla of toroidal field on axis; the divertor windings are made from super-conducting cables supportedmore » by steel structures and are designed to be maintainable; the particle collection assembly and auxiliary cryosorption vacuum pump are dual systems designed such that they can be reactivated alterntively to allow for continuous reactor operation; and the power requirement for energizing and operating the divertor is about 5 MW.« less

Self-actuating reactor shutdown system

DOEpatents

Barrus, Donald M.; Brummond, Willian A; Peterson, Leslie F.

1988-01-01

A control system for the automatic or self-actuated shutdown or "scram" of a nuclear reactor. The system is capable of initiating scram insertion by a signal from the plant protection system or by independent action directly sensing reactor conditions of low-flow or over-power. Self-actuation due to a loss of reactor coolant flow results from a decrease of pressure differential between the upper and lower ends of an absorber element. When the force due to this differential falls below the weight of the element, the element will fall by gravitational force to scram the reactor. Self-actuation due to high neutron flux is accomplished via a valve controlled by an electromagnet and a thermionic diode. In a reactor over-power, the diode will be heated to a change of state causing the electromagnet to be shorted thereby actuating the valve which provides the changed flow and pressure conditions required for scramming the absorber element.

71. ARAII. Construction progress at SL1 site near end of ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

71. ARA-II. Construction progress at SL-1 site near end of 1957. Buildings from right to left are guard house, support building, reactor building, water tank and pump house. Construction was 23 percent complete. December 20, 1957. Ineel photo no. 57-6224. Photographer: Jack L. Anderson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Damper mechanism for nuclear reactor control elements

DOEpatents

Taft, William Elwood

1976-01-01

A damper mechanism which provides a nuclear reactor control element decelerating function at the end of the scram stroke. The total damping function is produced by the combination of two assemblies, which operate in sequence. First, a tapered dashram assembly decelerates the control element to a lower velocity, after which a spring hydraulic damper assembly takes over to complete the final damping.

Cleanup Verification Package for the 118-C-1, 105-C Solid Waste Burial Ground

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

M. J. Appel and J. M. Capron

2007-07-25

This cleanup verification package documents completion of remedial action for the 118-C-1, 105-C Solid Waste Burial Ground. This waste site was the primary burial ground for general wastes from the operation of the 105-C Reactor and received process tubes, aluminum fuel spacers, control rods, reactor hardware, spent nuclear fuel and soft wastes.

Qualification of Daiichi Units 1, 2, and 3 Data for Severe Accident Evaluations - Process and Illustrative Examples from Prior TMI-2 Evaluations

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

Rempe, Joy Lynn; Knudson, Darrell Lee

2014-09-01

The accidents at the Three Mile Island Unit 2 (TMI-2) Pressurized Water Reactor (PWR) and the Daiichi Units 1, 2, and 3 Boiling Water Reactors (BWRs) provide unique opportunities to evaluate instrumentation exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during the TMI-2 accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2more » sensor survivability and data qualification efforts. This initial review focused on the set of sensors deemed most important by post-TMI-2 instrumentation evaluation programs. Instrumentation evaluation programs focused on data required by TMI-2 operators to assess the condition of the reactor and containment and the effect of mitigating actions taken by these operators. In addition, prior efforts focused on sensors providing data required for subsequent forensic evaluations and accident simulations. To encourage the potential for similar activities to be completed for qualifying data from Daiichi Units 1, 2, and 3, this report provides additional details related to the formal process used to develop a qualified TMI-2 data base and presents data qualification details for three parameters: primary system pressure; containment building temperature; and containment pressure. As described within this report, sensor evaluations and data qualification required implementation of various processes, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design to instruments easily removed from the TMI-2 plant for evaluations. As documented in this report, results from qualifying data for these parameters led to key insights related to TMI-2 accident progression. Hence, these selected examples illustrate the types of activities completed in the TMI-2 data qualification process and the importance of such a qualification effort. These details are documented in this report to facilitate implementation of similar process using data and examinations at the Daiichi Units 1, 2, and 3 reactors so that BWR-specific benefits can be obtained.« less

Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

NASA Astrophysics Data System (ADS)

Dixon, David D.; Hiatt, Matthew T.; Poston, David I.; Kapernick, Richard J.

2006-01-01

A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

Microbial Bioreactor Development in the ALS NSCORT

NASA Astrophysics Data System (ADS)

Mitchell, Cary; Whitaker, Dawn; Banks, M. Katherine; Heber, Albert J.; Turco, Ronald F.; Nies, Loring F.; Alleman, James E.; Sharvelle, Sybil E.; Li, Congna; Heller, Megan

The NASA Specialized Center of Research and Training in Advanced Life Support (the ALS NSCORT), a partnership of Alabama A & M, Howard, and Purdue Universities, was established by NASA in 2002 to develop technologies that will reduce the Equivalent System Mass (ESM) of regenerative processes within future space life-support systems. A key focus area of NSCORT research has been the development of efficient microbial bioreactors for treatment of human, crop, and food-process wastes while enabling resource recovery. The approach emphasizes optimizing the energy-saving advantages of hydrolytic enzymes for biomass degradation, with focus on treatment of solid wastes including crop residue, paper, food, and human metabolic wastes, treatment of greywater, cabin air, off-gases from other treatment systems, and habitat condensate. This summary includes important findings from those projects, status of technology development, and recommendations for next steps. The Plant-based Anaerobic-Aerobic Bioreactor-Linked Operation (PAABLO) system was developed to reduce crop residue while generating energy and/or food. Plant residues initially were added directly to the bioreactor, and recalcitrant residue was used as a substrate for growing plants or mushrooms. Subsequently, crop residue was first pretreated with fungi to hydrolyze polymers recalcitrant to bacteria, and leachate from the fungal beds was directed to the anaerobic digester. Exoenzymes from the fungi pre-soften fibrous plant materials, improving recovery of materials that are more easily biodegraded to methane that can be used for energy reclamation. An Autothermal Thermophilic Aerobic Digestion (ATAD) system was developed for biodegradable solid wastes. Objectives were to increase water and nutrient recovery, reduce waste volume, and inactivate pathogens. Operational parameters of the reactor were optimized for degradation and resource recovery while minimizing system requirements and footprint. The start-up behavior and recycling of effluent supernatant were evaluated to maximize degradation and minimize water input. The off-gases proceeded to a bioregenerative air-treatment reactor, and the sludge effluent was investigated for multiple downstream uses including dewatering by reed beds, use as a nutrient supplement for fish or mushroom growth, and as a growth medium and nutrient source for various crops. The Bio-Regenerative Environmental Air Treatment for Health (BREATHe I) reactor treated greywater and off-gases from the thermophilic aerobic digestion reactor which contained elevated levels of ammonia (NH3 ) and hydrogen sulfide (H2 S). BREATHe I development focused initially on removing greywater contaminants with clean air supplied to a biotrickling filter. Limited removal of organic carbon (70%) led to studies indicating that biodegradation metabolites of the surfactant disodium cocoamphodiacetate are recalcitrant. Subsequent studies showed that NH3 loaded at 150 mg/min and H2 S at 0.83 mg/min were removed completely, while removal of carbonaceous compounds from greywater remained constant. A BREATHe II reactor emphasized biofilters and biotrickling filters for removal of ersatz multicomponent gaseous waste streams representative of habitat air and atmospheric condensate. The model waste stream contained a mixture of acetone, n-butanol, methane, ethylene, and ammonia. Both biofilters and biotrickling filters packed with different media were able to achieve complete removal of easily soluble compounds such as acetone, n-butanol, and ammonia within a short startup period, whereas methane was not removed because of its extreme aqueous insolubility. Different packing media and bioreactor configurations were subsequently assessed, as well as the effect of influent ammonia concentration. Research sponsored in part by NASA grant NAG5-12686.

Kinetics of D-lactic acid production by Sporolactobacillus sp. strain CASD using repeated batch fermentation.

PubMed

Zhao, Bo; Wang, Limin; Li, Fengsong; Hua, Dongliang; Ma, Cuiqing; Ma, Yanhe; Xu, Ping

2010-08-01

D-lactic acid was produced by Sporolactobacillus sp. strain CASD in repeated batch fermentation with one- and two-reactor systems. The strain showed relatively high energy consumption in its growth-related metabolism in comparison with other lactic acid producers. When the fermentation was repeated with 10% (v/v) of previous culture to start a new batch, D-lactic acid production shifted from being cell-maintenance-dependent to cell-growth-dependent. In comparison with the one-reactor system, D-lactic acid production increased approximately 9% in the fourth batch of the two-reactor system. Strain CASD is an efficient D-lactic acid producer with increased growth rate at the early stage of repeated cycles, which explains the strain's physiological adaptation to repeated batch culture and improved performance in the two-reactor fermentation system. From a kinetic point of view, two-reactor fermentation system was shown to be an alternative for conventional one-reactor repeated batch operation. Copyright 2010 Elsevier Ltd. All rights reserved.

The results of systems tests of the 500 kV busbar controllable shunting reactor in the Tavricheskaya substation

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

Gusev, S. I.; Karpov, V. N.; Kiselev, A. N.

2009-09-15

The results of systems tests of the 500 kV busbar magnetization-controllable shunting reactor (CSR), set up in the Tavricheskaya substation, including measurements of the quality of the electric power, the harmonic composition of the network currents of the reactor for different values of the reactive power consumed, the determination of the regulating characteristics of the reactor, the speed of response of the shunting reactor in the current and voltage stabilization modes, and also the operation of the reactor under dynamic conditions for different perturbations, are presented. The results obtained are analyzed.

Hydrogasification reactor and method of operating same

DOEpatents

Hobbs, Raymond; Karner, Donald; Sun, Xiaolei; Boyle, John; Noguchi, Fuyuki

2013-09-10

The present invention provides a system and method for evaluating effects of process parameters on hydrogasification processes. The system includes a hydrogasification reactor, a pressurized feed system, a hopper system, a hydrogen gas source, and a carrier gas source. Pressurized carbonaceous material, such as coal, is fed to the reactor using the carrier gas and reacted with hydrogen to produce natural gas.

Zirconium Hydride Space Power Reactor design.

NASA Technical Reports Server (NTRS)

Asquith, J. G.; Mason, D. G.; Stamp, S.

1972-01-01

The Zirconium Hydride Space Power Reactor being designed and fabricated at Atomics International is intended for a wide range of potential applications. Throughout the program a series of reactor designs have been evaluated to establish the unique requirements imposed by coupling with various power conversion systems and for specific applications. Current design and development emphasis is upon a 100 kilowatt thermal reactor for application in a 5 kwe thermoelectric space power generating system, which is scheduled to be fabricated and ground tested in the mid 70s. The reactor design considerations reviewed in this paper will be discussed in the context of this 100 kwt reactor and a 300 kwt reactor previously designed for larger power demand applications.

Improved vortex reactor system

DOEpatents

Diebold, J.P.; Scahill, J.W.

1995-05-09

An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

The 5-kwe reactor thermoelectric system summary

NASA Technical Reports Server (NTRS)

Vanosdol, J. H. (Editor)

1973-01-01

Design of the 5-kwe reactor thermoelectric system was initiated in February 1972 and extended through the conceptual design phase into the preliminary design phase. Design effort was terminated in January, 1973. This report documents the system and component requirements, design approaches, and performance and design characteristics for the 5-kwe system. Included is summary information on the reactor, radiation shields, power conversion systems, thermoelectric pump, radiator/structure, liquid metal components, and the control system.

Exhaust system with emissions storage device and plasma reactor

DOEpatents

Hoard, John W.

1998-01-01

An exhaust system for a combustion system, comprising a storage device for collecting NO.sub.x, hydrocarbon, or particulate emissions, or mixture of these emissions, and a plasma reactor for destroying the collected emissions is described. After the emission is collected in by the storage device for a period of time, the emission is then destroyed in a non-thermal plasma generated by the plasma reactor. With respect to the direction of flow of the exhaust stream, the storage device must be located before the terminus of the plasma reactor, and it may be located wholly before, overlap with, or be contained within the plasma reactor.

Gaseous fuel reactors for power systems

NASA Technical Reports Server (NTRS)

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

1977-01-01

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

Nuclear electric propulsion reactor control systems status

NASA Technical Reports Server (NTRS)

Ferg, D. A.

1973-01-01

The thermionic reactor control system design studies conducted over the past several years for a nuclear electric propulsion system are described and summarized. The relevant reactor control system studies are discussed in qualitative terms, pointing out the significant advantages and disadvantages including the impact that the various control systems would have on the nuclear electric propulsion system design. A recommendation for the reference control system is made, and a program for future work leading to an engineering model is described.

The first PANDA tests

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

Dreier, J.; Huggenberger, M.; Aubert, C.

1996-08-01

The PANDA test facility at PSI in Switzerland is used to study the long-term Simplified Boiling Water Reactor (SBWR) Passive Containment Cooling System (PCCS) performance. The PANDA tests demonstrate performance on a larger scale than previous tests and examine the effects of any non-uniform spatial distributions of steam and non-condensables in the system. The PANDA facility has a 1:1 vertical scale, and 1:25 ``system`` scale (volume, power, etc.). Steady-state PCCS condenser performance tests and extensive facility characterization tests have been completed. Transient system behavior tests were conducted late in 1995; results from the first three transient tests (M3 series) aremore » reviewed. The first PANDA tests showed that the overall global behavior of the SBWR containment was globally repeatable and very favorable; the system exhibited great ``robustness.``« less

Novel duplex vapor: Electrochemical method for silicon solar cells. [chemical reactor for a silicon sodium reaction system

NASA Technical Reports Server (NTRS)

Nanis, L.; Sanjurjo, A.; Sancier, K.

1979-01-01

The scaled up chemical reactor for a SiF4-Na reaction system is examined for increased reaction rate and production rate. The reaction system which now produces 5 kg batches of mixed Si and NaF is evaluated. The reactor design is described along with an analysis of the increased capacity of the Na chip feeder. The reactor procedure is discussed and Si coalescence in the reaction products is diagnosed.

REACTOR CONTROL SYSTEM

DOEpatents

MacNeill, J.H.; Estabrook, J.Y.

1960-05-10

A reactor control system including a continuous tape passing through a first coolant passageway, over idler rollers, back through another parallel passageway, and over motor-driven rollers is described. Discrete portions of fuel or poison are carried on two opposed active sections of the tape. Driving the tape in forward or reverse directions causes both active sections to be simultaneously inserted or withdrawn uniformly, tending to maintain a more uniform flux within the reactor. The system is particularly useful in mobile reactors, where reduced inertial resistance to control rod movement is important.

78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-25

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0237] Cost-Benefit Analysis for Radwaste Systems for Light... (RG) 1.110, ``Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors... components for light water nuclear power reactors. ADDRESSES: Please refer to Docket ID NRC-2013-0237 when...

Application of a Self-Actuating Shutdown System (SASS) to a Gas-Cooled Fast Reactor (GCFR)

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

Germer, J.H.; Peterson, L.F.; Kluck, A.L.

1980-09-01

The application of a SASS (Self-Actuated Shutdown System) to a GCFR (Gas-Cooled Fast Reactor) is compared with similar systems designed for an LMFBR (Liquid Metal Fast Breeder Reactor). A comparison of three basic SASS concepts is given: hydrostatic holdup, fluidic control, and magnetic holdup.

78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-24

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling....79.1, ``Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors.'' This... emergency core cooling systems (ECCSs) for boiling- water reactors (BWRs) whose licenses are issued after...

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

Effects of glucose on the performance of enhanced biological phosphorus removal activated sludge enriched with acetate.

PubMed

Gebremariam, Seyoum Yami; Beutel, Marc W; Christian, David; Hess, Thomas F

2012-10-01

The effects of glucose on enhanced biological phosphorus removal (EBPR) activated sludge enriched with acetate was investigated using sequencing batch reactors. A glucose/acetate mixture was serially added to the test reactor in ratios of 25/75%, 50/50%, and 75/25% and the EBPR activity was compared to the control reactor fed with 100% acetate. P removal increased at a statistically significant level to a near-complete in the test reactor when the mixture increased to 50/50%. However, EBPR deteriorated when the glucose/acetate mixture increased to 75/25% in the test reactor and when the control reactor abruptly switched to 100% glucose. These results, in contrast to the EBPR conventional wisdom, suggest that the addition of glucose at moderate levels in wastewaters does not impede and may enhance EBPR, and that glucose waste products should be explored as an economical sustainable alternative when COD enhancement of EBPR is needed. Copyright © 2012 Elsevier Ltd. All rights reserved.

Anaerobic bioconversion of organic waste into biogas by hot water treatment at near-critical conditions: application in bioregenerative life support.

PubMed

Lissens, Geert; Verstraete, Willy; Albrecht, Tobias; Brunner, Gerd; Lasseur, Christophe

2003-01-01

The feasibility of nearly-complete conversion of lignocellulosic waste (70% food crops, 20% faecal matter and 10% green algae) into biogas was investigated in the context of a Life Support Project. The treatment comprised a series of processes, i.e. a mesophilic laboratory scale CSTR (continuously stirred tank reactor), an upflow biofilm reactor and a hydrothermolysis system in near-critical water. By the one-stage CSTR, a biogas yield of 75% with a specific biogas production of 0.37 l biogas g(-1) VSS (volatile suspended solids) added at a HRT (hydraulic retention time) of 20 d was obtained. Biogas yields further increased with 10-15% at HRT > 20 d, indicating the hydrolysis of lignocellulose to be the rate-limiting conversion step. The solids present in the CSTR-effluent were subsequently treated by hot water treatment (T approximately 310-350 degrees C, p approximately 240 bar), resulting in effective carbon liquefaction (50-60% without and 83% with carbon dioxide saturation) and complete hygienisation of the residue. Subsequent anaerobic digestion of the hydrolysate allowed further conversion of 48-60% on COD (chemical oxygen demand) basis. Thus, the total process yielded biogas corresponding with a COD conversion up to 90% of the original organic matter. It appears that mesophilic digestion in conjunction with hydrothermolysis at near-critical conditions offers interesting features for (nearly) complete, non-toxic and hygienic carbon and energy recovery from human waste in a bioregenerative life support context.

Direct energy recovery from primary and secondary sludges by supercritical water oxidation.

PubMed

Svanström, M; Modell, M; Tester, J

2004-01-01

Supercritical water oxidation (SCWO) oxidizes organic and biological materials virtually completely to benign products without the need for stack gas scrubbing. Heavy metals are recovered as stabilized solid, along with the sand and clay that is present in the feed. The technology has been under development for twenty years. The major obstacle to commercialization has been developing reactors that are not clogged by inorganic solid deposits. That problem has been solved by using tubular reactors with fluid velocities that are high enough to keep solids in suspension. Recently, system designs have been created that reduce the cost of processing sewage sludges below that of incineration. At 10 wt- % dry solids, sludge can be oxidized with virtually complete recovery of the sludge heating value as hot water or high-pressure steam. Liquid carbon dioxide of high purity can be recovered from the gaseous effluent and excess oxygen can be recovered for recycle. The net effect is to reduce the stack to a harmless vent with minimal flow rate of a clean gas. Complete simulations have been developed using physical property models that accurately simulate the thermodynamic properties of sub- and supercritical water in mixtures with O2, N2, CO2, and organics. Capital and operating cost estimates are given for sewage sludge treatment, which are less costly than incineration. The scenario of direct recovery of energy from sludges has inherent benefits compared to other gasification or liquefaction options.

Analysis of decommissioning costs for the AFRRI TRIGA reactor facility. Technical report

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

Forsbacka, M.; Moore, M.

1989-12-01

This report provides a cost analysis for decommissioning the Armed Forces Radiobiology Research Institute (AFRRI) TRIGA reactor facility. AFRRI is not suggesting that the AFRRI TRIGA reactor facility be decommissioned. This report was prepared in compliance with paragraph 50.33 of Title 10, Code of Federal Regulations, which requires that funding for the decommissioning of reactor facilities be available when licensed activities cease. The planned method of decommissioning is complete decontamination (DECON) of the AFRRI TRIGA reactor site to allow for restoration of the site to full public access. The cost of DECON in 1990 dollars is estimated to be $3,200,000.more » The anticipated ancillary costs of facility site demobilization and spent fuel shipment will be an additional $600,000. Thus, the total cost of terminating reactor operations at AFRRI will be about $3,800,000. The primary basis for developing this cost estimate was a study of the decommissioning costs of similar reactor facility performed by Battelle Pacific Northwest Laboratory, as provided in U.S. Nuclear Regulatory Commission publication NUREG/CR-1756. The data in this study were adapted to reflect the decommissioning requirements of the AFRRI TRIGA reactor facility.« less

A low-cost municipal sewage treatment system with a combination of UASB and the "fourth-generation" downflow hanging sponge reactors.

PubMed

Tandukar, M; Uemura, S; Machdar, I; Ohashi, A; Harada, H

2005-01-01

This paper presents an evaluation of the process performance of a pilot-scale "fourth generation" downflow hanging sponge (DHS) post-treatment system combined with a UASB pretreatment unit treating municipal wastewater. After the successful operation of the second- and third-generation DHS reactors, the fourth-generation DHS reactor was developed to overcome a few shortcomings of its predecessors. This reactor was designed to further enhance the treatment efficiency and simplify the construction process in real scale, especially for the application in developing countries. Configuration of the reactor was modified to enhance the dissolution of air into the wastewater and to avert the possible clogging of the reactor especially during sudden washout from the UASB reactor. The whole system was operated at a total hydraulic retention time (HRT) of 8 h (UASB: 6 h and DHS: 2 h) for a period of over 600 days. The combined system was able to remove 96% of unfiltered BOD with only 9 mg/L remaining in the final effluent. Likewise, F. coli were removed by 3.45 log with the final count of 10(3) to 10(4) MPN/100 ml. Nutrient removal by the system was also satisfactory.

Biodegradation of tech-hexachlorocyclohexane in a upflow anaerobic sludge blanket (UASB) reactor.

PubMed

Bhat, Praveena; Kumar, M Suresh; Mudliar, Sandeep N; Chakrabarti, T

2006-04-01

Biodegradability of technical grade hexachlorocyclohexane (tech-HCH) was studied in an upflow anaerobic sludge blanket reactor (UASB) under continuous mode of operation in concentration range of 100-200 mg/l and constant HRT of 48 h. At steady state operation more than 85% removal of tech-HCH (upto 175 mg/l concentration) and complete disappearance of beta-HCH was observed. Kinetic constants in terms of maximum specific tech-HCH utilization rate (k) and half saturation velocity constant (K(L)) were found to be 11.88 mg/g/day and 8.11 mg/g/day, respectively. The tech-HCH degrading seed preparation, UASB reactor startup and degradation in continuous mode of operation of the reactor is presented in this paper.

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.

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

Multi-reactor power system configurations for multimegawatt nuclear electric propulsion

NASA Technical Reports Server (NTRS)

George, Jeffrey A.

1991-01-01

A modular, multi-reactor power system and vehicle configuration for piloted nuclear electric propulsion (NEP) missions to Mars is presented. Such a design could provide enhanced system and mission reliability, allowing a comfortable safety margin for early manned flights, and would allow a range of piloted and cargo missions to be performed with a single power system design. Early use of common power modules for cargo missions would also provide progressive flight experience and validation of standardized systems for use in later piloted applications. System and mission analysis are presented to compare single and multi-reactor configurations for piloted Mars missions. A conceptual design for the Hydra modular multi-reactor NEP vehicle is presented.

Benchmark Simulation of Natural Circulation Cooling System with Salt Working Fluid Using SAM

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

Ahmed, K. K.; Scarlat, R. O.; Hu, R.

Liquid salt-cooled reactors, such as the Fluoride Salt-Cooled High-Temperature Reactor (FHR), offer passive decay heat removal through natural circulation using Direct Reactor Auxiliary Cooling System (DRACS) loops. The behavior of such systems should be well-understood through performance analysis. The advanced system thermal-hydraulics tool System Analysis Module (SAM) from Argonne National Laboratory has been selected for this purpose. The work presented here is part of a larger study in which SAM modeling capabilities are being enhanced for the system analyses of FHR or Molten Salt Reactors (MSR). Liquid salt thermophysical properties have been implemented in SAM, as well as properties ofmore » Dowtherm A, which is used as a simulant fluid for scaled experiments, for future code validation studies. Additional physics modules to represent phenomena specific to salt-cooled reactors, such as freezing of coolant, are being implemented in SAM. This study presents a useful first benchmark for the applicability of SAM to liquid salt-cooled reactors: it provides steady-state and transient comparisons for a salt reactor system. A RELAP5-3D model of the Mark-1 Pebble-Bed FHR (Mk1 PB-FHR), and in particular its DRACS loop for emergency heat removal, provides steady state and transient results for flow rates and temperatures in the system that are used here for code-to-code comparison with SAM. The transient studied is a loss of forced circulation with SCRAM event. To the knowledge of the authors, this is the first application of SAM to FHR or any other molten salt reactors. While building these models in SAM, any gaps in the code’s capability to simulate such systems are identified and addressed immediately, or listed as future improvements to the code.« less

The combined hybrid system: A symbiotic thermal reactor/fast reactor system for power generation and radioactive waste toxicity reduction

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

Hollaway, W.R.

1991-08-01

If there is to be a next generation of nuclear power in the United States, then the four fundamental obstacles confronting nuclear power technology must be overcome: safety, cost, waste management, and proliferation resistance. The Combined Hybrid System (CHS) is proposed as a possible solution to the problems preventing a vigorous resurgence of nuclear power. The CHS combines Thermal Reactors (for operability, safety, and cost) and Integral Fast Reactors (for waste treatment and actinide burning) in a symbiotic large scale system. The CHS addresses the safety and cost issues through the use of advanced reactor designs, the waste management issuemore » through the use of actinide burning, and the proliferation resistance issue through the use of an integral fuel cycle with co-located components. There are nine major components in the Combined Hybrid System linked by nineteen nuclear material mass flow streams. A computer code, CHASM, is used to analyze the mass flow rates CHS, and the reactor support ratio (the ratio of thermal/fast reactors), IFR of the system. The primary advantages of the CHS are its essentially actinide-free high-level radioactive waste, plus improved reactor safety, uranium utilization, and widening of the option base. The primary disadvantages of the CHS are the large capacity of IFRs required (approximately one MW{sub e} IFR capacity for every three MW{sub e} Thermal Reactor) and the novel radioactive waste streams produced by the CHS. The capability of the IFR to burn pure transuranic fuel, a primary assumption of this study, has yet to be proven. The Combined Hybrid System represents an attractive option for future nuclear power development; that disposal of the essentially actinide-free radioactive waste produced by the CHS provides an excellent alternative to the disposal of intact actinide-bearing Light Water Reactor spent fuel (reducing the toxicity based lifetime of the waste from roughly 360,000 years to about 510 years).« less

Analysis of space reactor system components: Investigation through simulation and non-nuclear testing

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.

The use of fission energy in space power and propulsion systems offers considerable advantages over chemical propulsion. Fission provides over six orders of magnitude higher energy density, which translates to higher vehicle specific impulse and lower specific mass. These characteristics enable ambitious space exploration missions. The natural space radiation environment provides an external source of protons and high energy, high Z particles that can result in the production of secondary neutrons through interactions in reactor structures. Applying the approximate proton source in geosynchronous orbit during a solar particle event, investigation using MCNPX 2.5.b for proton transport through the SAFE-400 heat pipe cooled reactor indicates an incoming secondary neutron current of (1.16 +/- 0.03) x 107 n/s at the core-reflector interface. This neutron current may affect reactor operation during low power maneuvers (e.g., start-up) and may provide a sufficient reactor start-up source. It is important that a reactor control system be designed to automatically adjust to changes in reactor power levels, maintaining nominal operation without user intervention. A robust, autonomous control system is developed and analyzed for application during reactor start-up, accounting for fluctuations in the radiation environment that result from changes in vehicle location or to temporal variations in the radiation field. Development of a nuclear reactor for space applications requires a significant amount of testing prior to deployment of a flight unit. High confidence in fission system performance can be obtained through relatively inexpensive non-nuclear tests performed in relevant environments, with the heat from nuclear fission simulated using electric resistance heaters. A series of non-nuclear experiments was performed to characterize various aspects of reactor operation. This work includes measurement of reactor core deformation due to material thermal expansion and implementation of a virtual reactivity feedback control loop; testing and thermal hydraulic characterization of the coolant flow paths for two space reactor concepts; and analysis of heat pipe operation during start-up and steady state operation.

An adaptive load-following control system for a space nuclear power system

NASA Astrophysics Data System (ADS)

Metzger, John D.; El-Genk, Mohamed S.

An adaptive load-following control system is proposed for a space nuclear power system. The conceptual design of the SP-100 space nuclear power system proposes operating the nuclear reactor at a base thermal power and accommodating changes in the electrical power demand with a shunt regulator. It is necessary to increase the reactor thermal power if the payload electrical demand exceeds the peak system electrical output for the associated reactor power. When it is necessary to change the nuclear reactor power to meet a change in the power demand, the power ascension or descension must be accomplished in a predetermined manner to avoid thermal stresses in the system and to achieve the desired reactor period. The load-following control system described has the ability to adapt to changes in the system and to changes in the satellite environment. The application is proposed of the model reference adaptive control (MRAC). The adaptive control system has the ability to control the dynamic response of nonlinear systems. Three basic subsets of adaptive control are: (1) gain scheduling, (2) self-tuning regulators, and (3) model reference adaptive control.

A Comparison of Fission Power System Options for Lunar and Mars Surface Applications

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2006-01-01

This paper presents a comparison of reactor and power conversion design options for 50 kWe class lunar and Mars surface power applications with scaling from 25 to 200 kWe. Design concepts and integration approaches are provided for three reactor-converter combinations: gas-cooled Brayton, liquid-metal Stirling, and liquid-metal thermoelectric. The study examines the mass and performance of low temperature, stainless steel based reactors and higher temperature refractory reactors. The preferred system implementation approach uses crew-assisted assembly and in-situ radiation shielding via installation of the reactor in an excavated hole. As an alternative, self-deployable system concepts that use earth-delivered, on-board radiation shielding are evaluated. The analyses indicate that among the 50 kWe stainless steel reactor options, the liquid-metal Stirling system provides the lowest mass at about 5300 kg followed by the gas-cooled Brayton at 5700 kg and the liquid-metal thermoelectric at 8400 kg. The use of a higher temperature, refractory reactor favors the gas-cooled Brayton option with a system mass of about 4200 kg as compared to the Stirling and thermoelectric options at 4700 and 5600 kg, respectively. The self-deployed concepts with on-board shielding result in a factor of two system mass increase as compared to the in-situ shielded concepts.

SP-100 Program: space reactor system and subsystem investigations

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

Harty, R.B.

1983-09-30

For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs.

SP-100 program: Space reactor system and subsystem investigations

NASA Astrophysics Data System (ADS)

Harty, R. B.

1983-09-01

For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. The nuclear safety review/approval process that is required for a space reactor system is summarized. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that is expected and to provide information that could be usable in future programs.

Integral Design Methodology of Photocatalytic Reactors for Air Pollution Remediation.

PubMed

Passalía, Claudio; Alfano, Orlando M; Brandi, Rodolfo J

2017-06-07

An integral reactor design methodology was developed to address the optimal design of photocatalytic wall reactors to be used in air pollution control. For a target pollutant to be eliminated from an air stream, the proposed methodology is initiated with a mechanistic derived reaction rate. The determination of intrinsic kinetic parameters is associated with the use of a simple geometry laboratory scale reactor, operation under kinetic control and a uniform incident radiation flux, which allows computing the local superficial rate of photon absorption. Thus, a simple model can describe the mass balance and a solution may be obtained. The kinetic parameters may be estimated by the combination of the mathematical model and the experimental results. The validated intrinsic kinetics obtained may be directly used in the scaling-up of any reactor configuration and size. The bench scale reactor may require the use of complex computational software to obtain the fields of velocity, radiation absorption and species concentration. The complete methodology was successfully applied to the elimination of airborne formaldehyde. The kinetic parameters were determined in a flat plate reactor, whilst a bench scale corrugated wall reactor was used to illustrate the scaling-up methodology. In addition, an optimal folding angle of the corrugated reactor was found using computational fluid dynamics tools.

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

Dynamic Response Testing in an Electrically Heated Reactor Test Facility

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.; Morton, T. J.

2006-01-01

Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG testing will utilize a higher fidelity point kinetics model to control core power transients, and reactivity feedback will be based on localized feedback coefficients and several independent temperature measurements taken within the core block. This paper presents preliminary test results and discusses the methodology that will be implemented in follow-on DDG testing and the additional instrumentation required to implement high fidelity dynamic testing.

Application of a fluidized bed reactor charged with aragonite for control of alkalinity, pH and carbon dioxide in marine recirculating aquaculture systems

USGS Publications Warehouse

Paul S Wills, PhD; Pfeiffer, Timothy; Baptiste, Richard; Watten, Barnaby J.

2016-01-01

Control of alkalinity, dissolved carbon dioxide (dCO2), and pH are critical in marine recirculating aquaculture systems (RAS) in order to maintain health and maximize growth. A small-scale prototype aragonite sand filled fluidized bed reactor was tested under varying conditions of alkalinity and dCO2 to develop and model the response of dCO2 across the reactor. A large-scale reactor was then incorporated into an operating marine recirculating aquaculture system to observe the reactor as the system moved toward equilibrium. The relationship between alkalinity dCO2, and pH across the reactor are described by multiple regression equations. The change in dCO2 across the small-scale reactor indicated a strong likelihood that an equilibrium alkalinity would be maintained by using a fluidized bed aragonite reactor. The large-scale reactor verified this observation and established equilibrium at an alkalinity of approximately 135 mg/L as CaCO3, dCO2 of 9 mg/L, and a pH of 7.0 within 4 days that was stable during a 14 day test period. The fluidized bed aragonite reactor has the potential to simplify alkalinity and pH control, and aid in dCO2 control in RAS design and operation. Aragonite sand, purchased in bulk, is less expensive than sodium bicarbonate and could reduce overall operating production costs.

NEUTRONIC REACTOR SYSTEM

DOEpatents

Goett, J.J.

1961-01-24

A system is described which includes a neutronic reactor containing a dispersion of fissionable material in a liquid moderator as fuel and a conveyor to which a portion of the dispersion may be passed and wherein the self heat of the slurry evaporates the moderator. Means are provided for condensing the liquid moderator and returning it to the reactor and for conveying the dried fissionable material away from the reactor.

The influence of pH adjustment on kinetics parameters in tapioca wastewater treatment using aerobic sequencing batch reactor system

NASA Astrophysics Data System (ADS)

Mulyani, Happy; Budianto, Gregorius Prima Indra; Margono, Kaavessina, Mujtahid

2018-02-01

The present investigation deals with the aerobic sequencing batch reactor system of tapioca wastewater treatment with varying pH influent conditions. This project was carried out to evaluate the effect of pH on kinetics parameters of system. It was done by operating aerobic sequencing batch reactor system during 8 hours in many tapioca wastewater conditions (pH 4.91, pH 7, pH 8). The Chemical Oxygen Demand (COD) and Mixed Liquor Volatile Suspended Solids (MLVSS) of the aerobic sequencing batch reactor system effluent at steady state condition were determined at interval time of two hours to generate data for substrate inhibition kinetics parameters. Values of the kinetics constants were determined using Monod and Andrews models. There was no inhibition constant (Ki) detected in all process variation of aerobic sequencing batch reactor system for tapioca wastewater treatment in this study. Furthermore, pH 8 was selected as the preferred aerobic sequencing batch reactor system condition in those ranging pH investigated due to its achievement of values of kinetics parameters such µmax = 0.010457/hour and Ks = 255.0664 mg/L COD.

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

CONTROL RODS FOR NUCLEAR REACTOR CORES

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

Bell, F.R.

1961-11-15

A reactor control rod is designed which has increased effectiveness as compared with the width of the aperture in the pressure vessel through which it passes. The control rod carries six fins, three on each side, and two of the fins are fixed while the other, being adjustable, is capable of movement from between the fixed fins to an extended position. Thus, the control rod assembly can be arranged so that the parts within the core form a substantially complete shell around the reactor central axis, while the apertures on the pressure vessel wall are well spaced for strength. (D.L.C.)

Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

NASA Technical Reports Server (NTRS)

Springborn, R. H.

1971-01-01

Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

Year One Summary of X-energy Pebble Fuel Development at ORNL

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

Helmreich, Grant W.; Hunn, John D.; McMurray, Jake W.

2017-06-01

The Advanced Reactor Concepts X-energy (ARC-Xe) Pebble Fuel Development project at Oak Ridge National Laboratory (ORNL) has successfully completed its first year, having made excellent progress in accomplishing programmatic objectives. The primary focus of research at ORNL in support of X-energy has been the training of X-energy fuel fabrication engineers and the establishment of US pebble fuel production capabilities able to supply the Xe-100 pebble-bed reactor. These efforts have been strongly supported by particle fuel fabrication and characterization expertise present at ORNL from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program.

Slow clean-up for fast reactor

NASA Astrophysics Data System (ADS)

Banks, Michael

2008-05-01

The year 2300 is so distant that one may be forgiven for thinking of it only in terms of science fiction. But this is the year that workers at the Dounreay power station in Northern Scotland - the UK's only centre for research into "fast" nuclear reactors - term as the "end point" by which time the site will be completely clear of radioactive material. More than 180 facilities - including the iconic dome that housed the Dounreay Fast Reactor (DFR) - were built at at the site since it opened in 1959, with almost 50 having been used to handle radioactive material.

Long lifetime fast spectrum reactor for lunar surface power system

NASA Astrophysics Data System (ADS)

Kambe, Mitsuru

1993-01-01

In the framework of innovative reactor research activities, a conceptual design study of fast spectrum reactor and primary system for 800 kWe lunar surface power system to be combined with potassium Rankine cycle power conversion has been conducted to meet the power requirements of the lunar base activities in the next century. The reactor subsystem is characterized by RAPID (Refueling by All Pins Integrated Design) concept to enhance inherent safety and to enable quick and simplifed refueling in every 10 years. RAPID concept affords power plant design lifetime of up to 30 years. Integrity of the reactor structure and replacement of failed primary circuits are also discussed. Substantial reduction in per-kWh cost on considering launch, emplacement, and final disposition can be expected by a long system lifetime.

Component and System Sensitivity Considerations for Design of a Lunar ISRU Oxygen Production Plant

NASA Technical Reports Server (NTRS)

Linne, Diane L.; Gokoglu, Suleyman; Hegde, Uday G.; Balasubramaniam, Ramaswamy; Santiago-Maldonado, Edgardo

2009-01-01

Component and system sensitivities of some design parameters of ISRU system components are analyzed. The differences between terrestrial and lunar excavation are discussed, and a qualitative comparison of large and small excavators is started. The effect of excavator size on the size of the ISRU plant's regolith hoppers is presented. Optimum operating conditions of both hydrogen and carbothermal reduction reactors are explored using recently developed analytical models. Design parameters such as batch size, conversion fraction, and maximum particle size are considered for a hydrogen reduction reactor while batch size, conversion fraction, number of melt zones, and methane flow rate are considered for a carbothermal reduction reactor. For both reactor types the effect of reactor operation on system energy and regolith delivery requirements is presented.

Microprocessor tester for the treat upgrade reactor trip system

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

Lenkszus, F.R.; Bucher, R.G.

1984-01-01

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety systemmore » is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations.« less

Liquid metal cooled nuclear reactors with passive cooling system

DOEpatents

Hunsbedt, Anstein; Fanning, Alan W.

1991-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of cooling medium flow circuits which cooperate to remove and carry heat away from the fuel core upon loss of the normal cooling flow circuit to areas external thereto.

10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

Code of Federal Regulations, 2012 CFR

2012-01-01

..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

Code of Federal Regulations, 2013 CFR

2013-01-01

..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

Code of Federal Regulations, 2014 CFR

2014-01-01

..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

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

NASA Technical Reports Server (NTRS)

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

1973-01-01

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

Demonstration of Robustness and Integrated Operation of a Series-Bosch System

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, J. Matthew; Barnett, Bill; Stanley, Christine M.; Junaedi, Christian; Vilekar, Saurabh A.; Kent, Ryan

2016-01-01

Manned missions beyond low Earth orbit will require highly robust, reliable, and maintainable life support systems that maximize recycling of water and oxygen. Bosch technology is one option to maximize oxygen recovery, in the form of water, from metabolically-produced carbon dioxide (CO2). A two stage approach to Bosch, called Series-Bosch, reduces metabolic CO2 with hydrogen (H2) to produce water and solid carbon using two reactors: a Reverse Water-Gas Shift (RWGS) reactor and a carbon formation (CF) reactor. Previous development efforts demonstrated the stand-alone performance of a RWGS reactor containing Incofoam(TradeMark) catalyst and designed for robustness against carbon formation, two membrane separators intended to maximize single pass conversion of reactants, and a batch CF reactor with both transit and surface catalysts. In the past year, Precision Combustion, Inc. (PCI) developed and delivered a RWGS reactor for testing at NASA. The reactor design was based on their patented Microlith(TradeMark) technology and was first evaluated under a Phase I Small Business Innovative Research (SBIR) effort in 2010. The Microlith(TradeMark) RWGS reactor was recently evaluated at NASA to compare its performance and operating conditions with the Incofoam(TradeMark) RWGS reactor. Separately, in 2015, a fully integrated demonstration of an S-Bosch system was conducted. In an effort to mitigate risk, a second integrated test was conducted to evaluate the effect of membrane failure on a closed-loop Bosch system. Here, we report and discuss the performance and robustness to carbon formation of both RWGS reactors. We report the results of the integrated operation of a Series-Bosch system and we discuss the technology readiness level. 1

JPRS Report, Science & Technology, China

DTIC Science & Technology

1992-09-24

Yuhong; YUHANG XUEBAO, No 3, Jul 92] 23 Improvement of Manufacturing Process and Analysis of Tensile Strength of SiC/Al Preform Wire [Wan Hong...centered on 600MW pressur- ized-water reactor nuclear power plants . Complete devel- opment of the 200MW nuclear low-temperature heat supply reactor...grain yields, substan- tially reduce the amounts of farm chemicals used; develop plant genetic atlas research, try to make major research

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

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

Davenport, Michael; Petti, D. A.

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

PRESSURIZED WATER REACTOR PROGRAM TECHNICAL PROGRESS REPORT FOR THE PERIOD MAY 5, 1955 TO JUNE 16, 1955

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

None

The current PWR plant and core parameters are listed. Resign requirements are briefly summarized for a radiation monitoring system, a fuel handling water system, a coolant purification system, an electrical power distribution system, and component shielding. Results of studies on thermal bowing and stressing of UO/sub 2/ are reported. A graph is presented of reactor power vs. reactor flow for various hot channel conditions. Development of U-- Mo and U-Nb alloys has been stopped because of the recent selection of UO/sub 2/ fuel material for the PWR core and blanket. The fabrication characteristics of UO/sub 2/ powders are being studied.more » Seamless Zircaloy-2 tubing has been tested to determine elastic limits, bursting pressures, and corrosion resistance. Fabrication techniques and tests for corrosion and defects in Zircaloy-clad U-Mo and UO/sub 2/ fuel rods are described. The preparation of UO/sub 2/ by various methods is being studied to determine which method produces a material most suitable for PWR fuel elements. The stability of UO/sub 2/ compacts in high temperature water and steam is being determined. Surface area and density measurements have been performed on samples of UO/sub 2/ powder prepared by various methods. Revelopment work on U-- Mo and U--Nb alloys has included studies of the effect on corrosion behavior of additions to the test water, additions to the alloys, homogenization of the alloys, annealing times, cladding, and fabrication techniques. Data are presented on relaxation in spring materials after exposure to a corrosive environment. Results are reported from loop and autoclave tests on fission product and crud deposition. Results of irradiation and corrosion testing of clad and unclad U--Mo and U-Nh alloys are described. The UO/sub 2/ irradiation program has included studies of dimensional changes, release of fission gases, and activity in the water surrounding the samples. A review of the methods of calculating reactor physics parameters has been completed, and the established procedures have been applied to determination of PWR reference design parameters. Critical experiments and primary loop shielding analyses are described. (D.E.B.)« less

JEN-1 Reactor Control System; SISTEMA DE CONTROL DEL REACTOR JEN-1

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

Cantillo, M.F.; Nuno, C.M.; Andreu, J.L.M.

1963-01-01

ABS>The JEN-1 3Mw power swimming pool reactor electrical control circuits are described. Start-up, power generation in the core, and shutdown are controlled by the reactor control system. This control system guarantees in each moment the safety conditions during reactor operation. Each circuit was represented by a scheme, complemented with a description of its function, components, and operation theory. Components described include: scram circuit; fission counter control circuit; servo control circuit; control circuit of safety sheets; control circuits of primary, secondary, and clean-up pump motors and tower fan motor; primary valve motor circuit; center cubicle alarm circuit; and process alarm circuit.more » (auth)« less

Method of producing gaseous products using a downflow reactor

DOEpatents

Cortright, Randy D; Rozmiarek, Robert T; Hornemann, Charles C

2014-09-16

Reactor systems and methods are provided for the catalytic conversion of liquid feedstocks to synthesis gases and other noncondensable gaseous products. The reactor systems include a heat exchange reactor configured to allow the liquid feedstock and gas product to flow concurrently in a downflow direction. The reactor systems and methods are particularly useful for producing hydrogen and light hydrocarbons from biomass-derived oxygenated hydrocarbons using aqueous phase reforming. The generated gases may find used as a fuel source for energy generation via PEM fuel cells, solid-oxide fuel cells, internal combustion engines, or gas turbine gensets, or used in other chemical processes to produce additional products. The gaseous products may also be collected for later use or distribution.

Synthesis of supported silver nano-spheres on zinc oxide nanorods for visible light photocatalytic applications

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

Saoud, Khaled; Alsoubaihi, Rola; Bensalah, Nasr

Highlights: • Synthesis of supported Ag NPs on ZnO nanorods using open vessel microwave reactor. • Use of the Ag/ZnO NPs as an efficient visible light photocatalyst. • Complete degradation of methylene blue in 1 h with 0.5 g/L Ag/ZnO NPs. - Abstract: We report the synthesis of silver (Ag) nano-spheres (NS) supported on zinc oxide (ZnO) nanorods through two step mechanism, using open vessel microwave reactor. Direct reduction of ZnO from zinc nitrates was followed by deposition precipitation of the silver on the ZnO nanorods. The supported Ag/ZnO nanoparticles were then characterized by electron microscopy, X-ray diffraction, FTIR, photoluminescencemore » and UV–vis spectroscopy. The visible light photocatalytic activity of Ag/ZnO system was investigated using a test contaminant, methylene blue (MB). Almost complete removal of MB in about 60 min for doses higher than 0.5 g/L of the Ag/ZnO photocatalyst was achieved. This significant improvement in the photocatalytic efficiency of Ag/ZnO photocatalyst under visible light irradiation can be attributed to the presence of Ag nanoparticles on the ZnO nanoparticles which greatly enhances absorption in the visible range of solar spectrum enabled by surface plasmon resonance effect from Ag nanoparticles.« less

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

Basher, A.M.H.

Poor control of steam generator water level of a nuclear power plant may lead to frequent nuclear reactor shutdowns. These shutdowns are more common at low power where the plant exhibits strong non-minimum phase characteristics and flow measurements at low power are unreliable in many instances. There is need to investigate this problem and systematically design a controller for water level regulation. This work is concerned with the study and the design of a suitable controller for a U-Tube Steam Generator (UTSG) of a Pressurized Water Reactor (PWR) which has time varying dynamics. The controller should be suitable for themore » water level control of UTSG without manual operation from start-up to full load transient condition. Some preliminary simulation results are presented that demonstrate the effectiveness of the proposed controller. The development of the complete control algorithm includes components such as robust output tracking, and adaptively estimating both the system parameters and state variables simultaneously. At the present time all these components are not completed due to time constraints. A robust tracking component of the controller for water level control is developed and its effectiveness on the parameter variations is demonstrated in this study. The results appear encouraging and they are only preliminary. Additional work is warranted to resolve other issues such as robust adaptive estimation.« less

Progress in space nuclear reactor power systems technology development - The SP-100 program

NASA Technical Reports Server (NTRS)

Davis, H. S.

1984-01-01

Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

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.

Small-volume, ultrahigh-vacuum-compatible high-pressure reaction cell for combined kinetic and in situ IR spectroscopic measurements on planar model catalysts

NASA Astrophysics Data System (ADS)

Zhao, Z.; Diemant, T.; Häring, T.; Rauscher, H.; Behm, R. J.

2005-12-01

We describe the design and performance of a high-pressure reaction cell for simultaneous kinetic and in situ infrared reflection (IR) spectroscopic measurements on model catalysts at elevated pressures, between 10-3 and 103mbars, which can be operated both as batch reactor and as flow reactor with defined gas flow. The cell is attached to an ultrahigh-vacuum (UHV) system, which is used for sample preparation and also contains facilities for sample characterization. Specific for this design is the combination of a small cell volume, which allows kinetic measurements with high sensitivity under batch or continuous flow conditions, the complete isolation of the cell from the UHV part during UHV measurements, continuous temperature control during both UHV and high-pressure operation, and rapid transfer between UHV and high-pressure stage. Gas dosing is performed by a designed gas-handling system, which allows operation as flow reactor with calibrated gas flows at adjustable pressures. To study the kinetics of reactions on the model catalysts, a quadrupole mass spectrometer is connected to the high-pressure cell. IR measurements are possible in situ by polarization-modulation infrared reflection-absorption spectroscopy, which also allows measurements at elevated pressures. The performance of the setup is demonstrated by test measurements on the kinetics for CO oxidation and the CO adsorption on a Au /TiO2/Ru(0001) model catalyst film at 1-50 mbar total pressure.

Game theoretic analysis of physical protection system design

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

Canion, B.; Schneider, E.; Bickel, E.

The physical protection system (PPS) of a fictional small modular reactor (SMR) facility have been modeled as a platform for a game theoretic approach to security decision analysis. To demonstrate the game theoretic approach, a rational adversary with complete knowledge of the facility has been modeled attempting a sabotage attack. The adversary adjusts his decisions in response to investments made by the defender to enhance the security measures. This can lead to a conservative physical protection system design. Since defender upgrades were limited by a budget, cost benefit analysis may be conducted upon security upgrades. One approach to cost benefitmore » analysis is the efficient frontier, which depicts the reduction in expected consequence per incremental increase in the security budget.« less

Study of reactor Brayton power systems for nuclear electric spacecraft

NASA Technical Reports Server (NTRS)

1979-01-01

The feasibility of using Brayton power systems for nuclear electric spacecraft was investigated. The primary performance parameters of systems mass and radiator area were determined for systems from 100 to 1000 kW sub e. Mathematical models of all system components were used to determine masses and volumes. Two completely independent systems provide propulsion power so that no single-point failure can jeopardize a mission. The waste heat radiators utilize armored heat pipes to limit meteorite puncture. The armor thickness was statistically determined to achieve the required probability of survival. A 400 kW sub e reference system received primary attention as required by the contract. The components of this system were defined and a conceptual layout was developed with encouraging results. An arrangement with redundant Brayton power systems having a 1500 K (2240 F) turbine inlet temperature was shown to be compatible with the dimensions of the space shuttle orbiter payload bay.

Small Reactor for Deep Space Exploration

ScienceCinema

none,

2018-06-06

This is the first demonstration of a space nuclear reactor system to produce electricity in the United States since 1965, and an experiment demonstrated the first use of a heat pipe to cool a small nuclear reactor and then harvest the heat to power a Stirling engine at the Nevada National Security Site's Device Assembly Facility confirms basic nuclear reactor physics and heat transfer for a simple, reliable space power system.

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.

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

Concepts and Tests for the Remote-Controlled Dismantling of the Biological Shield and Form work of the KNK Reactor - 13425

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

Neff, Sylvia; Graf, Anja; Petrick, Holger

The compact sodium-cooled nuclear reactor facility Karlsruhe (KNK), a prototype Fast Breeder, is currently in an advanced stage of dismantling. Complete dismantling is based on 10 partial licensing steps. In the frame of the 9. decommissioning permit, which is currently ongoing, the dismantling of the biological shield is foreseen. The biological shield consists of heavy reinforced concrete with built-in steel fitments, such as form-work of the reactor tank, pipe sleeves, ventilation channels, and measuring devices. Due to the activation of the inner part of the biological shield, dismantling has to be done remote-controlled. During a comprehensive basic design phase amore » practical dismantling strategy was developed. Necessary equipment and tools were defined. Preliminary tests revealed that hot wire plasma cutting is the most favorable cutting technology due to the geometrical boundary conditions, the varying distance between cutter and material, and the heavy concrete behind the steel form-work. The cutting devices will be operated remotely via a carrier system with an industrial manipulator. The carrier system has expandable claws to adjust to the varying diameter of the reactor shaft during dismantling progress. For design approval of this prototype development, interaction between manipulator and hot wire plasma cutting was tested in a real configuration. For the demolition of the concrete structure, an excavator with appropriate tools, such as a hydraulic hammer, was selected. Other mechanical cutting devices, such as a grinder or rope saw, were eliminated because of concrete containing steel spheres added to increase the shielding factor of the heavy concrete. Dismantling of the biological shield will be done in a ring-wise manner due to static reasons. During the demolition process, the excavator is positioned on its tripod in three concrete recesses made prior to the dismantling of the separate concrete rings. The excavator and the manipulator carrier system will be operated alternately. Main boundary condition for all the newly designed equipment is the decommissioning housing of limited space within the reactor building containment. To allow for a continuous removal of the concrete rubble, an additional opening on the lowest level of the reactor shaft will be made. All equipment and the interaction of the tools have to be tested before use in the controlled area. Therefore a full-scale model of the biological shield will be provided in a mock-up. The tests will be performed in early 2014. The dismantling of the biological shield is scheduled for 2015. (authors)« less

System Analysis for Decay Heat Removal in Lead-Bismuth Cooled Natural Circulated Reactors

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

Takaaki Sakai; Yasuhiro Enuma; Takashi Iwasaki

2002-07-01

Decay heat removal analyses for lead-bismuth cooled natural circulation reactors are described in this paper. A combined multi-dimensional plant dynamics code (MSG-COPD) has been developed to conduct the system analysis for the natural circulation reactors. For the preliminary study, transient analysis has been performed for a 100 MWe lead-bismuth-cooled reactor designed by Argonne National Laboratory (ANL). In addition, decay heat removal characteristics of a 400 MWe lead-bismuth-cooled natural circulation reactor designed by Japan Nuclear Cycle Development Institute (JNC) has been evaluated by using MSG-COPD. PRACS (Primary Reactor Auxiliary Cooling System) is prepared for the JNC's concept to get sufficient heatmore » removal capacity. During 2000 sec after the transient, the outlet temperature shows increasing tendency up to the maximum temperature of 430 Centigrade, because the buoyancy force in a primary circulation path is temporary reduced. However, the natural circulation is recovered by the PRACS system and the out let temperature decreases successfully. (authors)« less

System Analysis for Decay Heat Removal in Lead-Bismuth-Cooled Natural-Circulation Reactors

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

Sakai, Takaaki; Enuma, Yasuhiro; Iwasaki, Takashi

2004-03-15

Decay heat removal analyses for lead-bismuth-cooled natural-circulation reactors are described in this paper. A combined multidimensional plant dynamics code (MSG-COPD) has been developed to conduct the system analysis for the natural-circulation reactors. For the preliminary study, transient analysis has been performed for a 300-MW(thermal) lead-bismuth-cooled reactor designed by Argonne National Laboratory. In addition, decay heat removal characteristics of a 400-MW(electric) lead-bismuth-cooled natural-circulation reactor designed by the Japan Nuclear Cycle Development Institute (JNC) has been evaluated by using MSG-COPD. The primary reactor auxiliary cooling system (PRACS) is prepared for the JNC concept to get sufficient heat removal capacity. During 2000 smore » after the transient, the outlet temperature shows increasing tendency up to the maximum temperature of 430 deg. C because the buoyancy force in a primary circulation path is temporarily reduced. However, the natural circulation is recovered by the PRACS system, and the outlet temperature decreases successfully.« less

Estimates of power requirements for a Manned Mars Rover powered by a nuclear reactor

NASA Technical Reports Server (NTRS)

Morley, Nicholas J.; El-Genk, Mohamed S.; Cataldo, Robert; Bloomfield, Harvey

1991-01-01

This paper assesses the power requirement for a Manned Mars Rover vehicle. Auxiliary power needs are fulfilled using a hybrid solar photovoltaic/regenerative fuel cell system, while the primary power needs are meet using an SP-100 type reactor. The primary electric power needs, which include 30-kW(e) net user power, depend on the reactor thermal power and the efficiency of the power conversion system. Results show that an SP-100 type reactor coupled to a Free Piston Stirling Engine yields the lowest total vehicle mass and lowest specific mass for the power system. The second lowest mass was for a SP-100 reactor coupled to a Closed Brayton Cycle using He/Xe as the working fluid. The specific mass of the nuclear reactor power system, including a man-rated radiation shield, ranged from 150-kg/kW(e) to 190-kg/KW(e) and the total mass of the Rover vehicle varied depend upon the cruising speed.

CHARACTERISTIC QUALITIES OF SOME ATOMIC POWER STATIONS (in Hungarian)

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

Ligeti, G.

1962-04-01

Mostly as the result of economic factors, the current rate of construction of public atomic power stations has slowed down. The use of atomic energy is considered economical only in a few special cases, such as ship propulsion or supplying power to remote regions. For this reason, many reactors were designed especially for the construction of such midget'' power stations, operating at power levels ranging from 10 to 70 Mw. Technical details are given of such already-built or proposed systems, including the following: pressurized- water reactors such as the Babcock and Wilcox 60-Mw reactor, using 2.4% U/sup 235/ fuel; themore » Humphrey-Glasow Company's 20 Mw reactor; the gascooled system of the de Havilland Company; the organicmoderated reactor of the English Electric Company; the organic-moderated system of the Hawker-Siddeley Nuclear Power Company; the boiling-water reactor of the Mitchell Engineering Company and the steam-cooled, heavy-water reactor of the Rolls-Royce & Vickers Company. (TTT)« less

Inherently Safe Fission Power System for Lunar Outposts

NASA Astrophysics Data System (ADS)

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

2013-09-01

This paper presents the Solid Core-Sectored Compact Reactor (SC-SCoRe) and power system for future lunar outposts. The power system nominally provides 38 kWe continuously for 21 years, employs static components and has no single point failures in reactor cooling or power generation. The reactor core has six sectors, each has a separate pair of primary and secondary loops with liquid NaK-56 working fluid, thermoelectric (TE) power conversion and heat-pipes radiator panels. The electromagnetic (EM) pumps in the primary and secondary loops, powered with separate TE power units, ensure operation reliability and passive decay heat removal from the reactor after shutdown. The reactor poses no radiological concerns during launch, and remains sufficiently subcritical, with the radial reflector dissembled, when submerged in wet sand and the core flooded with seawater, following a launch abort accident. After 300 years of storage below grade on the Moon, the total radioactivity in the post-operation reactor drops below 164 Ci, a low enough radioactivity for a recovery and safe handling of the reactor.

A Methodology for the Integration of a Mechanistic Source Term Analysis in a Probabilistic Framework for Advanced Reactors

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

Grabaskas, Dave; Brunett, Acacia J.; Bucknor, Matthew

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level, the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of PRA methodologies to conduct a mechanistic source term (MST) analysis for event sequences that could result in the release ofmore » radionuclides. The MST analysis seeks to realistically model and assess the transport, retention, and release of radionuclides from the reactor to the environment. The MST methods developed during this project seek to satisfy the requirements of the Mechanistic Source Term element of the ASME/ANS Non-LWR PRA standard. The MST methodology consists of separate analysis approaches for risk-significant and non-risk significant event sequences that may result in the release of radionuclides from the reactor. For risk-significant event sequences, the methodology focuses on a detailed assessment, using mechanistic models, of radionuclide release from the fuel, transport through and release from the primary system, transport in the containment, and finally release to the environment. The analysis approach for non-risk significant event sequences examines the possibility of large radionuclide releases due to events such as re-criticality or the complete loss of radionuclide barriers. This paper provides details on the MST methodology, including the interface between the MST analysis and other elements of the PRA, and provides a simplified example MST calculation for a sodium fast reactor.« less

High Solid Fed-batch Butanol Fermentation with Simultaneous Product Recovery: Part II - Process Integration.

PubMed

Qureshi, Nasib; Klasson, K Thomas; Saha, Badal C; Liu, Siqing

2018-04-25

In these studies liquid hot water (LHW) pretreated and enzymatically hydrolyzed Sweet Sorghum Bagasse (SSB) hydrolyzates were fermented in a fed-batch reactor. As reported in the preceding paper, the culture was not able to ferment the hydrolyzate I in a batch process due to presence of high level of toxic chemicals, in particular acetic acid released from SSB during the hydrolytic process. To be able to ferment the hydrolyzate I obtained from 250 gL -1 SSB hydrolysis, a fed-batch reactor with in-situ butanol recovery was devised. The process was started with the hydrolyzate II and when good cell growth and vigorous fermentation were observed, the hydrolyzate I was slowly fed to the reactor. In this manner the culture was able to ferment all the sugars present in both the hydrolyzates to acetone butanol ethanol (ABE). In a control batch reactor in which ABE was produced from glucose, ABE productivity and yield of 0.42 gL -1 h -1 and 0.36 were obtained, respectively. In the fed-batch reactor fed with SSB hydrolyzates these productivity and yield values were 0.44 gL -1 h -1 and 0.45, respectively. ABE yield in the integrated system was high due to utilization of acetic acid to convert to ABE. In summary we were able to utilize both the hydrolyzates obtained from LHW pretreated and enzymatically hydrolyzed SSB (250 gL -1 ) and convert them to ABE. Complete fermentation was possible due to simultaneous recovery of ABE by vacuum. This article is protected by copyright. All rights reserved. © 2018 American Institute of Chemical Engineers.

Passive cooling system for top entry liquid metal cooled nuclear reactors

DOEpatents

Boardman, Charles E.; Hunsbedt, Anstein; Hui, Marvin M.

1992-01-01

A liquid metal cooled nuclear fission reactor plant having a top entry loop joined satellite assembly with a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during shutdown, or heat produced during a mishap. This satellite type reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary cooling system when rendered inoperative.

Solving Problems With SINDA/FLUINT

NASA Technical Reports Server (NTRS)

2002-01-01

SINDA/FLUINT, the NASA standard software system for thermohydraulic analysis, provides computational simulation of interacting thermal and fluid effects in designs modeled as heat transfer and fluid flow networks. The product saves time and money by making the user's design process faster and easier, and allowing the user to gain a better understanding of complex systems. The code is completely extensible, allowing the user to choose the features, accuracy and approximation levels, and outputs. Users can also add their own customizations as needed to handle unique design tasks or to automate repetitive tasks. Applications for SINDA/FLUINT include the pharmaceutical, petrochemical, biomedical, electronics, and energy industries. The system has been used to simulate nuclear reactors, windshield wipers, and human windpipes. In the automotive industry, it simulates the transient liquid/vapor flows within air conditioning systems.

The results of the investigations of Russian Research Center - {open_quotes}Kurchatov Institute{close_quotes} on molten salt applications to problems of nuclear energy systems

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

Novikov, V.M.

1995-10-01

The results of investigations on molten salt (MS) applications to problems of nuclear energy systems that have been conducted in Russian Research {open_quotes}Kurchatov Institute{close_quotes} are presented and discussed. The spectrum of these investigations is rather broad and covers the following items: physical characteristics of molten salt nuclear energy systems (MSNES); nuclear and radiation safety of MSNES; construction materials compatible with MS of different compositions; technological aspects of MS loops; in-reactor loop testing. It is shown that main findings of completed program support the conclusion that there are no physical nor technological obstacles on way of MS application to different nuclearmore » energy systems.« less

76 FR 55718 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Materials...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-09-08

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Materials, Metallurgy & Reactor Fuels The ACRS Subcommittee on Materials, Metallurgy & Reactor...'' for reactor coolant system (RCS) components, as mentioned in 10 CFR 50 Appendix A, GDC-4. The...

Application of Box-Wilson experimental design method for 2,4-dinitrotoluene treatment in a sequential anaerobic migrating blanket reactor (AMBR)/aerobic completely stirred tank reactor (CSTR) system.

PubMed

Kuşçu, Özlem Selçuk; Sponza, Delia Teresa

2011-03-15

A sequential aerobic completely stirred tank reactor (CSTR) following the anaerobic migrating blanket reactor (AMBR) was used to treat a synthetic wastewater containing 2,4-dinitrotoluene (2,4-DNT). A Box-Wilson statistical experiment design was used to determine the effects of 2,4-DNT and the hydraulic retention times (HRTs) on 2,4-DNT and COD removal efficiencies in the AMBR reactor. The 2,4-DNT concentrations in the feed (0-280 mg/L) and the HRT (0.5-10 days) were considered as the independent variables while the 2,4-DNT and chemical oxygen demand (COD) removal efficiencies, total and methane gas productions, methane gas percentage, pH, total volatile fatty acid (TVFA) and total volatile fatty acid/bicarbonate alkalinity (TVFA/Bic.Alk.) ratio were considered as the objective functions in the Box-Wilson statistical experiment design in the AMBR. The predicted data for the parameters given above were determined from the response functions by regression analysis of the experimental data and exhibited excellent agreement with the experimental results. The optimum HRT which gave the maximum COD (97.00%) and 2,4-DNT removal (99.90%) efficiencies was between 5 and 10 days at influent 2,4-DNT concentrations 1-280 mg/L in the AMBR. The aerobic CSTR was used for removals of residual COD remaining from the AMBR, and for metabolites of 2,4-DNT. The maximum COD removal efficiency was 99% at an HRT of 1.89 days at a 2,4-DNT concentration of 239 mg/L in the aerobic CSTR. It was found that 280 mg/L 2,4-DNT transformed to 2,4-diaminotoluene (2,4-DAT) via 2-amino-4-nitrotoluene (2-A-4-NT) and 4-amino-2-nitrotoluene (4-A-2-NT) in the AMBR. The maximum 2,4-DAT removal was 82% at an HRT of 8.61 days in the aerobic CSTR. The maximum total COD and 2,4-DNT removal efficiencies were 99.00% and 99.99%, respectively, at an influent 2,4-DNT concentration of 239 mg/L and at 1.89 days of HRT in the sequential AMBR/CSTR. Copyright © 2011 Elsevier B.V. All rights reserved.

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

NASA Technical Reports Server (NTRS)

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

1992-01-01

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

Assessing pretreatment reactor scaling through empirical analysis

DOE PAGES

Lischeske, James J.; Crawford, Nathan C.; Kuhn, Erik; ...

2016-10-10

Pretreatment is a critical step in the biochemical conversion of lignocellulosic biomass to fuels and chemicals. Due to the complexity of the physicochemical transformations involved, predictively scaling up technology from bench- to pilot-scale is difficult. This study examines how pretreatment effectiveness under nominally similar reaction conditions is influenced by pretreatment reactor design and scale using four different pretreatment reaction systems ranging from a 3 g batch reactor to a 10 dry-ton/d continuous reactor. The reactor systems examined were an Automated Solvent Extractor (ASE), Steam Explosion Reactor (SER), ZipperClave(R) reactor (ZCR), and Large Continuous Horizontal-Screw Reactor (LHR). To our knowledge, thismore » is the first such study performed on pretreatment reactors across a range of reaction conditions (time and temperature) and at different reactor scales. The comparative pretreatment performance results obtained for each reactor system were used to develop response surface models for total xylose yield after pretreatment and total sugar yield after pretreatment followed by enzymatic hydrolysis. Near- and very-near-optimal regions were defined as the set of conditions that the model identified as producing yields within one and two standard deviations of the optimum yield. Optimal conditions identified in the smallest-scale system (the ASE) were within the near-optimal region of the largest scale reactor system evaluated. A reaction severity factor modeling approach was shown to inadequately describe the optimal conditions in the ASE, incorrectly identifying a large set of sub-optimal conditions (as defined by the RSM) as optimal. The maximum total sugar yields for the ASE and LHR were 95%, while 89% was the optimum observed in the ZipperClave. The optimum condition identified using the automated and less costly to operate ASE system was within the very-near-optimal space for the total xylose yield of both the ZCR and the LHR, and was within the near-optimal space for total sugar yield for the LHR. This indicates that the ASE is a good tool for cost effectively finding near-optimal conditions for operating pilot-scale systems, which may be used as starting points for further optimization. Additionally, using a severity-factor approach to optimization was found to be inadequate compared to a multivariate optimization method. As a result, the ASE and the LHR were able to enable significantly higher total sugar yields after enzymatic hydrolysis relative to the ZCR, despite having similar optimal conditions and total xylose yields. This underscores the importance of incorporating mechanical disruption into pretreatment reactor designs to achieve high enzymatic digestibilities.« less

Assessing pretreatment reactor scaling through empirical analysis

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

Lischeske, James J.; Crawford, Nathan C.; Kuhn, Erik

Pretreatment is a critical step in the biochemical conversion of lignocellulosic biomass to fuels and chemicals. Due to the complexity of the physicochemical transformations involved, predictively scaling up technology from bench- to pilot-scale is difficult. This study examines how pretreatment effectiveness under nominally similar reaction conditions is influenced by pretreatment reactor design and scale using four different pretreatment reaction systems ranging from a 3 g batch reactor to a 10 dry-ton/d continuous reactor. The reactor systems examined were an Automated Solvent Extractor (ASE), Steam Explosion Reactor (SER), ZipperClave(R) reactor (ZCR), and Large Continuous Horizontal-Screw Reactor (LHR). To our knowledge, thismore » is the first such study performed on pretreatment reactors across a range of reaction conditions (time and temperature) and at different reactor scales. The comparative pretreatment performance results obtained for each reactor system were used to develop response surface models for total xylose yield after pretreatment and total sugar yield after pretreatment followed by enzymatic hydrolysis. Near- and very-near-optimal regions were defined as the set of conditions that the model identified as producing yields within one and two standard deviations of the optimum yield. Optimal conditions identified in the smallest-scale system (the ASE) were within the near-optimal region of the largest scale reactor system evaluated. A reaction severity factor modeling approach was shown to inadequately describe the optimal conditions in the ASE, incorrectly identifying a large set of sub-optimal conditions (as defined by the RSM) as optimal. The maximum total sugar yields for the ASE and LHR were 95%, while 89% was the optimum observed in the ZipperClave. The optimum condition identified using the automated and less costly to operate ASE system was within the very-near-optimal space for the total xylose yield of both the ZCR and the LHR, and was within the near-optimal space for total sugar yield for the LHR. This indicates that the ASE is a good tool for cost effectively finding near-optimal conditions for operating pilot-scale systems, which may be used as starting points for further optimization. Additionally, using a severity-factor approach to optimization was found to be inadequate compared to a multivariate optimization method. As a result, the ASE and the LHR were able to enable significantly higher total sugar yields after enzymatic hydrolysis relative to the ZCR, despite having similar optimal conditions and total xylose yields. This underscores the importance of incorporating mechanical disruption into pretreatment reactor designs to achieve high enzymatic digestibilities.« less

When Do Commercial Reactors Permanently Shut Down?

EIA Publications

2011-01-01

For those wishing to obtain current data, the following resources are available: U.S. reactors, go to the Energy Information Administration's nuclear reactor shutdown list. (Note: As of April 30, 2010, the last U.S. reactor to permanently shut down was Big Rock Point in 1997.) Foreign Reactors, go to the Power Reactor Information System (PRIS) on the International Atomic Energy Agency's website.

Implementation of New Reactivity Measurement System and New Reactor Noise Analysis Equipment in a VVER-440 Nuclear Power Plant

NASA Astrophysics Data System (ADS)

Vegh, János; Kiss, Sándor; Lipcsei, Sándor; Horvath, Csaba; Pos, István; Kiss, Gábor

2010-10-01

The paper deals with two recently developed, high-precision nuclear measurement systems installed at the VVER-440 units of the Hungarian Paks NPP. Both developments were motivated by the reactor power increase to 108%, and by the planned plant service time extension. The first part describes the RMR start-up reactivity measurement system with advanced services. High-precision picoampere meters were installed at each reactor unit and measured ionization chamber current signals are handled by a portable computer providing data acquisition and online reactivity calculation service. Detailed offline evaluation and analysis of reactor start-up measurements can be performed on the portable unit, too. The second part of the paper describes a new reactor noise diagnostics system using state-of-the-art data acquisition hardware and signal processing methods. Details of the new reactor noise measurement evaluation software are also outlined. Noise diagnostics at Paks NPP is a standard tool for core anomaly detection and for long-term noise trend monitoring. Regular application of these systems is illustrated by real plant data, e.g., results of standard reactivity measurements during a reactor startup session are given. Noise applications are also illustrated by real plant measurements; results of core anomaly detection are presented.