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

Expert systems for fault diagnosis in nuclear reactor control

NASA Astrophysics Data System (ADS)

Jalel, N. A.; Nicholson, H.

1990-11-01

An expert system for accident analysis and fault diagnosis for the Loss Of Fluid Test (LOFT) reactor, a small scale pressurized water reactor, was developed for a personal computer. The knowledge of the system is presented using a production rule approach with a backward chaining inference engine. The data base of the system includes simulated dependent state variables of the LOFT reactor model. Another system is designed to assist the operator in choosing the appropriate cooling mode and to diagnose the fault in the selected cooling system. The response tree, which is used to provide the link between a list of very specific accident sequences and a set of generic emergency procedures which help the operator in monitoring system status, and to differentiate between different accident sequences and select the correct procedures, is used to build the system knowledge base. Both systems are written in TURBO PROLOG language and can be run on an IBM PC compatible with 640k RAM, 40 Mbyte hard disk and color graphics.

A Basic LEGO Reactor Design for the Provision of Lunar Surface Power

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

John Darrell Bess

2008-06-01

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched with lunar shipments from Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, suchmore » as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides approximately 5 kWe. The overall envelope for a single subunit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. Six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network, subject to lunar base power demand. Improvements in reactor control methods, fuel form and matrix, shielding, as well as power conversion and heat rejection techniques can help generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for use on other extraterrestrial surfaces.« less

Small reactor power systems for manned planetary surface bases

NASA Technical Reports Server (NTRS)

Bloomfield, Harvey S.

1987-01-01

A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options.

Reactor protection system with automatic self-testing and diagnostic

DOEpatents

Gaubatz, Donald C.

1996-01-01

A reactor protection system having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically "identical" values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic.

Reactor protection system with automatic self-testing and diagnostic

DOEpatents

Gaubatz, D.C.

1996-12-17

A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically ``identical`` values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic. 16 figs.

A Performance-Based Training Qualification Guide/Checklist Developed for Reactor Operators at the High Flux Beam Reactor at Brookhaven National Laboratory.

ERIC Educational Resources Information Center

McNair, Robert C.

A Performance-Based Training (PBT) Qualification Guide/Checklist was developed that would enable a trainee to attain the skills, knowledge, and attitude required to operate the High Flux Beam Reactor at Brookhaven National Laboratory. Design of this guide/checklist was based on the Instructional System Design Model. The needs analysis identified…

Hydrodynamic study of an internal airlift reactor for microalgae culture.

PubMed

Rengel, Ana; Zoughaib, Assaad; Dron, Dominique; Clodic, Denis

2012-01-01

Internal airlift reactors are closed systems considered today for microalgae cultivation. Several works have studied their hydrodynamics but based on important solid concentrations, not with biomass concentrations usually found in microalgae cultures. In this study, an internal airlift reactor has been built and tested in order to clarify the hydrodynamics of this system, based on microalgae typical concentrations. A model is proposed taking into account the variation of air bubble velocity according to volumetric air flow rate injected into the system. A relationship between riser and downcomer gas holdups is established, which varied slightly with solids concentrations. The repartition of solids along the reactor resulted to be homogenous for the range of concentrations and volumetric air flow rate studied here. Liquid velocities increase with volumetric air flow rate, and they vary slightly when solids are added to the system. Finally, liquid circulation time found in each section of the reactor is in concordance with those employed in microalgae culture.

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

NASA Technical Reports Server (NTRS)

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

1991-01-01

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

78 FR 71675 - Update of the Office of Nuclear Reactor Regulation's Electronic Operating Reactor Correspondence

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-29

... correspondence to addressees and subscribers through a computer-based email distribution system. Since then, the... Electronic Operating Reactor Correspondence The U.S. Nuclear Regulatory Commission (NRC) is issuing this... available operating reactor licensing correspondence, effective December 9, 2013. Official agency records...

Update on Small Modular Reactors Dynamic System Modeling Tool: Web Application

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

Hale, Richard Edward; Cetiner, Sacit M.; Fugate, David L.

Previous reports focused on the development of component and system models as well as end-to-end system models using Modelica and Dymola for two advanced reactor architectures: (1) Advanced Liquid Metal Reactor and (2) fluoride high-temperature reactor (FHR). The focus of this report is the release of the first beta version of the web-based application for model use and collaboration, as well as an update on the FHR model. The web-based application allows novice users to configure end-to-end system models from preconfigured choices to investigate the instrumentation and controls implications of these designs and allows for the collaborative development of individualmore » component models that can be benchmarked against test systems for potential inclusion in the model library. A description of this application is provided along with examples of its use and a listing and discussion of all the models that currently exist in the library.« less

Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

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

Bucknor, Matthew D.; Grabaskas, David; Brunett, Acacia J.

2016-01-01

Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended due to deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologiesmore » for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Centering on an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive reactor cavity cooling system following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. While this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability for the reactor cavity cooling system (and the reactor system in general) to the postulated transient event.« less

Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

DOE PAGES

Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; ...

2017-01-24

We report that many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has beenmore » examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Lastly, although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.« less

Lunar in-core thermionic nuclear reactor power system conceptual design

NASA Technical Reports Server (NTRS)

Mason, Lee S.; Schmitz, Paul C.; Gallup, Donald R.

1991-01-01

This paper presents a conceptual design of a lunar in-core thermionic reactor power system. The concept consists of a thermionic reactor located in a lunar excavation with surface mounted waste heat radiators. The system was integrated with a proposed lunar base concept representative of recent NASA Space Exploration Initiative studies. The reference mission is a permanently-inhabited lunar base requiring a 550 kWe, 7 year life central power station. Performance parameters and assumptions were based on the Thermionic Fuel Element (TFE) Verification Program. Five design cases were analyzed ranging from conservative to advanced. The cases were selected to provide sensitivity effects on the achievement of TFE program goals.

An MFC-Based Online Monitoring and Alert System for Activated Sludge Process

PubMed Central

Xu, Gui-Hua; Wang, Yun-Kun; Sheng, Guo-Ping; Mu, Yang; Yu, Han-Qing

2014-01-01

In this study, based on a simple, compact and submersible microbial fuel cell (MFC), a novel online monitoring and alert system with self-diagnosis function was established for the activated sludge (AS) process. Such a submersible MFC utilized organic substrates and oxygen in the AS reactor as the electron donor and acceptor respectively, and could provide an evaluation on the status of the AS reactor and thus give a reliable early warning of potential risks. In order to evaluate the reliability and sensitivity of this online monitoring and alert system, a series of tests were conducted to examine the response of this system to various shocks imposed on the AS reactor. The results indicate that this online monitoring and alert system was highly sensitive to the performance variations of the AS reactor. The stability, sensitivity and repeatability of this online system provide feasibility of being incorporated into current control systems of wastewater treatment plants to real-time monitor, diagnose, alert and control the AS process. PMID:25345502

Multi-phase CFD modeling of solid sorbent carbon capture system

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

Ryan, E. M.; DeCroix, D.; Breault, R.

2013-07-01

Computational fluid dynamics (CFD) simulations are used to investigate a low temperature post-combustion carbon capture reactor. The CFD models are based on a small scale solid sorbent carbon capture reactor design from ADA-ES and Southern Company. The reactor is a fluidized bed design based on a silica-supported amine sorbent. CFD models using both Eulerian–Eulerian and Eulerian–Lagrangian multi-phase modeling methods are developed to investigate the hydrodynamics and adsorption of carbon dioxide in the reactor. Models developed in both FLUENT® and BARRACUDA are presented to explore the strengths and weaknesses of state of the art CFD codes for modeling multi-phase carbon capturemore » reactors. The results of the simulations show that the FLUENT® Eulerian–Lagrangian simulations (DDPM) are unstable for the given reactor design; while the BARRACUDA Eulerian–Lagrangian model is able to simulate the system given appropriate simplifying assumptions. FLUENT® Eulerian–Eulerian simulations also provide a stable solution for the carbon capture reactor given the appropriate simplifying assumptions.« less

Multi-Phase CFD Modeling of Solid Sorbent Carbon Capture System

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

Ryan, Emily M.; DeCroix, David; Breault, Ronald W.

2013-07-30

Computational fluid dynamics (CFD) simulations are used to investigate a low temperature post-combustion carbon capture reactor. The CFD models are based on a small scale solid sorbent carbon capture reactor design from ADA-ES and Southern Company. The reactor is a fluidized bed design based on a silica-supported amine sorbent. CFD models using both Eulerian-Eulerian and Eulerian-Lagrangian multi-phase modeling methods are developed to investigate the hydrodynamics and adsorption of carbon dioxide in the reactor. Models developed in both FLUENT® and BARRACUDA are presented to explore the strengths and weaknesses of state of the art CFD codes for modeling multi-phase carbon capturemore » reactors. The results of the simulations show that the FLUENT® Eulerian-Lagrangian simulations (DDPM) are unstable for the given reactor design; while the BARRACUDA Eulerian-Lagrangian model is able to simulate the system given appropriate simplifying assumptions. FLUENT® Eulerian-Eulerian simulations also provide a stable solution for the carbon capture reactor given the appropriate simplifying assumptions.« 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

An experimental study of ammonia borane based hydrogen storage systems

NASA Astrophysics Data System (ADS)

Deshpande, Kedaresh A.

2011-12-01

Hydrogen is a promising fuel for the future, capable of meeting the demands of energy storage and low pollutant emission. Chemical hydrides are potential candidates for chemical hydrogen storage, especially for automobile applications. Ammonia borane (AB) is a chemical hydride being investigated widely for its potential to realize the hydrogen economy. In this work, the yield of hydrogen obtained during neat AB thermolysis was quantified using two reactor systems. First, an oil bath heated glass reactor system was used with AB batches of 0.13 gram (+/- 0.001 gram). The rates of hydrogen generation were measured. Based on these experimental data, an electrically heated steel reactor system was designed and constructed to handle up to 2 grams of AB per batch. A majority of components were made of stainless-steel. The system consisted of an AB reservoir and feeder, a heated reactor, a gas processing unit and a system control and monitoring unit. An electronic data acquisition system was used to record experimental data. The performance of the steel reactor system was evaluated experimentally through batch reactions of 30 minutes each, for reaction temperatures in the range from 373 K to 430 K. The experimental data showed exothermic decomposition of AB accompanied by rapid generation of hydrogen during the initial period of the reaction. 90% of the hydrogen was generated during the initial 120 seconds after addition of AB to the reactor. At 430 K, the reaction produced 12 wt.% of hydrogen. The heat diffusion in the reactor system and the process of exothermic decomposition of AB were coupled in a two-dimensional model. Neat AB thermolysis was modeled as a global first order reactions based on Arrhenius theory. The values of equation constants were derived from curve fit of experimental data. The pre-exponential constant and the activation energy were estimated to be 4 s-1 (+/- 0.4 s-1) and 13000 J mol -1 s-1 (+/- 1050 J mol-1 s -1) respectively. The model was solved in COMSOL Multiphysics. The model was capable of simulating the transient response of the system and captured the observed trends such as the decrease in reactor temperature upon addition of AB and exothermic decomposition.

A Review of Gas-Cooled Reactor Concepts for SDI Applications

DTIC Science & Technology

1989-08-01

710 program .) Wire- Core Reactor (proposed by Rockwell). The wire- core reactor utilizes thin fuel wires woven between spacer wires to form an open...reactor is based on results of developmental studies of nuclear rocket propulsion systems. The reactor core is made up of annular fuel assemblies of...XE Addendum to Volume II. NERVA Fuel Development , Westinghouse Astronuclear Laboratory, TNR-230, July 15’ 1972. J I8- Rover Program Reactor Tests

Telescope-based cavity for negative ion beam neutralization in future fusion reactors.

PubMed

Fiorucci, Donatella; Hreibi, Ali; Chaibi, Walid

2018-03-01

In future fusion reactors, heating system efficiency is of the utmost importance. Photo-neutralization substantially increases the neutral beam injector (NBI) efficiency with respect to the foreseen system in the International Thermonuclear Experimental Reactor (ITER) based on a gaseous target. In this paper, we propose a telescope-based configuration to be used in the NBI photo-neutralizer cavity of the demonstration power plant (DEMO) project. This configuration greatly reduces the total length of the cavity, which likely solves overcrowding issues in a fusion reactor environment. Brought to a tabletop experiment, this cavity configuration is tested: a 4 mm beam width is obtained within a ≃1.5  m length cavity. The equivalent cavity g factor is measured to be 0.038(3), thus confirming the cavity stability.

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.

Reactor transient control in support of PFR/TREAT TUCOP experiments

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

Burrows, D.R.; Larsen, G.R.; Harrison, L.J.

1984-01-01

Unique energy deposition and experiment control requirements posed bythe PFR/TREAT series of transient undercooling/overpower (TUCOP) experiments resulted in equally unique TREAT reactor operations. New reactor control computer algorithms were written and used with the TREAT reactor control computer system to perform such functions as early power burst generation (based on test train flow conditions), burst generation produced by a step insertion of reactivity following a controlled power ramp, and shutdown (SCRAM) initiators based on both test train conditions and energy deposition. Specialized hardware was constructed to simulate test train inputs to the control computer system so that computer algorithms couldmore » be tested in real time without irradiating the experiment.« less

Analysis of reactor trips originating in balance of plant systems

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

Stetson, F.T.; Gallagher, D.W.; Le, P.T.

1990-09-01

This report documents the results of an analysis of balance-of-plant (BOP) related reactor trips at commercial US nuclear power plants of a 5-year period, from January 1, 1984, through December 31, 1988. The study was performed for the Plant Systems Branch, Office of Nuclear Reactor Regulation, US Nuclear Regulatory Commission. The objectives of the study were: to improve the level of understanding of BOP-related challenges to safety systems by identifying and categorizing such events; to prepare a computerized data base of BOP-related reactor trip events and use the data base to identify trends and patterns in the population of thesemore » events; to investigate the risk implications of BOP events that challenge safety systems; and to provide recommendations on how to address BOP-related concerns in regulatory context. 18 refs., 2 figs., 27 tabs.« less

Oxygen transport membrane reactor based method and system for generating electric power

DOEpatents

Kelly, Sean M.; Chakravarti, Shrikar; Li, Juan

2017-02-07

A carbon capture enabled system and method for generating electric power and/or fuel from methane containing sources using oxygen transport membranes by first converting the methane containing feed gas into a high pressure synthesis gas. Then, in one configuration the synthesis gas is combusted in oxy-combustion mode in oxygen transport membranes based boiler reactor operating at a pressure at least twice that of ambient pressure and the heat generated heats steam in thermally coupled steam generation tubes within the boiler reactor; the steam is expanded in steam turbine to generate power; and the carbon dioxide rich effluent leaving the boiler reactor is processed to isolate carbon. In another configuration the synthesis gas is further treated in a gas conditioning system configured for carbon capture in a pre-combustion mode using water gas shift reactors and acid gas removal units to produce hydrogen or hydrogen-rich fuel gas that fuels an integrated gas turbine and steam turbine system to generate power. The disclosed method and system can also be adapted to integrate with coal gasification systems to produce power from both coal and methane containing sources with greater than 90% carbon isolation.

Validation of light water reactor calculation methods and JEF-1-based data libraries by TRX and BAPL critical experiments

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

Paratte, J.M.; Pelloni, S.; Grimm, P.

1991-04-01

This paper analyzes the capability of various code systems and JEF-1-based nuclear data libraries to compute light water reactor lattices by comparing calculations with results from thermal reactor benchmark experiments TRX and BAPL and with previously published values. With the JEF-1 evaluation, eigenvalues are generally well predicted within 8 mk (1 mk = 0.001) or less by all code systems, and all methods give reasonable results for the measured reaction rate ratios within, or not too far from, the experimental uncertainty.

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.

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.

Research gaps and technology needs in development of PHM for passive AdvSMR components

NASA Astrophysics Data System (ADS)

Meyer, Ryan M.; Ramuhalli, Pradeep; Coble, Jamie B.; Hirt, Evelyn H.; Mitchell, Mark R.; Wootan, David W.; Berglin, Eric J.; Bond, Leonard J.; Henagar, Chuck H., Jr.

2014-02-01

Advanced small modular reactors (AdvSMRs), which are based on modularization of advanced reactor concepts, may provide a longer-term alternative to traditional light-water reactors and near-term small modular reactors (SMRs), which are based on integral pressurized water reactor (iPWR) concepts. SMRs are challenged economically because of losses in economy of scale; thus, there is increased motivation to reduce the controllable operations and maintenance costs through automation technologies including prognostics health management (PHM) systems. In this regard, PHM systems have the potential to play a vital role in supporting the deployment of AdvSMRs and face several unique challenges with respect to implementation for passive AdvSMR components. This paper presents a summary of a research gaps and technical needs assessment performed for implementation of PHM for passive AdvSMR components.

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.

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

Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.

We report that many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has beenmore » examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Lastly, although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.« less

The use of experimental data in an MTR-type nuclear reactor safety analysis

NASA Astrophysics Data System (ADS)

Day, Simon E.

Reactivity initiated accidents (RIAs) are a category of events required for research reactor safety analysis. A subset of this is unprotected RIAs in which mechanical systems or human intervention are not credited in the response of the system. Light-water cooled and moderated MTR-type ( i.e., aluminum-clad uranium plate fuel) reactors are self-limiting up to some reactivity insertion limit beyond which fuel damage occurs. This characteristic was studied in the Borax and Spert reactor tests of the 1950s and 1960s in the USA. This thesis considers the use of this experimental data in generic MTR-type reactor safety analysis. The approach presented herein is based on fundamental phenomenological understanding and uses correlations in the reactor test data with suitable account taken for differences in important system parameters. Specifically, a semi-empirical approach is used to quantify the relationship between the power, energy and temperature rise response of the system as well as parametric dependencies on void coefficient and the degree of subcooling. Secondary effects including the dependence on coolant flow are also examined. A rigorous curve fitting approach and error assessment is used to quantify the trends in the experimental data. In addition to the initial power burst stage of an unprotected transient, the longer term stability of the system is considered with a stylized treatment of characteristic power/temperature oscillations (chugging). A bridge from the HEU-based experimental data to the LEU fuel cycle is assessed and outlined based on existing simulation results presented in the literature. A cell-model based parametric study is included. The results are used to construct a practical safety analysis methodology for determining reactivity insertion safety limits for a light-water moderated and cooled MTR-type core.

Microchannel Reactor System for Catalytic Hydrogenation

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

Adeniyi Lawal; Woo Lee; Ron Besser

2010-12-22

We successfully demonstrated a novel process intensification concept enabled by the development of microchannel reactors, for energy efficient catalytic hydrogenation reactions at moderate temperature, and pressure, and low solvent levels. We designed, fabricated, evaluated, and optimized a laboratory-scale microchannel reactor system for hydrogenation of onitroanisole and a proprietary BMS molecule. In the second phase of the program, as a prelude to full-scale commercialization, we designed and developed a fully-automated skid-mounted multichannel microreactor pilot plant system for multiphase reactions. The system is capable of processing 1 – 10 kg/h of liquid substrate, and an industrially relevant immiscible liquid-liquid was successfully demonstratedmore » on the system. Our microreactor-based pilot plant is one-of-akind. We anticipate that this process intensification concept, if successfully demonstrated, will provide a paradigm-changing basis for replacing existing energy inefficient, cost ineffective, environmentally detrimental slurry semi-batch reactor-based manufacturing practiced in the pharmaceutical and fine chemicals industries.« less

ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

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

E. Blanford; E. Keldrauk; M. Laufer

2010-09-20

Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement,more » and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction using factory prefabricated structural modules, for application to external event shell and base isolated structures.« less

Expert system for surveillance and diagnosis of breach fuel elements

DOEpatents

Gross, K.C.

1988-01-21

An apparatus and method are disclosed for surveillance and diagnosis of breached fuel elements in a nuclear reactor. A delayed neutron monitoring system provides output signals indicating the delayed neutron activity and age and the equivalent recoil area of a breached fuel element. Sensors are used to provide outputs indicating the status of each component of the delayed neutron monitoring system. Detectors also generate output signals indicating the reactor power level and the primary coolant flow rate of the reactor. The outputs from the detectors and sensors are interfaced with an artificial intelligence-based knowledge system which implements predetermined logic and generates output signals indicating the operability of the reactor. 2 figs.

Expert system for surveillance and diagnosis of breach fuel elements

DOEpatents

Gross, Kenny C.

1989-01-01

An apparatus and method are disclosed for surveillance and diagnosis of breached fuel elements in a nuclear reactor. A delayed neutron monitoring system provides output signals indicating the delayed neutron activity and age and the equivalent recoil areas of a breached fuel element. Sensors are used to provide outputs indicating the status of each component of the delayed neutron monitoring system. Detectors also generate output signals indicating the reactor power level and the primary coolant flow rate of the reactor. The outputs from the detectors and sensors are interfaced with an artificial intelligence-based knowledge system which implements predetermined logic and generates output signals indicating the operability of the reactor.

Application of the Enabler to nuclear electric propulsion

NASA Astrophysics Data System (ADS)

Pierce, Bill L.

This paper describes a power system concept that provides the electric power for a baseline electric propulsion system for a piloted mission to Mars. A 10-MWe space power system is formed by coupling an Enabler reactor with a simple non-recuperated closed Brayton cycle. The Enabler reactor is a gas-cooled reactor based on proven reactor technology developed under the NERVA/Rover programs. The selected power cycle, which uses a helium-xenon mixture at 1920 K at the turbine inlet, is diagramed and described. The specific mass of the power system over the power range from 5 to 70 MWe is given. The impact of operating life on the specific mass of a 10-MWe system is also shown.

Real-time LMR control parameter generation using advanced adaptive synthesis

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

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

1990-01-01

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

Dynamic biogas upgrading based on the Sabatier process: thermodynamic and dynamic process simulation.

PubMed

Jürgensen, Lars; Ehimen, Ehiaze Augustine; Born, Jens; Holm-Nielsen, Jens Bo

2015-02-01

This study aimed to investigate the feasibility of substitute natural gas (SNG) generation using biogas from anaerobic digestion and hydrogen from renewable energy systems. Using thermodynamic equilibrium analysis, kinetic reactor modeling and transient simulation, an integrated approach for the operation of a biogas-based Sabatier process was put forward, which was then verified using a lab scale heterogenous methanation reactor. The process simulation using a kinetic reactor model demonstrated the feasibility of the production of SNG at gas grid standards using a single reactor setup. The Wobbe index, CO2 content and calorific value were found to be controllable by the H2/CO2 ratio fed the methanation reactor. An optimal H2/CO2 ratio of 3.45-3.7 was seen to result in a product gas with high calorific value and Wobbe index. The dynamic reactor simulation verified that the process start-up was feasible within several minutes to facilitate surplus electricity use from renewable energy systems. Copyright © 2014 Elsevier Ltd. All rights reserved.

Nuclear reactor power as applied to a space-based radar mission

NASA Technical Reports Server (NTRS)

Jaffe, L.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Fujita, T.; Grossman, M.; Bloomfield, H.; Heller, J.

1988-01-01

A space-based radar mission and spacecraft are examined to determine system requirements for a 300 kWe space nuclear reactor power system. The spacecraft configuration and its orbit, launch vehicle, and propulsion are described. Mission profiles are addressed, and storage in assembly orbit is considered. Dynamics and attitude control and the problems of nuclear and thermal radiation are examined.

Design of virtual SCADA simulation system for pressurized water reactor

NASA Astrophysics Data System (ADS)

Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

2016-02-01

The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

Sodium Based Heat Pipe Modules for Space Reactor Concepts: Stainless Steel SAFE-100 Core

NASA Technical Reports Server (NTRS)

Martin, James J.; Reid, Robert S.

2004-01-01

A heat pipe cooled reactor is one of several candidate reactor cores being considered for advanced space power and propulsion systems to support future space exploration applications. Long life heat pipe modules, with designs verified through a combination of theoretical analysis and experimental lifetime evaluations, would be necessary to establish the viability of any of these candidates, including the heat pipe reactor option. A hardware-based program was initiated to establish the infrastructure necessary to build heat pipe modules. This effort, initiated by Los Alamos National Laboratory and referred to as the Safe Affordable Fission Engine (SAFE) project, set out to fabricate and perform non-nuclear testing on a modular heat pipe reactor prototype that can provide 100 kilowatt from the core to an energy conversion system at 700 C. Prototypic heat pipe hardware was designed, fabricated, filled, closed-out and acceptance tested.

10 CFR 50.48 - Fire protection.

Code of Federal Regulations, 2011 CFR

2011-01-01

... suppression systems; and (iii) The means to limit fire damage to structures, systems, or components important...) Standard 805, “Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating... pressurized-water reactors (PWRs) is not permitted. (iv) Uncertainty analysis. An uncertainty analysis...

10 CFR 50.48 - Fire protection.

Code of Federal Regulations, 2010 CFR

2010-01-01

... suppression systems; and (iii) The means to limit fire damage to structures, systems, or components important...) Standard 805, “Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating... pressurized-water reactors (PWRs) is not permitted. (iv) Uncertainty analysis. An uncertainty analysis...

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.

RELAP-7 Software Verification and Validation Plan

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

Smith, Curtis L.; Choi, Yong-Joon; Zou, Ling

This INL plan comprehensively describes the software for RELAP-7 and documents the software, interface, and software design requirements for the application. The plan also describes the testing-based software verification and validation (SV&V) process—a set of specially designed software models used to test RELAP-7. The RELAP-7 (Reactor Excursion and Leak Analysis Program) code is a nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework – MOOSE (Multi-Physics Object-Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty yearsmore » of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s capability and extends the analysis capability for all reactor system simulation scenarios.« less

Liquid fuel molten salt reactors for thorium utilization

DOE PAGES

Gehin, Jess C.; Powers, Jeffrey J.

2016-04-08

Molten salt reactors (MSRs) represent a class of reactors that use liquid salt, usually fluoride- or chloride-based, as either a coolant with a solid fuel (such as fluoride salt-cooled high temperature reactors) or as a combined coolant and fuel with fuel dissolved in a carrier salt. For liquid-fuelled MSRs, the salt can be processed online or in a batch mode to allow for removal of fission products as well as introduction of fissile fuel and fertile materials during reactor operation. The MSR is most commonly associated with the 233U/thorium fuel cycle, as the nuclear properties of 233U combined with themore » online removal of parasitic absorbers allow for the ability to design a thermal-spectrum breeder reactor; however, MSR concepts have been developed using all neutron energy spectra (thermal, intermediate, fast, and mixed-spectrum zoned concepts) and with a variety of fuels including uranium, thorium, plutonium, and minor actinides. Early MSR work was supported by a significant research and development (R&D) program that resulted in two experimental systems operating at ORNL in the 1960s, the Aircraft Reactor Experiment and the Molten Salt Reactor Experiment. Subsequent design studies in the 1970s focusing on thermal-spectrum thorium-fueled systems established reference concepts for two major design variants: (1) a molten salt breeder reactor (MSBR), with multiple configurations that could breed additional fissile material or maintain self-sustaining operation; and (2) a denatured molten salt reactor (DMSR) with enhanced proliferation-resistance. T MSRs has been selected as one of six most promising Generation IV systems and development activities have been seen in fast-spectrum MSRs, waste-burning MSRs, MSRs fueled with low-enriched uranium (LEU), as well as more traditional thorium fuel cycle-based MSRs. This study provides an historical background of MSR R&D efforts, surveys and summarizes many of the recent development, and provides analysis comparing thorium-based MSRs.« less

Digital computer operation of a nuclear reactor

DOEpatents

Colley, R.W.

1982-06-29

A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

Digital computer operation of a nuclear reactor

DOEpatents

Colley, Robert W.

1984-01-01

A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

Safeguards Challenges for Pebble-Bed Reactors (PBRs):Peoples Republic of China (PRC)

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

Forsberg, Charles W.; Moses, David Lewis

2009-11-01

The Peoples Republic of China (PRC) is operating the HTR-10 pebble-bed reactor (PBR) and is in the process of building a prototype PBR plant with two modular reactors (250-MW(t) per reactor) feeding steam to a single turbine-generator. It is likely to be the first modular hightemperature reactor to be ready for commercial deployment in the world because it is a highpriority project for the PRC. The plant design features multiple modular reactors feeding steam to a single turbine generator where the number of modules determines the plant output. The design and commercialization strategy are based on PRC strengths: (1) amore » rapidly growing electric market that will support low-cost mass production of modular reactor units and (2) a balance of plant system based on economics of scale that uses the same mass-produced turbine-generator systems used in PRC coal plants. If successful, in addition to supplying the PRC market, this strategy could enable China to be the leading exporter of nuclear reactors to developing countries. The modular characteristics of the reactor match much of the need elsewhere in the world. PBRs have major safety advantages and a radically different fuel. The fuel, not the plant systems, is the primary safety system to prevent and mitigate the release of radionuclides under accident conditions. The fuel consists of small (6-cm) pebbles (spheres) containing coatedparticle fuel in a graphitized carbon matrix. The fuel loading per pebble is small (~9 grams of low-enriched uranium) and hundreds of thousands of pebbles are required to fuel a nuclear plant. The uranium concentration in the fuel is an order of magnitude less than in traditional nuclear fuels. These characteristics make the fuel significantly less attractive for illicit use (weapons production or dirty bomb); but, its unusual physical form may require changes in the tools used for safeguards. This report describes PBRs, what is different, and the safeguards challenges. A series of safeguards recommendations are made based on the assumption that the reactor is successfully commercialized and is widely deployed.« less

IECEC '83; Proceedings of the Eighteenth Intersociety Energy Conversion Engineering Conference, Orlando, FL, August 21-26, 1983. Volume 1 - Thermal energy systems

NASA Astrophysics Data System (ADS)

Among the topics discussed are the nuclear fuel cycle, advanced nuclear reactor designs, developments in central status power reactors, space nuclear reactors, magnetohydrodynamic devices, thermionic devices, thermoelectric devices, geothermal systems, solar thermal energy conversion systems, ocean thermal energy conversion (OTEC) developments, and advanced energy conversion concepts. Among the specific questions covered under these topic headings are a design concept for an advanced light water breeder reactor, energy conversion in MW-sized space power systems, directionally solidified cermet electrodes for thermionic energy converters, boron-based high temperature thermoelectric materials, geothermal energy commercialization, solar Stirling cycle power conversion, and OTEC production of methanol. For individual items see A84-30027 to A84-30055

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.

Systems aspects of a space nuclear reactor power system

NASA Technical Reports Server (NTRS)

Jaffe, L.; Fujita, T.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Grossman, M.; Bloomfield, H.; Heller, J.

1988-01-01

Various system aspects of a 300-kW nuclear reactor power system for spacecraft have been investigated. Special attention is given to the cases of a reusable OTV and a space-based radar. It is demonstrated that the stowed length of the power system is important to mission design, and that orbital storage for months to years may be needed for missions involving orbital assembly.

Design of virtual SCADA simulation system for pressurized water reactor

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

Wijaksono, Umar, E-mail: umar.wijaksono@student.upi.edu; Abdullah, Ade Gafar; Hakim, Dadang Lukman

The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles ofmore » energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.« less

Performance assessment of conventional and base-isolated nuclear power plants for earthquake and blast loadings

NASA Astrophysics Data System (ADS)

Huang, Yin-Nan

Nuclear power plants (NPPs) and spent nuclear fuel (SNF) are required by code and regulations to be designed for a family of extreme events, including very rare earthquake shaking, loss of coolant accidents, and tornado-borne missile impacts. Blast loading due to malevolent attack became a design consideration for NPPs and SNF after the terrorist attacks of September 11, 2001. The studies presented in this dissertation assess the performance of sample conventional and base isolated NPP reactor buildings subjected to seismic effects and blast loadings. The response of the sample reactor building to tornado-borne missile impacts and internal events (e.g., loss of coolant accidents) will not change if the building is base isolated and so these hazards were not considered. The sample NPP reactor building studied in this dissertation is composed of containment and internal structures with a total weight of approximately 75,000 tons. Four configurations of the reactor building are studied, including one conventional fixed-base reactor building and three base-isolated reactor buildings using Friction Pendulum(TM), lead rubber and low damping rubber bearings. The seismic assessment of the sample reactor building is performed using a new procedure proposed in this dissertation that builds on the methodology presented in the draft ATC-58 Guidelines and the widely used Zion method, which uses fragility curves defined in terms of ground-motion parameters for NPP seismic probabilistic risk assessment. The new procedure improves the Zion method by using fragility curves that are defined in terms of structural response parameters since damage and failure of NPP components are more closely tied to structural response parameters than to ground motion parameters. Alternate ground motion scaling methods are studied to help establish an optimal procedure for scaling ground motions for the purpose of seismic performance assessment. The proposed performance assessment procedure is used to evaluate the vulnerability of the conventional and base-isolated NPP reactor buildings. The seismic performance assessment confirms the utility of seismic isolation at reducing spectral demands on secondary systems. Procedures to reduce the construction cost of secondary systems in isolated reactor buildings are presented. A blast assessment of the sample reactor building is performed for an assumed threat of 2000 kg of TNT explosive detonated on the surface with a closest distance to the reactor building of 10 m. The air and ground shock waves produced by the design threat are generated and used for performance assessment. The air blast loading to the sample reactor building is computed using a Computational Fluid Dynamics code Air3D and the ground shock time series is generated using an attenuation model for soil/rock response. Response-history analysis of the sample conventional and base isolated reactor buildings to external blast loadings is performed using the hydrocode LS-DYNA. The spectral demands on the secondary systems in the isolated reactor building due to air blast loading are greater than those for the conventional reactor building but much smaller than those spectral demands associated with Safe Shutdown Earthquake shaking. The isolators are extremely effective at filtering out high acceleration, high frequency ground shock loading.

Nuclear fuel management optimization using genetic algorithms

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

DeChaine, M.D.; Feltus, M.A.

1995-07-01

The code independent genetic algorithm reactor optimization (CIGARO) system has been developed to optimize nuclear reactor loading patterns. It uses genetic algorithms (GAs) and a code-independent interface, so any reactor physics code (e.g., CASMO-3/SIMULATE-3) can be used to evaluate the loading patterns. The system is compared to other GA-based loading pattern optimizers. Tests were carried out to maximize the beginning of cycle k{sub eff} for a pressurized water reactor core loading with a penalty function to limit power peaking. The CIGARO system performed well, increasing the k{sub eff} after lowering the peak power. Tests of a prototype parallel evaluation methodmore » showed the potential for a significant speedup.« less

REACTOR - a Concept for establishing a System-of-Systems

NASA Astrophysics Data System (ADS)

Haener, Rainer; Hammitzsch, Martin; Wächter, Joachim

2014-05-01

REACTOR is a working title for activities implementing reliable, emergent, adaptive, and concurrent collaboration on the basis of transactional object repositories. It aims at establishing federations of autonomous yet interoperable systems (Systems-of-Systems), which are able to expose emergent behaviour. Following the principles of event-driven service-oriented architectures (SOA 2.0), REACTOR enables adaptive re-organisation by dynamic delegation of responsibilities and novel yet coherent monitoring strategies by combining information from different domains. Thus it allows collaborative decision-processes across system, discipline, and administrative boundaries. Interoperability is based on two approaches that implement interconnection and communication between existing heterogeneous infrastructures and information systems: Coordinated (orchestration-based) communication and publish/subscribe (choreography-based) communication. Choreography-based communication ensures the autonomy of the participating systems to the highest possible degree but requires the implementation of adapters, which provide functional access to information (publishing/consuming events) via a Message Oriented Middleware (MOM). Any interconnection of the systems (composition of service and message cascades) is established on the basis of global conversations that are enacted by choreographies specifying the expected behaviour of the participating systems with respect to agreed Service Level Agreements (SLA) required by e.g. national authorities. The specification of conversations, maintained in commonly available repositories also enables the utilisation of systems for purposes (evolving) other than initially intended. Orchestration-based communication additionally requires a central component that controls the information transfer via service requests or event processing and also takes responsibility of managing business processes. Commonly available transactional object repositories are well suited to establish brokers, which mediate metadata and semantic information about the resources of all involved systems. This concept has been developed within the project Collaborative, Complex, and Critical Decision-Support in Evolving Crises (TRIDEC) on the basis of semantic registries describing all facets of events and services utilisable for crisis management systems. The implementation utilises an operative infrastructure including an Enterprise Service Bus (ESB), adapters to proprietary sensor systems, a workflow engine, and a broker-based MOM. It also applies current technologies like actor-based frameworks for highly concurrent, distributed, and fault tolerant event-driven applications. Therefore REACTOR implementations are well suited to be hosted in a cloud that provides Infrastructure as a Service (IaaS). To provide low entry barriers for legacy and future systems, REACTOR adapts the principles of Design by Contract (DbC) as well as standardised and common information models like the Sensor Web Enablement (SWE) or the JavaScript Object Notation for geographic features (GeoJSON). REACTOR has been applied exemplarily within two different scenarios, Natural Crisis Management and Industrial Subsurface Development.

Research and proposal on selective catalytic reduction reactor optimization for industrial boiler.

PubMed

Yang, Yiming; Li, Jian; He, Hong

2017-08-24

The advanced computational fluid dynamics (CFD) software STAR-CCM+ was used to simulate a denitrification (De-NOx) project for a boiler in this paper, and the simulation result was verified based on a physical model. Two selective catalytic reduction (SCR) reactors were developed: reactor 1 was optimized and reactor 2 was developed based on reactor 1. Various indicators, including gas flow field, ammonia concentration distribution, temperature distribution, gas incident angle, and system pressure drop were analyzed. The analysis indicated that reactor 2 was of outstanding performance and could simplify developing greatly. Ammonia injection grid (AIG), the core component of the reactor, was studied; three AIGs were developed and their performances were compared and analyzed. The result indicated that AIG 3 was of the best performance. The technical indicators were proposed for SCR reactor based on the study. Flow filed distribution, gas incident angle, and temperature distribution are subjected to SCR reactor shape to a great extent, and reactor 2 proposed in this paper was of outstanding performance; ammonia concentration distribution is subjected to ammonia injection grid (AIG) shape, and AIG 3 could meet the technical indicator of ammonia concentration without mounting ammonia mixer. The developments above on the reactor and the AIG are both of great application value and social efficiency.

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.

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.

Microchannel Reactors for ISRU Applications

NASA Astrophysics Data System (ADS)

Carranza, Susana; Makel, Darby B.; Blizman, Brandon; Ward, Benjamin J.

2005-02-01

Affordable planning and execution of prolonged manned space missions depend upon the utilization of local resources and the waste products which are formed in manned spacecraft and surface bases. Successful in-situ resources utilization (ISRU) will require component technologies which provide optimal size, weight, volume, and power efficiency. Microchannel reactors enable the efficient chemical processing of in situ resources. The reactors can be designed for the processes that generate the most benefit for each mission. For instance, propellants (methane) can be produced from carbon dioxide from the Mars atmosphere using the Sabatier reaction and ethylene can be produced from the partial oxidation of methane. A system that synthesizes ethylene could be the precursor for systems to synthesize ethanol and polyethylene. Ethanol can be used as a nutrient for Astrobiology experiments, as well as the production of nutrients for human crew (e.g. sugars). Polyethylene can be used in the construction of habitats, tools, and replacement parts. This paper will present recent developments in miniature chemical reactors using advanced Micro Electro Mechanical Systems (MEMS) and microchannel technology to support ISRU of Mars and lunar missions. Among other applications, the technology has been demonstrated for the Sabatier process and for the partial oxidation of methane. Microchannel reactors were developed based on ceramic substrates as well as metal substrates. In both types of reactors, multiple layers coated with catalytic material are bonded, forming a monolithic structure. Such reactors are readily scalable with the incorporation of extra layers. In addition, this reactor structure minimizes pressure drop and catalyst settling, which are common problems in conventional packed bed reactors.

Vibration monitoring of Kraftwerk Union pressurized water reactors - Review, present status, and further development

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

Stolben, H.; Wehling, H.J.

Incipient damage to mechanical structure may be detected early in time by deviations from normal dynamic behavior. For vibration monitoring of coupled systems, only a small number of transducers are necessary, in general. On the basis, Kraftwerk Union has been involved in the development and construction of vibration monitoring systems for pressurized water reactors over the last 20 yr. The current state of the art permits vibration monitoring during normal operation by reactor personnel without expert assistance. The new SUS-86 microprocessor-based system allows further expansion toward an expert system.

Fault-tolerant reactor protection system

DOEpatents

Gaubatz, Donald C.

1997-01-01

A reactor protection system having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Each division performs independently of the others (asynchronous operation). All communications between the divisions are asynchronous. Each chassis substitutes its own spare sensor reading in the 2/3 vote if a sensor reading from one of the other chassis is faulty or missing. Therefore the presence of at least two valid sensor readings in excess of a set point is required before terminating the output to the hardware logic of a scram inhibition signal even when one of the four sensors is faulty or when one of the divisions is out of service.

Fault-tolerant reactor protection system

DOEpatents

Gaubatz, D.C.

1997-04-15

A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Each division performs independently of the others (asynchronous operation). All communications between the divisions are asynchronous. Each chassis substitutes its own spare sensor reading in the 2/3 vote if a sensor reading from one of the other chassis is faulty or missing. Therefore the presence of at least two valid sensor readings in excess of a set point is required before terminating the output to the hardware logic of a scram inhibition signal even when one of the four sensors is faulty or when one of the divisions is out of service. 16 figs.

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.

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2009-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, OH. This is a closed-cycle system that incorporates an electrically heated reactor core module, turbo alternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2010-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Nuclear system that burns its own wastes shows promise

NASA Technical Reports Server (NTRS)

Atchison, K.

1975-01-01

A nuclear fission energy system, capable of eliminating a significant amount of its radioactive wastes by burning them, is described. A theoretical investigation of this system conducted by computer analysis, is based on use of gaseous fuel nuclear reactors. Gaseous core reactors using a uranium plasma fuel are studied along with development for space propulsion.

Gas-phase optical fiber photocatalytic reactors for indoor air application: a preliminary study on performance indicators

NASA Astrophysics Data System (ADS)

Palmiste, Ü.; Voll, H.

2017-10-01

The development of advanced air cleaning technologies aims to reduce building energy consumption by reduction of outdoor air flow rates while keeping the indoor air quality at an acceptable level by air cleaning. Photocatalytic oxidation is an emerging technology for gas-phase air cleaning that can be applied in a standalone unit or a subsystem of a building mechanical ventilation system. Quantitative information on photocatalytic reactor performance is required to evaluate the technical and economic viability of the advanced air cleaning by PCO technology as an energy conservation measure in a building air conditioning system. Photocatalytic reactors applying optical fibers as light guide or photocatalyst coating support have been reported as an approach to address the current light utilization problems and thus, improve the overall efficiency. The aim of the paper is to present a preliminary evaluation on continuous flow optical fiber photocatalytic reactors based on performance indicators commonly applied for air cleaners. Based on experimental data, monolith-type optical fiber reactor performance surpasses annular-type optical fiber reactors in single-pass removal efficiency, clean air delivery rate and operating cost efficiency.

The Underwater Spectrometric System Based on CZT Detector for Survey of the Bottom of MR Reactor Pool - 13461

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

Potapov, Victor; Safronov, Alexey; Ivanov, Oleg

2013-07-01

The underwater spectrometer system for detection of irradiated nuclear fuel on the pool bottom of the reactor was elaborated. During the development process metrological studies of CdZnTe (CZT) detectors were conducted. These detectors are designed for spectrometric measurements in high radiation fields. A mathematical model based on the Monte Carlo method was created to evaluate the capability of such a system. A few experimental models were realized and the characteristics of the spectrometric system are represented. (authors)

Moon base reactor system

NASA Technical Reports Server (NTRS)

Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

1989-01-01

The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

The Cost-Effectiveness of Nuclear Power for Navy Surface Ships

DTIC Science & Technology

2011-05-01

shipbuilding plan. 1 All of the Navy’s aircraft car- riers (and submarines) are powered by nuclear reactors ; its other surface combatants are powered by...in whether the ships were powered by conventional systems that used petroleum-based fuels or by nuclear reactors . Estimates of the relative costs...would existing ships be retrofitted with nuclear reactors . 5. Those fuel -reduction findings are based on CBO’s analysis and on data provided to CBO by

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

Gehin, Jess C.; Powers, Jeffrey J.

Molten salt reactors (MSRs) represent a class of reactors that use liquid salt, usually fluoride- or chloride-based, as either a coolant with a solid fuel (such as fluoride salt-cooled high temperature reactors) or as a combined coolant and fuel with fuel dissolved in a carrier salt. For liquid-fuelled MSRs, the salt can be processed online or in a batch mode to allow for removal of fission products as well as introduction of fissile fuel and fertile materials during reactor operation. The MSR is most commonly associated with the 233U/thorium fuel cycle, as the nuclear properties of 233U combined with themore » online removal of parasitic absorbers allow for the ability to design a thermal-spectrum breeder reactor; however, MSR concepts have been developed using all neutron energy spectra (thermal, intermediate, fast, and mixed-spectrum zoned concepts) and with a variety of fuels including uranium, thorium, plutonium, and minor actinides. Early MSR work was supported by a significant research and development (R&D) program that resulted in two experimental systems operating at ORNL in the 1960s, the Aircraft Reactor Experiment and the Molten Salt Reactor Experiment. Subsequent design studies in the 1970s focusing on thermal-spectrum thorium-fueled systems established reference concepts for two major design variants: (1) a molten salt breeder reactor (MSBR), with multiple configurations that could breed additional fissile material or maintain self-sustaining operation; and (2) a denatured molten salt reactor (DMSR) with enhanced proliferation-resistance. T MSRs has been selected as one of six most promising Generation IV systems and development activities have been seen in fast-spectrum MSRs, waste-burning MSRs, MSRs fueled with low-enriched uranium (LEU), as well as more traditional thorium fuel cycle-based MSRs. This study provides an historical background of MSR R&D efforts, surveys and summarizes many of the recent development, and provides analysis comparing thorium-based MSRs.« less

Project Luna Succendo: The Lunar Evolutionary Growth-Optimized (LEGO) Reactor

NASA Astrophysics Data System (ADS)

Bess, John Darrell

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched within lunar shipments from the Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides 5 kWe using a free-piston Stirling space converter. The overall envelope for a single unit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. The subunits can be placed with centerline distances of approximately 0.6 m in a hexagonal-lattice pattern to provide sufficient neutronic coupling while allowing room for heat rejection and interstitial control. A lattice of six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network Future improvements include advances in reactor control methods, fuel form and matrix, determination of shielding requirements, as well as power conversion and heat rejection techniques to generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for use on other extraterrestrial surfaces such as Mars, other moons, and asteroids.

Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

NASA Astrophysics Data System (ADS)

Permana, Sidik

2017-07-01

A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

Nuclear thermal propulsion engine system design analysis code development

NASA Astrophysics Data System (ADS)

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

1992-01-01

A Nuclear Thermal Propulsion (NTP) Engine System Design Analyis Code has recently been developed to characterize key NTP engine system design features. Such a versatile, standalone NTP system performance and engine design code is required to support ongoing and future engine system and vehicle design efforts associated with proposed Space Exploration Initiative (SEI) missions of interest. Key areas of interest in the engine system modeling effort were the reactor, shielding, and inclusion of an engine multi-redundant propellant pump feed system design option. A solid-core nuclear thermal reactor and internal shielding code model was developed to estimate the reactor's thermal-hydraulic and physical parameters based on a prescribed thermal output which was integrated into a state-of-the-art engine system design model. The reactor code module has the capability to model graphite, composite, or carbide fuels. Key output from the model consists of reactor parameters such as thermal power, pressure drop, thermal profile, and heat generation in cooled structures (reflector, shield, and core supports), as well as the engine system parameters such as weight, dimensions, pressures, temperatures, mass flows, and performance. The model's overall analysis methodology and its key assumptions and capabilities are summarized in this paper.

System and method for the analysis of one or more compounds and/or species produced by a solution-based nuclear reactor

DOEpatents

Policke, Timothy A; Nygaard, Eric T

2014-05-06

The present invention relates generally to both a system and method for determining the composition of an off-gas from a solution nuclear reactor (e.g., an Aqueous Homogeneous Reactor (AHR)) and the composition of the fissioning solution from those measurements. In one embodiment, the present invention utilizes at least one quadrupole mass spectrometer (QMS) in a system and/or method designed to determine at least one or more of: (i) the rate of production of at least one gas and/or gas species from a nuclear reactor; (ii) the effect on pH by one or more nitrogen species; (iii) the rate of production of one or more fission gases; and/or (iv) the effect on pH of at least one gas and/or gas species other than one or more nitrogen species from a nuclear reactor.

SP-100 reactor with Brayton conversion for lunar surface applications

NASA Technical Reports Server (NTRS)

Mason, Lee S.; Rodriguez, Carlos D.; Mckissock, Barbara I.; Hanlon, James C.; Mansfield, Brian C.

1992-01-01

Examined here is the potential for integrating Brayton-cycle power conversion with the SP-100 reactor for lunar surface power system applications. Two designs were characterized and modeled. The first design integrates a 100-kWe SP-100 Brayton power system with a lunar lander. This system is intended to meet early lunar mission power needs while minimizing on-site installation requirements. Man-rated radiation protection is provided by an integral multilayer, cylindrical lithium hydride/tungsten (LiH/W) shield encircling the reactor vessel. Design emphasis is on ease of deployment, safety, and reliability, while utilizing relatively near-term technology. The second design combines Brayton conversion with the SP-100 reactor in a erectable 550-kWe powerplant concept intended to satisfy later-phase lunar base power requirements. This system capitalizes on experience gained from operating the initial 100-kWe module and incorporates some technology improvements. For this system, the reactor is emplaced in a lunar regolith excavation to provide man-rated shielding, and the Brayton engines and radiators are mounted on the lunar surface and extend radially from the central reactor. Design emphasis is on performance, safety, long life, and operational flexibility.

Control console replacement at the WPI Reactor

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

Not Available

1992-01-01

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

Liquid Metal Pump Technologies for Nuclear Surface Power

NASA Technical Reports Server (NTRS)

Polzin, Kurt A.

2007-01-01

Multiple liquid metal pump options are reviewed for the purpose of determining the technologies that are best suited for inclusion in a nuclear reactor thermal simulator intended to rest prototypical space nuclear surface power system components. Conduction, induction and thermoelectric electromagnetic pumps are evaluated based on their performance characteristics and the technical issues associated with incorporation into a reactor system. A thermoelectric electromagnetic pump is selected as the best option for use in NASA-MSFC's Fission Surface Power-Primary Test Circuit reactor simulator based on its relative simplicity, low power supply mass penalty, flight heritage, and the promise of increased pump efficiency over those earlier pump designs through the use of skutterudite thermoelectric elements.

Regenerative Carbonate-Based Thermochemical Energy Storage System for Concentrating Solar Power

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

Gangwal, Santosh; Muto, Andrew

Southern Research has developed a thermochemical energy storage (TCES) technology that utilizes the endothermic-exothermic reversible carbonation of calcium oxide (lime) to store thermal energy at high-temperatures, such as those achieved by next generation concentrating solar power (CSP) facilities. The major challenges addressed in the development of this system include refining a high capacity, yet durable sorbent material and designing a low thermal resistance low-cost heat exchanger reactor system to move heat between the sorbent and a heat transfer fluid under conditions relevant for CSP operation (e.g., energy density, reaction kinetics, heat flow). The proprietary stabilized sorbent was developed by Precisionmore » Combustion, Inc. (PCI). A factorial matrix of sorbent compositions covering the design space was tested using accelerated high throughput screening in a thermo-gravimetric analyzer. Several promising formulations were selected for more thorough evaluation and one formulation with high capacity (0.38 g CO 2/g sorbent) and durability (>99.7% capacity retention over 100 cycles) was chosen as a basis for further development of the energy storage reactor system. In parallel with this effort, a full range of currently available commercial and developmental heat exchange reactor systems and sorbent loading methods were examined through literature research and contacts with commercial vendors. Process models were developed to examine if a heat exchange reactor system and balance of plant can meet required TCES performance and cost targets, optimizing tradeoffs between thermal performance, exergetic efficiency, and cost. Reactor types evaluated included many forms, from microchannel reactor, to diffusion bonded heat exchanger, to shell and tube heat exchangers. The most viable design for application to a supercritical CO 2 power cycle operating at 200-300 bar pressure and >700°C was determined to be a combination of a diffusion bonded heat exchanger with a shell and tube reactor. A bench scale reactor system was then designed and constructed to test sorbent performance under more commercially relevant conditions. This system utilizes a tube-in tube reactor design containing approximately 250 grams sorbent and is able to operate under a wide range of temperature, pressure and flow conditions as needed to explore system performance under a variety of operating conditions. A variety of sorbent loading methods may be tested using the reactor design. Initial bench test results over 25 cycles showed very high sorbent stability (>99%) and sufficient capacity (>0.28 g CO 2/g sorbent) for an economical commercial-scale system. Initial technoeconomic evaluation of the proposed storage system show that the sorbent cost should not have a significant impact on overall system cost, and that the largest cost impacts come from the heat exchanger reactor and balance of plant equipment, including compressors and gas storage, due to the high temperatures for sCO 2 cycles. Current estimated system costs are $47/kWhth based on current material and equipment cost estimates.« less

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

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

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

Micro-reactors for characterization of nanostructure-based sensors.

PubMed

Savu, R; Silveira, J V; Flacker, A; Vaz, A R; Joanni, E; Pinto, A C; Gobbi, A L; Santos, T E A; Rotondaro, A L P; Moshkalev, S A

2012-05-01

Fabrication and testing of micro-reactors for the characterization of nanosensors is presented in this work. The reactors have a small volume (100 μl) and are equipped with gas input/output channels. They were machined from a single piece of kovar in order to avoid leaks in the system due to additional welding. The contact pins were electrically insulated from the body of the reactor using a borosilicate sealing glass and the reactor was hermetically sealed using a lid and an elastomeric o-ring. One of the advantages of the reactor lies in its simple assembly and ease of use with any vacuum/gas system, allowing the connection of more than one device. Moreover, the lid can be modified in order to fit a window for in situ optical characterization. In order to prove its versatility, carbon nanotube-based sensors were tested using this micro-reactor. The devices were fabricated by depositing carbon nanotubes over 1 μm thick gold electrodes patterned onto Si/SiO(2) substrates. The sensors were tested using oxygen and nitrogen atmospheres, in the pressure range between 10(-5) and 10(-1) mbar. The small chamber volume allowed the measurement of fast sensor characteristic times, with the sensors showing good sensitivity towards gas and pressure as well as high reproducibility.

Micro-reactors for characterization of nanostructure-based sensors

NASA Astrophysics Data System (ADS)

Savu, R.; Silveira, J. V.; Flacker, A.; Vaz, A. R.; Joanni, E.; Pinto, A. C.; Gobbi, A. L.; Santos, T. E. A.; Rotondaro, A. L. P.; Moshkalev, S. A.

2012-05-01

Fabrication and testing of micro-reactors for the characterization of nanosensors is presented in this work. The reactors have a small volume (100 μl) and are equipped with gas input/output channels. They were machined from a single piece of kovar in order to avoid leaks in the system due to additional welding. The contact pins were electrically insulated from the body of the reactor using a borosilicate sealing glass and the reactor was hermetically sealed using a lid and an elastomeric o-ring. One of the advantages of the reactor lies in its simple assembly and ease of use with any vacuum/gas system, allowing the connection of more than one device. Moreover, the lid can be modified in order to fit a window for in situ optical characterization. In order to prove its versatility, carbon nanotube-based sensors were tested using this micro-reactor. The devices were fabricated by depositing carbon nanotubes over 1 μm thick gold electrodes patterned onto Si/SiO2 substrates. The sensors were tested using oxygen and nitrogen atmospheres, in the pressure range between 10-5 and 10-1 mbar. The small chamber volume allowed the measurement of fast sensor characteristic times, with the sensors showing good sensitivity towards gas and pressure as well as high reproducibility.

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

Data acquisition system for segmented reactor antineutrino detector

NASA Astrophysics Data System (ADS)

Hons, Z.; Vlášek, J.

2017-01-01

This paper describes the data acquisition system used for data readout from the PMT channels of a segmented detector of reactor antineutrinos with active shielding. Theoretical approach to the data acquisition is described and two possible solutions using QDCs and digitizers are discussed. Also described are the results of the DAQ performance during routine data taking operation of DANSS. DANSS (Detector of the reactor AntiNeutrino based on Solid Scintillator) is a project aiming to measure a spectrum of reactor antineutrinos using inverse beta decay (IBD) in a plastic scintillator. The detector is located close to an industrial nuclear reactor core and is covered by passive and active shielding. It is expected to have about 15000 IBD interactions per day. Light from the detector is sensed by PMT and SiPM.

Space nuclear power systems; Proceedings of the 8th Symposium, Albuquerque, NM, Jan. 6-10, 1991. Pts. 1-3

NASA Technical Reports Server (NTRS)

El-Genk, Mohamed S. (Editor); Hoover, Mark D. (Editor)

1991-01-01

The present conference discusses NASA mission planning for space nuclear power, lunar mission design based on nuclear thermal rockets, inertial-electrostatic confinement fusion for space power, nuclear risk analysis of the Ulysses mission, the role of the interface in refractory metal alloy composites, an advanced thermionic reactor systems design code, and space high power nuclear-pumped lasers. Also discussed are exploration mission enhancements with power-beaming, power requirement estimates for a nuclear-powered manned Mars rover, SP-100 reactor design, safety, and testing, materials compatibility issues for fabric composite radiators, application of the enabler to nuclear electric propulsion, orbit-transfer with TOPAZ-type power sources, the thermoelectric properties of alloys, ruthenium silicide as a promising thermoelectric material, and innovative space-saving device for high-temperature piping systems. The second volume of this conference discusses engine concepts for nuclear electric propulsion, nuclear technologies for human exploration of the solar system, dynamic energy conversion, direct nuclear propulsion, thermionic conversion technology, reactor and power system control, thermal management, thermionic research, effects of radiation on electronics, heat-pipe technology, radioisotope power systems, and nuclear fuels for power reactors. The third volume discusses space power electronics, space nuclear fuels for propulsion reactors, power systems concepts, space power electronics systems, the use of artificial intelligence in space, flight qualifications and testing, microgravity two-phase flow, reactor manufacturing and processing, and space and environmental effects.

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

Isotopic signature of atmospheric xenon released from light water reactors.

PubMed

Kalinowski, Martin B; Pistner, Christoph

2006-01-01

A global monitoring system for atmospheric xenon radioactivity is being established as part of the International Monitoring System to verify compliance with the Comprehensive Nuclear-Test-Ban Treaty (CTBT). The isotopic activity ratios of (135)Xe, (133m)Xe, (133)Xe and (131m)Xe are of interest for distinguishing nuclear explosion sources from civilian releases. Simulations of light water reactor (LWR) fuel burn-up through three operational reactor power cycles are conducted to explore the possible xenon isotopic signature of nuclear reactor releases under different operational conditions. It is studied how ratio changes are related to various parameters including the neutron flux, uranium enrichment and fuel burn-up. Further, the impact of diffusion and mixing on the isotopic activity ratio variability are explored. The simulations are validated with reported reactor emissions. In addition, activity ratios are calculated for xenon isotopes released from nuclear explosions and these are compared to the reactor ratios in order to determine whether the discrimination of explosion releases from reactor effluents is possible based on isotopic activity ratios.

Nuclear reactor power as applied to a space-based radar mission

NASA Technical Reports Server (NTRS)

Jaffe, L.; Fujita, T.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Grossman, M.; Kia, T.; Nesmith, B.

1988-01-01

The SP-100 Project was established to develop and demonstrate feasibility of a space reactor power system (SRPS) at power levels of 10's of kilowatts to a megawatt. To help determine systems requirements for the SRPS, a mission and spacecraft were examined which utilize this power system for a space-based radar to observe moving objects. Aspects of the mission and spacecraft bearing on the power system were the primary objectives of this study; performance of the radar itself was not within the scope. The study was carried out by the Systems Design Audit Team of the SP-100 Project.

Reduction of the hydraulic retention time at constant high organic loading rate to reach the microbial limits of anaerobic digestion in various reactor systems.

PubMed

Ziganshin, Ayrat M; Schmidt, Thomas; Lv, Zuopeng; Liebetrau, Jan; Richnow, Hans Hermann; Kleinsteuber, Sabine; Nikolausz, Marcell

2016-10-01

The effects of hydraulic retention time (HRT) reduction at constant high organic loading rate on the activity of hydrogen-producing bacteria and methanogens were investigated in reactors digesting thin stillage. Stable isotope fingerprinting was additionally applied to assess methanogenic pathways. Based on hydA gene transcripts, Clostridiales was the most active hydrogen-producing order in continuous stirred tank reactor (CSTR), fixed-bed reactor (FBR) and anaerobic sequencing batch reactor (ASBR), but shorter HRT stimulated the activity of Spirochaetales. Further decreasing HRT diminished Spirochaetales activity in systems with biomass retention. Based on mcrA gene transcripts, Methanoculleus and Methanosarcina were the predominantly active in CSTR and ASBR, whereas Methanosaeta and Methanospirillum activity was more significant in stably performing FBR. Isotope values indicated the predominance of aceticlastic pathway in FBR. Interestingly, an increased activity of Methanosaeta was observed during shortening HRT in CSTR and ASBR despite high organic acids concentrations, what was supported by stable isotope data. Copyright © 2016 Elsevier Ltd. All rights reserved.

Control console replacement at the WPI Reactor. [Final report

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

Not Available

1992-12-31

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

NASA Astrophysics Data System (ADS)

Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

2018-02-01

Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

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

Soubies, B.; Henry, J.Y.; Le Meur, M.

1300 MWe pressurised water reactors (PWRs), like the 1400 MWe reactors, operate with microprocessor-based safety systems. This is particularly the case for the Digital Integrated Protection System (SPIN), which trips the reactor in an emergency and sets in action the safeguard functions. The softwares used in these systems must therefore be highly dependable in the execution of their functions. In the case of SPIN, three players are working at different levels to achieve this goal: the protection system manufacturer, Merlin Gerin; the designer of the nuclear steam supply system, Framatome; the operator of the nuclear power plants, Electricite de Francemore » (EDF), which is also responsible for the safety of its installations. Regulatory licenses are issued by the French safety authority, the Nuclear Installations Safety Directorate (French abbreviation DSIN), subsequent to a successful examination of the technical provisions adopted by the operator. This examination is carried out by the IPSN and the standing group on nuclear reactors. This communication sets out: the methods used by the manufacturer to develop SPIN software for the 1400 MWe PWRs (N4 series); the approach adopted by the IPSN to evaluate the safety software of the protection system for the N4 series of reactors.« less

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

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.; Sawicki, Jerzy T.

2003-01-01

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

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

NASA Astrophysics Data System (ADS)

Juhasz, Albert J.; Sawicki, Jerzy T.

2004-02-01

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

Safety features of subcritical fluid fueled systems

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

Bell, C.R.

1995-10-01

Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitativemore » in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved is very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible.« less

CONCEPTUAL DESIGN OF A LUNAR REGOLITH CLUSTERED-REACTOR SYSTEM

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

John Darrell Bess

2009-06-01

It is proposed that a fast-fission, heatpipe-cooled, lunar-surface power reactor system be divided into subcritical units that could be launched safely without the incorporation of additional spectral shift absorbers or other complex means of control. The reactor subunits are to be emplaced directly into the lunar regolith utilizing the regolith not just for shielding but as the reflector material to increase the neutron economy of the system. While a single subunit cannot achieve criticality by itself, coordinated placement of additional subunits will provide a critical reactor system for lunar surface power generation. A lunar regolith clustered-reactor system promotes reliability, safety,more » and ease of manufacture and testing at the cost of a slight increase in launch mass per rated power level and an overall reduction in neutron economy when compared to a single-reactor system. Additional subunits may be launched with future missions to increase the cluster size and power according to desired lunar base power demand and lifetime. The results address the potential uncertainties associated with the lunar regolith material and emplacement of the subunit systems. Physical distance between subunits within the clustered emplacement exhibits the most significant feedback regarding changes in overall system reactivity. Narrow, deep holes will be the most effective in reducing axial neutron leakage from the core. The variation in iron concentration in the lunar regolith can directly influence the overall system reactivity although its effects are less than the more dominant factors of subunit emplacement.« less

Chemical production processes and systems

DOEpatents

Holladay, Johnathan E.; Muzatko, Danielle S.; White, James F.; Zacher, Alan H.

2014-06-17

Hydrogenolysis systems are provided that can include a reactor housing an Ru-comprising hydrogenolysis catalyst and wherein the contents of the reactor is maintained at a neutral or acidic pH. Reactant reservoirs within the system can include a polyhydric alcohol compound and a base, wherein a weight ratio of the base to the compound is less than 0.05. Systems also include the product reservoir comprising a hydrogenolyzed polyhydric alcohol compound and salts of organic acids, and wherein the moles of base are substantially equivalent to the moles of salts or organic acids. Processes are provided that can include an Ru-comprising catalyst within a mixture having a neutral or acidic pH. A weight ratio of the base to the compound can be between 0.01 and 0.05 during exposing.

Chemical production processes and systems

DOEpatents

Holladay, Johnathan E; Muzatko, Danielle S; White, James F; Zacher, Alan H

2015-04-21

Hydrogenolysis systems are provided that can include a reactor housing an Ru-comprising hydrogenolysis catalyst and wherein the contents of the reactor is maintained at a neutral or acidic pH. Reactant reservoirs within the system can include a polyhydric alcohol compound and a base, wherein a weight ratio of the base to the compound is less than 0.05. Systems also include the product reservoir comprising a hydrogenolyzed polyhydric alcohol compound and salts of organic acids, and wherein the moles of base are substantially equivalent to the moles of salts or organic acids. Processes are provided that can include an Ru-comprising catalyst within a mixture having a neutral or acidic pH. A weight ratio of the base to the compound can be between 0.01 and 0.05 during exposing.

Definition of a Robust Supervisory Control Scheme for Sodium-Cooled Fast Reactors

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

Ponciroli, R.; Passerini, S.; Vilim, R. B.

In this work, an innovative control approach for metal-fueled Sodium-cooled Fast Reactors is proposed. With respect to the classical approach adopted for base-load Nuclear Power Plants, an alternative control strategy for operating the reactor at different power levels by respecting the system physical constraints is presented. In order to achieve a higher operational flexibility along with ensuring that the implemented control loops do not influence the system inherent passive safety features, a dedicated supervisory control scheme for the dynamic definition of the corresponding set-points to be supplied to the PID controllers is designed. In particular, the traditional approach based onmore » the adoption of tabulated lookup tables for the set-point definition is found not to be robust enough when failures of the implemented SISO (Single Input Single Output) actuators occur. Therefore, a feedback algorithm based on the Reference Governor approach, which allows for the optimization of reference signals according to the system operating conditions, is proposed.« less

System Concepts for Affordable Fission Surface Power

NASA Technical Reports Server (NTRS)

Mason, Lee; Poston, David; Qualls, Louis

2008-01-01

This paper presents an overview of an affordable Fission Surface Power (FSP) system that could be used for NASA applications on the Moon and Mars. The proposed FSP system uses a low temperature, uranium dioxide-fueled, liquid metal-cooled fission reactor coupled to free-piston Stirling converters. The concept was determined by a 12 month NASA/DOE study that examined design options and development strategies based on affordability and risk. The system is considered a low development risk based on the use of terrestrial-derived reactor technology, high efficiency power conversion, and conventional materials. The low-risk approach was selected over other options that could offer higher performance and/or lower mass.

3D Neutronic Analysis in MHD Calculations at ARIES-ST Fusion Reactors Systems

NASA Astrophysics Data System (ADS)

Hançerliogulları, Aybaba; Cini, Mesut

2013-10-01

In this study, we developed new models for liquid wall (FW) state at ARIES-ST fusion reactor systems. ARIES-ST is a 1,000 MWe fusion reactor system based on a low aspect ratio ST plasma. In this article, we analyzed the characteristic properties of magnetohydrodynamics (MHD) and heat transfer conditions by using Monte-Carlo simulation methods (ARIES Team et al. in Fusion Eng Des 49-50:689-695, 2000; Tillack et al. in Fusion Eng Des 65:215-261, 2003) . In fusion applications, liquid metals are traditionally considered to be the best working fluids. The working liquid must be a lithium-containing medium in order to provide adequate tritium that the plasma is self-sustained and that the fusion is a renewable energy source. As for Flibe free surface flows, the MHD effects caused by interaction with the mean flow is negligible, while a fairly uniform flow of thick can be maintained throughout the reactor based on 3-D MHD calculations. In this study, neutronic parameters, that is to say, energy multiplication factor radiation, heat flux and fissile fuel breeding were researched for fusion reactor with various thorium and uranium molten salts. Sufficient tritium amount is needed for the reactor to work itself. In the tritium breeding ratio (TBR) >1.05 ARIES-ST fusion model TBR is >1.1 so that tritium self-sufficiency is maintained for DT fusion systems (Starke et al. in Fusion Energ Des 84:1794-1798, 2009; Najmabadi et al. in Fusion Energ Des 80:3-23, 2006).

Liquid-Metal Pump Technologies for Nuclear Surface Power

NASA Technical Reports Server (NTRS)

Polzin, K. A.

2007-01-01

Multiple liquid-metal pump options are reviewed for the purpose of determining the technologies that are best suited for inclusion in a nuclear reactor thermal simulator intended to test prototypical space nuclear system components. Conduction, induction, and thermoelectric electromagnetic pumps are evaluated based on their performance characteristics and the technical issues associated with incorporation into a reactor system. The thermoelectric pump is recommended for inclusion in the planned system at NASA MSFC based on its relative simplicity, low power supply mass penalty, flight heritage, and the promise of increased pump efficiency over earlier flight pump designs through the use of skutterudite thermoelectric elements.

Calculation of Purification Systems of Hydrocarbonmoderated Reactors; CALCULO DE SISTEMAS DE PURIFICATION DE REACTORES MODERADOS FOR HIDROCARBUROS

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

Santos, A.A.

1958-01-01

culation of Purification Systems of Hydrocarbonmoderated Reactors). Agustin Alonso Santos. 1958. 23p. As as introduction to the calculation of the purification systems of bydrocarbon-moderated reactors, the effects of heat and radiation on the polyphenols are considered. The chemical, physical, and nuclear properties are tabulated. The formation velocity of the polymers and gases, pyrolysis, effects of heat on the polymer, and the activity accumulated in the moderator ars discussed. The calculation is based on the hypetheses that the radiation catalyzes the formation of polymers, the velocity of the polymerization reaction is constant, the polymer concentration is maintained at a limit whichmore » does not adversely affect the heat transfer properties, the velocity of the separation of polymers in the distillation column is in proportion to their concentration in the hydrocarbon and the pyrolysis causes gaseous products. Formulas are derived expressing the purified flow and the activities accumulated in the distillation residues. The results are applied to the parification system of the Organic Moderated Reactor Experiment (J.S.R.)« less

Multi-phase model development to assess RCIC system capabilities under severe accident conditions

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

Kirkland, Karen Vierow; Ross, Kyle; Beeny, Bradley

The Reactor Core Isolation Cooling (RCIC) System is a safety-related system that provides makeup water for core cooling of some Boiling Water Reactors (BWRs) with a Mark I containment. The RCIC System consists of a steam-driven Terry turbine that powers a centrifugal, multi-stage pump for providing water to the reactor pressure vessel. The Fukushima Dai-ichi accidents demonstrated that the RCIC System can play an important role under accident conditions in removing core decay heat. The unexpectedly sustained, good performance of the RCIC System in the Fukushima reactor demonstrates, firstly, that its capabilities are not well understood, and secondly, that themore » system has high potential for extended core cooling in accident scenarios. Better understanding and analysis tools would allow for more options to cope with a severe accident situation and to reduce the consequences. The objectives of this project were to develop physics-based models of the RCIC System, incorporate them into a multi-phase code and validate the models. This Final Technical Report details the progress throughout the project duration and the accomplishments.« less

Space nuclear power systems; Proceedings of the 8th Symposium, Albuquerque, NM, Jan. 6-10, 1991. Pts. 1-3

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.; Hoover, Mark D.

1991-07-01

The present conference discusses NASA mission planning for space nuclear power, lunar mission design based on nuclear thermal rockets, inertial-electrostatic confinement fusion for space power, nuclear risk analysis of the Ulysses mission, the role of the interface in refractory metal alloy composites, an advanced thermionic reactor systems design code, and space high power nuclear-pumped lasers. Also discussed are exploration mission enhancements with power-beaming, power requirement estimates for a nuclear-powered manned Mars rover, SP-100 reactor design, safety, and testing, materials compatibility issues for fabric composite radiators, application of the enabler to nuclear electric propulsion, orbit-transfer with TOPAZ-type power sources, the thermoelectric properties of alloys, ruthenium silicide as a promising thermoelectric material, and innovative space-saving device for high-temperature piping systems. The second volume of this conference discusses engine concepts for nuclear electric propulsion, nuclear technologies for human exploration of the solar system, dynamic energy conversion, direct nuclear propulsion, thermionic conversion technology, reactor and power system control, thermal management, thermionic research, effects of radiation on electronics, heat-pipe technology, radioisotope power systems, and nuclear fuels for power reactors. The third volume discusses space power electronics, space nuclear fuels for propulsion reactors, power systems concepts, space power electronics systems, the use of artificial intelligence in space, flight qualifications and testing, microgravity two-phase flow, reactor manufacturing and processing, and space and environmental effects. (For individual items see A93-13752 to A93-13937)

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

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.

2000-01-01

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

A 735 kV shunt reactors automatic switching system for Hydro-Quebec network

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

Bernard, S.; Trudel, G.; Scott, G.

1996-11-01

In recent years, Hydro-Quebec has undertaken a major program to upgrade the reliability of its transmission system. Much efforts have been directed toward increasing the system`s capacity to withstand extreme contingencies, usually caused by multiple incidents or the successive tripping of transmission lines. In order to counter such events, Hydro-Quebec has adopted a defensive scheme. Based entirely on automatic action, this scheme will mainly rely on: a 735 kV shunt reactor switching system (called MAIS); a generation rejection and/or remote load-shedding system (called RPTC); an underfrequency load-shedding system. The MAIS system, which is the subject of this paper, will bemore » implemented in 22 substations and is required to control voltage on the system after a severe event. Each MAIS system, acting locally, is entirely independent and will close or trip shunt reactors in response to local conditions.« less

Apparatus and systems for measuring elongation of objects, methods of measuring, and reactor

DOEpatents

Rempe, Joy L [Idaho Falls, ID; Knudson, Darrell L [Firth, ID; Daw, Joshua E [Idaho Falls, ID; Condie, Keith G [Idaho Falls, ID; Stoots, Carl M [Idaho Falls, ID

2011-11-29

Elongation measurement apparatuses and systems comprise at least two Linear Variable Differential Transformers (LVDTs) with a push rod coupled to each of the at least two LVDTs at one longitudinal end thereof. At least one push rod extends to a base and is coupled thereto at an opposing longitudinal end, and at least one other push rod extends to a location spaced apart from the base and is configured to receive a sample between an opposing longitudinal end of the at least one other push rod and the base. Nuclear reactors comprising such apparatuses and systems and methods of measuring elongation of a material are also disclosed.

Activation product transport in fusion reactors. [RAPTOR

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

Klein, A.C.

1983-01-01

Activated corrosion and neutron sputtering products will enter the coolant and/or tritium breeding material of fusion reactor power plants and experiments and cause personnel access problems. Radiation levels around plant components due to these products will cause difficulties with maintenance and repair operations throughout the plant. Similar problems are experienced around fission reactor systems. The determination of the transport of radioactive corrosion and neutron sputtering products through the system is achieved using the computer code RAPTOR. This code calculates the mass transfer of a number of activation products based on the corrosion and sputtering rates through the system, the depositionmore » and release characteristics of various plant components, the neturon flux spectrum, as well as other plant parameters. RAPTOR assembles a system of first order linear differential equations into a matrix equation based upon the reactor system parameters. Included in the transfer matrix are the deposition and erosion coefficients, and the decay and activation data for the various plant nodes and radioactive isotopes. A source vector supplies the corrosion and neutron sputtering source rates. This matrix equation is then solved using a matrix operator technique to give the specific activity distribution of each radioactive species throughout the plant. Once the amount of mass transfer is determined, the photon transport due to the radioactive corrosion and sputtering product sources can be evaluated, and dose rates around the plant components of interest as a function of time can be determined. This method has been used to estimate the radiation hazards around a number of fusion reactor system designs.« less

Enzymatic membrane reactors for biodegradation of recalcitrant compounds. Application to dye decolourisation.

PubMed

López, C; Mielgo, I; Moreira, M T; Feijoo, G; Lema, J M

2002-11-13

Membrane bioreactors are being increasingly used in enzymatic catalysed transformations. However, the application of enzymatic-based treatment systems in the environmental field is rather unusual. The aim of this paper is to overview the application of enzymatic membrane reactors to wastewater treatment, more specifically to dye decolourisation. Firstly, the basic aspects such as different configurations of enzymatic reactors, advantages and disadvantages associated to their utilisation are revised as well as the application of this technology to wastewater treatment. Secondly, dye decolourisation by white-rot fungi and their oxidative enzymes are discussed, presenting an overall view from for in vivo and in vitro systems. Finally, dye decolourisation by manganese peroxidase in an enzymatic membrane reactor in continuous operation is presented.

2007 Global Demilitarization Symposium and Exhibition

DTIC Science & Technology

2007-05-17

system! Defines reactive properties ! www.biorex.se Heavy metals • Water soluble • Easily spread • Elements • Not degradable • Hazardous • In high or low...Activated sludge Oat by-products Juniper fibre Sawdust Agro-based fibres Seaweed Grape stalk fibres Marine algae Husk of black gram www.biorex.se Sorbent...Description Reactor Illustration 18 inch diameter reactor 120 inches high Single diameter throughout entire reactor length 19 resistance heaters Alloy 600

Novel limiter pump topologies

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

Schultz, J.H.

1981-01-01

The use of limiter pumps as the principle plasma exhaust system of a magnetic confinement fusion device promises significant simplification, when compared to previously investigating divertor based systems. Further simplifications, such as the integration of the exhaust system with a radio frequency heating system and with the main reactor shield and structure are investigated below. The integrity of limiters in a reactor environment is threatened by many mechanisms, the most severe of which may be erosion by sputtering. Two novel topolgies are suggested which allow high erosion without limiter failure.

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...

High Efficiency Solar-based Catalytic Structure for CO 2 Reforming

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

Menkara, Hisham

Throughout this project, we developed and optimized various photocatalyst structures for CO 2 reforming into hydrocarbon fuels and various commodity chemical products. We also built several closed-loop and continuous fixed-bed photocatalytic reactor system prototypes for a larger-scale demonstration of CO 2 reforming into hydrocarbons, mainly methane and formic acid. The results achieved have indicated that with each type of reactor and structure, high reforming yields can be obtained by refining the structural and operational conditions of the reactor, as well as by using various sacrificial agents (hole scavengers). We have also demonstrated, for the first time, that an aqueous solutionmore » containing acid whey (a common bio waste) is a highly effective hole scavenger for a solar-based photocatalytic reactor system and can help reform CO 2 into several products at once. The optimization tasks performed throughout the project have resulted in efficiency increase in our conventional reactors from an initial 0.02% to about 0.25%, which is 10X higher than our original project goal. When acid whey was used as a sacrificial agent, the achieved energy efficiency for formic acid alone was ~0.4%, which is 16X that of our original project goal and higher than anything ever reported for a solar-based photocatalytic reactor. Therefore, by carefully selecting sacrificial agents, it should be possible to reach energy efficiency in the range of the photosynthetic efficiency of typical crop and biofuel plants (1-3%).« less

Approach to developing reliable space reactor power systems

NASA Technical Reports Server (NTRS)

Mondt, Jack F.; Shinbrot, Charles H.

1991-01-01

During Phase II, the Engineering Development Phase, the SP-100 Project has defined and is pursuing a new approach to developing reliable power systems. The approach to developing such a system during the early technology phase is described along with some preliminary examples to help explain the approach. Developing reliable components to meet space reactor power system requirements is based on a top-down systems approach which includes a point design based on a detailed technical specification of a 100-kW power system. The SP-100 system requirements implicitly recognize the challenge of achieving a high system reliability for a ten-year lifetime, while at the same time using technologies that require very significant development efforts. A low-cost method for assessing reliability, based on an understanding of fundamental failure mechanisms and design margins for specific failure mechanisms, is being developed as part of the SP-100 Program.

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.

Novel Anaerobic Wastewater Treatment System for Energy Generation at Forward Operating Bases

DTIC Science & Technology

2016-08-01

AnMBR) technology with clinoptilolite ion exchange and GreenBox™ ammonia electrolysis. The system generates both methane and hydrogen fuels...experimental setup. ................................................ 21 Figure 10. Methane phase semi batch experimental setup, a total of three reactors were...set up for PS + solid, Bioc and ADS methane phase reactors. .................... 21 Figure 11. Dried PS solid for the control, Bioc blend for the

Lunar electric power systems utilizing the SP-100 reactor coupled to dynamic conversion systems

NASA Technical Reports Server (NTRS)

Harty, Richard B.; Durand, Richard E.

1993-01-01

An integration study was performed by Rocketdyne under contract to NASA-LeRC. The study was concerned with coupling an SP-0100 reactor to either a Brayton or Stirling power conversion system. The application was for a surface power system to supply power requirements to a lunar base. A power level of 550 kWe was selected based on the NASA Space Exploration Initiative 90-day study. Reliability studies were initially performed to determine optimum power conversion redundancy. This study resulted in selecting three operating engines and one stand-by unit. Integration design studies indicated that either the Brayton or Stirling power conversion systems could be integrated with the PS-100 reactor. The Stirling system had an integration advantage because of smaller piping size and fewer components. The Stirling engine, however, is more complex and heavier than the Brayton rotating unit, which tends to off-set the Stirling integration advantage. From a performance consideration, the Brayton had a 9 percent mass advantage, and the Stirling had a 50 percent radiator advantage.

Developments and Tendencies in Fission Reactor Concepts

NASA Astrophysics Data System (ADS)

Adamov, E. O.; Fuji-Ie, Y.

This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC) - as an advanced and promising reactor system that offers solutions to the above problems. The difference (not confrontation) between the approaches to nuclear power development based on the principles of “inherent safety” and “natural safety” is demonstrated.

Solar photochemical process engineering for production of fuels and chemicals

NASA Technical Reports Server (NTRS)

Biddle, J. R.; Peterson, D. B.; Fujita, T.

1984-01-01

The engineering costs and performance of a nominal 25,000 scmd (883,000 scfd) photochemical plant to produce dihydrogen from water were studied. Two systems were considered, one based on flat-plate collector/reactors and the other on linear parabolic troughs. Engineering subsystems were specified including the collector/reactor, support hardware, field transport piping, gas compression equipment, and balance-of-plant (BOP) items. Overall plant efficiencies of 10.3 and 11.6% are estimated for the flat-plate and trough systems, respectively, based on assumed solar photochemical efficiencies of 12.9 and 14.6%. Because of the opposing effects of concentration ratio and operating temperature on efficiency, it was concluded that reactor cooling would be necessary with the trough system. Both active and passive cooling methods were considered. Capital costs and energy costs, for both concentrating and non-concentrating systems, were determined and their sensitivity to efficiency and economic parameters were analyzed. The overall plant efficiency is the single most important factor in determining the cost of the fuel.

Solar photochemical process engineering for production of fuels and chemicals

NASA Technical Reports Server (NTRS)

Biddle, J. R.; Peterson, D. B.; Fujita, T.

1985-01-01

The engineering costs and performance of a nominal 25,000 scmd (883,000 scfd) photochemical plant to produce dihydrogen from water were studied. Two systems were considered, one based on flat-plate collector/reactors and the other on linear parabolic troughs. Engineering subsystems were specified including the collector/reactor, support hardware, field transport piping, gas compression equipment, and balance-of-plant (BOP) items. Overall plant efficiencies of 10.3 and 11.6 percent are estimated for the flat-plate and trough systems, respectively, based on assumed solar photochemical efficiencies of 12.9 and 14.6 percent. Because of the opposing effects of concentration ratio and operating temperature on efficiency, it was concluded that reactor cooling would be necessary with the trough system. Both active and passive cooling methods were considered. Capital costs and energy costs, for both concentrating and non-concentrating systems, were determined and their sensitivity to efficiency and economic parameters were analyzed. The overall plant efficiency is the single most important factor in determining the cost of the fuel.

Nuclear reactor power for a space-based radar. SP-100 project

NASA Technical Reports Server (NTRS)

Bloomfield, Harvey; Heller, Jack; Jaffe, Leonard; Beatty, Richard; Bhandari, Pradeep; Chow, Edwin; Deininger, William; Ewell, Richard; Fujita, Toshio; Grossman, Merlin

1986-01-01

A space-based radar mission and spacecraft, using a 300 kWe nuclear reactor power system, has been examined, with emphasis on aspects affecting the power system. The radar antenna is a horizontal planar array, 32 X 64 m. The orbit is at 61 deg, 1088 km. The mass of the antenna with support structure is 42,000 kg; of the nuclear reactor power system, 8,300 kg; of the whole spacecraft about 51,000 kg, necessitating multiple launches and orbital assembly. The assembly orbit is at 57 deg, 400 km, high enough to provide the orbital lifetime needed for orbital assembly. The selected scenario uses six Shuttle launches to bring the spacecraft and a Centaur G upper-stage vehicle to assembly orbit. After assembly, the Centaur places the spacecraft in operational orbit, where it is deployed on radio command, the power system started, and the spacecraft becomes operational. Electric propulsion is an alternative and allows deployment in assembly orbit, but introduces a question of nuclear safety.

Space nuclear reactors — A post-operational disposal strategy

NASA Astrophysics Data System (ADS)

Angelo, Joseph A.; Buden, David

If 100-kWe and multimegawatt-electric class space nuclear reactors are to play a significant role in humanity's push into cislunar and heliocentric space in the next millennium, the obvious advantages of space nuclear power plants should not be denied to space mission planners due to a failure to develop internationally-acceptable post-operational disposal strategies for spent reactor cores. This is true whether the space reactor has shut down at the end of its normal mission lifetime or in response to an onboard system failure/emergency which causes a premature mission termination. Up until now the great majority of aerospace nuclear safety efforts have concentrated on prelaunch, launch and reactor startup activities. In fact, with the exception of the development of the "nuclear safe orbit" (NSO) concept, little technical attention has yet been given to the post-operational disposal of future space reactors. This paper describes the technical alternatives available for the safe, acceptable disposal of space reactors that could be used in a wide variety of space applications in the 21st Century. Post-operational core radioactivity levels for typical advanced design (hundred kWe-class) space reactors are presented as a function of decay time and contrasted to the spent core radionuclide inventory of the SNAP-10A system, the only nuclear reactor operated in space by the United States. The role of a permanent space station, smart robotic systems, and an operating lunar base in support of spent core disposal strategies is also presented, including use of a selected portion of the lunar surface as an internationally-designated spent reactor core repository.

Improved Nuclear Reactor and Shield Mass Model for Space Applications

NASA Technical Reports Server (NTRS)

Robb, Kevin

2004-01-01

New technologies are being developed to explore the distant reaches of the solar system. Beyond Mars, solar energy is inadequate to power advanced scientific instruments. One technology that can meet the energy requirements is the space nuclear reactor. The nuclear reactor is used as a heat source for which a heat-to-electricity conversion system is needed. Examples of such conversion systems are the Brayton, Rankine, and Stirling cycles. Since launch cost is proportional to the amount of mass to lift, mass is always a concern in designing spacecraft. Estimations of system masses are an important part in determining the feasibility of a design. I worked under Michael Barrett in the Thermal Energy Conversion Branch of the Power & Electric Propulsion Division. An in-house Closed Cycle Engine Program (CCEP) is used for the design and performance analysis of closed-Brayton-cycle energy conversion systems for space applications. This program also calculates the system mass including the heat source. CCEP uses the subroutine RSMASS, which has been updated to RSMASS-D, to estimate the mass of the reactor. RSMASS was developed in 1986 at Sandia National Laboratories to quickly estimate the mass of multi-megawatt nuclear reactors for space applications. In response to an emphasis for lower power reactors, RSMASS-D was developed in 1997 and is based off of the SP-100 liquid metal cooled reactor. The subroutine calculates the mass of reactor components such as the safety systems, instrumentation and control, radiation shield, structure, reflector, and core. The major improvements in RSMASS-D are that it uses higher fidelity calculations, is easier to use, and automatically optimizes the systems mass. RSMASS-D is accurate within 15% of actual data while RSMASS is only accurate within 50%. My goal this summer was to learn FORTRAN 77 programming language and update the CCEP program with the RSMASS-D model.

Sustainable Thorium Nuclear Fuel Cycles: A Comparison of Intermediate and Fast Neutron Spectrum Systems

DOE PAGES

Brown, Nicholas R.; Powers, Jeffrey J.; Feng, B.; ...

2015-05-21

This paper presents analyses of possible reactor representations of a nuclear fuel cycle with continuous recycling of thorium and produced uranium (mostly U-233) with thorium-only feed. The analysis was performed in the context of a U.S. Department of Energy effort to develop a compendium of informative nuclear fuel cycle performance data. The objective of this paper is to determine whether intermediate spectrum systems, having a majority of fission events occurring with incident neutron energies between 1 eV and 10 5 eV, perform as well as fast spectrum systems in this fuel cycle. The intermediate spectrum options analyzed include tight latticemore » heavy or light water-cooled reactors, continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. All options were modeled in reactor physics codes to calculate their lattice physics, spectrum characteristics, and fuel compositions over time. Based on these results, detailed metrics were calculated to compare the fuel cycle performance. These metrics include waste management and resource utilization, and are binned to accommodate uncertainties. The performance of the intermediate systems for this selfsustaining thorium fuel cycle was similar to a representative fast spectrum system. However, the number of fission neutrons emitted per neutron absorbed limits performance in intermediate spectrum systems.« less

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.

Modeling moving systems with RELAP5-3D

DOE PAGES

Mesina, G. L.; Aumiller, David L.; Buschman, Francis X.; ...

2015-12-04

RELAP5-3D is typically used to model stationary, land-based reactors. However, it can also model reactors in other inertial and accelerating frames of reference. By changing the magnitude of the gravitational vector through user input, RELAP5-3D can model reactors on a space station or the moon. The field equations have also been modified to model reactors in a non-inertial frame, such as occur in land-based reactors during earthquakes or onboard spacecraft. Transient body forces affect fluid flow in thermal-fluid machinery aboard accelerating crafts during rotational and translational accelerations. It is useful to express the equations of fluid motion in the acceleratingmore » frame of reference attached to the moving craft. However, careful treatment of the rotational and translational kinematics is required to accurately capture the physics of the fluid motion. Correlations for flow at angles between horizontal and vertical are generated via interpolation where no experimental studies or data exist. The equations for three-dimensional fluid motion in a non-inertial frame of reference are developed. As a result, two different systems for describing rotational motion are presented, user input is discussed, and an example is given.« less

Fuel Sustainability And Actinide Production Of Doping Minor Actinide In Water-Cooled Thorium Reactor

NASA Astrophysics Data System (ADS)

Permana, Sidik

2017-07-01

Fuel sustainability of nuclear energy is coming from an optimum fuel utilization of the reactor and fuel breeding program. Fuel cycle option becomes more important for fuel cycle utilization as well as fuel sustainability capability of the reactor. One of the important issues for recycle fuel option is nuclear proliferation resistance issue due to production plutonium. To reduce the proliferation resistance level, some barriers were used such as matrial barrier of nuclear fuel based on isotopic composition of even mass number of plutonium isotope. Analysis on nuclear fuel sustainability and actinide production composition based on water-cooled thorium reactor system has been done and all actinide composition are recycled into the reactor as a basic fuel cycle scheme. Some important parameters are evaluated such as doping composition of minor actinide (MA) and volume ratio of moderator to fuel (MFR). Some feasible parameters of breeding gains have been obtained by additional MA doping and some less moderation to fuel ratios (MFR). The system shows that plutonium and MA are obtained low compositions and it obtains some higher productions of even mass plutonium, which is mainly Pu-238 composition, as a control material to protect plutonium to be used as explosive devices.

An innovative online VFA monitoring system for the anerobic process, based on headspace gas chromatography.

PubMed

Boe, Kanokwan; Batstone, Damien John; Angelidaki, Irini

2007-03-01

A new method for online measurement of volatile fatty acids (VFA) in anerobic digesters has been developed based on headspace gas chromatography (HSGC). The method applies ex situ VFA stripping with variable headspace volume and gas analysis by gas chromatography-flame ionization detection (GC-FID). In each extraction, digester sample was acidified with H(3)PO(4) and NaHSO(4), then heated to strip the VFA into the gas phase. The gas was sampled in a low friction glass syringe before injected into the GC for measurement. The system has been tested for online monitoring of a lab-scale CSTR reactor treating manure for more than 6 months and has shown good agreement with off-line analysis. The system is capable of measuring individual VFA components. This is of advantage since specific VFA components such as propionic and butyric acid can give extra information about the process status. Another important advantage of this sensor is that there is no filtration, which makes possible application in high solids environments. The system can thus be easily applied in a full-scale biogas reactor by connecting the system to the liquid circulation loop to obtain fresh sample from the reactor. Local calibration is needed but automatic calibration is also possible using standard addition method. Sampling duration is 25-40 min, depending on the washing duration, and sensor response is 10 min. This is appropriate for full-scale reactors, since dynamics within most biogas reactors are of the order of several hours.

Integration of a photocatalytic multi-tube reactor for indoor air purification in HVAC systems: a feasibility study.

PubMed

van Walsem, Jeroen; Roegiers, Jelle; Modde, Bart; Lenaerts, Silvia; Denys, Siegfried

2018-04-24

This work is focused on an in-depth experimental characterization of multi-tube reactors for indoor air purification integrated in ventilation systems. Glass tubes were selected as an excellent photocatalyst substrate to meet the challenging requirements of the operating conditions in a ventilation system in which high flow rates are typical. Glass tubes show a low-pressure drop which reduces the energy demand of the ventilator, and additionally, they provide a large exposed surface area to allow interaction between indoor air contaminants and the photocatalyst. Furthermore, the performance of a range of P25-loaded sol-gel coatings was investigated, based on their adhesion properties and photocatalytic activities. Moreover, the UV light transmission and photocatalytic reactor performance under various operating conditions were studied. These results provide vital insights for the further development and scaling up of multi-tube reactors in ventilation systems which can provide a better comfort, improved air quality in indoor environments, and reduced human exposure to harmful pollutants.

Comparisons of selected laser beam power missions to conventionally powered missions

NASA Technical Reports Server (NTRS)

Bozek, John M.; Oleson, Steven R.; Landis, Geoffrey A.; Stavnes, Mark W.

1993-01-01

Earth-based laser sites beaming laser power to space assets have shown benefits over competing power system concepts for specific missions. Missions analyzed in this report that show benefits of laser beam power are low Earth orbit (LEO) to geosynchronous Earth orbit (GEO) transfer, LEO to low lunar orbit (LLO) cargo missions, and lunar-base power. Both laser- and solar-powered orbit-transfer vehicles (OTV's) make a 'tug' concept viable, which substantially reduces cumulative initial mass to LEO in comparison to chemical propulsion concepts. Lunar cargo missions utilizing laser electric propulsion from Earth-orbit to LLO show substantial mass saving to LEO over chemical propulsion systems. Lunar-base power system options were compared on a landed-mass basis. Photovoltaics with regenerative fuel cells, reactor-based systems, and laser-based systems were sized to meet a generic lunar-base power profile. A laser-based system begins to show landed mass benefits over reactor-based systems when proposed production facilities on the Moon require power levels greater than approximately 300 kWe. Benefit/cost ratios of laser power systems for an OTV, both to GEO and LLO, and for a lunar base were calculated to be greater than 1.

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

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

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

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.

Batch-reactor microfluidic device: first human use of a microfluidically produced PET radiotracer†

PubMed Central

Miraghaie, Reza; Kotta, Kishore; Ball, Carroll E.; Zhang, Jianzhong; Buchsbaum, Monte S.; Kolb, Hartmuth C.; Elizarov, Arkadij

2013-01-01

The very first microfluidic device used for the production of 18F-labeled tracers for clinical research is reported along with the first human Positron Emission Tomography scan obtained with a microfluidically produced radiotracer. The system integrates all operations necessary for the transformation of [18F]fluoride in irradiated cyclotron target water to a dose of radiopharmaceutical suitable for use in clinical research. The key microfluidic technologies developed for the device are a fluoride concentration system and a microfluidic batch reactor assembly. Concentration of fluoride was achieved by means of absorption of the fluoride anion on a micro ion-exchange column (5 μL of resin) followed by release of the radioactivity with 45 μL of the release solution (95 ± 3% overall efficiency). The reactor assembly includes an injection-molded reactor chip and a transparent machined lid press-fitted together. The resulting 50 μL cavity has a unique shape designed to minimize losses of liquid during reactor filling and liquid evaporation. The cavity has 8 ports for gases and liquids, each equipped with a 2-way on-chip mechanical valve rated for pressure up to 20.68 bar (300 psi). The temperature is controlled by a thermoelectric heater capable of heating the reactor up to 180 °C from RT in 150 s. A camera captures live video of the processes in the reactor. HPLC-based purification and reformulation units are also integrated in the device. The system is based on “split-box architecture”, with reagents loaded from outside of the radiation shielding. It can be installed either in a standard hot cell, or as a self-shielded unit. Along with a high level of integration and automation, split-box architecture allowed for multiple production runs without the user being exposed to radiation fields. The system was used to support clinical trials of [18F]fallypride, a neuroimaging radiopharmaceutical under IND Application #109,880. PMID:23135409

Analysis of JKT01 Neutron Flux Detector Measurements In RSG-GAS Reactor Using LabVIEW

NASA Astrophysics Data System (ADS)

Rokhmadi; Nur Rachman, Agus; Sujarwono; Taryo, Taswanda; Sunaryo, Geni Rina

2018-02-01

The RSG-GAS Reactor, one of the Indonesia research reactors and located in Serpong, is owned by the National Nuclear Energy Agency (BATAN). The RSG-GAS reactor has operated since 1987 and some instrumentation and control systems are considered to be degraded and ageing. It is therefore, necessary to evaluate the safety of all instrumentation and controls and one of the component systems to be evaluated is the performance of JKT01 neutron flux detector. Neutron Flux Detector JKT01 basically detects neutron fluxes in the reactor core and converts it into electrical signals. The electrical signal is then forwarded to the amplifier (Amplifier) to become the input of the reactor protection system. One output of it is transferred to the Main Control Room (RKU) showing on the analog meter as an indicator used by the reactor operator. To simulate all of this matter, a program to simulate the output of the JKT01 Neutron Flux Detector using LabVIEW was developed. The simulated data is estimated using a lot of equations also formulated in LabVIEW. The calculation results are also displayed on the interface using LabVIEW available in the PC. By using this simulation program, it is successful to perform anomaly detection experiments on the JKT01 detector of RSG-GAS Reactor. The simulation results showed that the anomaly JKT01 neutron flux using electrical-current-base are respectively, 1.5×,1.7× and 2.0×.

Design of neural network model-based controller in a fed-batch microbial electrolysis cell reactor for bio-hydrogen gas production

NASA Astrophysics Data System (ADS)

Azwar; Hussain, M. A.; Abdul-Wahab, A. K.; Zanil, M. F.; Mukhlishien

2018-03-01

One of major challenge in bio-hydrogen production process by using MEC process is nonlinear and highly complex system. This is mainly due to the presence of microbial interactions and highly complex phenomena in the system. Its complexity makes MEC system difficult to operate and control under optimal conditions. Thus, precise control is required for the MEC reactor, so that the amount of current required to produce hydrogen gas can be controlled according to the composition of the substrate in the reactor. In this work, two schemes for controlling the current and voltage of MEC were evaluated. The controllers evaluated are PID and Inverse neural network (NN) controller. The comparative study has been carried out under optimal condition for the production of bio-hydrogen gas wherein the controller output is based on the correlation of optimal current and voltage to the MEC. Various simulation tests involving multiple set-point changes and disturbances rejection have been evaluated and the performances of both controllers are discussed. The neural network-based controller results in fast response time and less overshoots while the offset effects are minimal. In conclusion, the Inverse neural network (NN)-based controllers provide better control performance for the MEC system compared to the PID controller.

Co3O4-based honeycombs as compact redox reactors/heat exchangers for thermochemical storage in the next generation CSP plants

NASA Astrophysics Data System (ADS)

Pagkoura, Chrysoula; Karagiannakis, George; Halevas, Eleftherios; Konstandopoulos, Athanasios G.

2016-05-01

Over the last years, several research groups have focused on developing efficient thermochemical heat storage (THS) systems, in-principle capable of being coupled with next generation high temperature Concentrated Solar Power plants. Among systems studied, the Co3O4/CoO redox system is a promising candidate. Currently, research efforts extend beyond basic level identification of promising materials to more application-oriented approaches aiming at validation of THS performance at pilot scale reactors. The present work focuses on the investigation of cobalt oxide based honeycomb structures as candidate reactors/heat exchangers to be employed for such purposes. In the evaluation conducted and presented here, cobalt oxide-based structures with different composition and geometrical characteristics were subjected to redox cycles in the temperature window between 800 and 1000°C under air flow. Basic aspects related to redox performance of each system are briefly discussed but the main focus lies on the evaluation of the segments structural stability after multi-cyclic operation. The latter is based on macroscopic visual observation and also supplemented by pre- (i.e. fresh samples) and post-characterization (i.e. after long term exposure) of extruded honeycombs via combined mercury porosimetry and SEM analysis.

Alternate fusion fuels workshop

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

Not Available

1981-06-01

The workshop was organized to focus on a specific confinement scheme: the tokamak. The workshop was divided into two parts: systems and physics. The topics discussed in the systems session were narrowly focused on systems and engineering considerations in the tokamak geometry. The workshop participants reviewed the status of system studies, trade-offs between d-t and d-d based reactors and engineering problems associated with the design of a high-temperature, high-field reactor utilizing advanced fuels. In the physics session issues were discussed dealing with high-beta stability, synchrotron losses and transport in alternate fuel systems. The agenda for the workshop is attached.

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

NASA Technical Reports Server (NTRS)

Srivastava, R.; Gould, R. K.

1979-01-01

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

Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

DOE PAGES

Cheng, Lap-Yan; Wei, Thomas Y. C.

2009-01-01

The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow weremore » evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.« less

Manned space flight nuclear system safety. Volume 3: Reactor system preliminary nuclear safety analysis. Part 2: Accident Model Document (AMD)

NASA Technical Reports Server (NTRS)

1972-01-01

The Accident Model Document is one of three documents of the Preliminary Safety Analysis Report (PSAR) - Reactor System as applied to a Space Base Program. Potential terrestrial nuclear hazards involving the zirconium hydride reactor-Brayton power module are identified for all phases of the Space Base program. The accidents/events that give rise to the hazards are defined and abort sequence trees are developed to determine the sequence of events leading to the hazard and the associated probabilities of occurence. Source terms are calculated to determine the magnitude of the hazards. The above data is used in the mission accident analysis to determine the most probable and significant accidents/events in each mission phase. The only significant hazards during the prelaunch and launch ascent phases of the mission are those which arise form criticality accidents. Fission product inventories during this time period were found to be very low due to very limited low power acceptance testing.

Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

NASA Astrophysics Data System (ADS)

Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.

2017-01-01

In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

Fuel development for gas-cooled fast reactors

NASA Astrophysics Data System (ADS)

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

2007-09-01

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

User Guide for VISION 3.4.7 (Verifiable Fuel Cycle Simulation) Model

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

Jacob J. Jacobson; Robert F. Jeffers; Gretchen E. Matthern

2011-07-01

The purpose of this document is to provide a guide for using the current version of the Verifiable Fuel Cycle Simulation (VISION) model. This is a complex model with many parameters and options; the user is strongly encouraged to read this user guide before attempting to run the model. This model is an R&D work in progress and may contain errors and omissions. It is based upon numerous assumptions. This model is intended to assist in evaluating 'what if' scenarios and in comparing fuel, reactor, and fuel processing alternatives at a systems level. The model is not intended as amore » tool for process flow and design modeling of specific facilities nor for tracking individual units of fuel or other material through the system. The model is intended to examine the interactions among the components of a fuel system as a function of time varying system parameters; this model represents a dynamic rather than steady-state approximation of the nuclear fuel system. VISION models the nuclear cycle at the system level, not individual facilities, e.g., 'reactor types' not individual reactors and 'separation types' not individual separation plants. Natural uranium can be enriched, which produces enriched uranium, which goes into fuel fabrication, and depleted uranium (DU), which goes into storage. Fuel is transformed (transmuted) in reactors and then goes into a storage buffer. Used fuel can be pulled from storage into either separation or disposal. If sent to separations, fuel is transformed (partitioned) into fuel products, recovered uranium, and various categories of waste. Recycled material is stored until used by its assigned reactor type. VISION is comprised of several Microsoft Excel input files, a Powersim Studio core, and several Microsoft Excel output files. All must be co-located in the same folder on a PC to function. You must use Powersim Studio 8 or better. We have tested VISION with the Studio 8 Expert, Executive, and Education versions. The Expert and Education versions work with the number of reactor types of 3 or less. For more reactor types, the Executive version is currently required. The input files are Excel2003 format (xls). The output files are macro-enabled Excel2007 format (xlsm). VISION 3.4 was designed with more flexibility than previous versions, which were structured for only three reactor types - LWRs that can use only uranium oxide (UOX) fuel, LWRs that can use multiple fuel types (LWR MF), and fast reactors. One could not have, for example, two types of fast reactors concurrently. The new version allows 10 reactor types and any user-defined uranium-plutonium fuel is allowed. (Thorium-based fuels can be input but several features of the model would not work.) The user identifies (by year) the primary fuel to be used for each reactor type. The user can identify for each primary fuel a contingent fuel to use if the primary fuel is not available, e.g., a reactor designated as using mixed oxide fuel (MOX) would have UOX as the contingent fuel. Another example is that a fast reactor using recycled transuranic (TRU) material can be designated as either having or not having appropriately enriched uranium oxide as a contingent fuel. Because of the need to study evolution in recycling and separation strategies, the user can now select the recycling strategy and separation technology, by year.« less

Impact of Bioreactor Environment and Recovery Method on the Profile of Bacterial Populations from Water Distribution Systems.

PubMed

Luo, Xia; Jellison, Kristen L; Huynh, Kevin; Widmer, Giovanni

2015-01-01

Multiple rotating annular reactors were seeded with biofilms flushed from water distribution systems to assess (1) whether biofilms grown in bioreactors are representative of biofilms flushed from the water distribution system in terms of bacterial composition and diversity, and (2) whether the biofilm sampling method affects the population profile of the attached bacterial community. Biofilms were grown in bioreactors until thickness stabilized (9 to 11 weeks) and harvested from reactor coupons by sonication, stomaching, bead-beating, and manual scraping. High-throughput sequencing of 16S rRNA amplicons was used to profile bacterial populations from flushed biofilms seeded into bioreactors as well as biofilms recovered from bioreactor coupons by different methods. β diversity between flushed and reactor biofilms was compared to β diversity between (i) biofilms harvested from different reactors and (ii) biofilms harvested by different methods from the same reactor. These analyses showed that average diversity between flushed and bioreactor biofilms was double the diversity between biofilms from different reactors operated in parallel. The diversity between bioreactors was larger than the diversity associated with different biofilm recovery methods. Compared to other experimental variables, the method used to recover biofilms had a negligible impact on the outcome of water biofilm analyses based on 16S amplicon sequencing. Results from this study show that biofilms grown in reactors over 9 to 11 weeks are not representative models of the microbial populations flushed from a distribution system. Furthermore, the bacterial population profile of biofilms grown in replicate reactors from the same flushed water are likely to diverge. However, four common sampling protocols, which differ with respect to disruption of bacterial cells, provide similar information with respect to the 16S rRNA population profile of the biofilm community.

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.

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.

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.

Scaling analysis for the direct reactor auxiliary cooling system for FHRs

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

Lv, Q.; Kim, I. H.; Sun, X.

2015-04-01

The Direct Reactor Auxiliary Cooling System (DRACS) is a passive residual heat removal system proposed for the Fluoride-salt-cooled High-temperature Reactor (FHR) that combines the coated particle fuel and graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three natural circulation/convection loops that rely on buoyancy as the driving force and are coupled via two heat exchangers, namely, the DRACS heat exchanger and the natural draft heat exchanger. A fluidic diode is employed to minimize the parasitic flow into the DRACS primary loop and correspondingly the heat loss to the DRACS during reactor normal operation, and tomore » activate the DRACS in accidents when the reactor is shut down. While the DRACS concept has been proposed, there are no actual prototypic DRACS systems for FHRs built or tested in the literature. In this paper, a detailed scaling analysis for the DRACS is performed, which will provide guidance for the design of scaled-down DRACS test facilities. Based on the Boussinesq assumption and one-dimensional flow formulation, the governing equations are non-dimensionalized by introducing appropriate dimensionless parameters. The key dimensionless numbers that characterize the DRACS system are obtained from the non-dimensional governing equations. Based on the dimensionless numbers and non-dimensional governing equations, similarity laws are proposed. In addition, a scaling methodology has been developed, which consists of a core scaling and a loop scaling. The consistency between the core and loop scaling is examined via the reference volume ratio, which can be obtained from both the core and loop scaling processes. The scaling methodology and similarity laws have been applied to obtain a scientific design of a scaled-down high-temperature DRACS test facility.« less

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.

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

Amitava Sarkar; James K. Neathery; Burtron H. Davis

A fundamental filtration study was started to investigate the separation of Fischer-Tropsch Synthesis (FTS) liquids from iron-based catalyst particles. Slurry-phase FTS in slurry bubble column reactor systems is the preferred mode of operation since the reaction is highly exothermic. Consequently, heavy wax products in one approach may be separated from catalyst particles before being removed from the reactor system. Achieving an efficient wax product separation from iron-based catalysts is one of the most challenging technical problems associated with slurry-phase iron-based FTS and is a key factor for optimizing operating costs. The separation problem is further compounded by attrition of ironmore » catalyst particles and the formation of ultra-fine particles.« less

Demand driven salt clean-up in a molten salt fast reactor - Defining a priority list.

PubMed

Merk, B; Litskevich, D; Gregg, R; Mount, A R

2018-01-01

The PUREX technology based on aqueous processes is currently the leading reprocessing technology in nuclear energy systems. It seems to be the most developed and established process for light water reactor fuel and the use of solid fuel. However, demand driven development of the nuclear system opens the way to liquid fuelled reactors, and disruptive technology development through the application of an integrated fuel cycle with a direct link to reactor operation. The possibilities of this new concept for innovative reprocessing technology development are analysed, the boundary conditions are discussed, and the economic as well as the neutron physical optimization parameters of the process are elucidated. Reactor physical knowledge of the influence of different elements on the neutron economy of the reactor is required. Using an innovative study approach, an element priority list for the salt clean-up is developed, which indicates that separation of Neodymium and Caesium is desirable, as they contribute almost 50% to the loss of criticality. Separating Zirconium and Samarium in addition from the fuel salt would remove nearly 80% of the loss of criticality due to fission products. The theoretical study is followed by a qualitative discussion of the different, demand driven optimization strategies which could satisfy the conflicting interests of sustainable reactor operation, efficient chemical processing for the salt clean-up, and the related economic as well as chemical engineering consequences. A new, innovative approach of balancing the throughput through salt processing based on a low number of separation process steps is developed. Next steps for the development of an economically viable salt clean-up process are identified.

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.

Gamma thermometer based reactor core liquid level detector

DOEpatents

Burns, Thomas J.

1983-01-01

A system is provided which employs a modified gamma thermometer for determining the liquid coolant level within a nuclear reactor core. The gamma thermometer which normally is employed to monitor local core heat generation rate (reactor power), is modified by thermocouple junctions and leads to obtain an unambiguous indication of the presence or absence of coolant liquid at the gamma thermometer location. A signal processor generates a signal based on the thermometer surface heat transfer coefficient by comparing the signals from the thermocouples at the thermometer location. The generated signal is a direct indication of loss of coolant due to the change in surface heat transfer when coolant liquid drops below the thermometer location. The loss of coolant indication is independent of reactor power at the thermometer location. Further, the same thermometer may still be used for the normal power monitoring function.

Double Retort System for Materials Compatibility Testing

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

V. Munne; EV Carelli

2006-02-23

With Naval Reactors (NR) approval of the Naval Reactors Prime Contractor Team (NRPCT) recommendation to develop a gas cooled reactor directly coupled to a Brayton power conversion system as the Space Nuclear Power Plant (SNPP) for Project Prometheus (References a and b) there was a need to investigate compatibility between the various materials to be used throughout the SNPP. Of particular interest was the transport of interstitial impurities from the nickel-base superalloys, which were leading candidates for most of the piping and turbine components to the refractory metal alloys planned for use in the reactor core. This kind of contaminationmore » has the potential to affect the lifetime of the core materials. This letter provides technical information regarding the assembly and operation of a double retort materials compatibility testing system and initial experimental results. The use of a double retort system to test materials compatibility through the transfer of impurities from a source to a sink material is described here. The system has independent temperature control for both materials and is far less complex than closed loops. The system is described in detail and the results of three experiments are presented.« less

Process of simultaneous hydrogen sulfide removal from biogas and nitrogen removal from swine wastewater.

PubMed

Deng, Liangwei; Chen, Huijuan; Chen, Ziai; Liu, Yi; Pu, Xiaodong; Song, Li

2009-12-01

The feasibility of a new flowchart describing simultaneous hydrogen sulfide removal from biogas and nitrogen removal from wastewater was investigated. It took 30 days for the reactor inoculated with aerobic sludge to attain a removal rate of 60% for H(2)S and NO(x)-N simultaneously. It took 34 and 48 days to attain the same removal rate for the reactor without inoculated sludge and the reactor inoculated with anaerobic sludge respectively. The reactor without inoculated sludge still operated successfully, despite requiring a slightly longer startup time. The packing material was capable of enhancing the removal efficiency of reactors. Based on the concentration of NO(x)-N and H(2)S in the effluent, the loading rate and the ability of the system to resist shock loading, the performance of the reactor filled with hollow plastic balls was greater than that of the reactor filled with elastic packing and the reactor filled with Pall rings.

A fuzzy-logic-based controller for methane production in anaerobic fixed-film reactors.

PubMed

Robles, A; Latrille, E; Ruano, M V; Steyer, J P

2017-01-01

The main objective of this work was to develop a controller for biogas production in continuous anaerobic fixed-bed reactors, which used effluent total volatile fatty acids (VFA) concentration as control input in order to prevent process acidification at closed loop. To this aim, a fuzzy-logic-based control system was developed, tuned and validated in an anaerobic fixed-bed reactor at pilot scale that treated industrial winery wastewater. The proposed controller varied the flow rate of wastewater entering the system as a function of the gaseous outflow rate of methane and VFA concentration. Simulation results show that the proposed controller is capable to achieve great process stability even when operating at high VFA concentrations. Pilot results showed the potential of this control approach to maintain the process working properly under similar conditions to the ones expected at full-scale plants.

Applications of power beaming from space-based nuclear power stations. [Laser beaming to airplanes; microwave beaming to ground

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

Powell, J.R.; Botts, T.E.; Hertzberg, A.

1981-01-01

Power beaming from space-based reactor systems is examined using an advanced compact, lightweight Rotating Bed Reactor (RBR). Closed Brayton power conversion efficiencies in the range of 30 to 40% can be achieved with turbines, with reactor exit temperatures on the order of 2000/sup 0/K and a liquid drop radiator to reject heat at temperatures of approx. 500/sup 0/K. Higher RBR coolant temperatures (up to approx. 3000/sup 0/K) are possible, but gains in power conversion efficiency are minimal, due to lower expander efficiency (e.g., a MHD generator). Two power beaming applications are examined - laser beaming to airplanes and microwave beamingmore » to fixed ground receivers. Use of the RBR greatly reduces system weight and cost, as compared to solar power sources. Payback times are a few years at present prices for power and airplane fuel.« less

A Comparison of Brayton and Stirling Space Nuclear Power Systems for Power Levels from 1 Kilowatt to 10 Megawatts

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2000-01-01

An analytical study was conducted to assess the performance and mass of Brayton and Stirling nuclear power systems for a wide range of future NASA space exploration missions. The power levels and design concepts were based on three different mission classes. Isotope systems, with power levels from 1 to 10 kW, were considered for planetary surface rovers and robotic science. Reactor power systems for planetary surface outposts and bases were evaluated from 10 to 500 kW. Finally, reactor power systems in the range from 100 kW to 10 mW were assessed for advanced propulsion applications. The analysis also examined the effect of advanced component technology on system performance. The advanced technologies included high temperature materials, lightweight radiators, and high voltage power management and distribution.

Evaluating the Cost, Safety, and Proliferation Risks of Small Floating Nuclear Reactors.

PubMed

Ford, Michael J; Abdulla, Ahmed; Morgan, M Granger

2017-11-01

It is hard to see how our energy system can be decarbonized if the world abandons nuclear power, but equally hard to introduce the technology in nonnuclear energy states. This is especially true in countries with limited technical, institutional, and regulatory capabilities, where safety and proliferation concerns are acute. Given the need to achieve serious emissions mitigation by mid-century, and the multidecadal effort required to develop robust nuclear governance institutions, we must look to other models that might facilitate nuclear plant deployment while mitigating the technology's risks. One such deployment paradigm is the build-own-operate-return model. Because returning small land-based reactors containing spent fuel is infeasible, we evaluate the cost, safety, and proliferation risks of a system in which small modular reactors are manufactured in a factory, and then deployed to a customer nation on a floating platform. This floating small modular reactor would be owned and operated by a single entity and returned unopened to the developed state for refueling. We developed a decision model that allows for a comparison of floating and land-based alternatives considering key International Atomic Energy Agency plant-siting criteria. Abandoning onsite refueling is beneficial, and floating reactors built in a central facility can potentially reduce the risk of cost overruns and the consequences of accidents. However, if the floating platform must be built to military-grade specifications, then the cost would be much higher than a land-based system. The analysis tool presented is flexible, and can assist planners in determining the scope of risks and uncertainty associated with different deployment options. © 2017 Society for Risk Analysis.

Fuel cycle for a fusion neutron source

NASA Astrophysics Data System (ADS)

Ananyev, S. S.; Spitsyn, A. V.; Kuteev, B. V.

2015-12-01

The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion-fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium-tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m3Pa/s, and temperature of reactor elements up to 650°C). The deuterium-tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.

Optical Sensors for Monitoring Gamma and Neutron Radiation

NASA Technical Reports Server (NTRS)

Boyd, Clark D.

2011-01-01

For safety and efficiency, nuclear reactors must be carefully monitored to provide feedback that enables the fission rate to be held at a constant target level via adjustments in the position of neutron-absorbing rods and moderating coolant flow rates. For automated reactor control, the monitoring system should provide calibrated analog or digital output. The sensors must survive and produce reliable output with minimal drift for at least one to two years, for replacement only during refueling. Small sensor size is preferred to enable more sensors to be placed in the core for more detailed characterization of the local fission rate and fuel consumption, since local deviations from the norm tend to amplify themselves. Currently, reactors are monitored by local power range meters (LPRMs) based on the neutron flux or gamma thermometers based on the gamma flux. LPRMs tend to be bulky, while gamma thermometers are subject to unwanted drift. Both electronic reactor sensors are plagued by electrical noise induced by ionizing radiation near the reactor core. A fiber optic sensor system was developed that is capable of tracking thermal neutron fluence and gamma flux in order to monitor nuclear reactor fission rates. The system provides near-real-time feedback from small- profile probes that are not sensitive to electromagnetic noise. The key novel feature is the practical design of fiber optic radiation sensors. The use of an actinoid element to monitor neutron flux in fiber optic EFPI (extrinsic Fabry-Perot interferometric) sensors is a new use of material. The materials and structure used in the sensor construction can be adjusted to result in a sensor that is sensitive to just thermal, gamma, or neutron stimulus, or any combination of the three. The tested design showed low sensitivity to thermal and gamma stimuli and high sensitivity to neutrons, with a fast response time.

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

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

Buden, D.

1993-08-11

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

Summary of space nuclear reactor power systems, 1983 - 1992

NASA Astrophysics Data System (ADS)

Buden, D.

1993-08-01

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

Validation of large-scale, monochromatic UV disinfection systems for drinking water using dyed microspheres.

PubMed

Blatchley, E R; Shen, C; Scheible, O K; Robinson, J P; Ragheb, K; Bergstrom, D E; Rokjer, D

2008-02-01

Dyed microspheres have been developed as a new method for validation of ultraviolet (UV) reactor systems. When properly applied, dyed microspheres allow measurement of the UV dose distribution delivered by a photochemical reactor for a given operating condition. Prior to this research, dyed microspheres had only been applied to a bench-scale UV reactor. The goal of this research was to extend the application of dyed microspheres to large-scale reactors. Dyed microsphere tests were conducted on two prototype large-scale UV reactors at the UV Validation and Research Center of New York (UV Center) in Johnstown, NY. All microsphere tests were conducted under conditions that had been used previously in biodosimetry experiments involving two challenge bacteriophage: MS2 and Qbeta. Numerical simulations based on computational fluid dynamics and irradiance field modeling were also performed for the same set of operating conditions used in the microspheres assays. Microsphere tests on the first reactor illustrated difficulties in sample collection and discrimination of microspheres against ambient particles. Changes in sample collection and work-up were implemented in tests conducted on the second reactor that allowed for improvements in microsphere capture and discrimination against the background. Under these conditions, estimates of the UV dose distribution from the microspheres assay were consistent with numerical simulations and the results of biodosimetry, using both challenge organisms. The combined application of dyed microspheres, biodosimetry, and numerical simulation offers the potential to provide a more in-depth description of reactor performance than any of these methods individually, or in combination. This approach also has the potential to substantially reduce uncertainties in reactor validation, thereby leading to better understanding of reactor performance, improvements in reactor design, and decreases in reactor capital and operating costs.

Corrosion and stress corrosion cracking in supercritical water

NASA Astrophysics Data System (ADS)

Was, G. S.; Ampornrat, P.; Gupta, G.; Teysseyre, S.; West, E. A.; Allen, T. R.; Sridharan, K.; Tan, L.; Chen, Y.; Ren, X.; Pister, C.

2007-09-01

Supercritical water (SCW) has attracted increasing attention since SCW boiler power plants were implemented to increase the efficiency of fossil-based power plants. The SCW reactor (SCWR) design has been selected as one of the Generation IV reactor concepts because of its higher thermal efficiency and plant simplification as compared to current light water reactors (LWRs). Reactor operating conditions call for a core coolant temperature between 280 °C and 620 °C at a pressure of 25 MPa and maximum expected neutron damage levels to any replaceable or permanent core component of 15 dpa (thermal reactor design) and 100 dpa (fast reactor design). Irradiation-induced changes in microstructure (swelling, radiation-induced segregation (RIS), hardening, phase stability) and mechanical properties (strength, thermal and irradiation-induced creep, fatigue) are also major concerns. Throughout the core, corrosion, stress corrosion cracking, and the effect of irradiation on these degradation modes are critical issues. This paper reviews the current understanding of the response of candidate materials for SCWR systems, focusing on the corrosion and stress corrosion cracking response, and highlights the design trade-offs associated with certain alloy systems. Ferritic-martensitic steels generally have the best resistance to stress corrosion cracking, but suffer from the worst oxidation. Austenitic stainless steels and Ni-base alloys have better oxidation resistance but are more susceptible to stress corrosion cracking. The promise of grain boundary engineering and surface modification in addressing corrosion and stress corrosion cracking performance is discussed.

Adaptive control method for core power control in TRIGA Mark II reactor

NASA Astrophysics Data System (ADS)

Sabri Minhat, Mohd; Selamat, Hazlina; Subha, Nurul Adilla Mohd

2018-01-01

The 1MWth Reactor TRIGA PUSPATI (RTP) Mark II type has undergone more than 35 years of operation. The existing core power control uses feedback control algorithm (FCA). It is challenging to keep the core power stable at the desired value within acceptable error bands to meet the safety demand of RTP due to the sensitivity of nuclear research reactor operation. Currently, the system is not satisfied with power tracking performance and can be improved. Therefore, a new design core power control is very important to improve the current performance in tracking and regulate reactor power by control the movement of control rods. In this paper, the adaptive controller and focus on Model Reference Adaptive Control (MRAC) and Self-Tuning Control (STC) were applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, adaptive controller model, and control rods selection programming. The mathematical models of the reactor core were based on point kinetics model, thermal hydraulic models, and reactivity models. The adaptive control model was presented using Lyapunov method to ensure stable close loop system and STC Generalised Minimum Variance (GMV) Controller was not necessary to know the exact plant transfer function in designing the core power control. The performance between proposed adaptive control and FCA will be compared via computer simulation and analysed the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

10 CFR Appendix E to Part 50 - Emergency Planning and Preparedness for Production and Utilization Facilities

Code of Federal Regulations, 2011 CFR

2011-01-01

... could communicate with a safety system. In this case, appropriate isolation devices would be required at..., feedwater flow, and reactor power; (2) Safety injection: Reactor core isolation cooling flow, high-pressure... data points identified in the ERDS Data Point Library 9 (site specific data base residing on the ERDS...

10 CFR Appendix E to Part 50 - Emergency Planning and Preparedness for Production and Utilization Facilities

Code of Federal Regulations, 2010 CFR

2010-01-01

... could communicate with a safety system. In this case, appropriate isolation devices would be required at..., feedwater flow, and reactor power; (2) Safety injection: Reactor core isolation cooling flow, high-pressure... data points identified in the ERDS Data Point Library 9 (site specific data base residing on the ERDS...

DEVELOPMENT OF A RATIONALLY BASED DESIGN PROTOCOL FOR THE ULTRAVIOLET LIGHT DISINFECTION PROCESS

EPA Science Inventory

A protocol is demonstrated for the design and evaluation of ultraviolet (UV) disinfection systems based on a mathematical model. The disinfection model incorporates the system's physical dimensions, the residence time distribution of the reactor and dispersion characteristics, th...

A modeling approach to describe ZVI-based anaerobic system.

PubMed

Xiao, Xiao; Sheng, Guo-Ping; Mu, Yang; Yu, Han-Qing

2013-10-15

Zero-valent iron (ZVI) is increasingly being added into anaerobic reactors to enhance the biological conversion of various less biodegradable pollutants (LBPs). Our study aimed to establish a new structure model based on the Anaerobic Digestion Model No. 1 (ADM1) to simulate such a ZVI-based anaerobic reactor. Three new processes, i.e., electron release from ZVI corrosion, H2 formation from ZVI corrosion, and transformation of LBPs, were integrated into ADM1. The established model was calibrated and tested using the experimental data from one published study, and validated using the data from another work. A good relationship between the predicted and measured results indicates that the proposed model was appropriate to describe the performance of the ZVI-based anaerobic system. Our model could provide more precise strategies for the design, development, and application of anaerobic systems especially for treating various LBPs-containing wastewaters. Copyright © 2013 Elsevier Ltd. All rights reserved.

The pre-conceptual design of the nuclear island of ASTRID

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

Saez, M.; Menou, S.; Uzu, B.

The CEA is involved in a substantial effort on the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) pre-conceptual design in cooperation with EDF, as experienced Sodium-cooled Fast Reactor (SFR) operator, AREVA, as experienced SFR Nuclear Island engineering company and components designer, ALSTOM POWER as energy conversion system designer and COMEX NUCLEAIRE as mechanical systems designer. The CEA is looking for other partnerships, in France and abroad. The ASTRID preliminary design is based on a sodium-cooled pool reactor of 1500 MWth generating about 600 MWe, which is required to guarantee the representativeness of the reactor core and the main componentsmore » with regard to future commercial reactors. ASTRID lifetime target is 60 years. Two Energy Conversion Systems are studied in parallel until the end of 2012: Rankine steam cycle or Brayton gas based energy conversion cycle. ASTRID design is guided by the following major objectives: improved safety, simplification of structures, improved In Service Inspection and Repair (ISIR), improved manufacturing conditions for cost reduction and increased quality, reduction of risks related to sodium fires and water/sodium reaction, and improved robustness against external hazards. The core is supported by a diagrid, which lay on a strong back to transfer the weight to the main vessel. AREVA is involved in a substantial effort in order to improve the core support structure in particular regarding the ISIR and the connection to primary pump. In the preliminary design, the primary system is formed by the main vessel and the upper closure comprising the reactor roof, two rotating plugs - used for fuel handling - and the components plugs located in the roof penetrations. The Above Core Structure deflects the sodium flow in the hot pool and provides support to core instrumentation and guidance of the control rod drive mechanisms. The number of the major components in the main vessel, primary pumps, Intermediate Heat Exchangers, and Decay Heat Exchangers are now under consideration. Under normal conditions, power release is achieved using the steam/water plant (in case of Rankine steam cycle) or the gas plant (in case of Brayton gas cycle). The diverse design and operating modes of Decay Heat Removal systems provide protection against common cause failures. A Decay Heat Removal system through the reactor vault is in particular studied with the objective to complement Direct Reactor Cooling systems. At this stage of the studies, the secondary system comprises four independent sodium loops (two and three sodium loops configurations are also investigated). Each loop includes one mechanical pump (or a large capacity Annular Linear Induction Electromagnetic Pump), and three modular Steam Generator Units characterized by once through straight tube units with a ferritic tube bundle; nevertheless, helical coil steam generator with tubes made of Alloy 800, and inverted type steam generator with a ferritic tube bundle are also investigated. The limited power of each modular Steam Generator Unit allows the whole secondary loop to withstand a large water/sodium reaction consecutive to the postulated simultaneous rupture of all the heat exchange tubes of one module. The arrangement of the components is based on the 'Regain' concept, in which the secondary pump is situated at a low level in the circuit; conventional arrangement, as SUPERPHENIX type, is a back-up option. Alternative arrangements based on gas cycles are also studied together with Na-gas heat exchanger design. This paper presents a status of the ASTRID pre-conceptual design. The most promising options are highlighted as well as less risky and back-up options. (authors)« 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.

New approach to control the methanogenic reactor of a two-phase anaerobic digestion system.

PubMed

von Sachs, Jürgen; Meyer, Ulrich; Rys, Paul; Feitkenhauer, Heiko

2003-03-01

A new control strategy for the methanogenic reactor of a two-phase anaerobic digestion system has been developed and successfully tested on the laboratory scale. The control strategy serves the purpose to detect inhibitory effects and to achieve good conversion. The concept is based on the idea that volatile fatty acids (VFA) can be measured in the influent of the methanogenic reactor by means of titration. Thus, information on the output (methane production) and input of the methanogenic reactor is available, and a (carbon) mass balance can be obtained. The control algorithm comprises a proportional/integral structure with the ratio of (a) the methane production rate measured online and (b) a maximum methane production rate expected (derived from the stoichiometry) as a control variable. The manipulated variable is the volumetric feed rate. Results are shown for an experiment with VFA (feed) concentration ramps and for experiments with sodium chloride as inhibitor.

Technical Requirements For Reactors To Be Deployed Internationally For the Global Nuclear Energy Partnership

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

Ingersoll, Daniel T

2007-01-01

Technical Requirements For Reactors To Be Deployed Internationally For the Global Nuclear Energy Partnership Robert Price U.S. Department of Energy, 1000 Independence Ave, SW, Washington, DC 20585, Daniel T. Ingersoll Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6162, INTRODUCTION The Global Nuclear Energy Partnership (GNEP) seeks to create an international regime to support large-scale growth in the worldwide use of nuclear energy. Fully meeting the GNEP vision may require the deployment of thousands of reactors in scores of countries, many of which do not use nuclear energy currently. Some of these needs will be met by large-scalemore » Generation III and III+ reactors (>1000 MWe) and Generation IV reactors when they are available. However, because many developing countries have small and immature electricity grids, the currently available Generation III(+) reactors may be unsuitable since they are too large, too expensive, and too complex. Therefore, GNEP envisions new types of reactors that must be developed for international deployment that are "right sized" for the developing countries and that are based on technologies, designs, and policies focused on reducing proliferation risk. The first step in developing such systems is the generation of technical requirements that will ensure that the systems meet both the GNEP policy goals and the power needs of the recipient countries. REQUIREMENTS Reactor systems deployed internationally within the GNEP context must meet a number of requirements similar to the safety, reliability, economics, and proliferation goals established for the DOE Generation IV program. Because of the emphasis on deployment to nonnuclear developing countries, the requirements will be weighted differently than with Generation IV, especially regarding safety and non-proliferation goals. Also, the reactors should be sized for market conditions in developing countries where energy demand per capita, institutional maturity and industrial infrastructure vary considerably, and must utilize fuel that is compatible with the fuel recycle technologies being developed by GNEP. Arrangements are already underway to establish Working Groups jointly with Japan and Russia to develop requirements for reactor systems. Additional bilateral and multilateral arrangements are expected as GNEP progresses. These Working Groups will be instrumental in establishing an international consensus on reactor system requirements. GNEP CERTIFICATION After establishing an accepted set of requirements for new reactors that are deployed internationally, a mechanism is needed that allows capable countries to continue to market their reactor technologies and services while assuring that they are compatible with GNEP goals and technologies. This will help to preserve the current system of open, commercial competition while steering the international community to meet common policy goals. The proposed vehicle to achieve this is the concept of GNEP Certification. Using objective criteria derived from the technical requirements in several key areas such as safety, security, non-proliferation, and safeguards, reactor designs could be evaluated and then certified if they meet the criteria. This certification would ensure that reactor designs meet internationally approved standards and that the designs are compatible with GNEP assured fuel services. SUMMARY New "right sized" power reactor systems will need to be developed and deployed internationally to fully achieve the GNEP vision of an expanded use of nuclear energy world-wide. The technical requirements for these systems are being developed through national and international Working Groups. The process is expected to culminate in a new GNEP Certification process that enables commercial competition while ensuring that the policy goals of GNEP are adequately met.« less

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

Filamentous bacteria existence in aerobic granular reactors.

PubMed

Figueroa, M; Val del Río, A; Campos, J L; Méndez, R; Mosquera-Corral, A

2015-05-01

Filamentous bacteria are associated to biomass settling problems in wastewater treatment plants. In systems based on aerobic granular biomass they have been proposed to contribute to the initial biomass aggregation process. However, their development on mature aerobic granular systems has not been sufficiently studied. In the present research work, filamentous bacteria were studied for the first time after long-term operation (up to 300 days) of aerobic granular systems. Chloroflexi and Sphaerotilus natans have been observed in a reactor fed with synthetic wastewater. These filamentous bacteria could only come from the inoculated sludge. Thiothrix and Chloroflexi bacteria were observed in aerobic granular biomass treating wastewater from a fish canning industry. Meganema perideroedes was detected in a reactor treating wastewater from a plant processing marine products. As a conclusion, the source of filamentous bacteria in these mature aerobic granular systems fed with industrial effluents was the incoming wastewater.

High Efficiency Nuclear Power Plants Using Liquid Fluoride Thorium Reactor Technology

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

2009-01-01

An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITs of 950 and 1200 K are presented. Power plant performance data were obtained for TITs ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo-generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

RELAP-7 Development Updates

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

Zhang, Hongbin; Zhao, Haihua; Gleicher, Frederick Nathan

RELAP-7 is a nuclear systems safety analysis code being developed at the Idaho National Laboratory, and is the next generation tool in the RELAP reactor safety/systems analysis application series. RELAP-7 development began in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway of the Light Water Reactor Sustainability (LWRS) program. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support nuclear power safety analysis. The code is beingmore » developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (the Multi-Physics Object-Oriented Simulation Environment). The initial development goal of the RELAP-7 approach focused primarily on the development of an implicit algorithm capable of strong (nonlinear) coupling of the dependent hydrodynamic variables contained in the 1-D/2-D flow models with the various 0-D system reactor components that compose various boiling water reactor (BWR) and pressurized water reactor nuclear power plants (NPPs). During Fiscal Year (FY) 2015, the RELAP-7 code has been further improved with expanded capability to support boiling water reactor (BWR) and pressurized water reactor NPPs analysis. The accumulator model has been developed. The code has also been coupled with other MOOSE-based applications such as neutronics code RattleSnake and fuel performance code BISON to perform multiphysics analysis. A major design requirement for the implicit algorithm in RELAP-7 is that it is capable of second-order discretization accuracy in both space and time, which eliminates the traditional first-order approximation errors. The second-order temporal is achieved by a second-order backward temporal difference, and the one-dimensional second-order accurate spatial discretization is achieved with the Galerkin approximation of Lagrange finite elements. During FY-2015, we have done numerical verification work to verify that the RELAP-7 code indeed achieves 2nd-order accuracy in both time and space for single phase models at the system level.« less

Analog to digital converter system for temperature monitoring -- B, C, D, DR, F, and H reactors

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

Ballowe, J.W.

1961-03-23

This document discusses a proposal that certain presently installed reactor process water outlet temperature data logging equipment in subject reactors to be replaced with new functionally simplified equipment of a more adequate design. The primary purpose of the proposed installation is to replace existing equipment which is obsolete and in three reactors is worn out to the point where the equipment is out of service frequently for periods of time up to 8 hours or more. The new equipment will provide reliable process tube temperature information for use in the functions of reactor control and product accountability. Based upon anticipatedmore » incremental production gains resulting from use of the new equipment, the amortization period for the project is calculated at 2.7 years.« less

A Boiling-Potassium Fluoride Reactor for an Artificial-Gravity NEP Vehicle

NASA Technical Reports Server (NTRS)

Sorensen, Kirk; Juhasz, Albert

2007-01-01

Several years ago a rotating manned spacecraft employing nuclear-electric propulsion was examined for Mars exploration. The reactor and its power conversion system essentially served as the counter-mass to an inflatable manned module. A solid-core boiling potassium reactor based on the MPRE concept of the 1960s was baselined in that study. This paper proposes the use of a liquid-fluoride reactor, employing direct boiling of potassium in the core, as a means to overcome some of the residual issues with the MPRE reactor concept. Several other improvements to the rotating Mars vehicle are proposed as well, such as Canfield joints to enable the electric engines to track the inertial thrust vector during rotation, and innovative "cold-ion" engine technologies to improve engine performance.

Structural materials issues for the next generation fission reactors

NASA Astrophysics Data System (ADS)

Chant, I.; Murty, K. L.

2010-09-01

Generation-IV reactor design concepts envisioned thus far cater to a common goal of providing safer, longer lasting, proliferation-resistant, and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-W reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses, and extremely corrosive environments, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This article addresses the material requirements for these advanced fission reactor types, specifically addressing structural materials issues depending on the specific application areas.

SP-100 flight qualification testing assessment

NASA Technical Reports Server (NTRS)

Jeanmougin, Nanette M.; Moore, Roger M.; Wait, David L.; Jacox, Michael G.

1988-01-01

The SP-100 is a compact space power system driven by a nuclear reactor that provides 100 kWe to the user at 200 VDC. The thermal energy generated by the nuclear reactor is converted into electrical energy by passive thermoelectric devices. Various options for tailoring the MIL-STD-1540B guidelines to the SP-100 nuclear power system are discussed. This study aids in selecting the appropriate qualification test program based on the cost, schedule, and test effectiveness of the various options.

Design criteria for a self-actuated shutdown system to ensure limitation of core damage. [LMFBR

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

Deane, N.A.; Atcheson, D.B.

1981-09-01

Safety-based functional requirements and design criteria for a self-actuated shutdown system (SASS) are derived in accordance with LOA-2 success criteria and reliability goals. The design basis transients have been defined and evaluated for the CDS Phase II design, which is a 2550 MWt mixed oxide heterogeneous core reactor. A partial set of reactor responses for selected transients is provided as a function of SASS characteristics such as reactivity worth, trip points, and insertion times.

Startup of RAPID-L Lunar Base Reactor by Lithium Release Module

NASA Astrophysics Data System (ADS)

Kambe, Mitsuru

The 200 kWe uranium-nitride fueled lithium cooled fast reactor concept RAPID-L to be combined with thermoelectric power conversion system for lunar base power system is demonstrated. Unique challenges in reactivity control systems design have been attempted in RAPID-L concept. The reactor involves the following innovative reactivity control systems: Lithium Expansion Modules (LEM) for inherent reactivity feedback, Lithium Injection Modules (LIM) for inherent ultimate shutdown, and Lithium Release Modules (LRM) for automated reactor startup. All these systems adopt lithium-6 as a liquid poison instead of conventional B4C rods or Be reflectors. These systems are effective independent of the magnitude and direction of the gravity force. In 2006, however, the following design amendment has been made. 1) B4C poison rods were added to ensure criticality safety in unintended positive reactivity insertion by LRMs due to fire in the launch phase accident; because LRM freeze seal melts at 800°C which result in positive reactivity insertion. 2) Lower hot standby temperature of 200°C was adopted instead of conventional 800°C to reduce the external power at the startup. In this paper, development of the LRM orifice which dominates the startup transient of RAPID-L is discussed. An attention was focused how to achieve sufficiently small flow rate of 6Li in the orifice because it enables moderate positive reactivity insertion rate. The LRM orifice performance has been confirmed using 0.5 mm diameter SUS316 orifice/lithium flow test setup in the glove box.

Advanced Plasma Pyrolysis Assembly (PPA) Reactor and Process Development

NASA Technical Reports Server (NTRS)

Wheeler, Richard R., Jr.; Hadley, Neal M.; Dahl, Roger W.; Abney, Morgan B.; Greenwood, Zachary; Miller, Lee; Medlen, Amber

2012-01-01

Design and development of a second generation Plasma Pyrolysis Assembly (PPA) reactor is currently underway as part of NASA's Atmosphere Revitalization Resource Recovery effort. By recovering up to 75% of the hydrogen currently lost as methane in the Sabatier reactor effluent, the PPA helps to minimize life support resupply costs for extended duration missions. To date, second generation PPA development has demonstrated significant technology advancements over the first generation device by doubling the methane processing rate while, at the same time, more than halving the required power. One development area of particular interest to NASA system engineers is fouling of the PPA reactor with carbonaceous products. As a mitigation plan, NASA MSFC has explored the feasibility of using an oxidative plasma based upon metabolic CO2 to regenerate the reactor window and gas inlet ports. The results and implications of this testing are addressed along with the advanced PPA reactor development.

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.

Design considerations for lunar base photovoltaic power systems

NASA Technical Reports Server (NTRS)

Hickman, J. Mark; Curtis, Henry B.; Landis, Geoffrey A.

1990-01-01

A survey was made of factors that may affect the design of photovoltaic arrays for a lunar base. These factors, which include the lunar environment and system design criteria, are examined. A photovoltaic power system design with a triangular array geometry is discussed and compared to a nuclear reactor power systems and a power system utilizing both nuclear and solar power sources.

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.

BOILING WATER REACTOR TECHNOLOGY STATUS OF THE ART REPORT. VOLUME II. WATER CHEMISTRY AND CORROSION

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

Breden, C.R.

1963-02-01

Information concerning the corrosive effects of water in power reactor moderator-coolant systems is presented. The information is based on investigations reported in the unclassified literature believed to be fairly complete to 1959, but less complete since then. The material is presented in sections on water decomposition, water chemistry, materials corrosion, corrosion product deposits, and radioactivity. It is noted that the report is presented as a part of a continuing program in development of less expensive materials for use in reactors. (J.R.D.)

Demand driven salt clean-up in a molten salt fast reactor – Defining a priority list

PubMed Central

Litskevich, D.; Gregg, R.; Mount, A. R.

2018-01-01

The PUREX technology based on aqueous processes is currently the leading reprocessing technology in nuclear energy systems. It seems to be the most developed and established process for light water reactor fuel and the use of solid fuel. However, demand driven development of the nuclear system opens the way to liquid fuelled reactors, and disruptive technology development through the application of an integrated fuel cycle with a direct link to reactor operation. The possibilities of this new concept for innovative reprocessing technology development are analysed, the boundary conditions are discussed, and the economic as well as the neutron physical optimization parameters of the process are elucidated. Reactor physical knowledge of the influence of different elements on the neutron economy of the reactor is required. Using an innovative study approach, an element priority list for the salt clean-up is developed, which indicates that separation of Neodymium and Caesium is desirable, as they contribute almost 50% to the loss of criticality. Separating Zirconium and Samarium in addition from the fuel salt would remove nearly 80% of the loss of criticality due to fission products. The theoretical study is followed by a qualitative discussion of the different, demand driven optimization strategies which could satisfy the conflicting interests of sustainable reactor operation, efficient chemical processing for the salt clean-up, and the related economic as well as chemical engineering consequences. A new, innovative approach of balancing the throughput through salt processing based on a low number of separation process steps is developed. Next steps for the development of an economically viable salt clean-up process are identified. PMID:29494604

Safety and Regulatory Issues of the Thorium Fuel Cycle

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

Ade, Brian; Worrall, Andrew; Powers, Jeffrey

2014-02-01

Thorium has been widely considered an alternative to uranium fuel because of its relatively large natural abundance and its ability to breed fissile fuel (233U) from natural thorium (232Th). Possible scenarios for using thorium in the nuclear fuel cycle include use in different nuclear reactor types (light water, high temperature gas cooled, fast spectrum sodium, molten salt, etc.), advanced accelerator-driven systems, or even fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based on concepts that mix thorium with uranium (UO2 + ThO2),more » add fertile thorium (ThO2) fuel pins to LWR fuel assemblies, or use mixed plutonium and thorium (PuO2 + ThO2) fuel assemblies. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts on the nuclear fuel. Thorium and its irradiation products have nuclear characteristics that are different from those of uranium. In addition, ThO2, alone or mixed with UO2 fuel, leads to different chemical and physical properties of the fuel. These aspects are key to reactor safety-related issues. The primary objectives of this report are to summarize historical, current, and proposed uses of thorium in nuclear reactors; provide some important properties of thorium fuel; perform qualitative and quantitative evaluations of both in-reactor and out-of-reactor safety issues and requirements specific to a thorium-based fuel cycle for current LWR reactor designs; and identify key knowledge gaps and technical issues that need to be addressed for the licensing of thorium LWR fuel in the United States.« less

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. .

Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

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

Liao, J.; Kucukboyaci, V. N.; Nguyen, L.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using themore » WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)« less

Multi-megawatt power system trade study

NASA Astrophysics Data System (ADS)

Longhurst, Glen R.; Schnitzler, Bruce G.; Parks, Benjamin T.

2002-01-01

A concept study was undertaken to evaluate potential multi-megawatt power sources for nuclear electric propulsion. The nominal electric power requirement was set at 15 MWe with an assumed mission profile of 120 days at full power, 60 days in hot standby, and another 120 days of full power, repeated several times for 7 years of service. Two configurations examined were (1) a gas-cooled reactor based on the NERVA Derivative design, operating a closed cycle Brayton power conversion system; and (2) a molten metal-cooled reactor based on SP-100 technology, driving a boiling potassium Rankine power conversion system. This study considered the relative merits of these two systems, seeking to optimize the specific mass. Conclusions were that either concept appeared capable of reaching the specific mass goal of 3-5 kg/kWe estimated to be needed for this class of mission, though neither could be realized without substantial development in reactor fuels technology, thermal radiator mass and volume efficiency, and power conversion and distribution electronics and systems capable of operating at high temperatures. The gas-Brayton system showed a specific mass advantage (3.17 vs 6.43 kg/kWe for the baseline cases) under the set of assumptions used and eliminated the need to deal with two-phase working fluid flows in the microgravity environment of space. .

Development of the reactor antineutrino detection technology within the iDream project

NASA Astrophysics Data System (ADS)

Gromov, M.; Kuznetsov, D.; Murchenko, A.; Novikova, G.; Obinyakov, B.; Oralbaev, A.; Plakitina, K.; Skorokhvatov, M.; Sukhotin, S.; Chepurnov, A.; Etenko, A.

2017-12-01

The iDREAM (industrial Detector for reactor antineutrino monitoring) project is aimed at remote monitoring of the operating modes of the atomic reactor on nuclear power plant to ensure a technical support of IAEA non-proliferation safeguards. The detector is a scintillator spectrometer. The sensitive volume (target) is filled with a liquid organic scintillator based on linear alkylbenzene where reactor antineutrinos will be detected via inverse beta-decay reaction. We present first results of laboratory tests after physical launch. The detector was deployed at sea level without background shielding. The number of calibrations with radioactive sources was conducted. All data were obtained by means of a slow control system which was put into operation.

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

NASA Astrophysics Data System (ADS)

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

2005-02-01

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

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

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

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

2005-02-06

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

Development of the Technology of Vortex Diagnostics to Improve the Safety of Operation of Nuclear Reactors

NASA Astrophysics Data System (ADS)

Mitrofanova, O. V.; Ivlev, O. A.; Pozdeeva, I. G.; Urtenov, D. S.

2017-11-01

The results of studies are aimed at developing theoretical foundations and instrumentation system to ensure a technology of vortex diagnostics of the state of flows of fluids for nuclear power installations with power water reactors and fast neutrons reactors with liquid-metal coolants. The technology of vortex diagnostics is based on the study of acoustic, magneto-hydrodynamic and resonant effects related to the formation of stable vortex structures. For creation a system of monitoring and diagnostics of the crisis phenomena due to hydrodynamics of the flow, it is proposed to use acoustic method to record the radiation of elastic waves in the fluids caused by the dynamic local rearrangement of its structure.

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.

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.

Chemical Looping Autothermal Reforming at a 120 kW Pilot Rig

NASA Astrophysics Data System (ADS)

Bofhàr-Nordenkampf, Johannes; Pröll, Tobias; Kolbitsch, Philipp; Hofbauer, Hermann

Chemical looping with selective oxygen transport allows two step combustion or autothermal reforming without mixing of fuel and air. The reactor system consists of two reactors, an air reactor and a fuel reactor with a suitable oxygen carrier that transports the necessary oxygen for operation. In the present study, a highly active nickel based oxygen carrier is tested in a novel dual circulating fluidized bed (DCFB) system at a scale of 120 kW fuel power. The mean particle size of the oxygen carrier is 120 μm and the pilot rig is fueled with natural gas. For the investigated oxygen carrier high CH4 conversion is achieved. Air/fuel ratio is varied at three different fuel reactor temperatures. For chemical looping reforming one can observe synthesis gas composition close to thermodynamic equilibrium. In spite of the fact that no additional steam has been added to the fuel besides the one present through steam fluidization of the loop seals, coke formation does not occur at global stoichiometric air/fuel ratios above 0.46.

Analysis of Process Gases and Trace Contaminants in Membrane-Aerated Gaseous Effluent Streams.

NASA Technical Reports Server (NTRS)

Coutts, Janelle L.; Lunn, Griffin Michael; Meyer, Caitlin E.

2015-01-01

In membrane-aerated biofilm reactors (MABRs), hollow fibers are used to supply oxygen to the biofilms and bulk fluid. A pressure and concentration gradient between the inner volume of the fibers and the reactor reservoir drives oxygen mass transport across the fibers toward the bulk solution, providing the fiber-adhered biofilm with oxygen. Conversely, bacterial metabolic gases from the bulk liquid, as well as from the biofilm, move opposite to the flow of oxygen, entering the hollow fiber and out of the reactor. Metabolic gases are excellent indicators of biofilm vitality, and can aid in microbial identification. Certain gases can be indicative of system perturbations and control anomalies, or potentially unwanted biological processes occurring within the reactor. In confined environments, such as those found during spaceflight, it is important to understand what compounds are being stripped from the reactor and potentially released into the crew cabin to determine the appropriateness or the requirement for additional mitigation factors. Reactor effluent gas analysis focused on samples provided from Kennedy Space Center's sub-scale MABRs, as well as Johnson Space Center's full-scale MABRs, using infrared spectroscopy and gas chromatography techniques. Process gases, such as carbon dioxide, oxygen, nitrogen, nitrogen dioxide, and nitrous oxide, were quantified to monitor reactor operations. Solid Phase Microextraction (SPME) GC-MS analysis was used to identify trace volatile compounds. Compounds of interest were subsequently quantified. Reactor supply air was examined to establish target compound baseline concentrations. Concentration levels were compared to average ISS concentration values and/or Spacecraft Maximum Allowable Concentration (SMAC) levels where appropriate. Based on a review of to-date results, current trace contaminant control systems (TCCS) currently on board the ISS should be able to handle the added load from bioreactor systems without the need for secondary mitigation.

Renewable Wood Pulp Paper Reactor with Hierarchical Micro/Nanopores for Continuous-Flow Nanocatalysis.

PubMed

Koga, Hirotaka; Namba, Naoko; Takahashi, Tsukasa; Nogi, Masaya; Nishina, Yuta

2017-06-22

Continuous-flow nanocatalysis based on metal nanoparticle catalyst-anchored flow reactors has recently provided an excellent platform for effective chemical manufacturing. However, there has been limited progress in porous structure design and recycling systems for metal nanoparticle-anchored flow reactors to create more efficient and sustainable catalytic processes. In this study, traditional paper is used for a highly efficient, recyclable, and even renewable flow reactor by tailoring the ultrastructures of wood pulp. The "paper reactor" offers hierarchically interconnected micro- and nanoscale pores, which can act as convective-flow and rapid-diffusion channels, respectively, for efficient access of reactants to metal nanoparticle catalysts. In continuous-flow, aqueous, room-temperature catalytic reduction of 4-nitrophenol to 4-aminophenol, a gold nanoparticle (AuNP)-anchored paper reactor with hierarchical micro/nanopores provided higher reaction efficiency than state-of-the-art AuNP-anchored flow reactors. Inspired by traditional paper materials, successful recycling and renewal of AuNP-anchored paper reactors were also demonstrated while high reaction efficiency was maintained. © 2017 The Authors. Published by Wiley-VCH Verlag GmbH & Co. KGaA.

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

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

Su'ud, Zaki; Anshari, Rio

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

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

NASA Astrophysics Data System (ADS)

Su'ud, Zaki; Anshari, Rio

2012-06-01

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

The diversity and unit of reactor noise theory

NASA Astrophysics Data System (ADS)

Kuang, Zhifeng

The study of reactor noise theory concerns questions about cause and effect relationships, and utilisation of random noise in nuclear reactor systems. The diversity of reactor noise theory arises from the variety of noise sources, the various mathematical treatments applied and various practical purposes. The neutron noise in zero- energy systems arises from the fluctuations in the number of neutrons per fission, the time between nuclear events, and the type of reactions. It can be used to evaluate system parameters. The mathematical treatment is based on the master equation of stochastic branching processes. The noise in power reactor systems is given rise by random processes of technological origin such as vibration of mechanical parts, boiling of the coolant, fluctuations of temperature and pressure. It can be used to monitor reactor behaviour with the possibility of detecting malfunctions at an early stage. The mathematical treatment is based on the Langevin equation. The unity of reactor noise theory arises from the fact that useful information from noise is embedded in the second moments of random variables, which lends the possibility of building up a unified mathematical description and analysis of the various reactor noise sources. Exploring such possibilities is the main subject among the three major topics reported in this thesis. The first subject is within the zero power noise in steady media, and we reported on the extension of the existing theory to more general cases. In Paper I, by use of the master equation approach, we have derived the most general Feynman- and Rossi-alpha formulae so far by taking the full joint statistics of the prompt and all the six groups of delayed neutron precursors, and a multiple emission source into account. The involved problems are solved with a combination of effective analytical techniques and symbolic algebra codes (Mathematica). Paper II gives a numerical evaluation of these formulae. An assessment of the contribution of the terms that are novel as compared to the traditional formulae has been made. The second subject treats a problem in power reactor noise with the Langevin formalism. With a very few exceptions, in all previous work the diffusion approximation was used. In order to extend the treatment to transport theory, in Paper III, we introduced a novel method, i.e. Padé approximation via Lanczos algorithm to calculate the transfer function of a finite slab reactor described by one-group transport equation. It was found that the local-global decomposition of the neutron noise, formerly only reproduced in at least 2- group theory, can be reconstructed. We have also showed the existence of a boundary layer of the neutron noise close to the boundary. Finally, we have explored the possibility of building up a unified theory to account for the coexistence of zero power and power reactor noise in a system. In Paper IV, a unified description of the neutron noise is given by the use of backward master equations in a model where the cross section fluctuations are given as a simple binary pseudorandom process. The general solution contains both the zero power and power reactor noise concurrently, and they can be extracted individually as limiting cases of the general solution. It justified the separate treatments of zero power and power reactor noise. The result was extended to the case including one group of delayed neutron precursors in Paper V.

Nuclear Thermal Rocket Simulation in NPSS

NASA Technical Reports Server (NTRS)

Belair, Michael L.; Sarmiento, Charles J.; Lavelle, Thomas M.

2013-01-01

Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic-metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

Nuclear Thermal Rocket Simulation in NPSS

NASA Technical Reports Server (NTRS)

Belair, Michael L.; Sarmiento, Charles J.; Lavelle, Thomas L.

2013-01-01

Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic- metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

Auto-thermal reforming using mixed ion-electronic conducting ceramic membranes for a small-scale H₂ production plant.

PubMed

Spallina, Vincenzo; Melchiori, Tommaso; Gallucci, Fausto; van Sint Annaland, Martin

2015-03-18

The integration of mixed ionic electronic conducting (MIEC) membranes for air separation in a small-to-medium scale unit for H2 production (in the range of 650-850 Nm3/h) via auto-thermal reforming of methane has been investigated in the present study. Membranes based on mixed ionic electronic conducting oxides such as Ba0.5Sr0.5Co0.8Fe0.2O3-δ (BSCF) give sufficiently high oxygen fluxes at temperatures above 800 °C with high purity (higher than 99%). Experimental results of membrane permeation tests are presented and used for the reactor design with a detailed reactor model. The assessment of the H2 plant has been carried out for different operating conditions and reactor geometry and an energy analysis has been carried out with the flowsheeting software Aspen Plus, including also the turbomachines required for a proper thermal integration. A micro-gas turbine is integrated in the system in order to supply part of the electricity required in the system. The analysis of the system shows that the reforming efficiency is in the range of 62%-70% in the case where the temperature at the auto-thermal reforming membrane reactor (ATR-MR) is equal to 900 °C. When the electric consumption and the thermal export are included the efficiency of the plant approaches 74%-78%. The design of the reactor has been carried out using a reactor model linked to the Aspen flowsheet and the results show that with a larger reactor volume the performance of the system can be improved, especially because of the reduced electric consumption. From this analysis it has been found that for a production of about 790 Nm3/h pure H2, a reactor with a diameter of 1 m and length of 1.8 m with about 1500 membranes of 2 cm diameter is required.

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

Integrated ‘omics analysis for studying the microbial community response to a pH perturbation of a cellulose-degrading bioreactor culture

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

Boaro, Amy A.; Kim, Young-Mo; Konopka, Allan E.

2015-01-05

Propionate accumulation is a common indicator of process imbalances in anaerobic bioreactor systems. The accumulation of propionate can occur due to low retention rates, hydrogen accumulation, or mechanical changes affecting the proximity between propionate oxidizers and partner species, thereby preventing necessary electron transfer. Few studies, however, have observed the changes in microbial community structure during propionate accumulation. We used 454 pyrosequencing of 16S rDNA to evaluate the community membership during propionate accumulations in replicate bioreactors with rumen based cultures. Half of the culture volume from a parent reactor was transferred to a sterile “daughter” reactor, and both systems were runmore » identically. Both reactors experienced a propionate accumulation after roughly 10 days, with the propionate accumulation being less pronounced in the parent reactor as compared to the daughter reactor. Non-metric multidimensional scaling (NMDS) was used to determine clustering patterns of the samples, and correlative methods were used to determine which OTUs were significantly associated with the movements of samples along the NMDS axes. The presence of Saccharofermentans characterized the position of early samples, whereas the presence of Ruminococcus and Succiniclasticum were more indicative of the positions of later samples. Hydrogen accumulation and low sequence counts indicated low methanogen activity. Although both reactor systems were closed to microbial inputs due to the sterilization of influent media, we recorded significant increases in reactor diversity over time. This suggests that changes in the abundances of dominant community members may affect the sequencing of rare taxa within samples.« less

Conceptual design study of Fusion Experimental Reactor (FY86 FER): Safety

NASA Astrophysics Data System (ADS)

Seki, Yasushi; Iida, Hiromasa; Honda, Tsutomu

1987-08-01

This report describes the study on safety for FER (Fusion Experimental Reactor) which has been designed as a next step machine to the JT-60. Though the final purpose of this study is to have an image of design base accident, maximum credible accident and to assess their risk or probability, etc., as FER plant system, the emphasis of this years study is placed on fuel-gas circulation system where the tritium inventory is maximum. The report consists of two chapters. The first chapter summarizes the FER system and describes FMEA (Failure Mode and Effect Analysis) and related accident progression sequence for FER plant system as a whole. The second chapter of this report is focused on fuel-gas circulation system including purification, isotope separation and storage. Probability of risk is assessed by the probabilistic risk analysis (PRA) procedure based on FMEA, ETA and FTA.

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.

Development of a Reduced-Order Three-Dimensional Flow Model for Thermal Mixing and Stratification Simulation during Reactor Transients

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

Hu, Rui

2017-09-03

Mixing, thermal-stratification, and mass transport phenomena in large pools or enclosures play major roles for the safety of reactor systems. Depending on the fidelity requirement and computational resources, various modeling methods, from the 0-D perfect mixing model to 3-D Computational Fluid Dynamics (CFD) models, are available. Each is associated with its own advantages and shortcomings. It is very desirable to develop an advanced and efficient thermal mixing and stratification modeling capability embedded in a modern system analysis code to improve the accuracy of reactor safety analyses and to reduce modeling uncertainties. An advanced system analysis tool, SAM, is being developedmore » at Argonne National Laboratory for advanced non-LWR reactor safety analysis. While SAM is being developed as a system-level modeling and simulation tool, a reduced-order three-dimensional module is under development to model the multi-dimensional flow and thermal mixing and stratification in large enclosures of reactor systems. This paper provides an overview of the three-dimensional finite element flow model in SAM, including the governing equations, stabilization scheme, and solution methods. Additionally, several verification and validation tests are presented, including lid-driven cavity flow, natural convection inside a cavity, laminar flow in a channel of parallel plates. Based on the comparisons with the analytical solutions and experimental results, it is demonstrated that the developed 3-D fluid model can perform very well for a wide range of flow problems.« less

Nuclear space power safety and facility guidelines study

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

Mehlman, W.F.

1995-09-11

This report addresses safety guidelines for space nuclear reactor power missions and was prepared by The Johns Hopkins University Applied Physics Laboratory (JHU/APL) under a Department of Energy grant, DE-FG01-94NE32180 dated 27 September 1994. This grant was based on a proposal submitted by the JHU/APL in response to an {open_quotes}Invitation for Proposals Designed to Support Federal Agencies and Commercial Interests in Meeting Special Power and Propulsion Needs for Future Space Missions{close_quotes}. The United States has not launched a nuclear reactor since SNAP 10A in April 1965 although many Radioisotope Thermoelectric Generators (RTGs) have been launched. An RTG powered system ismore » planned for launch as part of the Cassini mission to Saturn in 1997. Recently the Ballistic Missile Defense Office (BMDO) sponsored the Nuclear Electric Propulsion Space Test Program (NEPSTP) which was to demonstrate and evaluate the Russian-built TOPAZ II nuclear reactor as a power source in space. As of late 1993 the flight portion of this program was canceled but work to investigate the attributes of the reactor were continued but at a reduced level. While the future of space nuclear power systems is uncertain there are potential space missions which would require space nuclear power systems. The differences between space nuclear power systems and RTG devices are sufficient that safety and facility requirements warrant a review in the context of the unique features of a space nuclear reactor power 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.

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.

Hanford Site 100-N Area In Situ Bioremediation of UPR-100-N-17, Deep Petroleum Unplanned Release - 13245

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

Saueressig, Daniel G.

2013-07-01

In 1965 and 1966, approximately 303 m{sup 3} of Number 2 diesel fuel leaked from a pipeline used to support reactor operations at the Hanford Site's N Reactor. N Reactor was Hanford's longest operating reactor and served as the world's first dual purpose reactor for military and power production needs. The Interim Action Record of Decision for the 100-N Area identified in situ bioremediation as the preferred alternative to remediate the deep vadose zone contaminated by this release. A pilot project supplied oxygen into the vadose zone to stimulate microbial activity in the soil. The project monitored respiration rates asmore » an indicator of active biodegradation. Based on pilot study results, a full-scale system is being constructed and installed to remediate the vadose zone contamination. (authors)« less

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

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

Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.

2016-08-31

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF 2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. Themore » objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.« less

Reliability of digital reactor protection system based on extenics.

PubMed

Zhao, Jing; He, Ya-Nan; Gu, Peng-Fei; Chen, Wei-Hua; Gao, Feng

2016-01-01

After the Fukushima nuclear accident, safety of nuclear power plants (NPPs) is widespread concerned. The reliability of reactor protection system (RPS) is directly related to the safety of NPPs, however, it is difficult to accurately evaluate the reliability of digital RPS. The method is based on estimating probability has some uncertainties, which can not reflect the reliability status of RPS dynamically and support the maintenance and troubleshooting. In this paper, the reliability quantitative analysis method based on extenics is proposed for the digital RPS (safety-critical), by which the relationship between the reliability and response time of RPS is constructed. The reliability of the RPS for CPR1000 NPP is modeled and analyzed by the proposed method as an example. The results show that the proposed method is capable to estimate the RPS reliability effectively and provide support to maintenance and troubleshooting of digital RPS system.

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.

Pilot-scale investigation of sludge reduction in aerobic digestion system with endospore-forming bacteria.

PubMed

Seo, Kyu Won; Choi, Yong-Su; Gu, Man Bock; Kwon, Eilhann E; Tsang, Yiu Fai; Rinklebe, Jörg; Park, Chanhyuk

2017-11-01

A pilot-scale investigation of membrane-based aerobic digestion system dominated by endospore-forming bacteria was evaluated as one of the potential sludge treatment processes (STP). Most of the organic matter in the sludge was removed (90.1%) by the particular bacteria in the STP, which consisted of mixed liquor suspended solid (MLSS) contact reactor (MCR), MLSS oxidation reactor (MOR), and membrane bioreactor (MBR). The sludge was accumulated in the MBR without wasting, and then the effluent in STP was fed into the first step in water resource recovery facility (WRRF). According to the analysis of microbial communities in all reactors, various Bacillus species were present in the STP, mainly due to their intrinsic resistance to the extreme conditions. As the surviving Bacillus species might consume degraded microorganisms for their growth, these endospore-forming bacteria-based STP could be suitable for the sludge reduction when they operated for a long time. Copyright © 2017 Elsevier Ltd. All rights reserved.

A laser scanning system for metrology and viewing in ITER

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

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

1996-05-01

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

Ongoing Development of a Series Bosch Reactor System

NASA Technical Reports Server (NTRS)

Abney, Morgan; Mansell, Matt; DuMez, Sam; Thomas, John; Cooper, Charlie; Long, David

2013-01-01

Future manned missions to deep space or planetary surfaces will undoubtedly require highly robust, efficient, and regenerable life support systems that require minimal consumables. To meet this requirement, NASA continues to explore a Bosch-based carbon dioxide reduction system to recover oxygen from CO2. In order to improve the equivalent system mass of Bosch systems, we seek to design and test a "Series Bosch" system in which two reactors in series are optimized for the two steps of the reaction, as well as to explore the use of in situ materials as carbon deposition catalysts. Here we report recent developments in this effort including assembly and initial testing of a Reverse Water-Gas Shift reactor (RWGSr) and initial testing of two gas separation membranes. The RWGSr was sized to reduce CO2 produced by a crew of four to carbon monoxide as the first stage in a Series Bosch system. The gas separation membranes, necessary to recycle unreacted hydrogen and CO2, were similarly sized. Additionally, we report results of preliminary experiments designed to determine the catalytic properties of Martian and Lunar regolith simulant for the carbon deposition step.

Ongoing Development of a Series Bosch Reactor System

NASA Technical Reports Server (NTRS)

Abney, Morgan B; Mansell, J. Matthew; Stanley, Christine; Edmunson, Jennifer; DuMez, Samuel J.; Chen, Kevin

2013-01-01

Future manned missions to deep space or planetary surfaces will undoubtedly incorporate highly robust, efficient, and regenerable life support systems that require minimal consumables. To meet this requirement, NASA continues to explore a Bosch-based carbon dioxide reduction system to recover oxygen from CO2. In order to improve the equivalent system mass of Bosch systems, we seek to design and test a "Series Bosch" system in which two reactors in series are optimized for the two steps of the reaction, as well as to explore the use of in situ materials as carbon deposition catalysts. Here we report recent developments in this effort including assembly and initial testing of a Reverse Water-Gas Shift reactor (RWGSr) and initial testing of two gas separation membranes. The RWGSr was sized to reduce CO2 produced by a crew of four to carbon monoxide as the first stage in a Series Bosch system. The gas separation membranes, necessary to recycle unreacted hydrogen and CO2, were similarly sized. Additionally, we report results of preliminary experiments designed to determine the catalytic properties of Martian regolith simulant for the carbon formation step.

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.

Generating Breathable Air Through Dissociation of N2O

NASA Technical Reports Server (NTRS)

Zubrin, Robert; Frankie, Brian

2006-01-01

A nitrous oxide-based oxygen-supply system (NOBOSS) is an apparatus in which a breathable mixture comprising 2/3 volume parts of N2 and 1/3 volume part of O2 is generated through dissociation of N2O. The NOBOSS concept can be adapted to a variety of applications in which there are requirements for relatively compact, lightweight systems to supply breathable air. These could include air-supply systems for firefighters, divers, astronauts, and workers who must be protected against biological and chemical hazards. A NOBOSS stands in contrast to compressed-gas and cryogenic air-supply systems. Compressed-gas systems necessarily include massive tanks that can hold only relatively small amounts of gases. Alternatively, gases can be stored compactly in greater quantities and at low pressures when they are liquefied, but then cryogenic equipment is needed to maintain them in liquid form. Overcoming the disadvantages of both compressed-gas and cryogenic systems, the NOBOSS exploits the fact that N2O can be stored in liquid form at room temperature and moderate pressure. The mass of N2O that can be stored in a tank of a given mass is about 20 times the mass of compressed air that can be stored in a tank of equal mass. In a NOBOSS, N2O is exothermically dissociated to N2 and O2 in a main catalytic reactor. In order to ensure the dissociation of N2O to the maximum possible extent, the temperature of the reactor must be kept above 400 C. At the same time, to minimize concentrations of nitrogen oxides (which are toxic), it is necessary to keep the reactor temperature at or below 540 C. To keep the temperature within the required range throughout the reactor and, in particular, to prevent the formation of hot spots that would be generated by local concentrations of the exothermic dissociation reaction, the N2O is introduced into the reactor through an injector tube that features carefully spaced holes to distribute the input flow of N2O widely throughout the reactor. A NOBOSS includes one or more "destroyer" subsystems for removing any nitrogen oxides that remain downstream of the main N2O-dissociation reactor. A destroyer includes a carbon bed in series with a catalytic reactor, and is in thermal contact with the main N2O-dissociation reactor. The gas mixture that leaves the main reactor first goes through a carbon bed, which adsorbs all of the trace NO and most of the trace NO2. The gas mixture then goes through the destroyer catalytic reactor, wherein most or all of the remaining NO2 is dissociated. A NOBOSS can be designed to regulate its reactor temperature across a range of flow rates. One such system includes three destroyer loops; these loops act, in combination with a heat sink, to remove heat from the main N2O-dissociation reactor. In this system, the N2O and product gases play an additional role as coolants; thus, as needed, the coolant flow increases in proportion to the rate of generation of heat, helping to keep the main-reactor temperature below 540 C.

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.

Early Program Development

NASA Image and Video Library

2004-04-15

This artist's concept illustrates the NERVA (Nuclear Engine for Rocket Vehicle Application) engine's hot bleed cycle in which a small amount of hydrogen gas is diverted from the thrust nozzle, thus eliminating the need for a separate system to drive the turbine. The NERVA engine, based on KIWI nuclear reactor technology, would power a RIFT (Reactor-In-Flight-Test) nuclear stage, for which the Marshall Space Flight Center had development responsibility.

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.

Modeling and Design Optimization of Multifunctional Membrane Reactors for Direct Methane Aromatization

PubMed Central

Fouty, Nicholas J.; Carrasco, Juan C.; Lima, Fernando V.

2017-01-01

Due to the recent increase of natural gas production in the U.S., utilizing natural gas for higher-value chemicals has become imperative. Direct methane aromatization (DMA) is a promising process used to convert methane to benzene, but it is limited by low conversion of methane and rapid catalyst deactivation by coking. Past work has shown that membrane separation of the hydrogen produced in the DMA reactions can dramatically increase the methane conversion by shifting the equilibrium toward the products, but it also increases coke production. Oxygen introduction into the system has been shown to inhibit this coke production while not inhibiting the benzene production. This paper introduces a novel mathematical model and design to employ both methods in a multifunctional membrane reactor to push the DMA process into further viability. Multifunctional membrane reactors, in this case, are reactors where two different separations occur using two differently selective membranes, on which no systems studies have been found. The proposed multifunctional membrane design incorporates a hydrogen-selective membrane on the outer wall of the reaction zone, and an inner tube filled with airflow surrounded by an oxygen-selective membrane in the middle of the reactor. The design is shown to increase conversion via hydrogen removal by around 100%, and decrease coke production via oxygen addition by 10% when compared to a tubular reactor without any membranes. Optimization studies are performed to determine the best reactor design based on methane conversion, along with coke and benzene production. The obtained optimal design considers a small reactor (length = 25 cm, diameter of reaction tube = 0.7 cm) to subvert coke production and consumption of the product benzene as well as a high permeance (0.01 mol/s·m2·atm1/4) through the hydrogen-permeable membrane. This modeling and design approach sets the stage for guiding further development of multifunctional membrane reactor models and designs for natural gas utilization and other chemical reaction systems. PMID:28850068

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.

Manned space flight nuclear system safety. Volume 3: Reactor system preliminary nuclear safety analysis. Part 2A: Accident model document, appendix

NASA Technical Reports Server (NTRS)

1972-01-01

The detailed abort sequence trees for the reference zirconium hydride (ZrH) reactor power module that have been generated for each phase of the reference Space Base program mission are presented. The trees are graphical representations of causal sequences. Each tree begins with the phase identification and the dichotomy between success and failure. The success branch shows the mission phase objective as being achieved. The failure branch is subdivided, as conditions require, into various primary initiating abort conditions.

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

Glycerol Production and Transformation: A Critical Review with Particular Emphasis on Glycerol Reforming Reaction for Producing Hydrogen in Conventional and Membrane Reactors.

PubMed

Bagnato, Giuseppe; Iulianelli, Adolfo; Sanna, Aimaro; Basile, Angelo

2017-03-23

Glycerol represents an emerging renewable bio-derived feedstock, which could be used as a source for producing hydrogen through steam reforming reaction. In this review, the state-of-the-art about glycerol production processes is reviewed, with particular focus on glycerol reforming reactions and on the main catalysts under development. Furthermore, the use of membrane catalytic reactors instead of conventional reactors for steam reforming is discussed. Finally, the review describes the utilization of the Pd-based membrane reactor technology, pointing out the ability of these alternative fuel processors to simultaneously extract high purity hydrogen and enhance the whole performances of the reaction system in terms of glycerol conversion and hydrogen yield.

Glycerol Production and Transformation: A Critical Review with Particular Emphasis on Glycerol Reforming Reaction for Producing Hydrogen in Conventional and Membrane Reactors

PubMed Central

Bagnato, Giuseppe; Iulianelli, Adolfo; Sanna, Aimaro; Basile, Angelo

2017-01-01

Glycerol represents an emerging renewable bio-derived feedstock, which could be used as a source for producing hydrogen through steam reforming reaction. In this review, the state-of-the-art about glycerol production processes is reviewed, with particular focus on glycerol reforming reactions and on the main catalysts under development. Furthermore, the use of membrane catalytic reactors instead of conventional reactors for steam reforming is discussed. Finally, the review describes the utilization of the Pd-based membrane reactor technology, pointing out the ability of these alternative fuel processors to simultaneously extract high purity hydrogen and enhance the whole performances of the reaction system in terms of glycerol conversion and hydrogen yield. PMID:28333121

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.

Characteristics comparison between a cyclotron-based neutron source and KUR-HWNIF for boron neutron capture therapy

NASA Astrophysics Data System (ADS)

Tanaka, H.; Sakurai, Y.; Suzuki, M.; Masunaga, S.; Kinashi, Y.; Kashino, G.; Liu, Y.; Mitsumoto, T.; Yajima, S.; Tsutsui, H.; Maruhashi, A.; Ono, K.

2009-06-01

At Kyoto University Research Reactor Institute (KURRI), 275 clinical trials of boron neutron capture therapy (BNCT) have been performed as of March 2006, and the effectiveness of BNCT has been revealed. In order to further develop BNCT, it is desirable to supply accelerator-based epithermal-neutron sources that can be installed near the hospital. We proposed the method of filtering and moderating fast neutrons, which are emitted from the reaction between a beryllium target and 30-MeV protons accelerated by a cyclotron accelerator, using an optimum moderator system composed of iron, lead, aluminum and calcium fluoride. At present, an epithermal-neutron source is under construction from June 2008. This system consists of a cyclotron accelerator, beam transport system, neutron-yielding target, filter, moderator and irradiation bed. In this article, an overview of this system and the properties of the treatment neutron beam optimized by the MCNPX Monte Carlo neutron transport code are presented. The distribution of biological effect weighted dose in a head phantom compared with that of Kyoto University Research Reactor (KUR) is shown. It is confirmed that for the accelerator, the biological effect weighted dose for a deeply situated tumor in the phantom is 18% larger than that for KUR, when the limit dose of the normal brain is 10 Gy-eq. The therapeutic time of the cyclotron-based neutron sources are nearly one-quarter of that of KUR. The cyclotron-based epithermal-neutron source is a promising alternative to reactor-based neutron sources for treatments by BNCT.

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.

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

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

Boyack, B.E.

The PIUS reactor utilizes simplified, inherent, passive, or other innovative means to accomplish safety functions. Accordingly, the PIUS reactor is subject to the requirements of 10CFR52.47(b)(2)(i)(A). This regulation requires that the applicant adequately demonstrate the performance of each safety feature, interdependent effects among the safety features, and a sufficient data base on the safety features of the design to assess the analytical tools used for safety analysis. Los Alamos has assessed the quality and completeness of the existing and planned data bases used by Asea Brown Boveri to validate its safety analysis codes and other relevant data bases. Only amore » limited data base of separate effect and integral tests exist at present. This data base is not adequate to fulfill the requirements of 10CFR52.47(b)(2)(i)(A). Asea Brown Boveri has stated that it plans to conduct more separate effect and integral test programs. If appropriately designed and conducted, these test programs have the potential to satisfy most of the data base requirements of 10CFR52.47(b)(2)(i)(A) and remedy most of the deficiencies of the currently existing combined data base. However, the most important physical processes in PIUS are related to reactor shutdown because the PIUS reactor does not contain rodded shutdown and control systems. For safety-related reactor shutdown, PIUS relies on negative reactivity insertions from the moderator temperature coefficient and from boron entering the core from the reactor pool. Asea Brown Boveri has neither developed a direct experimental data base for these important processes nor provided a rationale for indirect testing of these key PIUS processes. This is assessed as a significant shortcoming. In preparing the conclusions of this report, test documentation and results have been reviewed for only one integral test program, the small-scale integral tests conducted in the ATLE facility.« less

Pratt & Whitney ESCORT derivative for mars surface power

NASA Astrophysics Data System (ADS)

Feller, Gerald J.; Joyner, Russell

1999-01-01

The purpose of this paper is to address the applicability of a common reactor system design from the Pratt & Whitney ESCORT nuclear thermal rocket engine concept to support current NASA mars surface-based power requirements. The ESCORT is a bimodal engine capable of supporting a wide range of propulsive thermal and vehicle electrical power requirements. The ESCORT engine is powered by a fast-spectrum beryllium-reflected CERMET-fueled nuclear reactor. In addition to an expander cycle propulsive mode, the ESCORT is capable of operating in an electrical power mode. In this mode, the reactor is used to heat a mixture of helium and xenon to drive a closed-loop Brayton cycle in order to generate electrical energy. Recent Design Reference Mission requirements (DRM) from NASA Johnson Space Center and NASA Lewis Research Center studies in 1997 and 1998 have detailed upgraded requirements for potential mars transfer missions. The current NASA DRM requires a nuclear thermal propulsion system capable of delivering total mission requirements of 200170 N (45000 lbf) thrust and 50 kWe of spacecraft electrical power. Additionally, these requirements detailed a surface power system capable of providing approximately 160 kW of electrical energy over an approximate 10 year period within a given weight and volume envelope. Current NASA studies use a SP-100 reactor (0.8 MT) and a NERVA derivative (1.6 MT) as baseline systems. A mobile power cart of approximate dimensions 1.7 m×4.5 m×4.4 m has been conceptualized to transport the reactor power system on the Mars Surface. The 63.25 cm diameter and 80.25 cm height of the ESCORT and its 1.3 MT of weight fit well within the current weight and volume target range of the NASA DRM requirements. The modifications required to the ESCORT reactor system to support this upgraded electrical power requirements along with operation in the Martian atmospheric conditions are addressed in this paper. Sufficient excess reactivity and burnup capability were designed into the ESCORT reactor system to support these upgraded requirements. Only slight modifications to reactor hardware were required to address any environmental considerations. These modifications involved sealing any refractory metal alloy components from the CO2 in the Martian Atmosphere. Also, the Brayton cycle Power Conversion Unit (PCU) hardware was modified to support the upgraded requirements. This paper discusses the design analysis performed and provides information on the final common reactor concept to be used on the Mars surface to support manned missions.

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.

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

A membrane-integrated fermentation reactor system: its effects in reducing the amount of sub-raw materials for D-lactic acid continuous fermentation by Sporolactobacillus laevolacticus.

PubMed

Mimitsuka, Takashi; Na, Kyungsu; Morita, Ken; Sawai, Hideki; Minegishi, Shinichi; Henmi, Masahiro; Yamada, Katsushige; Shimizu, Sakayu; Yonehara, Tetsu

2012-01-01

Continuous fermentation by retaining cells with a membrane-integrated fermentation reactor (MFR) system was found to reduce the amount of supplied sub-raw material. If the amount of sub-raw material can be reduced, continuous fermentation with the MFR system should become a more attractive process for industrialization, due to decreased material costs and loads during the refinement process. Our findings indicate that the production rate decreased when the amount of the sub-raw material was reduced in batch fermentation, but did not decrease during continuous fermentation with Sporolactobacillus laevolacticus. Moreover, continuous fermentation with a reduced amount of sub-raw material resulted in a productivity of 11.2 g/L/h over 800 h. In addition, the index of industrial process applicability used in the MFR system increased by 6.3-fold as compared with the conventional membrane-based fermentation reactor previously reported, suggesting a potential for the industrialization of this D-lactic acid continuous fermentation process.

High Efficiency Nuclear Power Plants using Liquid Fluoride Thorium Reactor Technology

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

2009-01-01

An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITS of 950 K and 1200 K are presented. Power plant performance data were obtained for TITS ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo -generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

Versatile Oxide Films Protect FeCrAl Alloys Under Normal Operation and Accident Conditions in Light Water Power Reactors

NASA Astrophysics Data System (ADS)

Rebak, Raul B.

2018-02-01

The US has currently a fleet of 99 nuclear power light water reactors which generate approximately 20% of the electricity consumed in the country. Near 90% of the reactors are at least 30 years old. There are incentives to make the existing reactors safer by using accident tolerant fuels (ATF). Compared to the standard UO2-zirconium-based system, ATF need to tolerate loss of active cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. Ferritic iron-chromium-aluminum (FeCrAl) alloys have been identified as an alternative to replace current zirconium alloys. They contain Fe (base) + 10-22 Cr + 4-6 Al and may contain smaller amounts of other elements such as molybdenum and traces of others. FeCrAl alloys offer outstanding resistance to attack by superheated steam by developing an alumina oxide on the surface in case of a loss of coolant accident like at Fukushima. FeCrAl alloys also perform well under normal operation conditions both in boiling water reactors and pressurized water reactors because they are protected by a thin oxide rich in chromium. Under normal operation condition, the key element is Cr and under accident conditions it is Al.

Treatment of petroleum refinery wastewater using a sequential anaerobic-aerobic moving-bed biofilm reactor system based on suspended ceramsite.

PubMed

Lu, Mang; Gu, Li-Peng; Xu, Wen-Hao

2013-01-01

In this study, a novel suspended ceramsite was prepared, which has high strength, optimum density (close to water), and high porosity. The ceramsite was used to feed a moving-bed biofilm reactor (MBBR) system with an anaerobic-aerobic (A/O) arrangement to treat petroleum refinery wastewater for simultaneous removal of chemical oxygen demand (COD) and ammonium. The hydraulic retention time (HRT) of the anaerobic-aerobic MBBR system was varied from 72 to 18 h. The anaerobic-aerobic system had a strong tolerance to shock loading. Compared with the professional emission standard of China, the effluent concentrations of COD and NH3-N in the system could satisfy grade I at HRTs of 72 and 36 h, and grade II at HRT of 18 h. The average sludge yield of the anaerobic reactor was estimated to be 0.0575 g suspended solid/g CODremoved. This work demonstrated that the anaerobic-aerobic MBBR system using the suspended ceramsite as bio-carrier could be applied to achieving high wastewater treatment efficiency.

Renewable Wood Pulp Paper Reactor with Hierarchical Micro/Nanopores for Continuous‐Flow Nanocatalysis

PubMed Central

Namba, Naoko; Takahashi, Tsukasa; Nogi, Masaya; Nishina, Yuta

2017-01-01

Abstract Continuous‐flow nanocatalysis based on metal nanoparticle catalyst‐anchored flow reactors has recently provided an excellent platform for effective chemical manufacturing. However, there has been limited progress in porous structure design and recycling systems for metal nanoparticle‐anchored flow reactors to create more efficient and sustainable catalytic processes. In this study, traditional paper is used for a highly efficient, recyclable, and even renewable flow reactor by tailoring the ultrastructures of wood pulp. The “paper reactor” offers hierarchically interconnected micro‐ and nanoscale pores, which can act as convective‐flow and rapid‐diffusion channels, respectively, for efficient access of reactants to metal nanoparticle catalysts. In continuous‐flow, aqueous, room‐temperature catalytic reduction of 4‐nitrophenol to 4‐aminophenol, a gold nanoparticle (AuNP)‐anchored paper reactor with hierarchical micro/nanopores provided higher reaction efficiency than state‐of‐the‐art AuNP‐anchored flow reactors. Inspired by traditional paper materials, successful recycling and renewal of AuNP‐anchored paper reactors were also demonstrated while high reaction efficiency was maintained. PMID:28394501

Direct-Drive Gas-Cooled Reactor Power System: Concept and Preliminary Testing

NASA Technical Reports Server (NTRS)

Wright, S. A.; Lipinski, R. J.; Godfroy, T. J.; Bragg-Sitton, S. M.; VanDyke, M. K.

2002-01-01

This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet- sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrically heated testing of simulated reactor components.

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

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

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

2014-01-01

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

Enzymatic conversion of sunflower oil to biodiesel in a solvent-free system: process optimization and the immobilized system stability.

PubMed

Ognjanovic, Nevena; Bezbradica, Dejan; Knezevic-Jugovic, Zorica

2009-11-01

The feasibility of using the commercial immobilized lipase from Candida antarctica (Novozyme 435) to synthesize biodiesel from sunflower oil in a solvent-free system has been proved. Using methanol as an acyl acceptor and the response surface methodology as an optimization technique, the optimal conditions for the transesterification has been found to be: 45 degrees C, 3% of enzyme based on oil weight, 3:1 methanol to oil molar ratio and with no added water in the system. Under these conditions, >99% of oil conversion to fatty acid methyl ester (FAME) has been achieved after 50 h of reaction, but the activity of the immobilized lipase decreased markedly over the course of repeated runs. In order to improve the enzyme stability, several alternative acyl acceptors have been tested for biodiesel production under solvent-free conditions. The use of methyl acetate seems to be of great interest, resulting in high FAME yield (95.65%) and increasing the half-life of the immobilized lipase by about 20.1 times as compared to methanol. The reaction has also been verified in the industrially feasible reaction system including both a batch stirred tank reactor and a packed bed reactor. Although satisfactory performance in the batch stirred tank reactor has been achieved, the kinetics in a packed bed reactor system seems to have a slightly better profile (93.6+/-3.75% FAME yield after 8-10 h), corresponding to the volumetric productivity of 48.5 g/(dm(3) h). The packed bed reactor has operated for up to 72 h with almost no loss in productivity, implying that the proposed process and the immobilized system could provide a promising solution for the biodiesel synthesis at the industrial scale.

SP-100 power system conceptual design for lunar base applications

NASA Technical Reports Server (NTRS)

Mason, Lee S.; Bloomfield, Harvey S.; Hainley, Donald C.

1989-01-01

A conceptual design is presented for a nuclear power system utilizing an SP-100 reactor and multiple Stirling cycle engines for operation on the lunar surface. Based on the results of this study, it was concluded that this power plant could be a viable option for an evolutionary lunar base. The design concept consists of a 2500 kWt (kilowatt thermal) SP-100 reactor coupled to eight free-piston Stirling engines. Two of the engines are held in reserve to provide conversion system redundancy. The remaining engines operate at 91.7 percent of their rated capacity of 150 kWe. The design power level for this system is 825 kWe. Each engine has a pumped heat-rejection loop connected to a heat pipe radiator. Power system performance, sizing, layout configurations, shielding options, and transmission line characteristics are described. System components and integration options are compared for safety, high performance, low mass, and ease of assembly. The power plant was integrated with a proposed human lunar base concept to ensure mission compatibility. This study should be considered a preliminary investigation; further studies are planned to investigate the effect of different technologies on this baseline design.

Reactor antineutrino detector iDREAM.

NASA Astrophysics Data System (ADS)

Gromov, M. B.; Lukyanchenko, G. A.; Novikova, G. J.; Obinyakov, B. A.; Oralbaev, A. Y.; Skorokhvatov, M. D.; Sukhotin, S. V.; Chepurnov, A. S.; Etenko, A. V.

2017-09-01

Industrial Detector for Reactor Antineutrino Monitoring (iDREAM) is a compact (≈ 3.5m 2) industrial electron antineutrino spectrometer. It is dedicated for remote monitoring of PWR reactor operational modes by neutrino method in real-time. Measurements of antineutrino flux from PWR allow to estimate a fuel mixture in active zone and to check the status of the reactor campaign for non-proliferation purposes. LAB-based gadolinium doped scintillator is exploited as a target. Multizone architecture of the detector with gamma-catcher surrounding fiducial volume and plastic muon veto above and below ensure high efficiency of IBD detection and background suppression. DAQ is based on Flash ADC with PSD discrimination algorithms while digital trigger is programmable and flexible due to FPGA. The prototype detector was started up in 2014. Preliminary works on registration Cerenkov radiation produced by cosmic muons were established with distilled water inside the detector in order to test electronic and slow control systems. Also in parallel a long-term measurements with different scintillator samples were conducted.

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.

Pulsed corona generation using a diode-based pulsed power generator

NASA Astrophysics Data System (ADS)

Pemen, A. J. M.; Grekhov, I. V.; van Heesch, E. J. M.; Yan, K.; Nair, S. A.; Korotkov, S. V.

2003-10-01

Pulsed plasma techniques serve a wide range of unconventional processes, such as gas and water processing, hydrogen production, and nanotechnology. Extending research on promising applications, such as pulsed corona processing, depends to a great extent on the availability of reliable, efficient and repetitive high-voltage pulsed power technology. Heavy-duty opening switches are the most critical components in high-voltage pulsed power systems with inductive energy storage. At the Ioffe Institute, an unconventional switching mechanism has been found, based on the fast recovery process in a diode. This article discusses the application of such a "drift-step-recovery-diode" for pulsed corona plasma generation. The principle of the diode-based nanosecond high-voltage generator will be discussed. The generator will be coupled to a corona reactor via a transmission-line transformer. The advantages of this concept, such as easy voltage transformation, load matching, switch protection and easy coupling with a dc bias voltage, will be discussed. The developed circuit is tested at both a resistive load and various corona reactors. Methods to optimize the energy transfer to a corona reactor have been evaluated. The impedance matching between the pulse generator and corona reactor can be significantly improved by using a dc bias voltage. At good matching, the corona energy increases and less energy reflects back to the generator. Matching can also be slightly improved by increasing the temperature in the corona reactor. More effective is to reduce the reactor pressure.

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.

Nuclear Hybrid Energy Systems FY16 Modeling Efforts at ORNL

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

Cetiner, Sacit M.; Greenwood, Michael Scott; Harrison, Thomas J.

A nuclear hybrid system uses a nuclear reactor as the basic power generation unit. The power generated by the nuclear reactor is utilized by one or more power customers as either thermal power, electrical power, or both. In general, a nuclear hybrid system will couple the nuclear reactor to at least one thermal power user in addition to the power conversion system. The definition and architecture of a particular nuclear hybrid system is flexible depending on local markets needs and opportunities. For example, locations in need of potable water may be best served by coupling a desalination plant to themore » nuclear system. Similarly, an area near oil refineries may have a need for emission-free hydrogen production. A nuclear hybrid system expands the nuclear power plant from its more familiar central power station role by diversifying its immediately and directly connected customer base. The definition, design, analysis, and optimization work currently performed with respect to the nuclear hybrid systems represents the work of three national laboratories. Idaho National Laboratory (INL) is the lead lab working with Argonne National Laboratory (ANL) and Oak Ridge National Laboratory. Each laboratory is providing modeling and simulation expertise for the integration of the hybrid system.« less

Phase 1 Space Fission Propulsion Energy Source Design

NASA Technical Reports Server (NTRS)

Houts, Mike; VanDyke, Melissa; Godfroy, Tom; Pedersen, Kevin; Martin, James; Dickens, Ricky; Salvail, Pat; Hrbud, Ivana; Carter, Robert; Rodgers, Stephen L. (Technical Monitor)

2002-01-01

Fission technology can enable rapid, affordable access to any point in the solar system. If fission propulsion systems are to be developed to their full potential; however, near-term customers must be identified and initial fission systems successfully developed, launched, and operated. Studies conducted in fiscal year 2001 (IISTP, 2001) show that fission electric propulsion (FEP) systems with a specific mass at or below 50 kg/kWjet could enhance or enable numerous robotic outer solar system missions of interest. At the required specific mass, it is possible to develop safe, affordable systems that meet mission requirements. To help select the system design to pursue, eight evaluation criteria were identified: system integration, safety, reliability, testability, specific mass, cost, schedule, and programmatic risk. A top-level comparison of four potential concepts was performed: a Testable, Passive, Redundant Reactor (TPRR), a Testable Multi-Cell In-Core Thermionic Reactor (TMCT), a Direct Gas Cooled Reactor (DGCR), and a Pumped Liquid Metal Reactor.(PLMR). Development of any of the four systems appears feasible. However, for power levels up to at least 500 kWt (enabling electric power levels of 125-175 kWe, given 25-35% power conversion efficiency) the TPRR has advantages related to several criteria and is competitive with respect to all. Hardware-based research and development has further increased confidence in the TPRR approach. Successful development and utilization of a "Phase I" fission electric propulsion system will enable advanced Phase 2 and Phase 3 systems capable of providing rapid, affordable access to any point in the solar system.

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.

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.

Control of algal production in a high rate algal pond: investigation through batch and continuous experiments.

PubMed

Derabe Maobe, H; Onodera, M; Takahashi, M; Satoh, H; Fukazawa, T

2014-01-01

For decades, arid and semi-arid regions in Africa have faced issues related to water availability for drinking, irrigation and livestock purposes. To tackle these issues, a laboratory scale greywater treatment system based on high rate algal pond (HRAP) technology was investigated in order to guide the operation of the pilot plant implemented in the 2iE campus in Ouagadougou (Burkina Faso). Because of the high suspended solids concentration generally found in effluents of this system, the aim of this study is to improve the performance of HRAPs in term of algal productivity and removal. To determine the selection mechanism of self-flocculated algae, three sets of sequencing batch reactors (SBRs) and three sets of continuous flow reactors (CFRs) were operated. Despite operation with the same solids retention time and the similarity of the algal growth rate found in these reactors, the algal productivity was higher in the SBRs owing to the short hydraulic retention time of 10 days in these reactors. By using a volume of CFR with twice the volume of our experimental CFRs, the algal concentration can be controlled during operation under similar physical conditions in both reactors.

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.

Verification of Modelica-Based Models with Analytical Solutions for Tritium Diffusion

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

Rader, Jordan D.; Greenwood, Michael Scott; Humrickhouse, Paul W.

Here, tritium transport in metal and molten salt fluids combined with diffusion through high-temperature structural materials is an important phenomenon in both magnetic confinement fusion (MCF) and molten salt reactor (MSR) applications. For MCF, tritium is desirable to capture for fusion fuel. For MSRs, uncaptured tritium potentially can be released to the environment. In either application, quantifying the time- and space-dependent tritium concentration in the working fluid(s) and structural components is necessary.Whereas capability exists specifically for calculating tritium transport in such systems (e.g., using TMAP for fusion reactors), it is desirable to unify the calculation of tritium transport with othermore » system variables such as dynamic fluid and structure temperature combined with control systems such as those that might be found in a system code. Some capability for radioactive trace substance transport exists in thermal-hydraulic systems codes (e.g., RELAP5-3D); however, this capability is not coupled to species diffusion through solids. Combined calculations of tritium transport and thermal-hydraulic solution have been demonstrated with TRIDENT but only for a specific type of MSR.Researchers at Oak Ridge National Laboratory have developed a set of Modelica-based dynamic system modeling tools called TRANsient Simulation Framework Of Reconfigurable Models (TRANSFORM) that were used previously to model advanced fission reactors and associated systems. In this system, the augmented TRANSFORM library includes dynamically coupled fluid and solid trace substance transport and diffusion. Results from simulations are compared against analytical solutions for verification.« less

Verification of Modelica-Based Models with Analytical Solutions for Tritium Diffusion

DOE PAGES

Rader, Jordan D.; Greenwood, Michael Scott; Humrickhouse, Paul W.

2018-03-20

Here, tritium transport in metal and molten salt fluids combined with diffusion through high-temperature structural materials is an important phenomenon in both magnetic confinement fusion (MCF) and molten salt reactor (MSR) applications. For MCF, tritium is desirable to capture for fusion fuel. For MSRs, uncaptured tritium potentially can be released to the environment. In either application, quantifying the time- and space-dependent tritium concentration in the working fluid(s) and structural components is necessary.Whereas capability exists specifically for calculating tritium transport in such systems (e.g., using TMAP for fusion reactors), it is desirable to unify the calculation of tritium transport with othermore » system variables such as dynamic fluid and structure temperature combined with control systems such as those that might be found in a system code. Some capability for radioactive trace substance transport exists in thermal-hydraulic systems codes (e.g., RELAP5-3D); however, this capability is not coupled to species diffusion through solids. Combined calculations of tritium transport and thermal-hydraulic solution have been demonstrated with TRIDENT but only for a specific type of MSR.Researchers at Oak Ridge National Laboratory have developed a set of Modelica-based dynamic system modeling tools called TRANsient Simulation Framework Of Reconfigurable Models (TRANSFORM) that were used previously to model advanced fission reactors and associated systems. In this system, the augmented TRANSFORM library includes dynamically coupled fluid and solid trace substance transport and diffusion. Results from simulations are compared against analytical solutions for verification.« less

Fiber Attachment Module Experiment (FAME): Using a Multiplexed Miniature Hollow Fiber Membrane Bioreactor Solution for Rapid Process Testing

NASA Astrophysics Data System (ADS)

Lunn, Griffin; Wheeler, Raymond; Hummerick, Mary; Birmele, Michele; Richards, Jeffrey; Coutts, Janelle; Koss, Lawrence; Spencer, Lashelle.; Johnsey, Marissa; Ellis, Ronald

Bioreactor research, even today, is mostly limited to continuous stirred-tank reactors (CSTRs). These are not an option for microgravity applications due to the lack of a gravity gradient to drive aeration as described by the Archimedes principle. This has led to testing of Hollow Fiber Membrane Bioreactors (HFMBs) for microgravity applications, including possible use for wastewater treatment systems for the International Space Station (ISS). Bioreactors and filtration systems for treating wastewater could avoid the need for harsh pretreatment chemicals and improve overall water recovery. However, the construction of these reactors is difficult and commercial off-the-shelf (COTS) versions do not exist in small sizes. We have used 1-L modular HFMBs in the past, but the need to perform rapid testing has led us to consider even smaller systems. To address this, we designed and built 125-mL, rectangular reactors, which we have called the Fiber Attachment Module Experiment (FAME) system. A polycarbonate rack of four square modules was developed with each module containing removable hollow fibers. Each FAME reactor is self-contained and can be easily plumbed with peristaltic and syringe pumps for continuous recycling of fluids and feeding, as well as fitted with sensors for monitoring pH, dissolved oxygen, and gas measurements similar to their larger counterparts. The first application tested in the FAME racks allowed analysis of over a dozen fiber surface treatments and three inoculation sources to achieve rapid reactor startup and biofilm attachment (based on carbon oxidation and nitrification of wastewater). With these miniature FAME reactors, data for this multi-factorial test were collected in duplicate over a six-month period; this greatly compressed time period required for gathering data needed to study and improve bioreactor performance.

Nuclear reactor construction with bottom supported reactor vessel

DOEpatents

Sharbaugh, John E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment structure base mat so as to insulate the reactor vessel bottom end wall from the containment structure base mat and allow the reactor vessel bottom end wall to freely expand as it heats up while providing continuous support thereof. Further, a deck is supported upon the side wall of the containment structure above the top open end of the reactor vessel, and a plurality of serially connected extendible and retractable annular bellows extend between the deck and the top open end of the reactor vessel and flexibly and sealably interconnect the reactor vessel at its top end to the deck. An annular guide ring is disposed on the containment structure and extends between its side wall and the top open end of the reactor vessel for providing lateral support of the reactor vessel top open end by limiting imposition of lateral loads on the annular bellows by the occurrence of a lateral seismic event.

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.

Nuclear power for space based systems

NASA Astrophysics Data System (ADS)

Livingston, J. M.; Ivanenok, Joseph F., III

1991-09-01

A 100 kWe closed Brayton cycle power conversion system utilizing a recuperator coupled to a NERVA derivative reactor for a lunar power plant is presented. Power plant mass versus recuperator effectiveness, compressor inlet temperature, and turbine pressure ratio are described.

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

NASA Technical Reports Server (NTRS)

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

2005-01-01

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

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

System-Level Heat Transfer Analysis, Thermal- Mechanical Cyclic Stress Analysis, and Environmental Fatigue Modeling of a Two-Loop Pressurized Water Reactor. A Preliminary Study

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

Mohanty, Subhasish; Soppet, William; Majumdar, Saurin

This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in April 2015 under the work package for environmentally assisted fatigue under DOE's Light Water Reactor Sustainability program. In this report, updates are discussed related to a system level preliminary finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent thermal-mechanical stress analysis were performed for typical design-basis thermal-mechanical fatigue cycles. The in-air fatigue lives of components, such as the hot and cold legs,more » were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in NUREG-6909. The discussed models and results are very preliminary. Further advancement of the discussed model is required for more accurate life prediction of reactor components. This report only presents the work related to finite element modelling activities. However, in between multiple tensile and fatigue tests were conducted. The related experimental results will be presented in the year-end report.« less

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.

Removal of gadolinium, a neutron poison from the moderator system of nuclear reactors.

PubMed

Rufus, A L; Kumar, Padma S; Jeena, K; Velmurugan, S

2018-01-15

Gadolinium as gadolinium nitrate is used as neutron poison in the moderator system for regulating and controlling the power generation of Pressurized Heavy Water Reactors (PHWR) and proposed to be used in Advanced Heavy Water Reactors (AHWR) owing to its high neutron absorption cross section. Removal of the added gadolinium nitrate (Gd 3+ and NO 3 - ) from the system after its intended use is done using ion exchange resins. In the present investigation, attempts have been made to optimize the ion exchange process for generation of low radioactive waste and maximize utilization of the ion exchange resins by employing different types of resins and different modes of operation. The investigations revealed that use of mixed bed (MB) resin column consisting of Strong Acid Cation (SAC) resin and Strong Base Anion (SBA) resin followed by SAC resin column is efficient in removing the Gd 3+ and NO 3 - from the system besides maintaining the pH of the moderator system in the desirable regime, where gadolinium does not get precipitated as its hydroxide. Copyright © 2017 Elsevier B.V. All rights reserved.

Applying design principles to fusion reactor configurations for propulsion in space

NASA Technical Reports Server (NTRS)

Carpenter, Scott A.; Deveny, Marc E.; Schulze, Norman R.

1993-01-01

The application of fusion power to space propulsion requires rethinking the engineering-design solution to controlled-fusion energy. Whereas the unit cost of electricity (COE) drives the engineering-design solution for utility-based fusion reactor configurations; initial mass to low earth orbit (IMLEO), specific jet power (kW(thrust)/kg(engine)), and reusability drive the engineering-design solution for successful application of fusion power to space propulsion. We applied three design principles (DP's) to adapt and optimize three candidate-terrestrial-fusion-reactor configurations for propulsion in space. The three design principles are: provide maximum direct access to space for waste radiation, operate components as passive radiators to minimize cooling-system mass, and optimize the plasma fuel, fuel mix, and temperature for best specific jet power. The three candidate terrestrial fusion reactor configurations are: the thermal barrier tandem mirror (TBTM), field reversed mirror (FRM), and levitated dipole field (LDF). The resulting three candidate space fusion propulsion systems have their IMLEO minimized and their specific jet power and reusability maximized. We performed a preliminary rating of these configurations and concluded that the leading engineering-design solution to space fusion propulsion is a modified TBTM that we call the Mirror Fusion Propulsion System (MFPS).

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions.

PubMed

Geslot, B; Vermeeren, L; Filliatre, P; Lopez, A Legrand; Barbot, L; Jammes, C; Bréaud, S; Oriol, L; Villard, J-F

2011-03-01

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 10(20) n∕cm(2). A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

NASA Astrophysics Data System (ADS)

Geslot, B.; Vermeeren, L.; Filliatre, P.; Lopez, A. Legrand; Barbot, L.; Jammes, C.; Bréaud, S.; Oriol, L.; Villard, J.-F.

2011-03-01

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 1020 n/cm2. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

Laser-based sensor for a coolant leak detection in a nuclear reactor

NASA Astrophysics Data System (ADS)

Kim, T.-S.; Park, H.; Ko, K.; Lim, G.; Cha, Y.-H.; Han, J.; Jeong, D.-Y.

2010-08-01

Currently, the nuclear industry needs strongly a reliable detection system to continuously monitor a coolant leak during a normal operation of reactors for the ensurance of nuclear safety. In this work, we propose a new device for the coolant leak detection based on tunable diode laser spectroscopy (TDLS) by using a compact diode laser. For the feasibility experiment, we established an experimental setup consisted of a near-IR diode laser with a wavelength of about 1392 nm, a home-made multi-pass cell and a sample injection system. The feasibility test was performed for the detection of the heavy water (D2O) leaks which can happen in a pressurized heavy water reactor (PWHR). As a result, the device based on the TDLS is shown to be operated successfully in detecting a HDO molecule, which is generated from the leaked heavy water by an isotope exchange reaction between D2O and H2O. Additionally, it is suggested that the performance of the new device, such as sensitivity and stability, can be improved by adapting a cavity enhanced absorption spectroscopy and a compact DFB diode laser. We presume that this laser-based leak detector has several advantages over the conventional techniques currently employed in the nuclear power plant, such as radiation monitoring, humidity monitoring and FT-IR spectroscopy.

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.

Thermally Simulated Testing of a Direct-Drive Gas-Cooled Nuclear Reactor

NASA Technical Reports Server (NTRS)

Godfroy, Thomas; Bragg-Sitton, Shannon; VanDyke, Melissa

2003-01-01

This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet-sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrical thermal simulation of reactor components and concepts.

Implicit time-integration method for simultaneous solution of a coupled non-linear system

NASA Astrophysics Data System (ADS)

Watson, Justin Kyle

Historically large physical problems have been divided into smaller problems based on the physics involved. This is no different in reactor safety analysis. The problem of analyzing a nuclear reactor for design basis accidents is performed by a handful of computer codes each solving a portion of the problem. The reactor thermal hydraulic response to an event is determined using a system code like TRAC RELAP Advanced Computational Engine (TRACE). The core power response to the same accident scenario is determined using a core physics code like Purdue Advanced Core Simulator (PARCS). Containment response to the reactor depressurization in a Loss Of Coolant Accident (LOCA) type event is calculated by a separate code. Sub-channel analysis is performed with yet another computer code. This is just a sample of the computer codes used to solve the overall problems of nuclear reactor design basis accidents. Traditionally each of these codes operates independently from each other using only the global results from one calculation as boundary conditions to another. Industry's drive to uprate power for reactors has motivated analysts to move from a conservative approach to design basis accident towards a best estimate method. To achieve a best estimate calculation efforts have been aimed at coupling the individual physics models to improve the accuracy of the analysis and reduce margins. The current coupling techniques are sequential in nature. During a calculation time-step data is passed between the two codes. The individual codes solve their portion of the calculation and converge to a solution before the calculation is allowed to proceed to the next time-step. This thesis presents a fully implicit method of simultaneous solving the neutron balance equations, heat conduction equations and the constitutive fluid dynamics equations. It discusses the problems involved in coupling different physics phenomena within multi-physics codes and presents a solution to these problems. The thesis also outlines the basic concepts behind the nodal balance equations, heat transfer equations and the thermal hydraulic equations, which will be coupled to form a fully implicit nonlinear system of equations. The coupling of separate physics models to solve a larger problem and improve accuracy and efficiency of a calculation is not a new idea, however implementing them in an implicit manner and solving the system simultaneously is. Also the application to reactor safety codes is new and has not be done with thermal hydraulics and neutronics codes on realistic applications in the past. The coupling technique described in this thesis is applicable to other similar coupled thermal hydraulic and core physics reactor safety codes. This technique is demonstrated using coupled input decks to show that the system is solved correctly and then verified by using two derivative test problems based on international benchmark problems the OECD/NRC Three mile Island (TMI) Main Steam Line Break (MSLB) problem (representative of pressurized water reactor analysis) and the OECD/NRC Peach Bottom (PB) Turbine Trip (TT) benchmark (representative of boiling water reactor analysis).

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.

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.

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…

KERENA safety concept in the context of the Fukushima accident

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

Zacharias, T.; Novotny, C.; Bielor, E.

Within the last three years AREVA NP and E.On KK finalized the basic design of KERENA which is a medium sized innovative boiling water reactor, based on the operational experience of German BWR nuclear power plants (NPPs). It is a generation III reactor design with a net electrical output of about 1250 MW. It combines active safety equipment of service-proven designs with new passive safety components, both safety classified. The passive systems utilize basic laws of physics, such as gravity and natural convection, enabling them to function without electric power. Even actuation of these systems is performed thanks to basicmore » physic laws. The degree of diversity in component and system design, achieved by combining active and passive equipment, results in a very low core damage frequency. The Fukushima accident enhanced the world wide discussion about the safety of operating nuclear power plants. World wide stress tests for operating nuclear power plants are being performed embracing both natural and man made hazards. Beside the assessment of existing power plants, also new designs are analyzed regarding the system response to beyond design base accidents. KERENA's optimal combination of diversified cooling systems (active and passive) allows passing efficiently such tests, with a high level of confidence. This paper describes the passive safety components and the KERENA reactor behavior after a Fukushima like accident. (authors)« less

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...

A CFD Model for High Pressure Liquid Poison Injection for CANDU-6 Shutdown System No. 2

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

Bo Wook Rhee; Chang Jun Jeong; Hye Jeong Yun

2002-07-01

In CANDU reactor one of the two reactor shutdown systems is the liquid poison injection system which injects the highly pressurized liquid neutron poison into the moderator tank via small holes on the nozzle pipes. To ensure the safe shutdown of a reactor it is necessary for the poison curtains generated by jets provide quick, and enough negative reactivity to the reactor during the early stage of the accident. In order to produce the neutron cross section necessary to perform this work, the poison concentration distribution during the transient is necessary. In this study, a set of models for analyzingmore » the transient poison concentration induced by this high pressure poison injection jet activated upon the reactor trip in a CANDU-6 reactor moderator tank has been developed and used to generate the poison concentration distribution of the poison curtains induced by the high pressure jets injected into the vacant region between the pressure tube banks. The poison injection rate through the jet holes drilled on the nozzle pipes is obtained by a 1-D transient hydrodynamic code called, ALITRIG, and this injection rate is used to provide the inlet boundary condition to a 3-D CFD model of the moderator tank based on CFX4.3, a CFD code, to simulate the formation of the poison jet curtain inside the moderator tank. For validation, an attempt was made to validate this model against a poison injection experiment performed at BARC. As conclusion this set of models is judged to be appropriate. (authors)« less

ORIGEN-based Nuclear Fuel Inventory Module for Fuel Cycle Assessment: Final Project Report

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

Skutnik, Steven E.

The goal of this project, “ORIGEN-based Nuclear Fuel Depletion Module for Fuel Cycle Assessment" is to create a physics-based reactor depletion and decay module for the Cyclus nuclear fuel cycle simulator in order to assess nuclear fuel inventories over a broad space of reactor operating conditions. The overall goal of this approach is to facilitate evaluations of nuclear fuel inventories for a broad space of scenarios, including extended used nuclear fuel storage and cascading impacts on fuel cycle options such as actinide recovery in used nuclear fuel, particularly for multiple recycle scenarios. The advantages of a physics-based approach (compared tomore » a recipe-based approach which has been typically employed for fuel cycle simulators) is in its inherent flexibility; such an approach can more readily accommodate the broad space of potential isotopic vectors that may be encountered under advanced fuel cycle options. In order to develop this flexible reactor analysis capability, we are leveraging the Origen nuclear fuel depletion and decay module from SCALE to produce a standalone “depletion engine” which will serve as the kernel of a Cyclus-based reactor analysis module. The ORIGEN depletion module is a rigorously benchmarked and extensively validated tool for nuclear fuel analysis and thus its incorporation into the Cyclus framework can bring these capabilities to bear on the problem of evaluating long-term impacts of fuel cycle option choices on relevant metrics of interest, including materials inventories and availability (for multiple recycle scenarios), long-term waste management and repository impacts, etc. Developing this Origen-based analysis capability for Cyclus requires the refinement of the Origen analysis sequence to the point where it can reasonably be compiled as a standalone sequence outside of SCALE; i.e., wherein all of the computational aspects of Origen (including reactor cross-section library processing and interpolation, input and output processing, and depletion/decay solvers) can be self-contained into a single executable sequence. Further, to embed this capability into other software environments (such as the Cyclus fuel cycle simulator) requires that Origen’s capabilities be encapsulated into a portable, self-contained library which other codes can then call directly through function calls, thereby directly accessing the solver and data processing capabilities of Origen. Additional components relevant to this work include modernization of the reactor data libraries used by Origen for conducting nuclear fuel depletion calculations. This work has included the development of new fuel assembly lattices not previously available (such as for CANDU heavy-water reactor assemblies) as well as validation of updated lattices for light-water reactors updated to employ modern nuclear data evaluations. The CyBORG reactor analysis module as-developed under this workscope is fully capable of dynamic calculation of depleted fuel compositions from all commercial U.S. reactor assembly types as well as a number of international fuel types, including MOX, VVER, MAGNOX, and PHWR CANDU fuel assemblies. In addition, the Origen-based depletion engine allows for CyBORG to evaluate novel fuel assembly and reactor design types via creation of Origen reactor data libraries via SCALE. The establishment of this new modeling capability affords fuel cycle modelers a substantially improved ability to model dynamically-changing fuel cycle and reactor conditions, including recycled fuel compositions from fuel cycle scenarios involving material recycle into thermal-spectrum systems.« less

Population dynamics in controlled unsteady-state systems: An application to the degradation of glyphosate in a sequencing batch reactor

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

Devarakonda, M.S.

1988-01-01

Control over population dynamics and organism selection in a biological waste treatment system provides an effective means of engineering process efficiency. Examples of applications of organism selection include control of filamentous organisms, biological nutrient removal, industrial waste treatment requiring the removal of specific substrates, and hazardous waste treatment. Inherently, full scale biological waste treatment systems are unsteady state systems due to the variations in the waste streams and mass flow rates of the substrates. Some systems, however, have the capacity to impose controlled selective pressures on the biological population by means of their operation. An example of such a systemmore » is the Sequencing Batch Reactor (SBR) which was the experimental system utilized in this research work. The concepts of organism selection were studied in detail for the biodegradation of a herbicide waste stream, with glyphosate as the target compound. The SBR provided a reactor configuration capable of exerting the necessary selective pressures to select and enrich for a glyphosate degrading population. Based on results for bench scale SBRs, a hypothesis was developed to explain population dynamics in glyphosate degrading systems.« 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.

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

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.

A new halogen-free chemical oscillator: the reaction between permanganate ion and ninhydrin in a continuously stirred tank reactor

NASA Astrophysics Data System (ADS)

Treindl, Ľudovít; Nagy, Arpád

1987-07-01

The reaction between permanganate ion and ninhydrin in the presence of phosphoric acid in aqueous solution shows sustained oscillations in a continuously stirred tank reactor (CSTR). It exhibits a kinetic bistability between an oscillatory and a stationary state. Our new oscillating system seems to be a second permanganate chemical oscillator, thus broadening the small group of non-halogen-based chemical oscillators.

Covariance Applications in Criticality Safety, Light Water Reactor Analysis, and Spent Fuel Characterization

DOE PAGES

Williams, M. L.; Wiarda, D.; Ilas, G.; ...

2014-06-15

Recently, we processed a new covariance data library based on ENDF/B-VII.1 for the SCALE nuclear analysis code system. The multigroup covariance data are discussed here, along with testing and application results for critical benchmark experiments. Moreover, the cross section covariance library, along with covariances for fission product yields and decay data, is used to compute uncertainties in the decay heat produced by a burned reactor fuel assembly.

Progress in the Production of JP-8 Based Hydrogen and Advanced Tactical Fuels for Military Applications

DTIC Science & Technology

2011-02-01

of a multi- year program to develop, optimize, and demonstrate the military viability of a technology for on-demand production of high...continuous reactor system used for kinetic rate data experiment 86 52 Schematic of a differential reactor. The catalyst bed is kept small , and...program to develop, optimize, and demonstrate the military viability of a technology for on-demand production of high-pressure hydrogen for fuel

Ultrafiltration membrane reactors for enzymatic resolution of amino acids: design model and optimization.

PubMed

Bódalo, A; Gómez, J L.; Gómez, E; Bastida, J; Máximo, M F.; Montiel, M C.

2001-03-08

In this paper the possibility of continuous resolution of DL-phenylalanine, catalyzed by L-aminoacylase in a ultrafiltration membrane reactor (UFMR) is presented. A simple design model, based on previous kinetic studies, has been demonstrated to be capable of describing the behavior of the experimental system. The model has been used to determine the optimal experimental conditions to carry out the asymmetrical hydrolysis of N-acetyl-DL-phenylalanine.

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.

Transmutation of Isotopes --- Ecological and Energy Production Aspects

NASA Astrophysics Data System (ADS)

Gudowski, Waclaw

2000-01-01

This paper describes principles of Accelerator-Driven Transmutation of Nuclear Wastes (ATW) and gives some flavour of the most important topics which are today under investigations in many countries. An assessment of the potential impact of ATW on a future of nuclear energy is also given. Nuclear reactors based on self-sustained fission reactions --- after spectacular development in fifties and sixties, that resulted in deployment of over 400 power reactors --- are wrestling today more with public acceptance than with irresolvable technological problems. In a whole spectrum of reasons which resulted in today's opposition against nuclear power few of them are very relevant for the nuclear physics community and they arose from the fact that development of nuclear power had been handed over to the nuclear engineers and technicians with some generically unresolved problems, which should have been solved properly by nuclear scientists. In a certain degree of simplification one can say, that most of the problems originate from very specific features of a fission phenomenon: self-sustained chain reaction in fissile materials and very strong radioactivity of fission products and very long half-life of some of the fission and activation products. And just this enormous concentration of radioactive fission products in the reactor core is the main problem of managing nuclear reactors: it requires unconditional guarantee for the reactor core integrity in order to avoid radioactive contamination of the environment; it creates problems to handle decay heat in the reactor core and finally it makes handling and/or disposal of spent fuel almost a philosophical issue, due to unimaginable long time scales of radioactive decay of some isotopes. A lot can be done to improve the design of conventional nuclear reactors (like Light Water Reactors); new, better reactors can be designed but it seems today very improbable to expect any radical change in the public perception of conventional nuclear power. In this context a lot of hopes and expectations have been expressed for novel systems called Accelerator-Driven Systems, Accelerator-Driven Transmutation of Waste or just Hybrid Reactors. All these names are used for description of the same nuclear system combining a powerful particle accelerator with a subcritical reactor. A careful analysis of possible environmental impact of ATW together with limitation of this technology is presented also in this paper.

A peculiar segmented flow microfluidics for isoquercitrin biosynthesis based on coupling of reaction and separation.

PubMed

Gong, An; Gu, Shuang-Shuang; Wang, Jun; Sheng, Sheng; Wu, Fu-An

2015-10-01

A segmented flow containing a buffer-ionic liquid/solvent in a micro-channel reactor was applied to synthesize isoquercitrin by the hesperidinase-catalyzed selective hydrolysis of rutin, based on a novel system of reaction coupling with separation. Within the developed microchannel reactor with one T-shaped inlet and outlet, the maximum isoquercitrin yield (101.7 ± 2.6%) was achieved in 20 min at 30 °C and 4 μL/min. Compared with a continuous-flow reactor, reaction rate was increased 4-fold due to a glycine-sodium hydroxide:[Bmim][BF4]/glycerol triacetate (1:1, v/v) system that formed a slug flow in microchannel and significantly increased mass transfer rates. The mass transfer coefficient significantly increased and exhibited a linear relationship with the flow rate. Hesperidinase could be efficiently reused at least 5 times, without losing any activity. The bonding mechanism and secondary structure of hesperidinase indicated that hesperidinase had a greater affinity to rutin at a production rate of 4 μL/min in this segmented flow microreactor. Copyright © 2015 Elsevier Ltd. All rights reserved.

Application of reliability-centered-maintenance to BWR ECCS motor operator valve performance

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

Feltus, M.A.; Choi, Y.A.

1993-01-01

This paper describes the application of reliability-centered maintenance (RCM) methods to plant probabilistic risk assessment (PRA) and safety analyses for four boiling water reactor emergency core cooling systems (ECCSs): (1) high-pressure coolant injection (HPCI); (2) reactor core isolation cooling (RCIC); (3) residual heat removal (RHR); and (4) core spray systems. Reliability-centered maintenance is a system function-based technique for improving a preventive maintenance program that is applied on a component basis. Those components that truly affect plant function are identified, and maintenance tasks are focused on preventing their failures. The RCM evaluation establishes the relevant criteria that preserve system function somore » that an RCM-focused approach can be flexible and dynamic.« less

ORNL experience and perspectives related to processing of thorium and 233U for nuclear fuel

DOE PAGES

Croff, Allen G.; Collins, Emory D.; Del Cul, G. D.; ...

2016-05-01

Thorium-based nuclear fuel cycles have received renewed attention in both research and public circles since about the year 2000. Much of the attention has been focused on nuclear fission energy production that utilizes thorium as a fertile element for producing fissionable 233U for recycle in thermal reactors, fast reactors, or externally driven systems. Here, lesser attention has been paid to other fuel cycle operations that are necessary for implementation of a sustainable thorium-based fuel cycle such as reprocessing and fabrication of recycle fuels containing 233U.

Development costs for a nuclear electric propulsion stage.

NASA Technical Reports Server (NTRS)

Mondt, J. F.; Prickett, W. Z.

1973-01-01

Development costs are presented for an unmanned nuclear electric propulsion (NEP) stage based upon a liquid metal cooled, in-core thermionic reactor. A total of 120 kWe are delivered to the thrust subsystem which employs mercury ion engines for electric propulsion. This study represents the most recent cost evaluation of the development of a reactor power system for a wide range of nuclear space power applications. These include geocentric, and outer planet and other deep space missions. The development program is described for the total NEP stage, based upon specific development programs for key NEP stage components and subsystems.

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

The Experimental Breeder Reactor II seismic probabilistic risk assessment

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

Roglans, J; Hill, D J

1994-02-01

The Experimental Breeder Reactor II (EBR-II) is a US Department of Energy (DOE) Category A research reactor located at Argonne National Laboratory (ANL)-West in Idaho. EBR-II is a 62.5 MW-thermal Liquid Metal Reactor (LMR) that started operation in 1964 and it is currently being used as a testbed in the Integral Fast Reactor (IFR) Program. ANL has completed a Level 1 Probabilistic Risk Assessment (PRA) for EBR-II. The Level 1 PRA for internal events and most external events was completed in June 1991. The seismic PRA for EBR-H has recently been completed. The EBR-II reactor building contains the reactor, themore » primary system, and the decay heat removal systems. The reactor vessel, which contains the core, and the primary system, consisting of two primary pumps and an intermediate heat exchanger, are immersed in the sodium-filled primary tank, which is suspended by six hangers from a beam support structure. Three systems or functions in EBR-II were identified as the most significant from the standpoint of risk of seismic-induced fuel damage: (1) the reactor shutdown system, (2) the structural integrity of the passive decay heat removal systems, and (3) the integrity of major structures, like the primary tank containing the reactor that could threaten both the reactivity control and decay heat removal functions. As part of the seismic PRA, efforts were concentrated in studying these three functions or systems. The passive safety response of EBR-II reactor -- both passive reactivity shutdown and passive decay heat removal, demonstrated in a series of tests in 1986 -- was explicitly accounted for in the seismic PRA as it had been included in the internal events assessment.« 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.

Weld monitor and failure detector for nuclear reactor system

DOEpatents

Sutton, Jr., Harry G.

1987-01-01

Critical but inaccessible welds in a nuclear reactor system are monitored throughout the life of the reactor by providing small aperture means projecting completely through the reactor vessel wall and also through the weld or welds to be monitored. The aperture means is normally sealed from the atmosphere within the reactor. Any incipient failure or cracking of the weld will cause the environment contained within the reactor to pass into the aperture means and thence to the outer surface of the reactor vessel where its presence is readily detected.

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.

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.

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

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.

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

Nuclear reactor cooling system decontamination reagent regeneration. [PWR; BWR

DOEpatents

Anstine, L.D.; James, D.B.; Melaika, E.A.; Peterson, J.P. Jr.

1980-06-06

An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution is described. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

Nuclear reactor cooling system decontamination reagent regeneration

DOEpatents

Anstine, Larry D.; James, Dean B.; Melaika, Edward A.; Peterson, Jr., John P.

1985-01-01

An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

Recovery of phosphorus and volatile fatty acids from wastewater and food waste with an iron-flocculation sequencing batch reactor and acidogenic co-fermentation.

PubMed

Li, Ruo-Hong; Li, Xiao-Yan

2017-12-01

A sequencing batch reactor-based system was developed for enhanced phosphorus (P) removal and recovery from municipal wastewater. The system consists of an iron-dosing SBR for P precipitation and a side-stream anaerobic reactor for sludge co-fermentation with food waste. During co-fermentation, sludge and food waste undergo acidogenesis, releasing phosphates under acidic conditions and producing volatile fatty acids (VFAs) into the supernatant. A few types of typical food waste were investigated for their effectiveness in acidogenesis and related enzymatic activities. The results show that approximately 96.4% of total P in wastewater was retained in activated sludge. Food waste with a high starch content favoured acidogenic fermentation. Around 55.7% of P from wastewater was recovered as vivianite, and around 66% of food waste loading was converted into VFAs. The new integration formed an effective system for wastewater treatment, food waste processing and simultaneous recovery of P and VFAs. Copyright © 2017 Elsevier Ltd. All rights reserved.

Application of a multi-channel system for continuous monitoring and an early warning system.

PubMed

Lee, J H; Song, C H; Kim, B C; Gu, M B

2006-01-01

A multi-channel continuous toxicity monitoring system developed in our laboratory, based on two-stage mini-bioreactors, was successfully implemented in the form of computer-based data acquisition. The multi-channel system consists of a series of a two-stage minibioreactor systems connected by a fiber optic probe to a luminometer, and uses genetically engineered bioluminescent bacteria for the detection of the potential toxicity from the soluble chemicals. This system can be stably and continuously operated due to the separation of the culture reactor from the test reactor and accomplish easy and long-term monitoring without system shut down by abrupt inflows of severe polluting chemicals. Four different recombinant bioluminescent bacteria were used in different channels so that the modes of the samples toxicities can be reasonably identified and evaluated based upon the response signature of each channel. The bioluminescent signatures were delivered from four channels by switching one at once, while the data is automatically logged to an IBM compatible computer. We also achieved the enhancement of the system through the manipulation of the dilution rate and the use of thermo-lux fusion strains. Finally, this system is now being implemented to a drinking water reservoir and river for remote sensing as an early warning system.

Development of on-line monitoring system for Nuclear Power Plant (NPP) using neuro-expert, noise analysis, and modified neural networks

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

Subekti, M.; Center for Development of Reactor Safety Technology, National Nuclear Energy Agency of Indonesia, Puspiptek Complex BO.80, Serpong-Tangerang, 15340; Ohno, T.

2006-07-01

The neuro-expert has been utilized in previous monitoring-system research of Pressure Water Reactor (PWR). The research improved the monitoring system by utilizing neuro-expert, conventional noise analysis and modified neural networks for capability extension. The parallel method applications required distributed architecture of computer-network for performing real-time tasks. The research aimed to improve the previous monitoring system, which could detect sensor degradation, and to perform the monitoring demonstration in High Temperature Engineering Tested Reactor (HTTR). The developing monitoring system based on some methods that have been tested using the data from online PWR simulator, as well as RSG-GAS (30 MW research reactormore » in Indonesia), will be applied in HTTR for more complex monitoring. (authors)« less

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.

Biological hydrogen production by Clostridium acetobutylicum in an unsaturated flow reactor.

PubMed

Zhang, Husen; Bruns, Mary Ann; Logan, Bruce E

2006-02-01

A mesophilic unsaturated flow (trickle bed) reactor was designed and tested for H2 production via fermentation of glucose. The reactor consisted of a column packed with glass beads and inoculated with a pure culture (Clostridium acetobutylicum ATCC 824). A defined medium containing glucose was fed at a flow rate of 1.6 mL/min (0.096 L/h) into the capped reactor, producing a hydraulic retention time of 2.1 min. Gas-phase H2 concentrations were constant, averaging 74 +/- 3% for all conditions tested. H2 production rates increased from 89 to 220 mL/hL of reactor when influent glucose concentrations were varied from 1.0 to 10.5 g/L. Specific H2 production rate ranged from 680 to 1270 mL/g glucose per liter of reactor (total volume). The H2 yield was 15-27%, based on a theoretical limit by fermentation of 4 moles of H2 from 1 mole of glucose. The major fermentation by-products in the liquid effluent were acetate and butyrate. The reactor rapidly (within 60-72 h) became clogged with biomass, requiring manual cleaning of the system. In order to make long-term operation of the reactor feasible, biofilm accumulation in the reactor will need to be controlled through some process such as backwashing. These tests using an unsaturated flow reactor demonstrate the feasibility of the process to produce high H2 gas concentrations in a trickle-bed type of reactor. A likely application of this reactor technology could be H2 gas recovery from pre-treatment of high carbohydrate-containing wastewaters.

Hydrodynamics of Packed Bed Reactor in Low Gravity

NASA Technical Reports Server (NTRS)

Motil, Brian J.; Nahra, Henry K.; Balakotaiah, Vemuri

2005-01-01

Packed bed reactors are well known for their vast and diverse applications in the chemical industry; from gas absorption, to stripping, to catalytic conversion. Use of this type of reactor in terrestrial applications has been rather extensive because of its simplicity and relative ease of operation. Developing similar reactors for use in microgravity is critical to many space-based advanced life support systems. However, the hydrodynamics of two-phase flow packed bed reactors in this new environment and the effects of one physiochemical process on another has not been adequately assessed. Surface tension or capillary forces play a much greater role which results in a shifting in flow regime transitions and pressure drop. Results from low gravity experiments related to flow regimes and two-phase pressure drop models are presented in this paper along with a description of plans for a flight experiment on the International Space Station (ISS). Understanding the packed bed hydrodynamics and its effects on mass transfer processes in microgravity is crucial for the design of packed bed chemical or biological reactors to be used for water reclamation and other life support processes involving water purification.

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.

Designing visual displays and system models for safe reactor operations based on the user`s perspective of the system

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

Brown-VanHoozer, S.A.

Most designers are not schooled in the area of human-interaction psychology and therefore tend to rely on the traditional ergonomic aspects of human factors when designing complex human-interactive workstations related to reactor operations. They do not take into account the differences in user information processing behavior and how these behaviors may affect individual and team performance when accessing visual displays or utilizing system models in process and control room areas. Unfortunately, by ignoring the importance of the integration of the user interface at the information process level, the result can be sub-optimization and inherently error- and failure-prone systems. Therefore, tomore » minimize or eliminate failures in human-interactive systems, it is essential that the designers understand how each user`s processing characteristics affects how the user gathers information, and how the user communicates the information to the designer and other users. A different type of approach in achieving this understanding is Neuro Linguistic Programming (NLP). The material presented in this paper is based on two studies involving the design of visual displays, NLP, and the user`s perspective model of a reactor system. The studies involve the methodology known as NLP, and its use in expanding design choices from the user`s ``model of the world,`` in the areas of virtual reality, workstation design, team structure, decision and learning style patterns, safety operations, pattern recognition, and much, much more.« less

DNA-Based Enzyme Reactors and Systems

PubMed Central

Linko, Veikko; Nummelin, Sami; Aarnos, Laura; Tapio, Kosti; Toppari, J. Jussi; Kostiainen, Mauri A.

2016-01-01

During recent years, the possibility to create custom biocompatible nanoshapes using DNA as a building material has rapidly emerged. Further, these rationally designed DNA structures could be exploited in positioning pivotal molecules, such as enzymes, with nanometer-level precision. This feature could be used in the fabrication of artificial biochemical machinery that is able to mimic the complex reactions found in living cells. Currently, DNA-enzyme hybrids can be used to control (multi-enzyme) cascade reactions and to regulate the enzyme functions and the reaction pathways. Moreover, sophisticated DNA structures can be utilized in encapsulating active enzymes and delivering the molecular cargo into cells. In this review, we focus on the latest enzyme systems based on novel DNA nanostructures: enzyme reactors, regulatory devices and carriers that can find uses in various biotechnological and nanomedical applications. PMID:28335267

Spectral measurements of direct and scattered gamma radiation at a boiling-water reactor site

NASA Astrophysics Data System (ADS)

Block, R. C.; Preiss, I. L.; Ryan, R. M.; Vargo, G. J.

1990-12-01

Quantitative surveys of direct and scattered gamma radiation emitted from the steam-power conversion systems of a boiling-water reactor and other on-site radiation sources were made using a directionally shielded HPGe gamma spectrometry system. The purpose of this study was to obtain data on the relative contributions and energy distributions of direct and scattered gamma radiation in the site environs. The principal radionuclide of concern in this study is 16N produced by the 16O(n,p) 16N reaction in the reactor coolant. Due to changes in facility operation resulting from the implementation of hydrogen water chemistry (HWC), the amount of 16N transported from the reactor to the main steam system under full power operation is excepted to increase by a factor of 1.2 to 5.0. This increase in the 16N source term in the nuclear steam must be considered in the design of new facilities to be constructed on site as well as the evaluation of existing facilities with repect to ALARA (As Low As Reasonably Achievable) dose limits in unrestricted areas. This study consisted of base-line measurements taken under normal BWR chemistry conditions in October, 1987 and a corresponding set taken under HWC conditions in July, 1988. Ground-level and elevated measurements, corresponding to second-story building height, were obtained. The primary conclusion of this study is that direct radiation from the steam-power conversion system is the predominant source of radiation in the site environs of this reactor and that air scattering (i.e. skyshine) does not appear to be significant.

Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

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

Corradin, Michael; Anderson, M.; Muci, M.

This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintainmore » similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.« less

Control of autothermal reforming reactor of diesel fuel

NASA Astrophysics Data System (ADS)

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

2016-05-01

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

MODELING THE AMBIENT CONDITION EFFECTS OF AN AIR-COOLED NATURAL CIRCULATION SYSTEM

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

Hu, Rui; Lisowski, Darius D.; Bucknor, Matthew

The Reactor Cavity Cooling System (RCCS) is a passive safety concept under consideration for the overall safety strategy of advanced reactors such as the High Temperature Gas-Cooled Reactor (HTGR). One such variant, air-cooled RCCS, uses natural convection to drive the flow of air from outside the reactor building to remove decay heat during normal operation and accident scenarios. The Natural convection Shutdown heat removal Test Facility (NSTF) at Argonne National Laboratory (“Argonne”) is a half-scale model of the primary features of one conceptual air-cooled RCCS design. The facility was constructed to carry out highly instrumented experiments to study the performancemore » of the RCCS concept for reactor decay heat removal that relies on natural convection cooling. Parallel modeling and simulation efforts were performed to support the design, operation, and analysis of the natural convection system. Throughout the testing program, strong influences of ambient conditions were observed in the experimental data when baseline tests were repeated under the same test procedures. Thus, significant analysis efforts were devoted to gaining a better understanding of these influences and the subsequent response of the NSTF to ambient conditions. It was determined that air humidity had negligible impacts on NSTF system performance and therefore did not warrant consideration in the models. However, temperature differences between the building exterior and interior air, along with the outside wind speed, were shown to be dominant factors. Combining the stack and wind effects together, an empirical model was developed based on theoretical considerations and using experimental data to correlate zero-power system flow rates with ambient meteorological conditions. Some coefficients in the model were obtained based on best fitting the experimental data. The predictive capability of the empirical model was demonstrated by applying it to the new set of experimental data. The empirical model was also implemented in the computational models of the NSTF using both RELAP5-3D and STARCCM+ codes. Accounting for the effects of ambient conditions, simulations from both codes predicted the natural circulation flow rates very well.« less

A Passive System Reliability Analysis for a Station Blackout

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

Brunett, Acacia; Bucknor, Matthew; Grabaskas, David

2015-05-03

The latest iterations of advanced reactor designs have included increased reliance on passive safety systems to maintain plant integrity during unplanned sequences. While these systems are advantageous in reducing the reliance on human intervention and availability of power, the phenomenological foundations on which these systems are built require a novel approach to a reliability assessment. Passive systems possess the unique ability to fail functionally without failing physically, a result of their explicit dependency on existing boundary conditions that drive their operating mode and capacity. Argonne National Laboratory is performing ongoing analyses that demonstrate various methodologies for the characterization of passivemore » system reliability within a probabilistic framework. Two reliability analysis techniques are utilized in this work. The first approach, the Reliability Method for Passive Systems, provides a mechanistic technique employing deterministic models and conventional static event trees. The second approach, a simulation-based technique, utilizes discrete dynamic event trees to treat time- dependent phenomena during scenario evolution. For this demonstration analysis, both reliability assessment techniques are used to analyze an extended station blackout in a pool-type sodium fast reactor (SFR) coupled with a reactor cavity cooling system (RCCS). This work demonstrates the entire process of a passive system reliability analysis, including identification of important parameters and failure metrics, treatment of uncertainties and analysis of results.« less

Probabilistic evaluation of seismic isolation effect with respect to siting of a fusion reactor facility

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

Takeda, Masatoshi; Komura, Toshiyuki; Hirotani, Tsutomu

1995-12-01

Annual failure probabilities of buildings and equipment were roughly evaluated for two fusion-reactor-like buildings, with and without seismic base isolation, in order to examine the effectiveness of the base isolation system regarding siting issues. The probabilities are calculated considering nonlinearity and rupture of isolators. While the probability of building failure for both buildings on the same site was almost equal, the function failures for equipment showed that the base-isolated building had higher reliability than the non-isolated building. Even if the base-isolated building alone is located on a higher seismic hazard area, it could compete favorably with the ordinary one inmore » reliability of equipment.« less

Performance and life cycle environmental benefits of recycling spent ion exchange brines by catalytic treatment of nitrate.

PubMed

Choe, Jong Kwon; Bergquist, Allison M; Jeong, Sangjo; Guest, Jeremy S; Werth, Charles J; Strathmann, Timothy J

2015-09-01

Salt used to make brines for regeneration of ion exchange (IX) resins is the dominant economic and environmental liability of IX treatment systems for nitrate-contaminated drinking water sources. To reduce salt usage, the applicability and environmental benefits of using a catalytic reduction technology to treat nitrate in spent IX brines and enable their reuse for IX resin regeneration were evaluated. Hybrid IX/catalyst systems were designed and life cycle assessment of process consumables are used to set performance targets for the catalyst reactor. Nitrate reduction was measured in a typical spent brine (i.e., 5000 mg/L NO3(-) and 70,000 mg/L NaCl) using bimetallic Pd-In hydrogenation catalysts with variable Pd (0.2-2.5 wt%) and In (0.0125-0.25 wt%) loadings on pelletized activated carbon support (Pd-In/C). The highest activity of 50 mgNO3(-)/(min - g(Pd)) was obtained with a 0.5 wt%Pd-0.1 wt%In/C catalyst. Catalyst longevity was demonstrated by observing no decrease in catalyst activity over more than 60 days in a packed-bed reactor. Based on catalyst activity measured in batch and packed-bed reactors, environmental impacts of hybrid IX/catalyst systems were evaluated for both sequencing-batch and continuous-flow packed-bed reactor designs and environmental impacts of the sequencing-batch hybrid system were found to be 38-81% of those of conventional IX. Major environmental impact contributors other than salt consumption include Pd metal, hydrogen (electron donor), and carbon dioxide (pH buffer). Sensitivity of environmental impacts of the sequencing-batch hybrid reactor system to sulfate and bicarbonate anions indicate the hybrid system is more sustainable than conventional IX when influent water contains <80 mg/L sulfate (at any bicarbonate level up to 100 mg/L) or <20 mg/L bicarbonate (at any sulfate level up to 100 mg/L) assuming 15 brine reuse cycles. The study showed that hybrid IX/catalyst reactor systems have potential to reduce resource consumption and improve environmental impacts associated with treating nitrate-contaminated water sources. Copyright © 2015 Elsevier Ltd. All rights reserved.

Current status of SPINNORs designs

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

Su'ud, Zaki

2010-06-22

This study discuss about the SPINNOR (Small Power Reactor, Indonesia, No On-site Refuelling) and the VSPINNOR (Very Small Power Reactor, Indonesia, No On-site Refuelling) which are small lead-bismuth cooled nuclear power reactors with fast neutron spectrum that could be operated for more than 10 or 15 years without on-site refuelling. They are based on the concept of a long-life core reactor developed in Indonesia since early 1990 in collaboration with the Research Laboratory for Nuclear Reactors of the Tokyo Institute of Technology (RLNR TITech). The reactor cores are designed to have near zero (less then one effective delayed neutron fraction)more » burn-up reactivity swing during the whole course of their operation to avoid a possibility of prompt criticality accident. The basic concept is that central region of the reactor core is filled with fertile (blanket) material. During the reactor operation fissile material accumulates in this central region, which helps to compensate fissile material loss in the peripheral core region and also contributes to negative coolant loss reactivity effect. A concept of high fuel volume fraction in the core is applied to achieve smaller size of a critical reactor. In this paper we consider to add Np-237 to the fuel to enhance non proliferation characteristics of the systems. The effect of Np-237 amount variation is discussed.« less

METHOD AND APPARATUS FOR HANDLING RADIOACTIVE PRODUCTS

DOEpatents

Nicoll, D.

1959-02-24

A device is described for handling fuel elements being discharged from a nuclear reactor. The device is adapted to be disposed beneath a reactor within the storage canal for spent fuel elements. The device is comprised essentially of a cylinder pivotally mounted to a base for rotational motion between a vertical position. where the mouth of the cylinder is in the top portion of the container for receiving a fuel element discharged from a reactor into the cylinder, and a horizontal position where the mouth of the cylinder is remote from the top portion of the container and the fuel element is discharged from the cylinder into the storage canal. The device is operated by hydraulic pressure means and is provided with a means to prevent contaminated primary liquid coolant in the reactor system from entering the storage canal with the spent fuel element.

Motor fuels and chemicals from coal via the Sasol Synthol route

NASA Astrophysics Data System (ADS)

Hoogendoorn, J. C.

1981-03-01

The production of synthetic motor fuels and chemicals from coal by the Sasol procedures is discussed. This process is based on the Fischer-Tropsch reaction by passing hydrogen and carbon monoxide in a specific ratio over iron catalysts at elevated temperatures and pressures. Two parallel reactor systems are discussed. The smaller system employs fixed-bed reactors, using a precipitated iron catalyst and produces predominantly heavy hydrocarbons of an aliphatic nature with carbon chains up to 100. These straight-chain hydrocarbons yield excellent waxes and high quality diesel oil. The larger system uses a powdered iron catalyst in a circulating fluid-bed reactor, a concept developed from American catalytic cracker technology. This system has the advantage of high production capacity and scale-up potential, and produces light olefins which can be used either as petrochemical feedstock or refined and added to the motor fuel pool, and ethylene which is augmented by ethane cracking. Analysis of product selectivities and values shows that co-production of chemicals and motor fuels from coal is profitable and efficient.

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.

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...

75 FR 13610 - Office of New Reactors; Interim Staff Guidance on Implementation of a Seismic Margin Analysis for...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-22

... Staff Guidance on Implementation of a Seismic Margin Analysis for New Reactors Based on Probabilistic... Seismic Margin Analysis for New Reactors Based on Probabilistic Risk Assessment,'' (Agencywide Documents.../COL-ISG-020 ``Implementation of a Seismic Margin Analysis for New Reactors Based on Probabilistic Risk...

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.

Gravity Scaling of a Power Reactor Water Shield

NASA Technical Reports Server (NTRS)

Reid, Robert S.; Pearson, J. Boise

2008-01-01

Water based reactor shielding is being considered as an affordable option for use on initial lunar surface power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxiliary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2007). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa(sup n). These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined.

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.

Total organic carbon analyzer

NASA Technical Reports Server (NTRS)

Godec, Richard G.; Kosenka, Paul P.; Smith, Brian D.; Hutte, Richard S.; Webb, Johanna V.; Sauer, Richard L.

1991-01-01

The development and testing of a breadboard version of a highly sensitive total-organic-carbon (TOC) analyzer are reported. Attention is given to the system components including the CO2 sensor, oxidation reactor, acidification module, and the sample-inlet system. Research is reported for an experimental reagentless oxidation reactor, and good results are reported for linearity, sensitivity, and selectivity in the CO2 sensor. The TOC analyzer is developed with gravity-independent components and is designed for minimal additions of chemical reagents. The reagentless oxidation reactor is based on electrolysis and UV photolysis and is shown to be potentially useful. The stability of the breadboard instrument is shown to be good on a day-to-day basis, and the analyzer is capable of 5 sample analyses per day for a period of about 80 days. The instrument can provide accurate TOC and TIC measurements over a concentration range of 20 ppb to 50 ppm C.

Molybdenum-base cermet fuel development

NASA Astrophysics Data System (ADS)

Pilger, James P.; Gurwell, William E.; Moss, Ronald W.; White, George D.; Seifert, David A.

Development of a multimegawatt (MMW) space nuclear power system requires identification and resolution of several technical feasibility issues before selecting one or more promising system concepts. Demonstration of reactor fuel fabrication technology is required for cermet-fueled reactor concepts. The MMW reactor fuel development activity at Pacific Northwest Laboratory (PNL) is focused on producing a molybdenum-matrix uranium-nitride (UN) fueled cermte. This cermet is to have a high matrix density (greater than or equal to 95 percent) for high strength and high thermal conductance coupled with a high particle (UN) porosity (approximately 25 percent) for retention of released fission gas at high burnup. Fabrication process development involves the use of porous TiN microspheres as surrogate fuel material until porous Un microspheres become available. Process development was conducted in the areas of microsphere synthesis, particle sealing/coating, and high-energy-rate forming (HERF) and the vacuum hot press consolidation techniques. This paper summarizes the status of these activities.

Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database

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

Wang, Jy-An John

Materials behaviors caused by neutron irradiation under fission and/or fusion environments can be little understood without practical examination. Easily accessible material information system with large material database using effective computers is necessary for design of nuclear materials and analyses or simulations of the phenomena. The developed Embrittlement Data Base (EDB) at ORNL is this comprehensive collection of data. EDB database contains power reactor pressure vessel surveillance data, the material test reactor data, foreign reactor data (through bilateral agreements authorized by NRC), and the fracture toughness data. The lessons learned from building EDB program and the associated database management activity regardingmore » Material Database Design Methodology, Architecture and the Embedded QA Protocol are described in this report. The development of IAEA International Database on Reactor Pressure Vessel Materials (IDRPVM) and the comparison of EDB database and IAEA IDRPVM database are provided in the report. The recommended database QA protocol and database infrastructure are also stated in the report.« 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.-

Algae-based biofilm productivity utilizing dairy wastewater: effects of temperature and organic carbon concentration.

PubMed

Fica, Zachary T; Sims, Ronald C

2016-01-01

Biofilm-based microalgal growth was determined as functions of organic chemical loading and water temperature utilizing dairy wastewater from a full-scale dairy farm. The dairy industry is a significant source of wastewater worldwide that could provide an inexpensive and nutrient rich feedstock for the cultivation of algae biomass for use in downstream processing of animal feed and aquaculture applications. Algal biomass was cultivated using a Rotating Algal Biofilm Reactor (RABR) system. The RABR is a biofilm-based technology that has been designed and used to remediate municipal wastewater and was applied to treat dairy wastewater through nutrient uptake, and simultaneously provide biomass for the production of renewable bioproducts. Aerial algal biofilm growth rates in dairy wastewater at 7 and 27 °C temperatures were shown to be 4.55 ± 0.17 g/m 2 -day and 7.57 ± 1.12 g/m 2 -day ash free dry weight (AFDW), respectively. Analysis of Variance (ANOVA) calculations indicated that both an increase in temperature of the wastewater and an increase in the level of organic carbon, from 300 to 1200 mg L -1 , contributed significantly to an increase in the rate of biomass growth in the system. However, ANOVA results indicated that the interaction of temperature and organic carbon content was not significantly related to the biofilm-based growth rate. A microalgae-based biofilm reactor was successfully used to treat turbid dairy wastewater. Temperature and organic carbon concentration had a statistically significant effect on algae-based biofilm productivity and treatment of dairy wastewater. The relationships between temperature, TOC, and productivity developed in this study may be used in the design and assessment of wastewater remediation systems and biomass production systems utilizing algae-based biofilm reactors for treating dairy wastes.

Building on knowledge base of sodium cooled fast spectrum reactors to develop materials technology for fusion reactors

NASA Astrophysics Data System (ADS)

Raj, Baldev; Rao, K. Bhanu Sankara

2009-04-01

The alloys 316L(N) and Mod. 9Cr-1Mo steel are the major structural materials for fabrication of structural components in sodium cooled fast reactors (SFRs). Various factors influencing the mechanical behaviour of these alloys and different modes of deformation and failure in SFR systems, their analysis and the simulated tests performed on components for assessment of structural integrity and the applicability of RCC-MR code for the design and validation of components are highlighted. The procedures followed for optimal design of die and punch for the near net shape forming of petals of main vessel of 500 MWe prototype fast breeder reactor (PFBR); the safe temperature and strain rate domains established using dynamic materials model for forming of 316L(N) and 9Cr-1Mo steels components by various industrial processes are illustrated. Weldability problems associated with 316L(N) and Mo. 9Cr-1Mo are briefly discussed. The utilization of artificial neural network models for prediction of creep rupture life and delta-ferrite in austenitic stainless steel welds is described. The usage of non-destructive examination techniques in characterization of deformation, fracture and various microstructural features in SFR materials is briefly discussed. Most of the experience gained on SFR systems could be utilized in developing science and technology for fusion reactors. Summary of the current status of knowledge on various aspects of fission and fusion systems with emphasis on cross fertilization of research is presented.

Modeling the competition between PHA-producing and non-PHA-producing bacteria in feast-famine SBR and staged CSTR systems.

PubMed

Marang, Leonie; van Loosdrecht, Mark C M; Kleerebezem, Robbert

2015-12-01

Although the enrichment of specialized microbial cultures for the production of polyhydroxyalkanoates (PHA) is generally performed in sequencing batch reactors (SBRs), the required feast-famine conditions can also be established using two or more continuous stirred-tank reactors (CSTRs) in series with partial biomass recirculation. The use of CSTRs offers several advantages, but will result in distributed residence times and a less strict separation between feast and famine conditions. The aim of this study was to investigate the impact of the reactor configuration, and various process and biomass-specific parameters, on the enrichment of PHA-producing bacteria. A set of mathematical models was developed to predict the growth of Plasticicumulans acidivorans-as a model PHA producer-in competition with a non-storing heterotroph. A macroscopic model considering lumped biomass and an agent-based model considering individual cells were created to study the effect of residence time distribution and the resulting distributed bacterial states. The simulations showed that in the 2-stage CSTR system the selective pressure for PHA-producing bacteria is significantly lower than in the SBR, and strongly affected by the chosen feast-famine ratio. This is the result of substrate competition based on both the maximum specific substrate uptake rate and substrate affinity. Although the macroscopic model overestimates the selective pressure in the 2-stage CSTR system, it provides a quick and fairly good impression of the reactor performance and the impact of process and biomass-specific parameters. © 2015 Wiley Periodicals, Inc.

Fuel-Cell Power Systems Incorporating Mg-Based H2 Generators

NASA Technical Reports Server (NTRS)

Kindler, Andrew; Narayan, Sri R.

2009-01-01

Two hydrogen generators based on reactions involving magnesium and steam have been proposed as means for generating the fuel (hydrogen gas) for such fuel-cell power systems as those to be used in the drive systems of advanced motor vehicles. The hydrogen generators would make it unnecessary to rely on any of the hydrogen storage systems developed thus far that are, variously, too expensive, too heavy, too bulky, and/or too unsafe to be practical. The two proposed hydrogen generators are denoted basic and advanced, respectively. In the basic hydrogen generator (see figure), steam at a temperature greater than or equals 330 C would be fed into a reactor charged with magnesium, wherein hydrogen would be released in the exothermic reaction Mg + H2O yields MgO + H2. The steam would be made in a flash boiler. To initiate the reaction, the boiler could be heated electrically by energy borrowed from a storage battery that would be recharged during normal operation of the associated fuel-cell subsystem. Once the reaction was underway, heat from the reaction would be fed to the boiler. If the boiler were made an integral part of the hydrogen-generator reactor vessel, then the problem of transfer of heat from the reactor to the boiler would be greatly simplified. A pump would be used to feed water from a storage tank to the boiler.

Gas Turbine Energy Conversion Systems for Nuclear Power Plants Applicable to LiFTR Liquid Fluoride Thorium Reactor Technology

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.

2014-01-01

This panel plans to cover thermal energy and electric power production issues facing our nation and the world over the next decades, with relevant technologies ranging from near term to mid-and far term.Although the main focus will be on ground based plants to provide baseload electric power, energy conversion systems (ECS) for space are also included, with solar- or nuclear energy sources for output power levels ranging tens of Watts to kilo-Watts for unmanned spacecraft, and eventual mega-Watts for lunar outposts and planetary surface colonies. Implications of these technologies on future terrestrial energy systems, combined with advanced fracking, are touched upon.Thorium based reactors, and nuclear fusion along with suitable gas turbine energy conversion systems (ECS) will also be considered by the panelists. The characteristics of the above mentioned ECS will be described, both in terms of their overall energy utilization effectiveness and also with regard to climactic effects due to exhaust emissions.

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.

LFR "Lead-Cooled Fast Reactor"

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

Cinotti, L; Fazio, C; Knebel, J

2006-05-11

The main purpose of this paper is to present the current status of development of the Lead-cooled Fast Reactor (LFR) in Generation IV (GEN IV), including the European contribution, to identify needed R&D and to present the corresponding GEN IV International Forum (GIF) R&D plan [1] to support the future development and deployment of lead-cooled fast reactors. The approach of the GIF plan is to consider the research priorities of each member country in proposing an integrated, coordinated R&D program to achieve common objectives, while avoiding duplication of effort. The integrated plan recognizes two principal technology tracks: (1) a small,more » transportable system of 10-100 MWe size that features a very long refuelling interval, and (2) a larger-sized system rated at about 600 MWe, intended for central station power generation. This paper provides some details of the important European contributions to the development of the LFR. Sixteen European organizations have, in fact, taken the initiative to present to the European Commission the proposal for a Specific Targeted Research and Training Project (STREP) devoted to the development of a European Lead-cooled System, known as the ELSY project; two additional organizations from the US and Korea have joined the project. Consequently, ELSY will constitute the reference system for the large lead-cooled reactor of GEN IV. The ELSY project aims to demonstrate the feasibility of designing a competitive and safe fast power reactor based on simple technical engineered features that achieves all of the GEN IV goals and gives assurance of investment protection. As far as new technology development is concerned, only a limited amount of R&D will be conducted in the initial phase of the ELSY project since the first priority is to define the design guidelines before launching a larger and expensive specific R&D program. In addition, the ELSY project is expected to benefit greatly from ongoing lead and lead-alloy technology development already being carried out in different institutes participating in this STREP. This is particularly true in Europe where a large R&D program associated with the development of Accelerator Driven Systems (ADS) is being actively pursued. The general objective of the ELSY project is to design an innovative lead-cooled fast reactor complemented by an analytical effort to assess the existing knowledge base in the field of lead-alloy coolants (i.e., lead-bismuth eutectic (LBE) and also lead/lithium) in order to extrapolate this knowledge base to pure lead. This analysis effort will be complemented with some limited R&D activities to acquire missing or confirmatory information about fundamental topics for ELSY that are not sufficiently covered in the ongoing European ADS program or elsewhere.« less

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

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.

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

Analysis of key safety metrics of thorium utilization in LWRs

DOE PAGES

Ade, Brian J.; Bowman, Stephen M.; Worrall, Andrew; ...

2016-04-08

Here, thorium has great potential to stretch nuclear fuel reserves because of its natural abundance and because it is possible to breed the 232Th isotope into a fissile fuel ( 233U). Various scenarios exist for utilization of thorium in the nuclear fuel cycle, including use in different nuclear reactor types (e.g., light water, high-temperature gas-cooled, fast spectrum sodium, and molten salt reactors), along with use in advanced accelerator-driven systems and even in fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based onmore » concepts that mix thorium with uranium (UO 2 + ThO 2) or that add fertile thorium (ThO 2) fuel pins to typical LWR fuel assemblies. Utilization of mixed fuel assemblies (PuO 2 + ThO 2) is also possible. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts to the nuclear fuel. Thorium and its irradiation products have different nuclear characteristics from those of uranium and its irradiation products. ThO 2, alone or mixed with UO 2 fuel, leads to different chemical and physical properties of the fuel. These key reactor safety–related issues have been studied at Oak Ridge National Laboratory and documented in “Safety and Regulatory Issues of the Thorium Fuel Cycle” (NUREG/CR-7176, U.S. Nuclear Regulatory Commission, 2014). Various reactor analyses were performed using the SCALE code system for comparison of key performance parameters of both ThO 2 + UO 2 and ThO 2 + PuO 2 against those of UO 2 and typical UO 2 + PuO 2 mixed oxide fuels, including reactivity coefficients and power sharing between surrounding UO 2 assemblies and the assembly of interest. The decay heat and radiological source terms for spent fuel after its discharge from the reactor are also presented. Based on this evaluation, potential impacts on safety requirements and identification of knowledge gaps that require additional analysis or research to develop a technical basis for the licensing of thorium fuel are identified.« less

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.

Development and Testing of Space Fission Technology at NASA-MSFC

NASA Technical Reports Server (NTRS)

Polzin, Kurt; Pearson, J. Boise; Houts, Michael

2008-01-01

The Early Flight Fission Test Facility (EFF-TF) at NASA-Marshall Space Flight Center (MSFC) provides a capability to perform hardware-directed activities to support multiple inspace nuclear reactor concepts by using a non-nuclear test methodology. This includes fabrication and testing at both the module/component level and near prototypic reactor configurations allowing for realistic thermal-hydraulic evaluations of systems. The EFF-TF is currently performing non-nuclear testing of hardware to support a technology development effort related to an affordable fission surface power (AFSP) system that could be deployed on the Lunar surface. The AFSP system is presently based on a pumped liquid metal-cooled reactor design, which builds on US and Russian space reactor technology as well as extensive US and international terrestrial liquid metal reactor experience. An important aspect of the current hardware development effort is the information and insight that can be gained from experiments performed in a relevant environment using realistic materials. This testing can often deliver valuable data and insights with a confidence that is not otherwise available or attainable. While the project is currently focused on potential fission surface power for the lunar surface, many of the present advances, testing capabilities, and lessons learned can be applied to the future development of a low-cost in-space fission power system. The potential development of such systems would be useful in fulfilling the power requirements for certain electric propulsion systems (magnetoplasmadynamic thruster, high-power Hall and ion thrusters). In addition, inspace fission power could be applied towards meeting spacecraft and propulsion needs on missions further from the Sun, where the usefulness of solar power is diminished. The affordable nature of the fission surface power system that NASA may decide to develop in the future might make derived systems generally attractive for powering spacecraft and propulsion systems in space. This presentation will discuss work on space nuclear systems that has been performed at MSFC's EFF-TF over the past 10 years. Emphasis will be place on both ongoing work related to FSP and historical work related to in-space systems potentially useful for powering electric propulsion systems.

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

Manned space flight nuclear system safety. Volume 5: Nuclear System safety guidelines. Part 1: Space base nuclear safety

NASA Technical Reports Server (NTRS)

1972-01-01

The design and operations guidelines and requirements developed in the study of space base nuclear system safety are presented. Guidelines and requirements are presented for the space base subsystems, nuclear hardware (reactor, isotope sources, dynamic generator equipment), experiments, interfacing vehicles, ground support systems, range safety and facilities. Cross indices and references are provided which relate guidelines to each other, and to substantiating data in other volumes. The guidelines are intended for the implementation of nuclear safety related design and operational considerations in future space programs.

Silicon carbide novel optical sensor for combustion systems and nuclear reactors

NASA Astrophysics Data System (ADS)

Lim, Geunsik; Kar, Aravinda

2014-09-01

Crystalline silicon carbide is a wide bandgap semiconductor material with excellent optical properties, chemical inertness, radiation hardness and high mechanical strength at high temperatures. It is an excellent material platform for sensor applications in harsh environments such as combustion systems and nuclear reactors. A laser doping technique is used to fabricate SiC sensors for different combustion gases such as CO2, CO, NO and NO2. The sensor operates based on the principle of semiconductor optics, producing optical signal in contrast to conventional electrical sensors that produces electrical signal. The sensor response is measured with a low power He-Ne or diode laser.

The Development of Neutron Radiography and Tomography on a SLOWPOKE-2 Reactor

NASA Astrophysics Data System (ADS)

Bennett, L. G. I.; Lewis, W. J.; Hungler, P. C.

Development of neutron radiography at the Royal Military College of Canada (RMC) started by trying to interest the Royal Canadian Air Force (RCAF) in this new non-destructive testing (NDT) technique. A Californium-252 based device was ordered and then installed at RMC for development of applicable techniques for aircraft by the first author. A second and transportable device was then designed, modified and used in trials at RCAF Bases and other locations for one year. This activity was the only foreign loan of the U.S. Californium Loan Program. Around this time, SLOWPOKE-2 reactors were being installed at four Canadian universities, while a new science and engineering building was being built at RMC. A reactor pool was incorporated and efforts to procure a reactor succeeded a decade later with a SLOWPOKE-2 reactor being installed at RMC. The only modification by the vendor for RMC was a thermal column replacing an irradiation site inside the reactor container for a later installation of a neutron beam tube (NBT). Development of a working NBT took several years, starting with the second author. A demonstration of the actual worth of neutron radiography took place with a CF-18 Hornet aircraft being neutron and X-radiographed at McClellan Air Force Base, Sacramento, CA. This inspection was followed by one of the rudders that had indications of water ingress being radiographed successfully at RMC just after the NBT became functional. The next step was to develop a neutron radioscopy system (NRS), initially employing film and then digital imaging, and is in use today for all flight control surfaces (FCS). With the third author, a technique capable of removing water from affected FCS was developed at RMC. Heating equipment and a vacuum system were utilized to carefully remove the water. This technique was proven using a sequence of near real time neutron images obtained during the drying process. The results of the drying process were correlated with a relative humidity gauge and an NDT technique that could be performed at Canadian Forces (CF) Bases was developed. In order to determine the structural integrity of the component having undergone this water removal, further research was required into the effect of water inside composite honeycomb structures. This need has led to the present development of neutron tomography on the reactor at RMC, which is capable of determining the exact location of water ingress inside composite components. This technique has been successfully applied to coupons as well as to complete rudders.

SP-100 ground engineering system test site description and progress update

NASA Astrophysics Data System (ADS)

Baxter, William F.; Burchell, Gail P.; Fitzgibbon, Davis G.; Swita, Walter R.

1991-01-01

The SP-100 Ground Engineering System Test Site will provide the facilities for the testing of an SP-100 reactor, which is technically prototypic of the generic design for producing 100 kilowatts of electricity. This effort is part of the program to develop a compact, space-based power system capable of producing several hundred kilowatts of electrical power. The test site is located on the U.S. Department of Energy's Hanford Site near Richland, Washington. The site is minimizing capital equipment costs by utilizing existing facilities and equipment to the maximum extent possible. The test cell is located in a decommissioned reactor containment building, and the secondary sodium cooling loop will use equipment from the Fast Flux Test Facility plant which has never been put into service. Modifications to the facility and special equipment are needed to accommodate the testing of the SP-100 reactor. Definitive design of the Ground Engineering System Test Site facility modifications and systems is in progress. The design of the test facility and the testing equipment will comply with the regulations and specifications of the U.S. Department of Energy and the State of Washington.

Nuclear Data Needs for Generation IV Nuclear Energy Systems

NASA Astrophysics Data System (ADS)

Rullhusen, Peter

2006-04-01

Nuclear data needs for generation IV systems. Future of nuclear energy and the role of nuclear data / P. Finck. Nuclear data needs for generation IV nuclear energy systems-summary of U.S. workshop / T. A. Taiwo, H. S. Khalil. Nuclear data needs for the assessment of gen. IV systems / G. Rimpault. Nuclear data needs for generation IV-lessons from benchmarks / S. C. van der Marck, A. Hogenbirk, M. C. Duijvestijn. Core design issues of the supercritical water fast reactor / M. Mori ... [et al.]. GFR core neutronics studies at CEA / J. C. Bosq ... [et al]. Comparative study on different phonon frequency spectra of graphite in GCR / Young-Sik Cho ... [et al.]. Innovative fuel types for minor actinides transmutation / D. Haas, A. Fernandez, J. Somers. The importance of nuclear data in modeling and designing generation IV fast reactors / K. D. Weaver. The GIF and Mexico-"everything is possible" / C. Arrenondo Sánchez -- Benmarks, sensitivity calculations, uncertainties. Sensitivity of advanced reactor and fuel cycle performance parameters to nuclear data uncertainties / G. Aliberti ... [et al.]. Sensitivity and uncertainty study for thermal molten salt reactors / A. Biduad ... [et al.]. Integral reactor physics benchmarks- The International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPHEP) / J. B. Briggs, D. W. Nigg, E. Sartori. Computer model of an error propagation through micro-campaign of fast neutron gas cooled nuclear reactor / E. Ivanov. Combining differential and integral experiments on [symbol] for reducing uncertainties in nuclear data applications / T. Kawano ... [et al.]. Sensitivity of activation cross sections of the Hafnium, Tanatalum and Tungsten stable isotopes to nuclear reaction mechanisms / V. Avrigeanu ... [et al.]. Generating covariance data with nuclear models / A. J. Koning. Sensitivity of Candu-SCWR reactors physics calculations to nuclear data files / K. S. Kozier, G. R. Dyck. The lead cooled fast reactor benchmark BREST-300: analysis with sensitivity method / V. Smirnov ... [et al.]. Sensitivity analysis of neutron cross-sections considered for design and safety studies of LFR and SFR generation IV systems / K. Tucek, J. Carlsson, H. Wider -- Experiments. INL capabilities for nuclear data measurements using the Argonne intense pulsed neutron source facility / J. D. Cole ... [et al.]. Cross-section measurements in the fast neutron energy range / A. Plompen. Recent measurements of neutron capture cross sections for minor actinides by a JNC and Kyoto University Group / H. Harada ... [et al.]. Determination of minor actinides fission cross sections by means of transfer reactions / M. Aiche ... [et al.] -- Evaluated data libraries. Nuclear data services from the NEA / H. Henriksson, Y. Rugama. Nuclear databases for energy applications: an IAEA perspective / R. Capote Noy, A. L. Nichols, A. Trkov. Nuclear data evaluation for generation IV / G. Noguère ... [et al.]. Improved evaluations of neutron-induced reactions on americium isotopes / P. Talou ... [et al.]. Using improved ENDF-based nuclear data for candu reactor calculations / J. Prodea. A comparative study on the graphite-moderated reactors using different evaluated nuclear data / Do Heon Kim ... [et al.].

Novel fabrication of silicon carbide based ceramics for nuclear applications

NASA Astrophysics Data System (ADS)

Singh, Abhishek Kumar

Advances in nuclear reactor technology and the use of gas-cooled fast reactors require the development of new materials that can operate at the higher temperatures expected in these systems. These materials include refractory alloys based on Nb, Zr, Ta, Mo, W, and Re; ceramics and composites such as SiC--SiCf; carbon--carbon composites; and advanced coatings. Besides the ability to handle higher expected temperatures, effective heat transfer between reactor components is necessary for improved efficiency. Improving thermal conductivity of the fuel can lower the center-line temperature and, thereby, enhance power production capabilities and reduce the risk of premature fuel pellet failure. Crystalline silicon carbide has superior characteristics as a structural material from the viewpoint of its thermal and mechanical properties, thermal shock resistance, chemical stability, and low radioactivation. Therefore, there have been many efforts to develop SiC based composites in various forms for use in advanced energy systems. In recent years, with the development of high yield preceramic precursors, the polymer infiltration and pyrolysis (PIP) method has aroused interest for the fabrication of ceramic based materials, for various applications ranging from disc brakes to nuclear reactor fuels. The pyrolysis of preceramic polymers allow new types of ceramic materials to be processed at relatively low temperatures. The raw materials are element-organic polymers whose composition and architecture can be tailored and varied. The primary focus of this study is to use a pyrolysis based process to fabricate a host of novel silicon carbide-metal carbide or oxide composites, and to synthesize new materials based on mixed-metal silicocarbides that cannot be processed using conventional techniques. Allylhydridopolycarbosilane (AHPCS), which is an organometal polymer, was used as the precursor for silicon carbide. Inert gas pyrolysis of AHPCS produces near-stoichiometric amorphous silicon carbide (a-SiC) at 900--1150 °C. Results indicated that this processing technique can be effectively used to fabricate various silicon carbide composites with UC or UO2 as the nuclear component.

Co-digestion of concentrated black water and kitchen refuse in an accumulation system within the DESAR (decentralized sanitation and reuse) concept.

PubMed

Kujawa-Roeleveld, K; Elmitwalli, T; Gaillard, A; van Leeuwen, M; Zeeman, G

2003-01-01

Co-digestion of concentrated black water and kitchen refuse within the DESAR concept was the objective of this pilot research. The digestion took place in two, non-mixed accumulation reactors (AC1 and AC2) inoculated with digested primary sludge from a WWTP at a temperature of 20 degrees C for a period of around 150 days. Reactor AC1 was fed with a mixture of faeces, urine and kitchen refuse in the equivalent amount that one individual generates per day. The AC2 was fed with a mixture of faeces and kitchen refuse in the equivalent amount that two individuals produce per day. Some contribution of urine to AC2 was not to be avoided. Detailed characterisation of waste(water) was performed. The performance of the stratified reactor was followed by monitoring the reactor content for several reactors' heights as well as being based on the biogas production. In general the system exposed good process stability. The methanisation of 34 and 61% was obtained for AC1 and AC2 respectively. The biogas yield was 26.5 and 50.8 L/p/d for the respective reactors. Proper choice of inoculum as well as good buffering capacity did not lead to accumulation of VFA and an inhibitive effect due to relatively high ammonium concentration. The chosen process is a promising technology showing good process stability especially for high strength influent.

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.

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

Safety approach to the selection of design criteria for the CRBRP reactor refueling system

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

Meisl, C J; Berg, G E; Sharkey, N F

1979-01-01

The selection of safety design criteria for Liquid Metal Fast Breeder Reactor (LMFBR) refueling systems required the extrapolation of regulations and guidelines intended for Light Water Reactor refueling systems and was encumbered by the lack of benefit from a commercially licensed predecessor other than Fermi. The overall approach and underlying logic are described for developing safety design criteria for the reactor refueling system (RRS) of the Clinch River Breeder Reactor Plant (CRBRP). The complete selection process used to establish the criteria is presented, from the definition of safety functions to the finalization of safety design criteria in the appropriate documents.more » The process steps are illustrated by examples.« 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.

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

PubMed

Elmitwalli, Tarek; Otterpohl, Ralf

2011-01-01

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

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

Development of a Rational Design Procedure for Overland Flow Systems,

DTIC Science & Technology

1982-02-01

to pre- report is based on reactor kinetics, a concept dict the hydraulic detention time. familiar to most environmental engineers. In the 2...Determine the removal kinetics for BOD, TSS, case of overland flow, the reactor is the soil surface NH3 -N and total P. where various physical, biological and...water along the top of each section, and a bed of down. The amount of time needed to reach hydraulic crushed stone placed beneath the pipe helped to

Development of a three-dimensional core dynamics analysis program for commercial boiling water reactors

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

Bessho, Yasunori; Yokomizo, Osamu; Yoshimoto, Yuichiro

1997-03-01

Development and qualification results are described for a three-dimensional, time-domain core dynamics analysis program for commercial boiling water reactors (BWRs). The program allows analysis of the reactor core with a detailed mesh division, which eliminates calculational ambiguity in the nuclear-thermal-hydraulic stability analysis caused by reactor core regional division. During development, emphasis was placed on high calculational speed and large memory size as attained by the latest supercomputer technology. The program consists of six major modules, namely a core neutronics module, a fuel heat conduction/transfer module, a fuel channel thermal-hydraulic module, an upper plenum/separator module, a feedwater/recirculation flow module, and amore » control system module. Its core neutronics module is based on the modified one-group neutron kinetics equation with the prompt jump approximation and with six delayed neutron precursor groups. The module is used to analyze one fuel bundle of the reactor core with one mesh (region). The fuel heat conduction/transfer module solves the one-dimensional heat conduction equation in the radial direction with ten nodes in the fuel pin. The fuel channel thermal-hydraulic module is based on separated three-equation, two-phase flow equations with the drift flux correlation, and it analyzes one fuel bundle of the reactor core with one channel to evaluate flow redistribution between channels precisely. Thermal margin is evaluated by using the GEXL correlation, for example, in the module.« less

Development of the Packed Bed Reactor ISS Flight Experiment

NASA Technical Reports Server (NTRS)

Patton, Martin O.; Bruzas, Anthony E.; Rame, Enrique; Motil, Brian J.

2012-01-01

Packed bed reactors are compact, require minimum power and maintenance to operate, and are highly reliable. These features make this technology a leading candidate as a potential unit operation in support of long duration human space exploration. On earth, this type of reactor accounts for approximately 80% of all the reactors used in the chemical process industry today. Development of this technology for space exploration is truly crosscutting with many other potential applications (e.g., in-situ chemical processing of planetary materials and transport of nutrients through soil). NASA is developing an ISS experiment to address this technology with particular focus on water reclamation and air revitalization. Earlier research and development efforts funded by NASA have resulted in two hydrodynamic models which require validation with appropriate instrumentation in an extended microgravity environment. The first model developed by Motil et al., (2003) is based on a modified Ergun equation. This model was demonstrated at moderate gas and liquid flow rates, but extension to the lower flow rates expected in many advanced life support systems must be validated. The other model, developed by Guo et al., (2004) is based on Darcy s (1856) law for two-phase flow. This model has been validated for a narrow range of flow parameters indirectly (without full instrumentation) and included test points where the flow was not fully developed. The flight experiment presented will be designed with removable test sections to test the hydrodynamic models. The experiment will provide flexibility to test additional beds with different types of packing in the future. One initial test bed is based on the VRA (Volatile Removal Assembly), a packed bed reactor currently on ISS whose behavior in micro-gravity is not fully understood. Improving the performance of this system through an accurate model will increase our ability to purify water in the space environment.

Neutronics Analysis of SMART Small Modular Reactor using SRAC 2006 Code

NASA Astrophysics Data System (ADS)

Ramdhani, Rahmi N.; Prastyo, Puguh A.; Waris, Abdul; Widayani; Kurniadi, Rizal

2017-07-01

Small modular reactors (SMRs) are part of a new generation of nuclear reactor being developed worldwide. One of the advantages of SMR is the flexibility to adopt the advanced design concepts and technology. SMART (System integrated Modular Advanced ReacTor) is a small sized integral type PWR with a thermal power of 330 MW that has been developed by KAERI (Korea Atomic Energy Research Institute). SMART core consists of 57 fuel assemblies which are based on the well proven 17×17 array that has been used in Korean commercial PWRs. SMART is soluble boron free, and the high initial reactivity is mainly controlled by burnable absorbers. The goal of this study is to perform neutronics evaluation of SMART core with UO2 as main fuel. Neutronics calculation was performed by using PIJ and CITATION modules of SRAC 2006 code with JENDL 3.3 as nuclear data library.

Integrated Risk-Informed Decision-Making for an ALMR PRISM

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

Muhlheim, Michael David; Belles, Randy; Denning, Richard S.

Decision-making is the process of identifying decision alternatives, assessing those alternatives based on predefined metrics, selecting an alternative (i.e., making a decision), and then implementing that alternative. The generation of decisions requires a structured, coherent process, or a decision-making process. The overall objective for this work is that the generalized framework is adopted into an autonomous decision-making framework and tailored to specific requirements for various applications. In this context, automation is the use of computing resources to make decisions and implement a structured decision-making process with limited or no human intervention. The overriding goal of automation is to replace ormore » supplement human decision makers with reconfigurable decision-making modules that can perform a given set of tasks rationally, consistently, and reliably. Risk-informed decision-making requires a probabilistic assessment of the likelihood of success given the status of the plant/systems and component health, and a deterministic assessment between plant operating parameters and reactor protection parameters to prevent unnecessary trips and challenges to plant safety systems. The probabilistic portion of the decision-making engine of the supervisory control system is based on the control actions associated with an ALMR PRISM. Newly incorporated into the probabilistic models are the prognostic/diagnostic models developed by Pacific Northwest National Laboratory. These allow decisions to incorporate the health of components into the decision–making process. Once the control options are identified and ranked based on the likelihood of success, the supervisory control system transmits the options to the deterministic portion of the platform. The deterministic portion of the decision-making engine uses thermal-hydraulic modeling and components for an advanced liquid-metal reactor Power Reactor Inherently Safe Module. The deterministic multi-attribute decision-making framework uses various sensor data (e.g., reactor outlet temperature, steam generator drum level) and calculates its position within the challenge state, its trajectory, and its margin within the controllable domain using utility functions to evaluate current and projected plant state space for different control decisions. The metrics that are evaluated are based on reactor trip set points. The integration of the deterministic calculations using multi-physics analyses and probabilistic safety calculations allows for the examination and quantification of margin recovery strategies. This also provides validation of the control options identified from the probabilistic assessment. Thus, the thermalhydraulics analyses are used to validate the control options identified from the probabilistic assessment. Future work includes evaluating other possible metrics and computational efficiencies, and developing a user interface to mimic display panels at a modern nuclear power plant.« less

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.

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.

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

Fuel Processor Development for a Soldier-Portable Fuel Cell System

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

Palo, Daniel R.; Holladay, Jamie D.; Rozmiarek, Robert T.

2002-01-01

Battelle is currently developing a soldier-portable power system for the U.S. Army that will continuously provide 15 W (25 W peak) of base load electric power for weeks or months using a micro technology-based fuel processor. The fuel processing train consists of a combustor, two vaporizers, and a steam-reforming reactor. This paper describes the concept and experimental progress to date.

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.

Measurement and Analysis of Structural Integrity of Reactor Core Support Structure in Pressurized Water Reactor (PWR) Plant

NASA Astrophysics Data System (ADS)

Ansari, Saleem A.; Haroon, Muhammad; Rashid, Atif; Kazmi, Zafar

2017-02-01

Extensive calculation and measurements of flow-induced vibrations (FIV) of reactor internals were made in a PWR plant to assess the structural integrity of reactor core support structure against coolant flow. The work was done to meet the requirements of the Fukushima Response Action Plan (FRAP) for enhancement of reactor safety, and the regulatory guide RG-1.20. For the core surveillance measurements the Reactor Internals Vibration Monitoring System (IVMS) has been developed based on detailed neutron noise analysis of the flux signals from the four ex-core neutron detectors. The natural frequencies, displacement and mode shapes of the reactor core barrel (CB) motion were determined with the help of IVMS. The random pressure fluctuations in reactor coolant flow due to turbulence force have been identified as the predominant cause of beam-mode deflection of CB. The dynamic FIV calculations were also made to supplement the core surveillance measurements. The calculational package employed the computational fluid dynamics, mode shape analysis, calculation of power spectral densities of flow & pressure fields and the structural response to random flow excitation forces. The dynamic loads and stiffness of the Hold-Down Spring that keeps the core structure in position against upward coolant thrust were also determined by noise measurements. Also, the boron concentration in primary coolant at any time of the core cycle has been determined with the IVMS.

On-Site Determination and Monitoring of Real-Time Fluence Delivery for an Operating UV Reactor Based on a True Fluence Rate Detector.

PubMed

Li, Mengkai; Li, Wentao; Qiang, Zhimin; Blatchley, Ernest R

2017-07-18

At present, on-site fluence (distribution) determination and monitoring of an operating UV system represent a considerable challenge. The recently developed microfluorescent silica detector (MFSD) is able to measure the approximate true fluence rate (FR) at a fixed position in a UV reactor that can be compared with a FR model directly. Hence it has provided a connection between model calculation and real-time fluence determination. In this study, an on-site determination and monitoring method of fluence delivery for an operating UV reactor was developed. True FR detectors, a UV transmittance (UVT) meter, and a flow rate meter were used for fundamental measurements. The fluence distribution, as well as reduction equivalent fluence (REF), 10th percentile dose in the UV fluence distribution (F 10 ), minimum fluence (F min ), and mean fluence (F mean ) of a test reactor, was calculated in advance by the combined use of computational fluid dynamics and FR field modeling. A field test was carried out on the test reactor for disinfection of a secondary water supply. The estimated real-time REF, F 10 , F min , and F mean decreased 73.6%, 71.4%, 69.6%, and 72.9%, respectively, during a 6-month period, which was attributable to lamp output attenuation and sleeve fouling. The results were analyzed with synchronous data from a previously developed triparameter UV monitoring system and water temperature sensor. This study allowed demonstration of an accurate method for on-site, real-time fluence determination which could be used to enhance the security and public confidence of UV-based water treatment processes.

Oxygen transport membrane based advanced power cycle with low pressure synthesis gas slip stream

DOEpatents

Kromer, Brian R.; Litwin, Michael M.; Kelly, Sean M.

2016-09-27

A method and system for generating electrical power in which a high pressure synthesis gas stream generated in a gasifier is partially oxidized in an oxygen transport membrane based reactor, expanded and thereafter, is combusted in an oxygen transport membrane based boiler. A low pressure synthesis gas slip stream is split off downstream of the expanders and used as the source of fuel in the oxygen transport membrane based partial oxidation reactors to allow the oxygen transport membrane to operate at low fuel pressures with high fuel utilization. The combustion within the boiler generates heat to raise steam to in turn generate electricity by a generator coupled to a steam turbine. The resultant flue gas can be purified to produce a carbon dioxide product.

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

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.

Evaluation of power density on the bioethanol production using mesoscale oscillatory baffled reactor and stirred tank reactor

NASA Astrophysics Data System (ADS)

Yussof, H. W.; Bahri, S. S.; Mazlan, N. A.

2018-03-01

A recent development in oscillatory baffled reactor technology is down-scaling the reactor, so that it can be used for production of small-scale bioproduct. In the present study, a mesoscale oscillatory baffled reactor (MOBR) with central baffle system was developed. The reactor performance of the MOBR was compared with conventional stirred tank reactor (STR) to evaluate the performance of bioethanol fermentation using Saccharomyces cerevisiae. Evaluation was made at similar power density of 24.21, 57.38, 112.35 and 193.67 Wm-3 by varying frequency (f), amplitude (xo) and agitation speed (rpm). It was found that the MOBR improved the mixing intensity resulted in lower glucose concentration (0.988 gL-1) and higher bioethanol concentration (38.98 gL-1) after 12 hours fermentation at power density of 193.67 Wm-3. Based on the results, the bioethanol yield obtained using MOBR was 39% higher than the maximum achieved in STR. Bioethanol production using MOBR proved to be feasible as it is not only able to compete with conventional STR but also offers advantages of straight-forward scale-up, whereas it is complicated and difficult in STR. Overall, MOBR offers great prospective over the conventional STR.

Creep Strength of Nb-1Zr for SP-100 Applications

NASA Astrophysics Data System (ADS)

Horak, James A.; Egner, Larry K.

1994-07-01

Power systems that are used to provide electrical power in space are designed to optimize conversion of thermal energy to electrical energy and to minimize the mass and volume that must be launched. Only refractory metals and their alloys have sufficient long-term strength for several years of uninterrupted operation at the required temperatures of 1200 K and above. The high power densities and temperatures at which these reactors must operate require the use of liquid-metal coolants. The alloy Nb-1 wt % Zr (Nb-lZr), which exhibits excellent corrosion resistance to alkali liquid-metals at high temperatures, is being considered for the fuel cladding, reactor structural, and heat-transport systems for the SP-100 reactor system. Useful lifetime of this system is limited by creep deformation in the reactor core. Nb-lZr sheet procured to American Society for Testing and Materials (ASTM) specifications for reactor grade and commercial grade has been processed by several different cold work and annealing treatments to attempt to produce the grain structure (size, shape, and distribution of sizes) that provides the maximum creep strength of this alloy at temperatures from 1250 to 1450 K. The effects of grain size, differences in oxygen concentrations, tungsten concentrations, and electron beam and gas tungsten arc weldments on creep strength were studied. Grain size has a large effect on creep strength at 1450 K but only material with a very large grain size (150 μm) exhibits significantly higher creep strength at 1350 K. Differences in oxygen or tungsten concentrations did not affect creep strength, and the creep strengths of weldments were equal to, or greater than, those for base metal.

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

Brown, Nicholas R.; Powers, Jeffrey J.; Mueller, Don

In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF 2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy researchmore » and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in 7Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.« less

Designing a SCADA system simulator for fast breeder reactor

NASA Astrophysics Data System (ADS)

Nugraha, E.; Abdullah, A. G.; Hakim, D. L.

2016-04-01

SCADA (Supervisory Control and Data Acquisition) system simulator is a Human Machine Interface-based software that is able to visualize the process of a plant. This study describes the results of the process of designing a SCADA system simulator that aims to facilitate the operator in monitoring, controlling, handling the alarm, accessing historical data and historical trend in Nuclear Power Plant (NPP) type Fast Breeder Reactor (FBR). This research used simulation to simulate NPP type FBR Kalpakkam in India. This simulator was developed using Wonderware Intouch software 10 and is equipped with main menu, plant overview, area graphics, control display, set point display, alarm system, real-time trending, historical trending and security system. This simulator can properly simulate the principle of energy flow and energy conversion process on NPP type FBR. This SCADA system simulator can be used as training media for NPP type FBR prospective operators.

On-Line Model-Based System For Nuclear Plant Monitoring

NASA Astrophysics Data System (ADS)

Tsoukalas, Lefteri H.; Lee, G. W.; Ragheb, Magdi; McDonough, T.; Niziolek, F.; Parker, M.

1989-03-01

A prototypical on-line model-based system, LASALLE1, developed at the University of Illinois in collaboration with the Illinois Department of Nuclear Safety (IDNS) is described. Its main purpose is to interpret about 300 signals, updated every two minutes at IDNS from the LaSalle Nuclear Power Plant, and to diagnose possible abnormal conditions. It is written in VAX/VMS OPS5 and operates on both the on-line and testing modes. In its knowledge base, operator and plant actions pertaining to the Emergency Operating Procedure(EOP) A-01, are encoded. This is a procedure driven by a reactor's coolant level and pressure signals; with the purpose of shutting down the reactor, maintaining adequate core cooling and reducing the reactor pressure and temperature to cold shutdown conditions ( about 90 to 200 °F). The monitoring of the procedure is performed from the perspective of Emergency Preparedness. Two major issues are addressed in this system. First, the management of the short-term or working memory of the system. LASALLE1 must reach its inferences, display its conclusion and update a message file every two minutes before a new set of data arrives from the plant. This was achieved by superimposing additional layers of control over the inferencing strategies inherent in OPS5, and developing special rules for the management of the used or outdated information. The second issue is the representation of information and its uncertainty. The concepts of information granularity and performance-level, which are based on a coupling of Probability Theory and the theory of Fuzzy Sets, are used for this purpose. The estimation of the performance-level incorporates a mathematical methodology which accounts for two types of uncertainty encountered in monitoring physical systems: Random uncertainty, in the form of of probability density functions generated by observations, measurements and sensors data and fuzzy uncertainty represented by membership functions based on symbolic , stochastic or numerical models estimating the "plausible", "possible" or "expected" values of the system parameters. Examples from both the on-line mode and the testing mode of the system will be discussed to illustrate the present methodology.

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

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.

Safety analysis

NASA Technical Reports Server (NTRS)

Knight, John C.

1995-01-01

We are engaged in a research program in safety-critical computing that is based on two case studies. We use these case studies to provide application-specific details of the various research issues, and as targets for evaluation of research ideas. The first case study is the Magnetic Stereotaxis System (MSS), an investigational device for performing human neurosurgery being developed in a joint effort between the Department of Physics at the University of Virginia and the Department of Neurosurgery at the University of Iowa. The system operates by manipulating a small permanent magnet (known as a 'seed') within the brain using an externally applied magnetic field. By varying the magnitude and gradient of the external magnetic field, the seed can be moved along a non-linear path and positioned at a site requiring therapy, e.g., a tumor. The magnetic field required for movement through brain tissue is extremely high, and is generated by a set of six superconducting magnets located in a housing surrounding the patient's head. The system uses two X-ray cameras positioned at right angles to detect in real time the locations of the seed and of X-ray opaque markers affixed to the patient's skull. the X-ray images are used to locate the objects of interest in a canonical frame of reference. the second case study is the University of Virginia Research Nuclear Reactor (UVAR). It is a 2 MW thermal, concrete-walled pool reactor. The system operates using 20 to 25 plate-type fuel assemblies placed on a rectangular grid plate. There are three scramable safety rods, and one non-scramable regulating rod that can be put in automatic mode. It was originally constructed in 1959 as a 1 MW system, and it was upgraded to 2 MW in 1973. Though only a research reactor rather than a power reactor, the issues raised are significant and can be related to the problems faced by full-scale reactor systems.

Fixture For Sampling Volatile Materials In Containers

NASA Technical Reports Server (NTRS)

Melton, Donald; Pratz, Earl Howard

1995-01-01

Fixture based on T-connector enables mass-spectrometric analysis of volatile contents of cylindrical containers without exposing contents to ambient conditions. Used to sample volatile contents of pressurized containers, contents of such enclosed processing systems as gas-phase reactors, gases in automotive emission systems, and gas in hostile environments.

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.

Impact of the volume of gaseous phase in closed reactors on ANC results and modelling

NASA Astrophysics Data System (ADS)

Drapeau, Clémentine; Delolme, Cécile; Lassabatere, Laurent; Blanc, Denise

2016-04-01

The understanding of the geochemical behavior of polluted solid materials is often challenging and requires huge expenses of time and money. Nevertheless, given the increasing amounts of polluted solid materials and related risks for the environment, it is more and more crucial to understand the leaching of majors and trace metals elements from these matrices. In the designs of methods to quantify pollutant solubilization, the combination of experimental procedures with modeling approaches has recently gained attention. Among usual methods, some rely on the association of ANC and geochemical modeling. ANC experiments - Acid Neutralization Capacity - consists in adding known quantities of acid or base to a mixture of water and contaminated solid materials at a given liquid / solid ratio in closed reactors. Reactors are agitated for 48h and then pH, conductivity, redox potential, carbon, majors and heavy metal solubilized are quantified. However, in most cases, the amounts of matrix and water do not reach the total volume of reactors, leaving some space for air (gaseous phase). Despite this fact, no clear indication is given in standard procedures about the effect of this gaseous phase. Even worse, the gaseous phase is never accounted for when exploiting or modeling ANC data. The gaseous phase may exchange CO2 with the solution, which may, in turn, impact both pH and element release. This study lies within the most general framework for the use of geochemical modeling for the prediction of ANC results for the case of pure phases to real phase assemblages. In this study, we focus on the effect of the gaseous phase on ANC experiments on different mineral phases through geochemical modeling. To do so, we use PHREEQC code to model the evolution of pH and element release (including majors and heavy metals) when several matrices are put in contact with acid or base. We model the following scenarios for the gaseous phase: no gas, contact with the atmosphere (open system) and real reactors conditions (semi-closed systems). The solid phases tested are pure phases (calcite, sulfides, etc.) and synthetic assemblages mimicking complex polluted matrices. The modeling clearly shows that the systems are sensitive to the opening to the atmosphere. If the open system and the system with no gas are entirely different, "real" reactors also differ significantly from the other systems. Apparently, the presence of the gaseous phase in reactors greatly impacts pH and element release. This parameter should be accounted for in ANC experimental procedures and modeling. In addition to this numerical study, experimental results, previously obtained for urban polluted sediments, are analyzed in lights of the findings of the numerical study. This step allows us to strengthen conclusions and to pinpoint at the necessity to account for the gaseous phase when performing and modeling ANC experiments.

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.

HOMOGENEOUS NUCLEAR POWER REACTOR

DOEpatents

King, L.D.P.

1959-09-01

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

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...

75 FR 36698 - Draft Regulatory Guide: Issuance, Availability

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-28

... information based on the likelihood of pipe breaks of different sizes. The rule would divide all coolant... to and including a ``transition break size,'' and breaks larger than the transition size up to the largest pipe in the reactor coolant system. Selection of the transition size was based upon pipe break...

Efficiency of wastewater treatment in SBR and IFAS-MBSBBR systems in specified technological conditions.

PubMed

Sytek-Szmeichel, K; Podedworna, J; Zubrowska-Sudol, M

2016-01-01

The objective of this study is to compare wastewater treatment effectiveness in sequencing batch reactor (SBR) and integrated fixed-film activated sludge-moving-bed sequencing batch biofilm reactor (IFAS-MBSBBR) systems in specific technological conditions. The comparison of these two technologies was based on the following assumptions, shared by both series, I and II: the reactor's active volume was 28 L; 8-hour cycle of reactor's work, with the same sequence and duration of its consecutive phases; and the dissolved oxygen concentration in the aerobic phases was maintained at a level of 3.0 mg O2/L. For both experimental series (I and II), comparable effectiveness of organic compound (chemical oxygen demand (COD)) removal, nitrification and biological phosphorus removal has been obtained at levels of 95.1%, 97% and 99%, respectively. The presence of the carrier improved the efficiency of total nitrogen removal from 86.3% to 91.7%. On the basis of monitoring tests, it has been found that the ratio of simultaneous denitrification in phases with aeration to the total efficiency of denitrification in the cycle was 1.5 times higher for IFAS-MBSBBR.

Removal of gasoline volatile organic compounds via air biofiltration

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

Miller, R.S.; Saberiyan, A.G.; Esler, C.T.

1995-12-31

Volatile organic compounds (VOCs) generated by vapor extraction and air-stripping systems can be biologically treated in an air biofiltration unit. An air biofilter consists of one or more beds of packing material inoculated with heterotrophic microorganisms capable of degrading the organic contaminant of concern. Waste gases and oxygen are passed through the inoculated packing material, where the microorganisms will degrade the contaminant and release CO{sub 2} + H{sub 2}O. Based on data obtained from a treatability study, a full-scale unit was designed and constructed to be used for treating gasoline vapors generated by a vapor-extraction and groundwater-treatment system at amore » site in California. The unit is composed of two cylindrical reactors with a total packing volume of 3 m{sup 3}. Both reactors are packed with sphagnum moss and inoculated with hydrocarbon-degrading microorganisms of Pseudomonas and Arthrobacter spp. The two reactors are connected in series for air-flow passage. Parallel lines are used for injection of water, nutrients, and buffer to each reactor. Data collected during the startup program have demonstrated an air biofiltration unit with high organic-vapor-removal efficiency.« less

78 FR 47010 - Proposed Safety Evaluation for Plant-Specific

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-02

... to the existing SR on the reactor core isolation cooling system to maintain consistency within the... TS Bases are revised to reflect the change to the SRs. The proposed change captures the on-going...

Temperature measuring analysis of the nuclear reactor fuel assembly

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

Urban, F., E-mail: jozef.bereznai@stuba.sk, E-mail: zdenko.zavodny@stuba.sk; Kučák, L., E-mail: jozef.bereznai@stuba.sk, E-mail: zdenko.zavodny@stuba.sk; Bereznai, J., E-mail: jozef.bereznai@stuba.sk, E-mail: zdenko.zavodny@stuba.sk

2014-08-06

Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuelmore » assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.« less

Zero valent iron simultaneously enhances methane production and sulfate reduction in anaerobic granular sludge reactors.

PubMed

Liu, Yiwen; Zhang, Yaobin; Ni, Bing-Jie

2015-05-15

Zero valent iron (ZVI) packed anaerobic granular sludge reactors have been developed for improved anaerobic wastewater treatment. In this work, a mathematical model is developed to describe the enhanced methane production and sulfate reduction in anaerobic granular sludge reactors with the addition of ZVI. The model is successfully calibrated and validated using long-term experimental data sets from two independent ZVI-enhanced anaerobic granular sludge reactors with different operational conditions. The model satisfactorily describes the chemical oxygen demand (COD) removal, sulfate reduction and methane production data from both systems. Results show ZVI directly promotes propionate degradation and methanogenesis to enhance methane production. Simultaneously, ZVI alleviates the inhibition of un-dissociated H2S on acetogens, methanogens and sulfate reducing bacteria (SRB) through buffering pH (Fe(0) + 2H(+) = Fe(2+) + H2) and iron sulfide precipitation, which improve the sulfate reduction capacity, especially under deterioration conditions. In addition, the enhancement of ZVI on methane production and sulfate reduction occurs mainly at relatively low COD/ [Formula: see text] ratio (e.g., 2-4.5) rather than high COD/ [Formula: see text] ratio (e.g., 16.7) compared to the reactor without ZVI addition. The model proposed in this work is expected to provide support for further development of a more efficient ZVI-based anaerobic granular system. Copyright © 2015 Elsevier Ltd. All rights reserved.

Room temperature micro-hydrogen-generator

NASA Astrophysics Data System (ADS)

Gervasio, Don; Tasic, Sonja; Zenhausern, Frederic

A new compact and cost-effective hydrogen-gas generator has been made that is well suited for supplying hydrogen to a fuel-cell for providing base electrical power to hand-carried appliances. This hydrogen-generator operates at room temperature, ambient pressure and is orientation-independent. The hydrogen-gas is generated by the heterogeneous catalytic hydrolysis of aqueous alkaline borohydride solution as it flows into a micro-reactor. This reactor has a membrane as one wall. Using the membrane keeps the liquid in the reactor, but allows the hydrogen-gas to pass out of the reactor to a fuel-cell anode. Aqueous alkaline 30 wt% borohydride solution is safe and promotes long application life, because this solution is non-toxic, non-flammable, and is a high energy-density (≥2200 W-h per liter or per kilogram) hydrogen-storage solution. The hydrogen is released from this storage-solution only when it passes over the solid catalyst surface in the reactor, so controlling the flow of the solution over the catalyst controls the rate of hydrogen-gas generation. This allows hydrogen generation to be matched to hydrogen consumption in the fuel-cell, so there is virtually no free hydrogen-gas during power generation. A hydrogen-generator scaled for a system to provide about 10 W electrical power is described here. However, the technology is expected to be scalable for systems providing power spanning from 1 W to kW levels.

Structural materials for Gen-IV nuclear reactors: Challenges and opportunities

NASA Astrophysics Data System (ADS)

Murty, K. L.; Charit, I.

2008-12-01

Generation-IV reactor design concepts envisioned thus far cater toward a common goal of providing safer, longer lasting, proliferation-resistant and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-IV reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses and extremely corrosive environment, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This paper presents a summary of various Gen-IV reactor concepts, with emphasis on the structural materials issues depending on the specific application areas. This paper also discusses the challenges involved in using the existing materials under both service and off-normal conditions. Tasks become increasingly complex due to the operation of various fundamental phenomena like radiation-induced segregation, radiation-enhanced diffusion, precipitation, interactions between impurity elements and radiation-produced defects, swelling, helium generation and so forth. Further, high temperature capability (e.g. creep properties) of these materials is a critical, performance-limiting factor. It is demonstrated that novel alloy and microstructural design approaches coupled with new materials processing and fabrication techniques may mitigate the challenges, and the optimum system performance may be achieved under much demanding conditions.

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.

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

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.

Comparison of simple, small, full-scale sewage treatment systems in Brazil: UASB-maturation ponds-coarse filter; UASB-horizontal subsurface-flow wetland; vertical-flow wetland (first stage of French system).

PubMed

von Sperling, M

2015-01-01

This paper presents a comparison between three simple sewage treatment lines involving natural processes: (a) upflow anaerobic sludge blanket (UASB) reactor-three maturation ponds in series-coarse rock filter; (b) UASB reactor-horizontal subsurface-flow constructed wetland; and (c) vertical-flow constructed wetlands treating raw sewage (first stage of the French system). The evaluation was based on several years of practical experience with three small full-scale plants receiving the same influent wastewater (population equivalents of 220, 60 and 100 inhabitants) in the city of Belo Horizonte, Brazil. The comparison included interpretation of concentrations and removal efficiencies based on monitoring data (organic matter, solids, nitrogen, phosphorus, coliforms and helminth eggs), together with an evaluation of practical aspects, such as land and volume requirements, sludge production and handling, plant management, clogging and others. Based on an integrated evaluation of all aspects involved, it is worth emphasizing that each system has its own specificities, and no generalization can be made on the best option. The overall conclusion is that the three lines are suitable for sewage treatment in small communities in warm-climate regions.

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.

Numerical modeling study on the epitaxial growth of silicon from dichlorosilane

NASA Astrophysics Data System (ADS)

Zaidi, Imama; Jang, Yeon-Ho; Ko, Dong Guk; Im, Ik-Tae

2018-02-01

Computer simulations play an important role in determining the optimal design parameters for chemical vapor deposition (CVD) reactors, such as flow rates, positions of the inlet and outlet orifices, and rotational rates, etc. Reliability of the results of these simulations depends on the set of chemical reaction used to represent the process of deposition in the reactor. Aim of the present work is to validate the simple empirical reaction to model the epitaxial growth of silicon for a Dichlorosilane-H2 (DCS)-H2 system. Governing equations for continuity, momentum, energy, and reacting species are solved numerically using the finite volume method. The agreement between experimental and predicted growth rates for various DCS flow rates is shown to be satisfactory. The increase in growth rate with the increase in pressure is in accordance with the available data. Based on the validated chemical reaction model, a study was carried out to analyze the uniformity of the silicon layer thickness for two different flow rates in a planetary reactor. It was concluded that, based on the operating conditions, the uniformity of the silicon layer over the wafer is independent of the satellite rotational rate in the reactor.

Lunar Surface Reactor Shielding Study

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

Kang, Shawn; McAlpine, William; Lipinski, Ronald

A nuclear reactor system could provide power to support long term human exploration of the moon. Such a system would require shielding to protect astronauts from its emitted radiations. Shielding studies have been performed for a Gas Cooled Reactor system because it is considered to be the most suitable nuclear reactor system available for lunar exploration, based on its tolerance of oxidizing lunar regolith and its good conversion efficiency. The goals of the shielding studies were to determine a material shielding configuration that reduces the dose (rem) to the required level in order to protect astronauts, and to estimate themore » mass of regolith that would provide an equivalent protective effect if it were used as the shielding material. All calculations were performed using MCNPX, a Monte Carlo transport code. Lithium hydride must be kept between 600 K and 700 K to prevent excessive swelling from large amounts of gamma or neutron irradiation. The issue is that radiation damage causes separation of the lithium and the hydrogen, resulting in lithium metal and hydrogen gas. The proposed design uses a layer of B4C to reduce the combined neutron and gamma dose to below 0.5Grads before the LiH is introduced. Below 0.5Grads the swelling in LiH is small (less than about 1%) for all temperatures. This approach causes the shield to be heavier than if the B4C were replaced by LiH, but it makes the shield much more robust and reliable.« less

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.

Optical Waveguide Solar Energy System for Lunar Materials Processing

NASA Technical Reports Server (NTRS)

Nakamura, T.; Case, J. A.; Senior, C. L.

1997-01-01

This paper discusses results of our work on development of the Optical Waveguide (OW) Solar Energy System for Lunar Materials Processing. In the OW system as shown, solar radiation is collected by the concentrator which transfers the concentrated solar radiation to the OW transmission line consisting of low-loss optical fibers. The OW line transmits the solar radiation to the thermal reactor of the lunar materials processing plant. The feature of the OW system are: (1) Highly concentrated solar radiation (up to 104 suns) can be transmitted via flexible OW lines directly into the thermal reactor for materials processing: (2) Solar radiation intensity or spectra can be tailored to specific materials processing steps; (3) Provide solar energy to locations or inside of enclosures that would not otherwise have an access to solar energy; and (4) The system can be modularized and can be easily transported to and deployed at the lunar base.

The hard start phenomena in hypergolic engines. Volume 4: The chemistry of hydrazine fuels and nitrogen tetroxide propellant systems

NASA Technical Reports Server (NTRS)

Miron, Y.; Perlee, H. E.

1974-01-01

The various chemical reactions that occur and that could possibly occur in the RCS engines utilizing hydrazine-type fuel/nitrogen tetroxide propellant systems, prior to ignition (preignition), during combustion, and after combustion (postcombustion), and endeavors to relate the hard-start phenomenon to some of these reactions are discussed. The discussion is based on studies utilizing a variety of experimental techniques and apparatus as well as current theories of chemical reactions and reaction kinetics. The chemical reactions were studied in low pressure gas flow reactors, low temperature homogeneous- and heterogeneous-phase reactors, simulated two-dimensional (2-D) engines, and scaled and full size engines.

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

Application of reliability-centered maintenance to boiling water reactor emergency core cooling systems fault-tree analysis

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

Choi, Y.A.; Feltus, M.A.

1995-07-01

Reliability-centered maintenance (RCM) methods are applied to boiling water reactor plant-specific emergency core cooling system probabilistic risk assessment (PRA) fault trees. The RCM is a technique that is system function-based, for improving a preventive maintenance (PM) program, which is applied on a component basis. Many PM programs are based on time-directed maintenance tasks, while RCM methods focus on component condition-directed maintenance tasks. Stroke time test data for motor-operated valves (MOVs) are used to address three aspects concerning RCM: (a) to determine if MOV stroke time testing was useful as a condition-directed PM task; (b) to determine and compare the plant-specificmore » MOV failure data from a broad RCM philosophy time period compared with a PM period and, also, compared with generic industry MOV failure data; and (c) to determine the effects and impact of the plant-specific MOV failure data on core damage frequency (CDF) and system unavailabilities for these emergency systems. The MOV stroke time test data from four emergency core cooling systems [i.e., high-pressure coolant injection (HPCI), reactor core isolation cooling (RCIC), low-pressure core spray (LPCS), and residual heat removal/low-pressure coolant injection (RHR/LPCI)] were gathered from Philadelphia Electric Company`s Peach Bottom Atomic Power Station Units 2 and 3 between 1980 and 1992. The analyses showed that MOV stroke time testing was not a predictor for eminent failure and should be considered as a go/no-go test. The failure data from the broad RCM philosophy showed an improvement compared with the PM-period failure rates in the emergency core cooling system MOVs. Also, the plant-specific MOV failure rates for both maintenance philosophies were shown to be lower than the generic industry estimates.« less

Human Factors and Technical Considerations for a Computerized Operator Support System Prototype

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

Ulrich, Thomas Anthony; Lew, Roger Thomas; Medema, Heather Dawne

2015-09-01

A prototype computerized operator support system (COSS) has been developed in order to demonstrate the concept and provide a test bed for further research. The prototype is based on four underlying elements consisting of a digital alarm system, computer-based procedures, PI&D system representations, and a recommender module for mitigation actions. At this point, the prototype simulates an interface to a sensor validation module and a fault diagnosis module. These two modules will be fully integrated in the next version of the prototype. The initial version of the prototype is now operational at the Idaho National Laboratory using the U.S. Departmentmore » of Energy’s Light Water Reactor Sustainability (LWRS) Human Systems Simulation Laboratory (HSSL). The HSSL is a full-scope, full-scale glass top simulator capable of simulating existing and future nuclear power plant main control rooms. The COSS is interfaced to the Generic Pressurized Water Reactor (gPWR) simulator with industry-typical control board layouts. The glass top panels display realistic images of the control boards that can be operated by touch gestures. A section of the simulated control board was dedicated to the COSS human-system interface (HSI), which resulted in a seamless integration of the COSS into the normal control room environment. A COSS demonstration scenario has been developed for the prototype involving the Chemical & Volume Control System (CVCS) of the PWR simulator. It involves a primary coolant leak outside of containment that would require tripping the reactor if not mitigated in a very short timeframe. The COSS prototype presents a series of operator screens that provide the needed information and soft controls to successfully mitigate the event.« less

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...

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

Not Available

The Department of Energy (DOE) has contracted with Asea Brown Boveri-Combustion Engineering (ABB-CE) to provide information on the capability of ABB-CE`s System 80 + Advanced Light Water Reactor (ALWR) to transform, through reactor burnup, 100 metric tonnes (MT) of weapons grade plutonium (Pu) into a form which is not readily useable in weapons. This information is being developed as part of DOE`s Plutonium Disposition Study, initiated by DOE in response to Congressional action. This document, Volume 1, presents a technical description of the various elements of the System 80 + Standard Plant Design upon which the Plutonium Disposition Study wasmore » based. The System 80 + Standard Design is fully developed and directly suited to meeting the mission objectives for plutonium disposal. The bass U0{sub 2} plant design is discussed here.« less

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

NASA Technical Reports Server (NTRS)

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

1973-01-01

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

Scaling Studies for Advanced High Temperature Reactor Concepts, Final Technical Report: October 2014—December 2017

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

Woods, Brian; Gutowska, Izabela; Chiger, Howard

Computer simulations of nuclear reactor thermal-hydraulic phenomena are often used in the design and licensing of nuclear reactor systems. In order to assess the accuracy of these computer simulations, computer codes and methods are often validated against experimental data. This experimental data must be of sufficiently high quality in order to conduct a robust validation exercise. In addition, this experimental data is generally collected at experimental facilities that are of a smaller scale than the reactor systems that are being simulated due to cost considerations. Therefore, smaller scale test facilities must be designed and constructed in such a fashion tomore » ensure that the prototypical behavior of a particular nuclear reactor system is preserved. The work completed through this project has resulted in scaling analyses and conceptual design development for a test facility capable of collecting code validation data for the following high temperature gas reactor systems and events— 1. Passive natural circulation core cooling system, 2. pebble bed gas reactor concept, 3. General Atomics Energy Multiplier Module reactor, and 4. prismatic block design steam-water ingress event. In the event that code validation data for these systems or events is needed in the future, significant progress in the design of an appropriate integral-type test facility has already been completed as a result of this project. Where applicable, the next step would be to begin the detailed design development and material procurement. As part of this project applicable scaling analyses were completed and test facility design requirements developed. Conceptual designs were developed for the implementation of these design requirements at the Oregon State University (OSU) High Temperature Test Facility (HTTF). The original HTTF is based on a ¼-scale model of a high temperature gas reactor concept with the capability for both forced and natural circulation flow through a prismatic core with an electrical heat source. The peak core region temperature capability is 1400°C. As part of this project, an inventory of test facilities that could be used for these experimental programs was completed. Several of these facilities showed some promise, however, upon further investigation it became clear that only the OSU HTTF had the power and/or peak temperature limits that would allow for the experimental programs envisioned herein. Thus the conceptual design and feasibility study development focused on examining the feasibility of configuring the current HTTF to collect validation data for these experimental programs. In addition to the scaling analyses and conceptual design development, a test plan was developed for the envisioned modified test facility. This test plan included a discussion on an appropriate shakedown test program as well as the specific matrix tests. Finally, a feasibility study was completed to determine the cost and schedule considerations that would be important to any test program developed to investigate these designs and events.« less

Low acetate concentrations favor polyphosphate-accumulating organisms over glycogen-accumulating organisms in enhanced biological phosphorus removal from wastewater.

PubMed

Tu, Yunjie; Schuler, Andrew J

2013-04-16

Glycogen-accumulating organisms (GAOs) are thought to compete with polyphosphate-accumulating organisms (PAOs) in enhanced biological phosphorus removal (EBPR) wastewater treatment systems. A laboratory sequencing batch reactor (SBR) was operated for one year to test the hypothesis that PAOs have a competitive advantage at low acetate concentrations, with a focus on low pH conditions previously shown to favor GAOs. PAOs dominated the system under conventional SBR operation with rapid acetate addition (producing high in-reactor concentrations) and pH values of 7.4-8.4. GAOs dominated when the pH was decreased (6.4-7.0). Decreasing the acetate addition rate led to very low reactor acetate concentrations, and PAOs recovered, supporting the study hypothesis. When the acetate feed rate was increased, EBPR failed again. Dominant PAOs and GAOs were Candidatus Accumulibacter phosphatis and Defluviicoccus Cluster 2, respectively, according to fluorescent in situ hybridization and 454 pyrosequencing. Surprisingly, GAOs were not the immediate causes of PAO failures, based on functional and population measurements. Pyrosequencing results suggested Dechloromonas and Tetrasphaera spp. may have also been PAOs, and additional potential GAOs were also identified. Full-scale systems typically have lower in-reactor acetate concentrations than laboratory SBRs, and so, previous laboratory studies may have overestimated the practical importance of GAOs as causes of EBPR failure.

Methods and apparatuses for deoxygenating pyrolysis oil

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

Baird, Lance Awender; Brandvold, Timothy A.; Frey, Stanley Joseph

Methods and apparatuses are provided for deoxygenating pyrolysis oil. A method includes contacting a pyrolysis oil with a deoxygenation catalyst in a first reactor at deoxygenation conditions to produce a first reactor effluent. The first reactor effluent has a first oxygen concentration and a first hydrogen concentration, based on hydrocarbons in the first reactor effluent, and the first reactor effluent includes an aromatic compound. The first reactor effluent is contacted with a dehydrogenation catalyst in a second reactor at conditions that deoxygenate the first reactor effluent while preserving the aromatic compound to produce a second reactor effluent. The second reactormore » effluent has a second oxygen concentration lower than the first oxygen concentration and a second hydrogen concentration that is equal to or lower than the first hydrogen concentration, where the second oxygen concentration and the second hydrogen concentration are based on the hydrocarbons in the second reactor effluent.« 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.

Analysis on Reactor Criticality Condition and Fuel Conversion Capability Based on Different Loaded Plutonium Composition in FBR Core

NASA Astrophysics Data System (ADS)

Permana, Sidik; Saputra, Geby; Suzuki, Mitsutoshi; Saito, Masaki

2017-01-01

Reactor criticality condition and fuel conversion capability are depending on the fuel arrangement schemes, reactor core geometry and fuel burnup process as well as the effect of different fuel cycle and fuel composition. Criticality condition of reactor core and breeding ratio capability have been investigated in this present study based on fast breeder reactor (FBR) type for different loaded fuel compositions of plutonium in the fuel core regions. Loaded fuel of Plutonium compositions are based on spent nuclear fuel (SNF) of light water reactor (LWR) for different fuel burnup process and cooling time conditions of the reactors. Obtained results show that different initial fuels of plutonium gives a significant chance in criticality conditions and fuel conversion capability. Loaded plutonium based on higher burnup process gives a reduction value of criticality condition or less excess reactivity. It also obtains more fuel breeding ratio capability or more breeding gain. Some loaded plutonium based on longer cooling time of LWR gives less excess reactivity and in the same time, it gives higher breeding ratio capability of the reactors. More composition of even mass plutonium isotopes gives more absorption neutron which affects to decresing criticality or less excess reactivity in the core. Similar condition that more absorption neutron by fertile material or even mass plutonium will produce more fissile material or odd mass plutonium isotopes to increase the breeding gain of the reactor.

Design of an Experimental Facility for Passive Heat Removal in Advanced Nuclear Reactors

NASA Astrophysics Data System (ADS)

Bersano, Andrea

With reference to innovative heat exchangers to be used in passive safety system of Gen- eration IV nuclear reactors and Small Modular Reactors it is necessary to study the natural circulation and the efficiency of heat removal systems. Especially in safety systems, as the decay heat removal system of many reactors, it is increasing the use of passive components in order to improve their availability and reliability during possible accidental scenarios, reducing the need of human intervention. Many of these systems are based on natural circulation, so they require an intense analysis due to the possible instability of the related phenomena. The aim of this thesis work is to build a scaled facility which can reproduce, in a simplified way, the decay heat removal system (DHR2) of the lead-cooled fast reactor ALFRED and, in particular, the bayonet heat exchanger, which transfers heat from lead to water. Given the thermal power to be removed, the natural circulation flow rate and the pressure drops will be studied both experimentally and numerically using the code RELAP5 3D. The first phase of preliminary analysis and project includes: the calculations to design the heat source and heat sink, the choice of materials and components and CAD drawings of the facility. After that, the numerical study is performed using the thermal-hydraulic code RELAP5 3D in order to simulate the behavior of the system. The purpose is to run pretest simulations of the facility to optimize the dimensioning setting the operative parameters (temperature, pressure, etc.) and to chose the most adequate measurement devices. The model of the system is continually developed to better simulate the system studied. High attention is dedicated to the control logic of the system to obtain acceptable results. The initial experimental tests phase consists in cold zero power tests of the facility in order to characterize and to calibrate the pressure drops. In future works the experimental results will be compared to the values predicted by the system code and differences will be discussed with the ultimate goal to qualify RELAP5-3D for the analysis of decay heat removal systems in natural circulation. The numerical data will be also used to understand the key parameters related to the heat transfer in natural circulation and to optimize the operation of the system.

Grey water characteristics and treatment options for rural areas in Jordan.

PubMed

Halalsheh, M; Dalahmeh, S; Sayed, M; Suleiman, W; Shareef, M; Mansour, M; Safi, M

2008-09-01

Low water consumption in rural areas in Jordan had resulted in the production of concentrated grey water. Average COD, BOD and TSS values were 2568mg/l, 1056mg/l and 845mg/l, respectively. The average grey water generation was measured to be 14L/c.d. Three different treatment options were selected based on certain criterions, and discussed in this article. The examined treatment systems are septic tank followed by intermittent sand filter; septic tank followed by wetlands; and UASB-hybrid reactor. Advantages and disadvantages of each system are presented. It was concluded that UASB-hybrid reactor would be the most suitable treatment option in terms of compactness and simplicity in operation. The volume of UASB-hybrid reactor was calculated to be 0.268m(3) with a surface area of 0.138m(2) for each house having 10 inhabitants on average. Produced effluent is expected to meet Jordanian standards set for reclaimed water reuse in irrigating fruit trees.

Fuel processing for PEM fuel cells: transport and kinetic issues of system design

NASA Astrophysics Data System (ADS)

Zalc, J. M.; Löffler, D. G.

In light of the distribution and storage issues associated with hydrogen, efficient on-board fuel processing will be a significant factor in the implementation of PEM fuel cells for automotive applications. Here, we apply basic chemical engineering principles to gain insight into the factors that limit performance in each component of a fuel processor. A system consisting of a plate reactor steam reformer, water-gas shift unit, and preferential oxidation reactor is used as a case study. It is found that for a steam reformer based on catalyst-coated foils, mass transfer from the bulk gas to the catalyst surface is the limiting process. The water-gas shift reactor is expected to be the largest component of the fuel processor and is limited by intrinsic catalyst activity, while a successful preferential oxidation unit depends on strict temperature control in order to minimize parasitic hydrogen oxidation. This stepwise approach of sequentially eliminating rate-limiting processes can be used to identify possible means of performance enhancement in a broad range of applications.

Separations in the STATS report

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

Choppin, G.R.

1996-12-31

The Separations Technology and Transmutation Systems (STATS) Committee formed a Subcommittee on Separations. This subcommittee was charged with evaluating the separations proposed for the several reactor and accelerator transmutation systems. It was also asked to review the processing options for the safe management of high-level waste generated by the defense programs, in particular, the special problems involved in dealing with the waste at the U.S. Department of Energy (DOE) facility in Hanford, Washington. Based on the evaluations from the Subcommittee on Separations, the STATS Committee concluded that for the reactor transmutation programs, aqueous separations involving a combination of PUREX andmore » TRUEX solvent extraction processes could be used. However, additional research and development (R&D) would be required before full plant-scale use of the TRUEX technology could be employed. Alternate separations technology for the reactor transmutation program involves pyroprocessing. This process would require a significant amount of R&D before its full-scale application can be evaluated.« less