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.

Apparatus and method for thermal power generation

DOEpatents

Cohen, Paul; Redding, Arnold H.

1978-01-01

An improved thermal power plant and method of power generation which minimizes thermal stress and chemical impurity buildup in the vaporizing component, particularly beneficial under loss of normal feed fluid and startup conditions. The invention is particularly applicable to a liquid metal fast breeder reactor plant.

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

NASA Technical Reports Server (NTRS)

Hissam, David Andy; Stewart, Eric T.

2006-01-01

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

Nuclear Power as a Basis for Future Electricity Generation

NASA Astrophysics Data System (ADS)

Pioro, Igor; Buruchenko, Sergey

2017-12-01

It is well known that electrical-power generation is the key factor for advances in industry, agriculture, technology and the level of living. Also, strong power industry with diverse energy sources is very important for country independence. In general, electrical energy can be generated from: 1) burning mined and refined energy sources such as coal, natural gas, oil, and nuclear; and 2) harnessing energy sources such as hydro, biomass, wind, geothermal, solar, and wave power. Today, the main sources for electrical-energy generation are: 1) thermal power - primarily using coal and secondarily - natural gas; 2) “large” hydro power from dams and rivers and 3) nuclear power from various reactor designs. The balance of the energy sources is from using oil, biomass, wind, geothermal and solar, and have visible impact just in some countries. In spite of significant emphasis in the world on using renewables sources of energy, in particular, wind and solar, they have quite significant disadvantages compared to “traditional” sources for electricity generation such as thermal, hydro, and nuclear. These disadvantages include low density of energy, which requires large areas to be covered with wind turbines or photovoltaic panels or heliostats, and dependence of these sources on Mother Nature, i.e., to be unreliable ones and to have low (20 - 40%) or very low (5 - 15%) capacity factors. Fossil-fueled power plants represent concentrated and reliable source of energy. Also, they operate usually as “fast-response” plants to follow rapidly changing electrical-energy consumption during a day. However, due to combustion process they emit a lot of carbon dioxide, which contribute to the climate change in the world. Moreover, coal-fired power plants, as the most popular ones, create huge amount of slag and ash, and, eventually, emit other dangerous and harmful gases. Therefore, Nuclear Power Plants (NPPs), which are also concentrated and reliable source of energy

Medical aspects of power generation, present and future.

PubMed

Linnemann, R E

1979-01-01

It can be seen that the radiation emissions of nuclear power plants are small indeed, compared to natural background radiation and other man-made sources of radiation. For example, the poulation is exposed to 100 times more radiation from television sets than from nuclear power reactors. The assumed risks to the people in this country from nuclear power reactors are also small compared to the normal risks which are tolerated in this society. The complete elimination of all hazards is a most difficult if not impossible task. If we need and desire a certain level of electrical energy, if we must choose between alternative sourves of the energy, then it is apparent that the total impact on our health from nuclear power generation of electricity, under normal operations and in consideration of catastrophic accident probabilities, is significantly less than that of continuing or increasing use of fossil fuels to generate electricity.

EXPERIMENTAL MOLTEN-SALT-FUELED 30-Mw POWER REACTOR

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

Alexander, L.G.; Kinyon, B.W.; Lackey, M.E.

1960-03-24

A preliminary design study was made of an experimental molten-salt- fueled power reactor. The reactor considered is a single-region homogeneous burner coupled with a Loeffler steam-generating cycle. Conceptual plant layouts, basic information on the major fuel circuit components, a process flowsheet, and the nuclear characteristics of the core are presented. The design plant electrical output is 10 Mw, and the total construction cost is estimated to be approximately ,000,000. (auth)

Addition of acetate improves stability of power generation using microbial fuel cells treating domestic wastewater.

PubMed

Stager, Jennifer L; Zhang, Xiaoyuan; Logan, Bruce E

2017-12-01

Power generation using microbial fuel cells (MFCs) must provide stable, continuous conversion of organic matter in wastewaters into electricity. However, when relatively small diameter (0.8cm) graphite fiber brush anodes were placed close to the cathodes in MFCs, power generation was unstable during treatment of low strength domestic wastewater. One reactor produced 149mW/m 2 before power generation failed, while the other reactor produced 257mW/m 2 , with both reactors exhibiting severe power overshoot in polarization tests. Using separators or activated carbon cathodes did not result in stable operation as the reactors continued to exhibit power overshoot based on polarization tests. However, adding acetate (1g/L) to the wastewater produced stable performance during fed batch and continuous flow operation, and there was no power overshoot in polarization tests. These results highlight the importance of wastewater strength and brush anode size for producing stable and continuous power in compact MFCs. Copyright © 2017 Elsevier B.V. All rights reserved.

Power consumption analysis DBD plasma ozone generator

NASA Astrophysics Data System (ADS)

Nur, M.; Restiwijaya, M.; Muchlisin, Z.; Susan, I. A.; Arianto, F.; Widyanto, S. A.

2016-11-01

Studies on the consumption of energy by an ozone generator with various constructions electrodes of dielectric barrier discharge plasma (DBDP) reactor has been carried out. This research was done to get the configuration of the reactor, that is capable to produce high ozone concentrations with low energy consumption. BDBP reactors were constructed by spiral- cylindrical configuration, plasma ozone was generated by high voltage AC voltage up to 25 kV and maximum frequency of 23 kHz. The reactor consists of an active electrode in the form of a spiral-shaped with variation diameter Dc, and it was made by using copper wire with diameter Dw. In this research, we variated number of loops coil windings N as well as Dc and Dw. Ozone concentrations greater when the wire's diameter Dw and the diameter of the coil windings applied was greater. We found that impedance greater will minimize the concentration of ozone, in contrary to the greater capacitance will increase the concentration of ozone. The ozone concentrations increase with augmenting of power. Maximum power is effective at DBD reactor spiral-cylinder is on the Dc = 20 mm, Dw = 1.2 mm, and the number of coil windings N = 10 loops with the resulting concentration is greater than 20 ppm and it consumes energy of 177.60 watts

POWER GENERATING NEUTRONIC REACTOR SYSTEM

DOEpatents

Vernon, H.C.

1958-03-01

This patent relates to reactor systems of the type wherein the cooiing medium is a liquid which is converted by the heat of the reaction to steam which is conveyed directly to a pnime mover such as a steam turbine driving a generatore after which it is condensed and returred to the coolant circuit. In this design, the reactor core is disposed within a tank for containing either a slurry type fuel or an aggregation of solid fuel elements such as elongated rods submerged in a liquid moderator such as heavy water. The top of the tank is provided with a nozzle which extends into an expansion chamber connected with the upper end of the tank, the coolant being maintained in the expansion chamber at a level above the nozzle and the steam being formed in the expansion chamber.

Deployment history and design considerations for space reactor power systems

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.

2009-05-01

The history of the deployment of nuclear reactors in Earth orbits is reviewed with emphases on lessons learned and the operation and safety experiences. The former Soviet Union's "BUK" power systems, with SiGe thermoelectric conversion and fast neutron energy spectrum reactors, powered a total of 31 Radar Ocean Reconnaissance Satellites (RORSATs) from 1970 to 1988 in 260 km orbit. Two of the former Soviet Union's TOPAZ reactors, with in-core thermionic conversion and epithermal neutron energy spectrum, powered two Cosmos missions launched in 1987 in ˜800 km orbit. The US' SNAP-10A system, with SiGe energy conversion and a thermal neutron energy spectrum reactor, was launched in 1965 in 1300 km orbit. The three reactor systems used liquid NaK-78 coolant, stainless steel structure and highly enriched uranium fuel (90-96 wt%) and operated at a reactor exit temperature of 833-973 K. The BUK reactors used U-Mo fuel rods, TOPAZ used UO 2 fuel rods and four ZrH moderator disks, and the SNAP-10A used moderated U-ZrH fuel rods. These low power space reactor systems were designed for short missions (˜0.5 kW e and ˜1 year for SNAP-10A, <3.0 kW e and <6 months for BUK, and ˜5.5 kW e and up to 1 year for TOPAZ). The deactivated BUK reactors at the end of mission, which varied in duration from a few hours to ˜4.5 months, were boosted into ˜800 km storage orbit with a decay life of more than 600 year. The ejection of the last 16 BUK reactor fuel cores caused significant contamination of Earth orbits with NaK droplets that varied in sizes from a few microns to 5 cm. Power systems to enhance or enable future interplanetary exploration, in-situ resources utilization on Mars and the Moon, and civilian missions in 1000-3000 km orbits would generate significantly more power of 10's to 100's kW e for 5-10 years, or even longer. A number of design options to enhance the operation reliability and safety of these high power space reactor power systems are presented and discussed.

Underground nuclear power station using self-regulating heat-pipe controlled reactors

DOEpatents

Hampel, Viktor E.

1989-01-01

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

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.

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.

NEUTRONIC REACTOR POWER PLANT

DOEpatents

Metcalf, H.E.

1962-12-25

This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

Background radiation measurements at high power research reactors

NASA Astrophysics Data System (ADS)

Ashenfelter, J.; Balantekin, B.; Baldenegro, C. X.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bowden, N. S.; Bryan, C. D.; Cherwinka, J. J.; Chu, R.; Classen, T.; Davee, D.; Dean, D.; Deichert, G.; Dolinski, M. J.; Dolph, J.; Dwyer, D. A.; Fan, S.; Gaison, J. K.; Galindo-Uribarri, A.; Gilje, K.; Glenn, A.; Green, M.; Han, K.; Hans, S.; Heeger, K. M.; Heffron, B.; Jaffe, D. E.; Kettell, S.; Langford, T. J.; Littlejohn, B. R.; Martinez, D.; McKeown, R. D.; Morrell, S.; Mueller, P. E.; Mumm, H. P.; Napolitano, J.; Norcini, D.; Pushin, D.; Romero, E.; Rosero, R.; Saldana, L.; Seilhan, B. S.; Sharma, R.; Stemen, N. T.; Surukuchi, P. T.; Thompson, S. J.; Varner, R. L.; Wang, W.; Watson, S. M.; White, B.; White, C.; Wilhelmi, J.; Williams, C.; Wise, T.; Yao, H.; Yeh, M.; Yen, Y.-R.; Zhang, C.; Zhang, X.; Prospect Collaboration

2016-01-01

Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. The general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

ENGINEERING AND CONSTRUCTING THE HALLAM NUCLEAR POWER FACILITY REACTOR STRUCTURE

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

Mahlmeister, J E; Haberer, W V; Casey, D F

1960-12-15

The Hallam Nuclear Power Facility reactor structure, including the cavity liner, is described, and the design philosophy and special design requirements which were developed during the preliminary and final engineering phases of the project are explained. The structure was designed for 600 deg F inlet and 1000 deg F outlet operating sodium temperatures and fabricated of austenitic and ferritic stainless steels. Support for the reactor core components and adequate containment for biological safeguards were readily provided even though quite conservative design philosophy was used. The calculated operating characteristics, including heat generation, temperature distributions and stress levels for full-power operation, aremore » summarized. Ship fabrication and field installation experiences are also briefly related. Results of this project have established that the sodium graphite reactor permits practical and economical fabrication and field erection procedures; considerably higher operating design temperatures are believed possible without radical design changes. Also, larger reactor structures can be similarly constructed for higher capacity (300 to 1000 Mwe) nuclear power plants. (auth)« less

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

An underground nuclear power station using self-regulating heat-pipe controlled reactors

DOEpatents

Hampel, V.E.

1988-05-17

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast- acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor. 5 figs.

Thermionic reactors for space nuclear power

NASA Technical Reports Server (NTRS)

Homeyer, W. G.; Merrill, M. H.; Holland, J. W.; Fisher, C. R.; Allen, D. T.

1985-01-01

Thermionic reactor designs for a variety of space power applications spanning the range from 5 kWe to 3 MWe are described. In all of these reactors, nuclear heat is converted directly to electrical energy in thermionic fuel elements (TFEs). A circulating reactor coolant carries heat from the core of TFEs directly to a heat rejection radiator system. The recent design of a thermionic reactor to meet the SP-100 requirements is emphasized. Design studies of reactors at other power levels show that the same TFE can be used over a broad range in power, and that design modifications can extend the range to many megawatts. The design of the SP-100 TFE is similar to that of TFEs operated successfully in test reactors, but with design improvements to extend the operating lifetime to seven years.

Study Neutronic of Small Pb-Bi Cooled Non-Refuelling Nuclear Power Plant Reactor (SPINNOR) with Hexagonal Geometry Calculation

NASA Astrophysics Data System (ADS)

Nur Krisna, Dwita; Su'ud, Zaki

2017-01-01

Nuclear reactor technology is growing rapidly, especially in developing Nuclear Power Plant (NPP). The utilization of nuclear energy in power generation systems has been progressing phase of the first generation to the fourth generation. This final project paper discusses the analysis neutronic one-cooled fast reactor type Pb-Bi, which is capable of operating up to 20 years without refueling. This reactor uses Thorium Uranium Nitride as fuel and operating on power range 100-500MWtNPPs. The method of calculation used a computer simulation program utilizing the SRAC. SPINNOR reactor is designed with the geometry of hexagonal shaped terrace that radially divided into three regions, namely the outermost regions with highest percentage of fuel, the middle regions with medium percentage of fuel, and most in the area with the lowest percentage. SPINNOR fast reactor operated for 20 years with variations in the percentage of Uranium-233 by 7%, 7.75%, and 8.5%. The neutronic calculation and analysis show that the design can be optimized in a fast reactor for thermal power output SPINNOR 300MWt with a fuel fraction 60% and variations of Uranium-233 enrichment of 7%-8.5%.

Background radiation measurements at high power research reactors

DOE PAGES

Ashenfelter, J.; Yeh, M.; Balantekin, B.; ...

2015-10-23

Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the backgroundmore » fields encountered. Furthermore, the general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.« less

Waste Generated from LMR-AMTEC Reactor Concept

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

Hasan, Ahmed; Mohamed, Yasser, T.; Mohammaden, Tarek, F.

2003-02-25

The candidate Liquid Metal Reactor-Alkali Metal Thermal -to- Electric Converter (LMR-AMTEC) is considered to be the first reactor that would use pure liquid potassium as a secondary coolant, in which potassium vapor aids in the conversion of thermal energy to electric energy. As with all energy production, the thermal generation of electricity produces wastes. These wastes must be managed in ways which safeguard human health and minimize their impact on the environment. Nuclear power is the only energy industry, which takes full responsibility for all its wastes. Based on the candidate design of the LMR-AMTEC components and the coolant types,more » different wastes will be generated from LMR. These wastes must be classified and characterized according to the U.S. Code of Federal Regulation, CFR. This paper defines the waste generation and waste characterization from LMR-AMTEC and reviews the applicable U.S. regulations that govern waste transportation, treatment, storage and final disposition. The wastes generated from LMR-AMTEC are characterized as: (1) mixed waste which is generated from liquid sodium contaminated by fission products and activated corrosion products; (2) hazardous waste which is generated from liquid potassium contaminated by corrosion products; (3) spent nuclear fuel; and (4) low-level radioactive waste which is generated from the packing materials (e.g. activated carbon in cold trap and purification units). The regulations and management of these wastes are summarized in this paper.« less

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.

Reactor power system deployment and startup

NASA Technical Reports Server (NTRS)

Wetch, J. R.; Nelin, C. J.; Britt, E. J.; Klein, G.

1985-01-01

This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems.

Small and medium power reactors 1987

NASA Astrophysics Data System (ADS)

1987-12-01

This TECDOC follows the publication of TECDOC-347: Small and Medium Power Reactors (SMPR) Project Initiation Study, Phase 1, published in 1985 and TECDOC-376: Small and Medium Power Reactors 1985 published in 1986. It is mainly intended for decision makers in Developing Member States interested in embarking on a nuclear power program. It consists of two parts: (1) guidelines for the introduction of small and medium power reactors in developing countries. These Guidelines were established during the Advisory Group Meeting held in Vienna from 11 to 15 May 1987. Their purpose is to review key aspects relating to the introduction of small and medium power reactors in developing countries; (2) up-dated information on SMPR Concepts Contributed by Supplier Industries. According to the recommendations of the Second Technical Committee Meeting on SMPRs held in Vienna in March 1985, this part contains the up-dated information formerly published in Annex 1 of the above mentioned TECDOC-347.

Autonomous Control Capabilities for Space Reactor Power Systems

NASA Astrophysics Data System (ADS)

Wood, Richard T.; Neal, John S.; Brittain, C. Ray; Mullens, James A.

2004-02-01

The National Aeronautics and Space Administration's (NASA's) Project Prometheus, the Nuclear Systems Program, is investigating a possible Jupiter Icy Moons Orbiter (JIMO) mission, which would conduct in-depth studies of three of the moons of Jupiter by using a space reactor power system (SRPS) to provide energy for propulsion and spacecraft power for more than a decade. Terrestrial nuclear power plants rely upon varying degrees of direct human control and interaction for operations and maintenance over a forty to sixty year lifetime. In contrast, an SRPS is intended to provide continuous, remote, unattended operation for up to fifteen years with no maintenance. Uncertainties, rare events, degradation, and communications delays with Earth are challenges that SRPS control must accommodate. Autonomous control is needed to address these challenges and optimize the reactor control design. In this paper, we describe an autonomous control concept for generic SRPS designs. The formulation of an autonomous control concept, which includes identification of high-level functional requirements and generation of a research and development plan for enabling technologies, is among the technical activities that are being conducted under the U.S. Department of Energy's Space Reactor Technology Program in support of the NASA's Project Prometheus. The findings from this program are intended to contribute to the successful realization of the JIMO mission.

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.

Compact power reactor

DOEpatents

Wetch, Joseph R.; Dieckamp, Herman M.; Wilson, Lewis A.

1978-01-01

There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

100-kWe lunar/Mars surface power utilizing the SP-100 reactor with dynamic conversion

NASA Technical Reports Server (NTRS)

Harty, Richard B.; Mason, Lee S.

1992-01-01

Results are presented from a study of the coupling of an SP-100 nuclear reactor with either a Stirling or Brayton power system, at the 100 kWe level, for a power generating system suitable for operation in the lunar and Martian surface environments. In the lunar environment, the reactor and primary coolant loop would be contained in a guard vessel to protect from a loss of primary loop containment. For Mars, all refractory components, including the reactor, coolant, and power conversion components will be contained in a vacuum vessel for protection against the CO2 environment.

Small reactor power system for space application

NASA Technical Reports Server (NTRS)

Shirbacheh, M.

1987-01-01

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

Space and Terrestrial Power System Integration Optimization Code BRMAPS for Gas Turbine Space Power Plants With Nuclear Reactor Heat Sources

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.

2007-01-01

In view of the difficult times the US and global economies are experiencing today, funds for the development of advanced fission reactors nuclear power systems for space propulsion and planetary surface applications are currently not available. However, according to the Energy Policy Act of 2005 the U.S. needs to invest in developing fission reactor technology for ground based terrestrial power plants. Such plants would make a significant contribution toward drastic reduction of worldwide greenhouse gas emissions and associated global warming. To accomplish this goal the Next Generation Nuclear Plant Project (NGNP) has been established by DOE under the Generation IV Nuclear Systems Initiative. Idaho National Laboratory (INL) was designated as the lead in the development of VHTR (Very High Temperature Reactor) and HTGR (High Temperature Gas Reactor) technology to be integrated with MMW (multi-megawatt) helium gas turbine driven electric power AC generators. However, the advantages of transmitting power in high voltage DC form over large distances are also explored in the seminar lecture series. As an attractive alternate heat source the Liquid Fluoride Reactor (LFR), pioneered at ORNL (Oak Ridge National Laboratory) in the mid 1960's, would offer much higher energy yields than current nuclear plants by using an inherently safe energy conversion scheme based on the Thorium --> U233 fuel cycle and a fission process with a negative temperature coefficient of reactivity. The power plants are to be sized to meet electric power demand during peak periods and also for providing thermal energy for hydrogen (H2) production during "off peak" periods. This approach will both supply electric power by using environmentally clean nuclear heat which does not generate green house gases, and also provide a clean fuel H2 for the future, when, due to increased global demand and the decline in discovering new deposits, our supply of liquid fossil fuels will have been used up. This is

POWER REACTOR

DOEpatents

Zinn, W.H.

1958-07-01

A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.

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

Neutron radiation characteristics of the IVth generation reactor spent fuel

NASA Astrophysics Data System (ADS)

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

2018-03-01

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

Mini-cavity plasma core reactors for dual-mode space nuclear power/propulsion systems. M.S. Thesis

NASA Technical Reports Server (NTRS)

Chow, S.

1976-01-01

A mini-cavity plasma core reactor is investigated for potential use in a dual-mode space power and propulsion system. In the propulsive mode, hydrogen propellant is injected radially inward through the reactor solid regions and into the cavity. The propellant is heated by both solid driver fuel elements surrounding the cavity and uranium plasma before it is exhausted out the nozzle. The propellant only removes a fraction of the driver power, the remainder is transferred by a coolant fluid to a power conversion system, which incorporates a radiator for heat rejection. Neutronic feasibility of dual mode operation and smaller reactor sizes than those previously investigated are shown to be possible. A heat transfer analysis of one such reactor shows that the dual-mode concept is applicable when power generation mode thermal power levels are within the same order of magnitude as direct thrust mode thermal power levels.

Automated determinations of selenium in thermal power plant wastewater by sequential hydride generation and chemiluminescence detection.

PubMed

Ezoe, Kentaro; Ohyama, Seiichi; Hashem, Md Abul; Ohira, Shin-Ichi; Toda, Kei

2016-02-01

After the Fukushima disaster, power generation from nuclear power plants in Japan was completely stopped and old coal-based power plants were re-commissioned to compensate for the decrease in power generation capacity. Although coal is a relatively inexpensive fuel for power generation, it contains high levels (mgkg(-1)) of selenium, which could contaminate the wastewater from thermal power plants. In this work, an automated selenium monitoring system was developed based on sequential hydride generation and chemiluminescence detection. This method could be applied to control of wastewater contamination. In this method, selenium is vaporized as H2Se, which reacts with ozone to produce chemiluminescence. However, interference from arsenic is of concern because the ozone-induced chemiluminescence intensity of H2Se is much lower than that of AsH3. This problem was successfully addressed by vaporizing arsenic and selenium individually in a sequential procedure using a syringe pump equipped with an eight-port selection valve and hot and cold reactors. Oxidative decomposition of organoselenium compounds and pre-reduction of the selenium were performed in the hot reactor, and vapor generation of arsenic and selenium were performed separately in the cold reactor. Sample transfers between the reactors were carried out by a pneumatic air operation by switching with three-way solenoid valves. The detection limit for selenium was 0.008 mg L(-1) and calibration curve was linear up to 1.0 mg L(-1), which provided suitable performance for controlling selenium in wastewater to around the allowable limit (0.1 mg L(-1)). This system consumes few chemicals and is stable for more than a month without any maintenance. Wastewater samples from thermal power plants were collected, and data obtained by the proposed method were compared with those from batchwise water treatment followed by hydride generation-atomic fluorescence spectrometry. Copyright © 2015 Elsevier B.V. All rights

Power flattening on modified CANDLE small long life gas-cooled fast reactor

NASA Astrophysics Data System (ADS)

Monado, Fiber; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Ariani, Menik; Sekimoto, Hiroshi

2014-09-01

Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

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.

77 FR 60039 - Non-Power Reactor License Renewal

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-02

... NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 [NRC-2011-0087] RIN 3150-AI96 Non-Power Reactor... the final regulatory basis for rulemaking to streamline non-power reactor license renewal. This final... Reactor (RTR) License Renewal Process. This contemplated rulemaking also recommends conforming changes to...

77 FR 38742 - Non-Power Reactor License Renewal

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-29

...-0087] RIN 3150-AI96 Non-Power Reactor License Renewal AGENCY: Nuclear Regulatory Commission. ACTION... reactors. This contemplated rulemaking would also make conforming changes to address technical issues in existing non-power reactor regulations. The NRC is seeking input from the public, licensees, certificate...

Estimates of power requirements for a Manned Mars Rover powered by a nuclear reactor

NASA Technical Reports Server (NTRS)

Morley, Nicholas J.; El-Genk, Mohamed S.; Cataldo, Robert; Bloomfield, Harvey

1991-01-01

This paper assesses the power requirement for a Manned Mars Rover vehicle. Auxiliary power needs are fulfilled using a hybrid solar photovoltaic/regenerative fuel cell system, while the primary power needs are meet using an SP-100 type reactor. The primary electric power needs, which include 30-kW(e) net user power, depend on the reactor thermal power and the efficiency of the power conversion system. Results show that an SP-100 type reactor coupled to a Free Piston Stirling Engine yields the lowest total vehicle mass and lowest specific mass for the power system. The second lowest mass was for a SP-100 reactor coupled to a Closed Brayton Cycle using He/Xe as the working fluid. The specific mass of the nuclear reactor power system, including a man-rated radiation shield, ranged from 150-kg/kW(e) to 190-kg/KW(e) and the total mass of the Rover vehicle varied depend upon the cruising speed.

Transmutation of actinides in power reactors.

PubMed

Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

2005-01-01

Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides.

MC 2 -3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

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

Lee, Changho; Yang, Won Sik

This paper presents the methods and performance of the MC2 -3 code, which is a multigroup cross-section generation code for fast reactor analysis, developed to improve the resonance self-shielding and spectrum calculation methods of MC2 -2 and to simplify the current multistep schemes generating region-dependent broad-group cross sections. Using the basic neutron data from ENDF/B data files, MC2 -3 solves the consistent P1 multigroup transport equation to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (2082) or hyperfine (~400more » 000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified temperatures. The pointwise cross sections are directly used in the hyperfine group calculation, whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for a two-dimensional whole-core problem to generate region-dependent broad-group cross sections. Verification tests have been performed using the benchmark problems for various fast critical experiments including Los Alamos National Laboratory critical assemblies; Zero-Power Reactor, Zero-Power Physics Reactor, and Bundesamt für Strahlenschutz experiments; Monju start-up core; and Advanced Burner Test Reactor. Verification and validation results with ENDF/B-VII.0 data indicated that eigenvalues from MC2 -3/DIF3D agreed well with Monte Carlo N-Particle5 MCNP5 or VIM Monte Carlo solutions within 200 pcm and regionwise one-group fluxes were in good agreement with Monte Carlo

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.

78 FR 64028 - Decommissioning of Nuclear Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-25

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0035] Decommissioning of Nuclear Power Reactors AGENCY... Commission (NRC) is issuing Revision 1 of regulatory guide (RG) 1.184 ``Decommissioning of Nuclear Power... the NRC's regulations relating to the decommissioning process for nuclear power reactors. The revision...

The Dynomak: An advanced spheromak reactor system with imposed-dynamo current drive and next-generation nuclear power technologies

NASA Astrophysics Data System (ADS)

Sutherland, D. A.; Jarboe, T. R.; Marklin, G.; Morgan, K. D.; Nelson, B. A.

2013-10-01

A high-beta spheromak reactor system has been designed with an overnight capital cost that is competitive with conventional power sources. This reactor system utilizes recently discovered imposed-dynamo current drive (IDCD) and a molten salt blanket system for first wall cooling, neutron moderation and tritium breeding. Currently available materials and ITER developed cryogenic pumping systems were implemented in this design on the basis of technological feasibility. A tritium breeding ratio of greater than 1.1 has been calculated using a Monte Carlo N-Particle (MCNP5) neutron transport simulation. High-temperature superconducting tapes (YBCO) were used for the equilibrium coil set, substantially reducing the recirculating power fraction when compared to previous spheromak reactor studies. Using zirconium hydride for neutron shielding, a limiting equilibrium coil lifetime of at least thirty full-power years has been achieved. The primary FLiBe loop was coupled to a supercritical carbon dioxide Brayton cycle due to attractive economics and high thermal efficiencies. With these advancements, an electrical output of 1000 MW from a thermal output of 2486 MW was achieved, yielding an overall plant efficiency of approximately 40%. A paper concerning the Dynomak reactor design is currently being reviewed for publication.

Design of megawatt power level heat pipe reactors

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

Mcclure, Patrick Ray; Poston, David Irvin; Dasari, Venkateswara Rao

An important niche for nuclear energy is the need for power at remote locations removed from a reliable electrical grid. Nuclear energy has potential applications at strategic defense locations, theaters of battle, remote communities, and emergency locations. With proper safeguards, a 1 to 10-MWe (megawatt electric) mobile reactor system could provide robust, self-contained, and long-term power in any environment. Heat pipe-cooled fast-spectrum nuclear reactors have been identified as a candidate for these applications. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than “traditional” reactors.more » The goal of this project was to develop a scalable conceptual design for a compact reactor and to identify scaling issues for compact heat pipe cooled reactors in general. Toward this goal two detailed concepts were developed, the first concept with more conventional materials and a power of about 2 MWe and a the second concept with less conventional materials and a power level of about 5 MWe. A series of more qualitative advanced designs were developed (with less detail) that show power levels can be pushed to approximately 30 MWe.« less

Mars, the Moon, and the Ends of the Earth: Autonomy for Small Reactor Power Systems

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

Wood, Richard Thomas

2008-01-01

In recent years, the National Aeronautics and Space Administration (NASA) has been considering deep space missions that utilize a small-reactor power system (SRPS) to provide energy for propulsion and spacecraft power. Additionally, application of SRPS modules as a planetary power source is being investigated to enable a continuous human presence for nonpolar lunar sites and on Mars. A SRPS can supply high-sustained power for space and surface applications that is both reliable and mass efficient. The use of small nuclear reactors for deep space or planetary missions presents some unique challenges regarding the operations and control of the power system.more » Current-generation terrestrial nuclear reactors employ varying degrees of human control and decision-making for operations and benefit from periodic human interaction for maintenance. In contrast, the control system of a SRPS employed for deep space missions must be able to accommodate unattended operations due to communications delays and periods of planetary occlusion while adapting to evolving or degraded conditions with no opportunity for repair or refurbishment. While surface power systems for planetary outposts face less extreme delays and periods of isolation and may benefit from limited maintenance capabilities, considerations such as human safety, resource limitations and usage priorities, and economics favor minimizing direct, continuous human interaction with the SRPS for online, dedicated power system management. Thus, a SRPS control system for space or planetary missions must provide capabilities for operational autonomy. For terrestrial reactors, large-scale power plants remain the preferred near-term option for nuclear power generation. However, the desire to reduce reliance on carbon-emitting power sources in developing countries may lead to increased consideration of SRPS modules for local power generation in remote regions that are characterized by emerging, less established

Gas-core reactor power transient analysis

NASA Technical Reports Server (NTRS)

Kascak, A. F.

1972-01-01

The gas core reactor is a proposed device which features high temperatures. It has applications in high specific impulse space missions, and possibly in low thermal pollution MHD power plants. The nuclear fuel is a ball of uranium plasma radiating thermal photons as opposed to gamma rays. This thermal energy is picked up before it reaches the solid cavity liner by an inflowing seeded propellant stream and convected out through a rocket nozzle. A wall-burnout condition will exist if there is not enough flow of propellant to convect the energy back into the cavity. A reactor must therefore operate with a certain amount of excess propellant flow. Due to the thermal inertia of the flowing propellant, the reactor can undergo power transients in excess of the steady-state wall burnout power for short periods of time. The objective of this study was to determine how long the wall burnout power could be exceeded without burning out the cavity liner. The model used in the heat-transfer calculation was one-dimensional, and thermal radiation was assumed to be a diffusion process.

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.

Formulation and experimental evaluation of closed-form control laws for the rapid maneuvering of reactor neutronic power

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

Bernard, J.A.

1989-09-01

This report describes both the theoretical development and the experimental evaluation of a novel, robust methodology for the time-optimal adjustment of a reactor's neutronic power under conditions of closed-loop digital control. Central to the approach are the MIT-SNL Period-Generated Minimum Time Control Laws' which determine the rate at which reactivity should be changed in order to cause a reactor's neutronic power to conform to a specified trajectory. Using these laws, reactor power can be safely raised by five to seven orders of magnitude in a few seconds. The MIT-SNL laws were developed to facilitate rapid increases of neutronic power onmore » spacecraft reactors operating in an SDI environment. However, these laws are generic and have other applications including the rapid recovery of research and test reactors subsequent to an unanticipated shutdown, power increases following the achievement of criticality on commercial reactors, power adjustments on commercial reactors so as to minimize thermal stress, and automated startups. The work reported here was performed by the Massachusetts Institute of Technology under contract to the Sandia National Laboratories. Support was also provided by the US Department of Energy's Division of University and Industry Programs. The work described in this report is significant in that a novel solution to the problem of time-optimal control of neutronic power was identified, in that a rigorous description of a reactor's dynamics was derived in that the rate of change of reactivity was recognized as the proper control signal, and in that extensive experimental trials were conducted of these newly developed concepts on actual nuclear reactors. 43 refs., 118 figs., 11 tabs.« less

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.

A small, 1400 K, reactor for Brayton space power systems.

NASA Technical Reports Server (NTRS)

Lantz, E.; Mayo, W.

1972-01-01

An investigation was conducted to determine minimum dimensions and minimum weight obtainable in a design for a reactor using uranium-233 nitride or plutonium-239 nitride as fuel. Such a reactor had been considered by Krasner et al. (1971). Present space power status is discussed, together with questions of reactor design and power distribution in the reactor. The characteristics of various reactor types are compared, giving attention also to a zirconium hydride reactor.

Optimization of power-cycle arrangements for Supercritical Water cooled Reactors (SCWRs)

NASA Astrophysics Data System (ADS)

Lizon-A-Lugrin, Laure

The world energy demand is continuously rising due to the increase of both the world population and the standard of life quality. Further, to assure both a healthy world economy as well as adequate social standards, in a relatively short term, new energy-conversion technologies are mandatory. Within this framework, a Generation IV International Forum (GIF) was established by the participation of 10 countries to collaborate for developing nuclear power reactors that will replace the present technology by 2030. The main goals of these nuclear-power reactors are: economic competitiveness, sustainability, safety, reliability and resistance to proliferation. As a member of the GIF, Canada has decided to orient its efforts towards the design of a CANDU-type Super Critical Water-cooled Reactor (SCWR). Such a system must run at a coolant outlet temperature of about 625°C and at a pressure of 25 MPa. It is obvious that at such conditions the overall efficiency of this kind of Nuclear Power Plant (NPP) will compete with actual supercritical water-power boilers. In addition, from a heat-transfer viewpoint, the use of a supercritical fluid allows the limitation imposed by Critical Heat Flux (CHF) conditions, which characterize actual technologies, to be removed. Furthermore, it will be also possible to use direct thermodynamic cycles where the supercritical fluid expands right away in a turbine without the necessity of using intermediate steam generators and/or separators. This work presents several thermodynamic cycles that could be appropriate to run SCWR power plants. Improving both thermal efficiency and mechanical power constitutes a multi-objective optimization problem and requires specific tools. To this aim, an efficient and robust evolutionary algorithm, based on genetic algorithm, is used and coupled to an appropriate power plant thermodynamic simulation model. The results provide numerous combinations to achieve a thermal efficiency higher than 50% with a

Experimental Evaluation of a Water Shield for a Surface Power Reactor

NASA Technical Reports Server (NTRS)

Pearson, J. Boise; Reid, Robert S.

2006-01-01

As part of the Vision for Space Exploration the end of the next decade will bring man back to the surface of the moon. One of the most critical issues for the establishment of human presence on the moon will be the availability of compact power sources. The establishment of man on the moon will require power from greater than 10's of kWt's in follow on years. Nuclear reactors are extremely we11 suited to meet the needs for power generation on the lunar or Martian surface. reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), Boron Carbide, and others. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to remove the potential for radiation streaming paths. The water shield concept relies on predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. MSFC has developed the experience and fac necessary to do this evaluation in the Early Flight Fission - Test Facility (EFF-TF).

Water chemistry of the secondary circuit at a nuclear power station with a VVER power reactor

NASA Astrophysics Data System (ADS)

Tyapkov, V. F.; Erpyleva, S. F.

2017-05-01

Results of implementation of the secondary circuit organic amine water chemistry at Russian nuclear power plant (NPP) with VVER-1000 reactors are presented. The requirements for improving the reliability, safety, and efficiency of NPPs and for prolonging the service life of main equipment items necessitate the implementation of new technologies, such as new water chemistries. Data are analyzed on the chemical control of power unit coolant for quality after the changeover to operation with the feed of higher amines, such as morpholine and ethanolamine. Power units having equipment containing copper alloy components were converted from the all-volatile water chemistry to the ethanolamine or morpholine water chemistry with no increase in pH of the steam generator feedwater. This enables the iron content in the steam generator feedwater to be decreased from 6-12 to 2.0-2.5 μg/dm3. It is demonstrated that pH of high-temperature water is among the basic factors controlling erosion and corrosion wear of the piping and the ingress of corrosion products into NPP steam generators. For NPP power units having equipment whose construction material does not include copper alloys, the water chemistries with elevated pH of the secondary coolant are adopted. Stable dosing of correction chemicals at these power units maintains pH25 of 9.5 to 9.7 in the steam generator feedwater with a maximum iron content of 2 μg/dm3 in the steam generator feedwater.

Discharge Characteristics of Series Surface/Packed-Bed Discharge Reactor Diven by Bipolar Pulsed Power

NASA Astrophysics Data System (ADS)

Hu, Jian; Jiang, Nan; Li, Jie; Shang, Kefeng; Lu, Na; Wu, Yan; Mizuno, Akira

2016-03-01

The discharge characteristics of the series surface/packed-bed discharge (SSPBD) reactor driven by bipolar pulse power were systemically investigated in this study. In order to evaluate the advantages of the SSPBD reactor, it was compared with traditional surface discharge (SD) reactor and packed-bed discharge (PBD) reactor in terms of the discharge voltage, discharge current, and ozone formation. The SSPBD reactor exhibited a faster rising time and lower tail voltage than the SD and PBD reactors. The distribution of the active species generated in different discharge regions of the SSPBD reactor was analyzed by optical emission spectra and ozone analysis. It was found that the packed-bed discharge region (3.5 mg/L), rather than the surface discharge region (1.3 mg/L) in the SSPBD reactor played a more important role in ozone generation. The optical emission spectroscopy analysis indicated that more intense peaks of the active species (e.g. N2 and OI) in the optical emission spectra were observed in the packed-bed region. supported by National Natural Science Foundation of China (No. 51177007), the Joint Funds of National Natural Science Foundation of China (No. U1462105), and Dalian University of Technology Fundamental Research Fund of China (No. DUT15RC(3)030)

Static and dynamic high power, space nuclear electric generating systems

NASA Technical Reports Server (NTRS)

Wetch, J. R.; Begg, L. L.; Koester, J. K.

1985-01-01

Space nuclear electric generating systems concepts have been assessed for their potential in satisfying future spacecraft high power (several megawatt) requirements. Conceptual designs have been prepared for reactor power systems using the most promising static (thermionic) and the most promising dynamic conversion processes. Component and system layouts, along with system mass and envelope requirements have been made. Key development problems have been identified and the impact of the conversion process selection upon thermal management and upon system and vehicle configuration is addressed.

Power monitoring in space nuclear reactors using silicon carbide radiation detectors

NASA Technical Reports Server (NTRS)

Ruddy, Frank H.; Patel, Jagdish U.; Williams, John G.

2005-01-01

Space reactor power monitors based on silicon carbide (SiC) semiconductor neutron detectors are proposed. Detection of fast leakage neutrons using SiC detectors in ex-core locations could be used to determine reactor power: Neutron fluxes, gamma-ray dose rates and ambient temperatures have been calculated as a function of distance from the reactor core, and the feasibility of power monitoring with SiC detectors has been evaluated at several ex-core locations. Arrays of SiC diodes can be configured to provide the required count rates to monitor reactor power from startup to full power Due to their resistance to temperature and the effects of neutron and gamma-ray exposure, SiC detectors can be expected to provide power monitoring information for the fill mission of a space reactor.

Thermionic reactor power conditioner design for nuclear electric propulsion.

NASA Technical Reports Server (NTRS)

Jacobsen, A. S.; Tasca, D. M.

1971-01-01

Consideration of the effects of various thermionic reactor parameters and requirements upon spacecraft power conditioning design. A basic spacecraft is defined using nuclear electric propulsion, requiring approximately 120 kWe. The interrelationships of reactor operating characteristics and power conditioning requirements are discussed and evaluated, and the effects on power conditioner design and performance are presented.

Aging management program of the reactor building concrete at Point Lepreau Generating Station

NASA Astrophysics Data System (ADS)

Aldea, C.-M.; Shenton, B.; Demerchant, M. M.; Gendron, T.

2011-04-01

In order for New Brunswick Power Nuclear (NBPN) to control the risks of degradation of the concrete reactor building at the Point Lepreau Generating Station (PLGS) the development of an aging management plan (AMP) was initiated. The intention of this plan was to determine the requirements for specific structural components of concrete of the reactor building that require regular inspection and maintenance to ensure the safe and reliable operation of the plant. The document is currently in draft form and presents an integrated methodology for the application of an AMP for the concrete of the reactor building. The current AMP addresses the reactor building structure and various components, such as joint sealant and liners that are integral to the structure. It does not include internal components housed within the structure. This paper provides background information regarding the document developed and the strategy developed to manage potential degradation of the concrete of the reactor building, as well as specific programs and preventive and corrective maintenance activities initiated.

Station Blackout Analysis of HTGR-Type Experimental Power Reactor

NASA Astrophysics Data System (ADS)

Syarip; Zuhdi, Aliq; Falah, Sabilul

2018-01-01

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

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.

STEAM GENERATOR FOR NUCLEAR REACTOR

DOEpatents

Kinyon, B.W.; Whitman, G.D.

1963-07-16

The steam generator described for use in reactor powergenerating systems employs a series of concentric tubes providing annular passage of steam and water and includes a unique arrangement for separating the steam from the water. (AEC)

System and method for generating steady state confining current for a toroidal plasma fusion reactor

DOEpatents

Bers, Abraham

1981-01-01

A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave RF energy is injected into said plasma to estalish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected RF energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected RF energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range .DELTA.. The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width .DELTA. in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated inthe plasma.

PR-EDB: Power Reactor Embrittlement Database - Version 3

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

Wang, Jy-An John; Subramani, Ranjit

2008-03-01

The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industrymore » standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access

Consumption of the electric power inside silent discharge reactors

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

Yehia, Ashraf, E-mail: yehia30161@yahoo.com

An experimental study was made in this paper to investigate the relation between the places of the dielectric barriers, which cover the surfaces of the electrodes in the coaxial cylindrical reactors, and the rate of change of the electric power that is consumed in forming silent discharges. Therefore, silent discharges have been formed inside three coaxial cylindrical reactors. The dielectric barriers in these reactors were pasted on both the internal surface of the outer electrode in the first reactor and the external surface of the inner electrode in the second reactor as well as the surfaces of the two electrodesmore » in the third reactor. The reactor under study has been fed by atmospheric air that flowed inside it with a constant rate at normal temperature and pressure, in parallel with the application of a sinusoidal ac voltage between the electrodes of the reactor. The electric power consumed in forming the silent discharges inside the three reactors was measured as a function of the ac peak voltage. The validity of the experimental results was investigated by applying Manley's equation on the same discharge conditions. The results have shown that the rate of consumption of the electric power relative to the ac peak voltage per unit width of the discharge gap improves by a ratio of either 26.8% or 80% or 128% depending on the places of the dielectric barriers that cover the surfaces of the electrodes inside the three reactors.« less

The effectiveness of using the combined-cycle technology in a nuclear power plant unit equipped with an SVBR-100 reactor

NASA Astrophysics Data System (ADS)

Kasilov, V. F.; Dudolin, A. A.; Gospodchenkov, I. V.

2015-05-01

The design of a modular SVBR-100 reactor with a lead-bismuth alloy liquid-metal coolant is described. The basic thermal circuit of a power unit built around the SVBR-100 reactor is presented together with the results of its calculation. The gross electrical efficiency of the turbine unit driven by saturated steam at a pressure of 6.7 MPa is estimated at η{el/gr} = 35.5%. Ways for improving the efficiency of this power unit and increasing its power output by applying gas-turbine and combined-cycle technologies are considered. With implementing a combined-cycle power-generating system comprising two GE-6101FA gas-turbine units with a total capacity of 140 MW, it becomes possible to obtain the efficiency of the combined-cycle plant equipped with the SVBR-100 reactor η{el/gr} = 45.39% and its electrical power output equal to 328 MW. The heat-recovery boiler used as part of this power installation generates superheated steam with a temperature of 560°C, due to which there is no need to use a moisture separator/steam reheater in the turbine unit thermal circuit.

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

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

Kelley, D.B.

1995-12-31

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

From the first nuclear power plant to fourth-generation nuclear power installations [on the 60th anniversary of the World's First nuclear power plant

NASA Astrophysics Data System (ADS)

Rachkov, V. I.; Kalyakin, S. G.; Kukharchuk, O. F.; Orlov, Yu. I.; Sorokin, A. P.

2014-05-01

Successful commissioning in the 1954 of the World's First nuclear power plant constructed at the Institute for Physics and Power Engineering (IPPE) in Obninsk signaled a turn from military programs to peaceful utilization of atomic energy. Up to the decommissioning of this plant, the AM reactor served as one of the main reactor bases on which neutron-physical investigations and investigations in solid state physics were carried out, fuel rods and electricity generating channels were tested, and isotope products were bred. The plant served as a center for training Soviet and foreign specialists on nuclear power plants, the personnel of the Lenin nuclear-powered icebreaker, and others. The IPPE development history is linked with the names of I.V. Kurchatov, A.I. Leipunskii, D.I. Blokhintsev, A.P. Aleksandrov, and E.P. Slavskii. More than 120 projects of various nuclear power installations were developed under the scientific leadership of the IPPE for submarine, terrestrial, and space applications, including two water-cooled power units at the Beloyarsk NPP in Ural, the Bilibino nuclear cogeneration station in Chukotka, crawler-mounted transportable TES-3 power station, the BN-350 reactor in Kazakhstan, and the BN-600 power unit at the Beloyarsk NPP. Owing to efforts taken on implementing the program for developing fast-neutron reactors, Russia occupied leading positions around the world in this field. All this time, IPPE specialists worked on elaborating the principles of energy supertechnologies of the 21st century. New large experimental installations have been put in operation, including the nuclear-laser setup B, the EGP-15 accelerator, the large physical setup BFS, the high-pressure setup SVD-2; scientific, engineering, and technological schools have been established in the field of high- and intermediate-energy nuclear physics, electrostatic accelerators of multicharge ions, plasma processes in thermionic converters and nuclear-pumped lasers, physics of compact

Assessment of nuclear reactor concepts for low power space applications

NASA Technical Reports Server (NTRS)

Klein, Andrew C.; Gedeon, Stephen R.; Morey, Dennis C.

1988-01-01

The results of a preliminary small reactor concepts feasibility and safety evaluation designed to provide a first order validation of the nuclear feasibility and safety of six small reactor concepts are given. These small reactor concepts have potential space applications for missions in the 1 to 20 kWe power output range. It was concluded that low power concepts are available from the U.S. nuclear industry that have the potential for meeting both the operational and launch safety space mission requirements. However, each design has its uncertainties, and further work is required. The reactor concepts must be mated to a power conversion technology that can offer safe and reliable operation.

Analysis of UF6 breeder reactor power plants

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.

1976-01-01

Gaseous UF6 fueled breeder reactor design and technical applications of such concepts are summarized. Special attention was given to application in nuclear power plants and to reactor efficiency and safety factors.

10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear... requirements for immediate notification of the NRC by licensed operating nuclear power reactors are contained...

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.

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors

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

M. L. Grossbeck J-P.A. Renier Tim Bigelow

2003-09-30

Burnable poisons are used in nuclear reactors to produce a more level distribution of power in the reactor core and to reduce to necessity for a large control system. An ideal burnable poison would burn at the same rate as the fuel. In this study, separation of neutron-absorbing isotopes was investigated in order to eliminate isotopes that remain as absorbers at the end of fuel life, thus reducing useful fuel life. The isotopes Gd-157, Dy-164, and Er-167 were found to have desirable properties. These isotopes were separated from naturally occurring elements by means of plasma separation to evaluate feasibility andmore » cost. It was found that pure Gd-157 could save approximately $6 million at the end of four years. However, the cost of separation, using the existing facility, made separation cost- ineffective. Using a magnet with three times the field strength is expected to reduce the cost by a factor of ten, making isotopically separated burnable poisons a favorable method of increasing fuel life in commercial reactors, in particular Generation-IV reactors. The project also investigated various burnable poison configurations, and studied incorporation of metallic burnable poisons into fuel cladding.« less

COST FUNCTION STUDIES FOR POWER REACTORS

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

Heestand, J.; Wos, L.T.

1961-11-01

A function to evaluate the cost of electricity produced by a nuclear power reactor was developed. The basic equation, revenue = capital charges + profit + operating expenses, was expanded in terms of various cost parameters to enable analysis of multiregion nuclear reactors with uranium and/or plutonium for fuel. A corresponding IBM 704 computer program, which will compute either the price of electricity or the value of plutonium, is presented in detail. (auth)

Thermoelectric power generator for variable thermal power source

DOEpatents

Bell, Lon E; Crane, Douglas Todd

2015-04-14

Traditional power generation systems using thermoelectric power generators are designed to operate most efficiently for a single operating condition. The present invention provides a power generation system in which the characteristics of the thermoelectrics, the flow of the thermal power, and the operational characteristics of the power generator are monitored and controlled such that higher operation efficiencies and/or higher output powers can be maintained with variably thermal power input. Such a system is particularly beneficial in variable thermal power source systems, such as recovering power from the waste heat generated in the exhaust of combustion engines.

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

DynMo: Dynamic Simulation Model for Space Reactor Power Systems

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed; Tournier, Jean-Michel

2005-02-01

A Dynamic simulation Model (DynMo) for space reactor power systems is developed using the SIMULINK® platform. DynMo is modular and could be applied to power systems with different types of reactors, energy conversion, and heat pipe radiators. This paper presents a general description of DynMo-TE for a space power system powered by a Sectored Compact Reactor (SCoRe) and that employs off-the-shelf SiGe thermoelectric converters. SCoRe is liquid metal cooled and designed for avoidance of a single point failure. The reactor core is divided into six equal sectors that are neutronically, but not thermal-hydraulically, coupled. To avoid a single point failure in the power system, each reactor sector has its own primary and secondary loops, and each loop is equipped with an electromagnetic (EM) pump. A Power Conversion assembly (PCA) and a Thermoelectric Conversion Assembly (TCA) of the primary and secondary EM pumps thermally couple each pair of a primary and a secondary loop. The secondary loop transports the heat rejected by the PCA and the pumps TCA to a rubidium heat pipes radiator panel. The primary loops transport the thermal power from the reactor sector to the PCAs for supplying a total of 145-152 kWe to the load at 441-452 VDC, depending on the selections of the primary and secondary liquid metal coolants. The primary and secondary coolant combinations investigated are lithium (Li)/Li, Li/sodium (Na), Na-Na, Li/NaK-78 and Na/NaK-78, for which the reactor exit temperature is kept below 1250 K. The results of a startup transient of the system from an initial temperature of 500 K are compared and discussed.

10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear power reactor licensee licensed under §§ 50.21(b) or 50.22 holding an operating license under this part...

10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear power reactor licensee licensed under §§ 50.21(b) or 50.22 holding an operating license under this part...

Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

NASA Astrophysics Data System (ADS)

Verma, V.; Barbot, L.; Filliatre, P.; Hellesen, C.; Jammes, C.; Svärd, S. Jacobsson

2017-07-01

Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment.

System and method for generating steady state confining current for a toroidal plasma fusion reactor

DOEpatents

Fisch, Nathaniel J.

1981-01-01

A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave RF energy is injected into said plasma to establish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected RF energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected RF energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range .DELTA.. The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width .DELTA. in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated in the plasma.

A Compact Torus Fusion Reactor Utilizing a Continuously Generated Strings of CT's. The CT String Reactor, CTSR.

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

Hartman, C W; Reisman, D B; McLean, H S

2007-05-30

A fusion reactor is described in which a moving string of mutually repelling compact toruses (alternating helicity, unidirectional Btheta) is generated by repetitive injection using a magnetized coaxial gun driven by continuous gun current with alternating poloidal field. An injected CT relaxes to a minimum magnetic energy equilibrium, moves into a compression cone, and enters a conducting cylinder where the plasma is heated to fusion-producing temperature. The CT then passes into a blanketed region where fusion energy is produced and, on emergence from the fusion region, the CT undergoes controlled expansion in an exit cone where an alternating poloidal fieldmore » opens the flux surfaces to directly recover the CT magnetic energy as current which is returned to the formation gun. The CT String Reactor (CTSTR) reactor satisfies all the necessary MHD stability requirements and is based on extrapolation of experimentally achieved formation, stability, and plasma confinement. It is supported by extensive 2D, MHD calculations. CTSTR employs minimal external fields supplied by normal conductors, and can produce high fusion power density with uniform wall loading. The geometric simplicity of CTSTR acts to minimize initial and maintenance costs, including periodic replacement of the reactor first wall.« less

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.

77 FR 8902 - Draft Regulatory Guide: Issuance, Availability Decommissioning of Nuclear Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-15

... Decommissioning of Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide... draft regulatory guide (DG) DG-1271 ``Decommissioning of Nuclear Power Reactors.'' This guide describes... Regulatory Guide 1.184, ``Decommissioning of Nuclear Power Reactors,'' dated July 2000. This proposed...

Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

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

Gougar, Hans David

2015-10-01

The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each ofmore » the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.« less

78 FR 73898 - Operator Licensing Examination Standards for Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-09

... Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft NUREG; request for comment. SUMMARY: The U.S..., Revision 10, ``Operator Licensing Examination Standards for Power Reactors.'' DATES: Submit comments [email protected] . Both of the Office of New Reactors; or Timothy Kolb, Office of Nuclear Reactor Regulation, U...

Future trends in power generation cost by power resource

NASA Astrophysics Data System (ADS)

1992-08-01

The Japan Energy Economy Research Institute has been evaluating power generation cost by each power resource every year focusing on nuclear power generation. The Institute is surveying the cost evaluations by power resources in France, Britain and the U.S.A., the nuclear generation advanced nations. The OECD is making power generation cost estimation using a hypothesis which uniforms basically the conditions varying in different member countries. In model power generation cost calculations conducted by the Ministry of International Trade and Industry of Japan, nuclear power generation is the most economical system in any fiscal year. According to recent calculations performed by the Japan Energy Economy Research Institute, the situation is such that it is difficult to distinguish the economical one from others among the power generation systems in terms of generation costs except for thermal power generation. Economic evaluations are given on estimated power generation costs based on construction costs for nuclear and thermal power plants, nuclear fuel cycling cost, and fuel cost data on petroleum, LNG and coal. With regard to the future trends, scenario analyses are made on generation costs, that assume fluctuations in fuel prices and construction costs, the important factors to give economic influence on power generation.

The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues

NASA Astrophysics Data System (ADS)

Dautray, Robert

2011-06-01

, etc., radioprotection measures and treatment for the "transuranic" elements. For a long period of time, France was in the forefront of nuclear breeder power generation science, technological research and also in the knowledge base related to breeder reactors. It is in the country's interest to pursue these efforts and this could per se constitute one of the national priorities. Nous sommes naturellement bien conscients de l'énorme problème qui se pose au Japon actuellement comme suite au tremblement de terre et au tsunami de mars 2011 et leurs conséquences, notamment sur des installations électronucléaires. Le texte que nous présentons concerne des conditions totalement générales, indépendantes des problèmes spécifiques de sûreté qu'il faudra, de toute façon, traiter dans le cadre d'un développement éventuel de l'énergie nucléaire.We are aware, of course, of the huge problem that Japan has to deal with the aftermath of the quake and tsunami of March 2011 and their consequences on electronuclear power plants. The text that we present here concerns general physical topics independent of the specific safety problems, general physical topics which will have to be solved in the case of a contingent development of electronuclear power plants.

Hydrogen production by reforming of liquid hydrocarbons in a membrane reactor for portable power generation-Experimental studies

NASA Astrophysics Data System (ADS)

Damle, Ashok S.

One of the most promising technologies for lightweight, compact, portable power generation is proton exchange membrane (PEM) fuel cells. PEM fuel cells, however, require a source of pure hydrogen. Steam reforming of hydrocarbons in an integrated membrane reactor has potential to provide pure hydrogen in a compact system. Continuous separation of product hydrogen from the reforming gas mixture is expected to increase the yield of hydrogen significantly as predicted by model simulations. In the laboratory-scale experimental studies reported here steam reforming of liquid hydrocarbon fuels, butane, methanol and Clearlite ® was conducted to produce pure hydrogen in a single step membrane reformer using commercially available Pd-Ag foil membranes and reforming/WGS catalysts. All of the experimental results demonstrated increase in hydrocarbon conversion due to hydrogen separation when compared with the hydrocarbon conversion without any hydrogen separation. Increase in hydrogen recovery was also shown to result in corresponding increase in hydrocarbon conversion in these studies demonstrating the basic concept. The experiments also provided insight into the effect of individual variables such as pressure, temperature, gas space velocity, and steam to carbon ratio. Steam reforming of butane was found to be limited by reaction kinetics for the experimental conditions used: catalysts used, average gas space velocity, and the reactor characteristics of surface area to volume ratio. Steam reforming of methanol in the presence of only WGS catalyst on the other hand indicated that the membrane reactor performance was limited by membrane permeation, especially at lower temperatures and lower feed pressures due to slower reconstitution of CO and H 2 into methane thus maintaining high hydrogen partial pressures in the reacting gas mixture. The limited amount of data collected with steam reforming of Clearlite ® indicated very good match between theoretical predictions and

The Effect of Flow Swirling on the Safety and Reliability of Nuclear Power Installations of New Generation

NASA Astrophysics Data System (ADS)

Mitrofanova, O. V.; Ivlev, O. A.; Urtenov, D. S.

2018-03-01

Hydrodynamics and heat exchange in the elements of thermal hydraulic tracts of ship nuclear reactors of the new generation were numerically simulated in this work. Parts of the coolant circuit in the collector and piping systems with geometries that may lead to generation of stable large-scale vortexes, causing a wide range of acoustic oscillations of the coolant, were selected as modeling objects. The purpose of the research is to develop principles of physical and mathematical modeling for scientific substantiation of optimal layout solutions that ensure enhanced operational life of icebreaker’s nuclear power installations of new generation with reactors of integral type.

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.

Core follow calculation with the nTRACER numerical reactor and verification using power reactor measurement data

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

Jung, Y. S.; Joo, H. G.; Yoon, J. I.

The nTRACER direct whole core transport code employing the planar MOC solution based 3-D calculation method, the subgroup method for resonance treatment, the Krylov matrix exponential method for depletion, and a subchannel thermal/hydraulic calculation solver was developed for practical high-fidelity simulation of power reactors. Its accuracy and performance is verified by comparing with the measurement data obtained for three pressurized water reactor cores. It is demonstrated that accurate and detailed multi-physic simulation of power reactors is practically realizable without any prior calculations or adjustments. (authors)

Xenon-induced power oscillations in a generic small modular reactor

NASA Astrophysics Data System (ADS)

Kitcher, Evans Damenortey

As world demand for energy continues to grow at unprecedented rates, the world energy portfolio of the future will inevitably include a nuclear energy contribution. It has been suggested that the Small Modular Reactor (SMR) could play a significant role in the spread of civilian nuclear technology to nations previously without nuclear energy. As part of the design process, the SMR design must be assessed for the threat to operations posed by xenon-induced power oscillations. In this research, a generic SMR design was analyzed with respect to just such a threat. In order to do so, a multi-physics coupling routine was developed with MCNP/MCNPX as the neutronics solver. Thermal hydraulic assessments were performed using a single channel analysis tool developed in Python. Fuel and coolant temperature profiles were implemented in the form of temperature dependent fuel cross sections generated using the SIGACE code and reactor core coolant densities. The Power Axial Offset (PAO) and Xenon Axial Offset (XAO) parameters were chosen to quantify any oscillatory behavior observed. The methodology was benchmarked against results from literature of startup tests performed at a four-loop PWR in Korea. The developed benchmark model replicated the pertinent features of the reactor within ten percent of the literature values. The results of the benchmark demonstrated that the developed methodology captured the desired phenomena accurately. Subsequently, a high fidelity SMR core model was developed and assessed. Results of the analysis revealed an inherently stable SMR design at beginning of core life and end of core life under full-power and half-power conditions. The effect of axial discretization, stochastic noise and convergence of the Monte Carlo tallies in the calculations of the PAO and XAO parameters was investigated. All were found to be quite small and the inherently stable nature of the core design with respect to xenon-induced power oscillations was confirmed. Finally, a

Fuel supply of nuclear power industry with the introduction of fast reactors

NASA Astrophysics Data System (ADS)

Muraviev, E. V.

2014-12-01

The results of studies conducted for the validation of the updated development strategy for nuclear power industry in Russia in the 21st century are presented. Scenarios with different options for the reprocessing of spent fuel of thermal reactors and large-scale growth of nuclear power industry based on fast reactors of inherent safety with a breeding ratio of ˜1 in a closed nuclear fuel cycle are considered. The possibility of enhanced fuel breeding in fast reactors is also taken into account in the analysis. The potential to establish a large-scale nuclear power industry that covers 100% of the increase in electric power requirements in Russia is demonstrated. This power industry may be built by the end of the century through the introduction of fast reactors (replacing thermal ones) with a gross uranium consumption of up to ˜1 million t and the termination of uranium mining even if the reprocessing of spent fuel of thermal reactors is stopped or suffers a long-term delay.

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.

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.

Power Smoothing and MPPT for Grid-connected Wind Power Generation with Doubly Fed Induction Generator

NASA Astrophysics Data System (ADS)

Kai, Takaaki; Tanaka, Yuji; Kaneda, Hirotoshi; Kobayashi, Daichi; Tanaka, Akio

Recently, doubly fed induction generator (DFIG) and synchronous generator are mostly applied for wind power generation, and variable speed control and power factor control are executed for high efficiently for wind energy capture and high quality for power system voltage. In variable speed control, a wind speed or a generator speed is used for maximum power point tracking. However, performances of a wind generation power fluctuation due to wind speed variation have not yet investigated for those controls. The authors discuss power smoothing by those controls for the DFIG inter-connected to 6.6kV distribution line. The performances are verified using power system simulation software PSCAD/EMTDC for actual wind speed data and are examined from an approximate equation of wind generation power fluctuation for wind speed variation.

Electrical power generating system

NASA Technical Reports Server (NTRS)

Nola, F. J. (Inventor)

1983-01-01

A power generating system for adjusting coupling an induction motor, as a generator, to an A.C. power line wherein the motor and power line are connected through a triac is described. The triac is regulated to normally turn on at a relatively late point in each half cycle of its operation, whereby at less than operating speed, and thus when the induction motor functions as a motor rather than as a generator, power consumption from the line is substantially reduced.

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

Generation III reactors safety requirements and the design solutions

NASA Astrophysics Data System (ADS)

Felten, P.

2009-03-01

Nuclear energy's public acceptance, and hence its development, depends on its safety. As a reactor designer, we will first briefly remind the basic safety principles of nuclear reactors' design. We will then show how the industry, and in particular Areva with its EPR, made design evolution in the wake of the Three Miles Island accident in 1979. In particular, for this new generation of reactors, severe accidents are taken into account beyond the standard design basis accidents. Today, Areva's EPR meets all so-called "generation III" safety requirements and was licensed by several nuclear safety authorities in the world. Many innovative solutions are integrated in the EPR, some of which will be introduced here.

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.

10 CFR 50.44 - Combustible gas control for nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere with...

10 CFR 50.44 - Combustible gas control for nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere with...

Power Peaking Effect of OTTO Fuel Scheme Pebble Bed Reactor

NASA Astrophysics Data System (ADS)

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

2018-02-01

Pebble Bed Reactor (PBR) type of Hight Temperature Gas-cooled Reactor (HTGR) is a very interesting nuclear reactor design to fulfill the growing electricity and heat demand with a superior passive safety features. Effort to introduce the PBR design to the market can be strengthen by simplifying its system with the Once-through-then-out (OTTO) cycle PBR in which the pebble fuel only pass the core once. Important challenge in the OTTO fuel scheme is the power peaking effect which limit the maximum nominal power or burnup of the design. Parametric survey is perform in this study to investigate the contribution of different design parameters to power peaking effect of OTTO cycle PBR. PEBBED code is utilized in this study to perform the equilibrium PBR core analysis for different design parameter and fuel scheme. The parameters include its core diameter, height-per-diameter (H/D), power density, and core nominal power. Results of this study show that diameter and H/D effectsare stronger compare to the power density and nominal core power. Results of this study might become an importance guidance for design optimization of OTTO fuel scheme PBR.

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.

Locating hot and cold-legs in a nuclear powered steam generation system

DOEpatents

Ekeroth, D.E.; Corletti, M.M.

1993-11-16

A nuclear reactor steam generator includes a reactor vessel for heating water and a steam generator with a pump casing at the lowest point on the steam generator. A cold-leg pipe extends horizontally between the steam generator and the reactor vessel to return water from the steam generator to the reactor vessel. The bottom of the cold-leg pipe is at a first height above the bottom of the reactor vessel. A hot-leg pipe with one end connected to the steam generator and a second end connected to the reactor vessel has a first pipe region extending downwardly from the steam generator to a location between the steam generator and the reactor vessel at which a bottom of the hot-leg pipe is at a second height above the bottom of the reactor vessel. A second region extends from that location in a horizontal direction at the second height to the point at which the hot-leg pipe connects to the reactor vessel. A pump is attached to the casing at a location below the first and second heights and returns water from the steam generator to the reactor vessel over the cold-leg. The first height is greater than the second height and the bottom of the steam generator is at a height above the bottom of the reactor vessel that is greater than the first and second heights. A residual heat recovery pump is below the hot-leg and has an inlet line from the hot-leg that slopes down continuously to the pump inlet. 2 figures.

Locating hot and cold-legs in a nuclear powered steam generation system

DOEpatents

Ekeroth, Douglas E.; Corletti, Michael M.

1993-01-01

A nuclear reactor steam generator includes a reactor vessel for heating water and a steam generator with a pump casing at the lowest point on the steam generator. A cold-leg pipe extends horizontally between the steam generator and the reactor vessel to return water from the steam generator to the reactor vessel. The bottom of the cold-leg pipe is at a first height above the bottom of the reactor vessel. A hot-leg pipe with one end connected to the steam generator and a second end connected to the reactor vessel has a first pipe region extending downwardly from the steam generator to a location between the steam generator and the reactor vessel at which a bottom of the hot-leg pipe is at a second height above the bottom of the reactor vessel. A second region extends from that location in a horizontal direction at the second height to the point at which the hot-leg pipe connects to the reactor vessel. A pump is attached to the casing at a location below the first and second heights and returns water from the steam generator to the reactor vessel over the cold-leg. The first height is greater than the second height and the bottom of the steam generator is at a height above the bottom of the reactor vessel that is greater than the first and second heights. A residual heat recovery pump is below the hot-leg and has an inlet line from the hot-leg that slopes down continuously to the pump inlet.

A facility for testing 10 to 100-kWe space power reactors

NASA Astrophysics Data System (ADS)

Carlson, William F.; Bitten, Ernest J.

1993-01-01

This paper describes an existing facility that could be used in a cost-effective manner to test space power reactors in the 10 to 100-kWe range before launch. The facility has been designed to conduct full power tests of 100-kWe SP-100 reactor systems and already has the structural features that would be required for lower power testing. The paper describes a reasonable scenario starting with the acceptance at the test site of the unfueled reactor assembly and the separately shipped nuclear fuel. After fueling the reactor and installing it in the facility, cold critical tests are performed, and the reactor is then shipped to the launch site. The availability of this facility represents a cost-effective means of performing the required prelaunch test program.

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.

Power Control of New Wind Power Generation System with Induction Generator Excited by Voltage Source Converter

NASA Astrophysics Data System (ADS)

Morizane, Toshimitsu; Kimura, Noriyuki; Taniguchi, Katsunori

This paper investigates advantages of new combination of the induction generator for wind power and the power electronic equipment. Induction generator is popularly used for the wind power generation. The disadvantage of it is impossible to generate power at the lower rotor speed than the synchronous speed. To compensate this disadvantage, expensive synchronous generator with the permanent magnets is sometimes used. In proposed scheme, the diode rectifier is used to convert the real power from the induction generator to the intermediate dc voltage, while only the reactive power necessary to excite the induction generator is supplied from the voltage source converter (VSC). This means that the rating of the expensive VSC is minimized and total cost of the wind power generation system is decreased compared to the system with synchronous generator. Simulation study to investigate the control strategy of proposed system is performed. The results show the reduction of the VSC rating is prospective.

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power reactor...

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power reactor...

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power reactor...

FALCON nuclear-reactor-pumped laser program and wireless power transmission

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

Lipinski, R.J.; Pickard, P.S.

1992-12-31

FALCON is a high-power, reactor-pumped laser concept. The major strengths of a reactor-pumped laser are (1) simple, modular construction, (2) long-duration, closed-cycle capability, (3) self-contained power, (4) compact size, and (5) a variety of wavelengths (from visible to infrared). Reactor-pumped lasing has been demonstrated experimentally in various mixtures of xenon, argon, neon, and helium at wavelengths of 585, 703, 725, 1271, 1733, 1792, 2032, 2630, 2650, and 3370 nm with intrinsic efficiency as high as 2.5%. Powers up to 300 W for 2 ms have been demonstrated. Projected beam quality for FALCON is good enough that frequency doubling at reasonablemore » efficiencies could be expected to yield wavelengths at 353, 363, 636, 867, 896, 1016, 1315, 1325, and 1685 nm. Appropriate missions for FALCON are described and include power beaming to satellites, the moon, and unmanned surveillance planes; lunar mapping; space debris removal; and laser propulsion.« less

FALCON nuclear-reactor-pumped laser program and wireless power transmission

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

Lipinski, R.J.; Pickard, P.S.

1992-01-01

FALCON is a high-power, reactor-pumped laser concept. The major strengths of a reactor-pumped laser are (1) simple, modular construction, (2) long-duration, closed-cycle capability, (3) self-contained power, (4) compact size, and (5) a variety of wavelengths (from visible to infrared). Reactor-pumped lasing has been demonstrated experimentally in various mixtures of xenon, argon, neon, and helium at wavelengths of 585, 703, 725, 1271, 1733, 1792, 2032, 2630, 2650, and 3370 nm with intrinsic efficiency as high as 2.5%. Powers up to 300 W for 2 ms have been demonstrated. Projected beam quality for FALCON is good enough that frequency doubling at reasonablemore » efficiencies could be expected to yield wavelengths at 353, 363, 636, 867, 896, 1016, 1315, 1325, and 1685 nm. Appropriate missions for FALCON are described and include power beaming to satellites, the moon, and unmanned surveillance planes; lunar mapping; space debris removal; and laser propulsion.« less

Preliminary Evaluation of Convective Heat Transfer in a Water Shield for a Surface Power Reactor

NASA Technical Reports Server (NTRS)

Pearson J. Boise; Reid, Robert S.

2007-01-01

As part of the Vision for Space Exploration, the end of the next decade will bring man back to the surface of the moon. A crucial issue for the establishment of human presence on the moon will be the availability of compact power sources. This presence could require greater than 10's of kWt's in follow on years. Nuclear reactors are well suited to meet the needs for power generation on the lunar or Martian surface. Radiation shielding is a key component of any surface power reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), and boron carbide. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to fix the location of any vapor that could form radiation streaming paths. The water shield concept relies on the predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. NASA Marshall Space Flight Center has developed the experience and facilities necessary to do this evaluation in its Early Flight Fission - Test Facility (EFF-TF).

The WPI reactor-readying for the next generation

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

Bobek, L.M.

1993-01-01

Built in 1959, the 10-kW open-pool nuclear training reactor at Worcester Polytechnic Institute (WPI) was one of the first such facilities in the nation located on a university campus. Since then, the reactor and its related facilities have been used to train two generations of nuclear engineers and scientists for the nuclear industry. With the use of nuclear technology playing an increasing role in many segments of the economy, WPI with its nuclear reactor facility is committed to continuing its mission of training future nuclear engineers and scientists. The WPI reactor includes a 6-in. beam port, graphite thermal column, andmore » in-core sample facility. The reactor, housed in an open 8000-gal tank of water, is designed so that the core is readily accessible. Both the control console and the peripheral counting equipment used for student projects and laboratory exercises are located in the reactor room. This arrangement provides convenience and flexibility in using the reactor for foil activations in neutron flux measurements, diffusion measurements, radioactive decay measurements, and the neutron activation of samples for analysis. In 1988, the reactor was successfully converted to low-enriched uranium fuel.« less

Next generation fuel irradiation capability in the High Flux Reactor Petten

NASA Astrophysics Data System (ADS)

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

2009-07-01

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

STEAM GENERATOR FOR GAS COOLED NUCLEAR REACTORS

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

None

1960-03-14

A steam generator for a gas-cooled nuclear reactor is disposed inside the same pressure vessel as the reactor and has a tube system heated by the gas circulating through the reactor; the pressure vessel is double-walled, and the interspace between these two walls is filled with concrete serving as radiation shielding. The steam generator has a cylindricaIly shaped vertical casing, through which the heating gas circulates, while the tubes are arranged in a plurality of parallel horizontal planes and each of them have the shape of an involute of a circle. The tubes are uniformly distributed over the available surfacemore » in the plane, all the tubes of the same plane being connected in parallel. The exterior extremities of these involute-shaped tubes are each connected with similar tubes disposed in the adjacent lower situated plane, while the interior extremities are connected with tubes in the adjacent higher situated plane. The alimentation of the tubes is performed over annular headers. The tube system is self-supporting, the tubes being joined together by welded spacers. The fluid flow in the tubes is performed by forced circulation. (NPO)« less

Design and Build of Reactor Simulator for Fission Surface Power Technology Demonstrator Unit

NASA Technical Reports Server (NTRS)

Godfroy, Thomas; Dickens, Ricky; Houts, Michael; Pearson, Boise; Webster, Kenny; Gibson, Marc; Qualls, Lou; Poston, Dave; Werner, Jim; Radel, Ross

2011-01-01

The Nuclear Systems Team at NASA Marshall Space Flight Center (MSFC) focuses on technology development for state of the art capability in non-nuclear testing of nuclear system and Space Nuclear Power for fission reactor systems for lunar and Mars surface power generation as well as radioisotope power systems for both spacecraft and surface applications. Currently being designed and developed is a reactor simulator (RxSim) for incorporation into the Technology Demonstrator Unit (TDU) for the Fission Surface Power System (FSPS) Program, which is supported by multiple national laboratories and NASA centers. The ultimate purpose of the RxSim is to provide heated NaK to a pair of Stirling engines in the TDU. The RxSim includes many different systems, components, and instrumentation that have been developed at MSFC while working with pumped NaK systems and in partnership with the national laboratories and NASA centers. The main components of the RxSim are a core, a pump, a heat exchanger (to mimic the thermal load of the Stirling engines), and a flow meter for tests at MSFC. When tested at NASA Glenn Research Center (GRC) the heat exchanger will be replaced with a Stirling power conversion engine. Additional components include storage reservoirs, expansion volumes, overflow catch tanks, safety and support hardware, instrumentation (temperature, pressure, flow) for data collection, and power supplies. This paper will discuss the design and current build status of the RxSim for delivery to GRC in early 2012.

Design and Build of Reactor Simulator for Fission Surface Power Technology Demonstrator Unit

NASA Astrophysics Data System (ADS)

Godfroy, T.; Dickens, R.; Houts, M.; Pearson, B.; Webster, K.; Gibson, M.; Qualls, L.; Poston, D.; Werner, J.; Radel, R.

The Nuclear Systems Team at Marshall Space Flight Center (MSFC) focuses on technology development for state of the art capability in non-nuclear testing of nuclear system and Space Nuclear Power for fission reactor systems for lunar and mars surface power generation as well as radioisotope power systems for both spacecraft and surface applications. Currently being designed and developed is a reactor simulator (RxSim) for incorporation into the Technology Demonstrator Unit (TDU) for the Fission Surface Power System (FSPS) Program which is supported by multiple national laboratories and NASA centers. The ultimate purpose of the RxSim is to provide heated NaK to a pair of Stirling engines in the TDU. The RxSim includes many different systems, components, and instrumentation that have been developed at MSFC while working with pumped NaK systems and in partnership with the national laboratories and NASA centers. The main components of the RxSim are a core, a pump, a heat exchanger (to mimic the thermal load of the Stirling engines), and a flow meter when being tested at MSFC. When tested at GRC the heat exchanger will be replaced with a Stirling power conversion engine. Additional components include storage reservoirs, expansion volumes, overflow catch tanks, safety and support hardware, instrumenta- tion (temperature, pressure, flow) data collection, and power supplies. This paper will discuss the design and current build status of the RxSim for delivery to GRC in early 2012.

10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

Code of Federal Regulations, 2010 CFR

2010-01-01

... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which the...

10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

Code of Federal Regulations, 2011 CFR

2011-01-01

... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which the...

Chapter 11: Marine and Hydrokinetic Power Generation and Power Plants

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

Muljadi, Eduard; Yu, Yi-Hsiang

Marine and hydrokinetic (MHK) power generation is a relatively new type of renewable generation. Predecessors such as wind power generation, hydropower plant generation, geothermal generation, photovoltaic generation, and solar thermal generation have gained a lot of attention because of their successful implementation. The successful integration of renewable generation into the electric power grid has energized the power system global communities to take the lessons learned, innovations, and market structure to focus on the large potential of MHK to also contribute to the pool of renewable energy generation. This chapter covers the broad spectrum of MHK generation. The state-of-the-art power takeoffmore » methods will be discussed. Types of electrical generators will be presented, and the options for implementation will be presented.« less

Gaseous-fuel nuclear reactor research for multimegawatt power in space

NASA Technical Reports Server (NTRS)

Thom, K.; Schneider, R. T.; Helmick, H. H.

1977-01-01

In the gaseous-fuel reactor concept, the fissile material is contained in a moderator-reflector cavity and exists in the form of a flowing gas or plasma separated from the cavity walls by means of fluid mechanical forces. Temperatures in excess of structural limitations are possible for low-specific-mass power and high-specific-impulse propulsion in space. Experiments have been conducted with a canister filled with enriched UF6 inserted into a beryllium-reflected cavity. A theoretically predicted critical mass of 6 kg was measured. The UF6 was also circulated through this cavity, demonstrating stable reactor operation with the fuel in motion. Because the flowing gaseous fuel can be continuously processed, the radioactive waste in this type of reactor can be kept small. Another potential of fissioning gases is the possibility of converting the kinetic energy of fission fragments directly into coherent electromagnetic radiation, the nuclear pumping of lasers. Numerous nuclear laser experiments indicate the possibility of transmitting power in space directly from fission energy. The estimated specific mass of a multimegawatt gaseous-fuel reactor power system is from 1 to 5 kg/kW while the companion laser-power receiver station would be much lower in specific mass.

NACA Zero Power Reactor Facility Hazards Summary

NASA Technical Reports Server (NTRS)

1957-01-01

The Lewis Flight Propulsion Laboratory of the National Advisory Committee for Aeronautics proposes to build a zero power research reactor facility which will be located in the laboratory grounds near Clevelaurd, Ohio. The purpose of this report is to inform the Advisory Commit tee on Reactor Safeguards of the U. S. Atomic Energy Commission in re gard to the design of the reactor facility, the cha,acteristics of th e site, and the hazards of operation at this location, The purpose o f this reactor is to perform critical experiments, to measure reactiv ity effects, to serve as a neutron source, and to serve as a training tool. The reactor facility is described. This is followed by a discu ssion of the nuclear characteristics and the control system. Site cha racteristics are then discussed followed by a discussion of the exper iments which may be conducted in the facility. The potential hazards of the facility are then considered, particularly, the maximum credib le accident. Finally, the administrative procedure is discussed.

Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

DOEpatents

Lasche, G.P.

1983-09-29

The invention is a laser or particle-beam-driven fusion reactor system which takes maximum advantage of both the very short pulsed nature of the energy release of inertial confinement fusion (ICF) and the very small volumes within which the thermonuclear burn takes place. The pulsed nature of ICF permits dynamic direct energy conversion schemes such as magnetohydrodynamic (MHD) generation and magnetic flux compression; the small volumes permit very compact blanket geometries. By fully exploiting these characteristics of ICF, it is possible to design a fusion reactor with exceptionally high power density, high net electric efficiency, and low neutron-induced radioactivity. The invention includes a compact blanket design and method and apparatus for obtaining energy utilizing the compact blanket.

10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Immediate notification requirements for operating nuclear power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear...

10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Immediate notification requirements for operating nuclear power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear...

Isolated Power Generation System Using Permanent Magnet Synchronous Generator with Improved Power Quality

NASA Astrophysics Data System (ADS)

Arya, Sabha Raj; Patel, Ashish; Giri, Ashutosh

2018-03-01

This paper deals wind energy based power generation system using Permanent Magnet Synchronous Generator (PMSG). It is controlled using advanced enhanced phase-lock loop for power quality features using distribution static compensator to eliminate the harmonics and to provide KVAR compensation as well as load balancing. It also manages rated potential at the point of common interface under linear and non-linear loads. In order to have better efficiency and reliable operation of PMSG driven by wind turbine, it is necessary to analyze the governing equation of wind based turbine and PMSG under fixed and variable wind speed. For handling power quality problems, power electronics based shunt connected custom power device is used in three wire system. The simulations in MATLAB/Simulink environment have been carried out in order to demonstrate this model and control approach used for the power quality enhancement. The performance results show the adequate performance of PMSG based power generation system and control algorithm.

Isolated Power Generation System Using Permanent Magnet Synchronous Generator with Improved Power Quality

NASA Astrophysics Data System (ADS)

Arya, Sabha Raj; Patel, Ashish; Giri, Ashutosh

2018-06-01

This paper deals wind energy based power generation system using Permanent Magnet Synchronous Generator (PMSG). It is controlled using advanced enhanced phase-lock loop for power quality features using distribution static compensator to eliminate the harmonics and to provide KVAR compensation as well as load balancing. It also manages rated potential at the point of common interface under linear and non-linear loads. In order to have better efficiency and reliable operation of PMSG driven by wind turbine, it is necessary to analyze the governing equation of wind based turbine and PMSG under fixed and variable wind speed. For handling power quality problems, power electronics based shunt connected custom power device is used in three wire system. The simulations in MATLAB/Simulink environment have been carried out in order to demonstrate this model and control approach used for the power quality enhancement. The performance results show the adequate performance of PMSG based power generation system and control algorithm.

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

NASA Technical Reports Server (NTRS)

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

1976-01-01

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

Safety and Environment aspects of Tokamak- type Fusion Power Reactor- An Overview

NASA Astrophysics Data System (ADS)

Doshi, Bharat; Reddy, D. Chenna

2017-04-01

Naturally occurring thermonuclear fusion reaction (of light atoms to form a heavier nucleus) in the sun and every star in the universe, releases incredible amounts of energy. Demonstrating the controlled and sustained reaction of deuterium-tritium plasma should enable the development of fusion as an energy source here on Earth. The promising fusion power reactors could be operated on the deuterium-tritium fuel cycle with fuel self-sufficiency. The potential impact of fusion power on the environment and the possible risks associated with operating large-scale fusion power plants is being studied by different countries. The results show that fusion can be a very safe and sustainable energy source. A fusion power plant possesses not only intrinsic advantages with respect to safety compared to other sources of energy, but also a negligible long term impact on the environment provided certain precautions are taken in its design. One of the important considerations is in the selection of low activation structural materials for reactor vessel. Selection of the materials for first wall and breeding blanket components is also important from safety issues. It is possible to fully benefit from the advantages of fusion energy if safety and environmental concerns are taken into account when considering the conceptual studies of a reactor design. The significant safety hazards are due to the tritium inventory and energetic neutron fluence induced activity in the reactor vessel, first wall components, blanket system etc. The potential of release of radioactivity under operational and accident conditions needs attention while designing the fusion reactor. Appropriate safety analysis for the quantification of the risk shall be done following different methods such as FFMEA (Functional Failure Modes and Effects Analysis) and HAZOP (Hazards and operability). Level of safety and safety classification such as nuclear safety and non-nuclear safety is very important for the FPR (Fusion

76 FR 74630 - Making Changes to Emergency Plans for Nuclear Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-01

... NUCLEAR REGULATORY COMMISSION 10 CFR Parts 50 and 52 RIN 3150-AI10 [NRC-2008-0122] Making Changes to Emergency Plans for Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission. ACTION... guide (RG) 1.219, ``Guidance on Making Changes to Emergency Plans for Nuclear Power Reactors.'' This...

Steam generator for liquid metal fast breeder reactor

DOEpatents

Gillett, James E.; Garner, Daniel C.; Wineman, Arthur L.; Robey, Robert M.

1985-01-01

Improvements in the design of internal components of J-shaped steam generators for liquid metal fast breeder reactors. Complex design improvements have been made to the internals of J-shaped steam generators which improvements are intended to reduce tube vibration, tube jamming, flow problems in the upper portion of the steam generator, manufacturing complexities in tube spacer attachments, thermal stripping potentials and difficulties in the weld fabrication of certain components.

Aircraft Photovoltaic Power-Generating System.

NASA Astrophysics Data System (ADS)

Doellner, Oscar Leonard

Photovoltaic cells, appropriately cooled and operating in the combustion-created high radiant-intensity environment of gas-turbine and jet engines, may replace the conventional (gearbox-driven) electrical power generators aboard jet aircraft. This study projects significant improvements not only in aircraft electrical power-generating-system performance, but also in overall aircraft performance. Jet -engine design modifications incorporating this concept not only save weight (and thus fuel), but are--in themselves --favorable to jet-engine performance. The dissertation concentrates on operational, constructional, structural, thermal, optical, radiometrical, thin-film, and solid-state theoretical aspects of the overall project. This new electrical power-generating system offers solid-state reliability with electrical power-output capability comparable to that of existing aircraft electromechanical power-generating systems (alternators and generators). In addition to improvements in aircraft performance, significant aircraft fuel- and weight-saving advantages are projected.

Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

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

Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.

The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, andmore » other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.« less

75 FR 34776 - Florida Power & Light Company; Turkey Point Nuclear Generating Plant, Units 3 and 4...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-18

... changes to the reactor, fuel, plant, structures, support structures, water, or land at the Turkey Point... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-250 and 50-251; NRC-2010-0212] Florida Power & Light Company; Turkey Point Nuclear Generating Plant, Units 3 and 4; Environmental Assessment and Finding of No...

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

Low-power lead-cooled fast reactor loaded with MOX-fuel

NASA Astrophysics Data System (ADS)

Sitdikov, E. R.; Terekhova, A. M.

2017-01-01

Fast reactor for the purpose of implementation of research, education of undergraduate and doctoral students in handling innovative fast reactors and training specialists for atomic research centers and nuclear power plants (BRUTs) was considered. Hard neutron spectrum achieved in the fast reactor with compact core and lead coolant. Possibility of prompt neutron runaway of the reactor is excluded due to the low reactivity margin which is less than the effective fraction of delayed neutrons. The possibility of using MOX fuel in the BRUTs reactor was examined. The effect of Keff growth connected with replacement of natural lead coolant to 208Pb coolant was evaluated. The calculations and reactor core model were performed using the Serpent Monte Carlo code.

Investigation of materials for fusion power reactors

NASA Astrophysics Data System (ADS)

Bouhaddane, A.; Slugeň, V.; Sojak, S.; Veterníková, J.; Petriska, M.; Bartošová, I.

2014-06-01

The possibility of application of nuclear-physical methods to observe radiation damage to structural materials of nuclear facilities is nowadays a very actual topic. The radiation damage to materials of advanced nuclear facilities, caused by extreme radiation stress, is a process, which significantly limits their operational life as well as their safety. In the centre of our interest is the study of the radiation degradation and activation of the metals and alloys for the new nuclear facilities (Generation IV fission reactors, fusion reactors ITER and DEMO). The observation of the microstructure changes in the reactor steels is based on experimental investigation using the method of positron annihilation spectroscopy (PAS). The experimental part of the work contains measurements focused on model reactor alloys and ODS steels. There were 12 model reactor steels and 3 ODS steels. We were investigating the influence of chemical composition on the production of defects in crystal lattice. With application of the LT 9 program, the spectra of specimen have been evaluated and the most convenient samples have been determined.

Computer optimization of reactor-thermoelectric space power systems

NASA Technical Reports Server (NTRS)

Maag, W. L.; Finnegan, P. M.; Fishbach, L. H.

1973-01-01

A computer simulation and optimization code that has been developed for nuclear space power systems is described. The results of using this code to analyze two reactor-thermoelectric systems are presented.

REVIEW OF POWER AND HEAT REACTOR DESIGNS. Domestic and Foreign

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

Appleby, E.R., comp

1963-10-01

Unclassified information from domestic and foreign literature from January 1952 through September 1963 is compiled. Design characteristics and current information on the status of the individual designs are given, along with references for the associated literature. SNAP systems, proposed reactors, and chemonuclear and test reactors with characteristics similar to power reactors are included. The designs are indexed by name, location, type, and some special characteristics. (D.C.W.)

Nuclear reactor power for an electrically powered orbital transfer vehicle

NASA Technical Reports Server (NTRS)

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

1987-01-01

To help determine the systems requirements for a 300-kWe space nuclear reactor power system, a mission and spacecraft have been examined which utilize electric propulsion and this nuclear reactor power for multiple transfers of cargo between low earth orbit (LEO) and geosynchronous earth orbit (GEO). A propulsion system employing ion thrusters and xenon propellant was selected. Propellant and thrusters are replaced after each sortie to GEO. The mass of the Orbital Transfer Vehicle (OTV), empty and dry, is 11,000 kg; nominal propellant load is 5000 kg. The OTV operates between a circular orbit at 925 km altitude, 28.5 deg inclination, and GEO. Cargo is brought to the OTV by Shuttle and an Orbital Maneuvering Vehicle (OMV); the OTV then takes it to GEO. The OTV can also bring cargo back from GEO, for transfer by OMV to the Shuttle. OTV propellant is resupplied and the ion thrusters are replaced by the OMV before each trip to GEO. At the end of mission life, the OTV's electric propulsion is used to place it in a heliocentric orbit so that the reactor will not return to earth. The nominal cargo capability to GEO is 6000 kg with a transit time of 120 days; 1350 kg can be transferred in 90 days, and 14,300 kg in 240 days. These capabilities can be considerably increased by using separate Shuttle launches to bring up propellant and cargo, or by changing to mercury propellant.

Evaluation on nitrogen oxides and nanoparticle removal and nitrogen monoxide generation using a wet-type nonthermal plasma reactor

NASA Astrophysics Data System (ADS)

Takehana, Kotaro; Kuroki, Tomoyuki; Okubo, Masaaki

2018-05-01

Nitrogen oxides (NOx) emitted from power plants and combustion sources cause air pollution problems. Selective catalytic reduction technology is remarkably useful for NOx removal. However, there are several drawbacks such as preparation of reducing agents, usage of harmful heavy metals, and higher cost. On the other hand, trace NO is a vasodilator agent and employed in inhalation therapies for treating pulmonary hypertension in humans. Considering these factors, in the present study, a wet-type nonthermal plasma reactor, which can control NOx and nanoparticle emissions and generate NO, is investigated. The fundamental characteristics of the reactor are investigated. First, the experiment of nanoparticle removal is carried out. Collection efficiencies of over 99% are achieved for nanoparticles at 50 and 100 ml min‑1 of liquid flow rates. Second, experiments of NOx removal under air atmosphere and NOx generation under nitrogen atmosphere are carried out. NOx-removal efficiencies of over 95% under the air plasma are achieved in 50–200 ml min‑1 liquid flow rates. Moreover, under nitrogen plasma, NOx is generated, of which the major portion is NO. For example, NO concentration is 25 ppm, while NOx concentration is 31 ppm at 50 ml min‑1 liquid flow rate. Finally, experiments of NO generation under the nitrogen atmosphere with or without flowing water are carried out. When water flows on the inner surface of the reactor, approximately 14 ppm of NO is generated. Therefore, NO generation requires flowing water. It is considered that the reaction of N and OH, which is similar to the extended Zeldovich mechanism, could occur to induce NO formation. From these results, it is verified that the wet-type plasma reactor is useful for NOx removal and NO generation under nitrogen atmosphere with flowing water.

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.

Automated power control system for reactor TRIGA PUSPATI

NASA Astrophysics Data System (ADS)

Ghazali, Anith Khairunnisa; Minhat, Mohd Sabri; Hassan, Mohd Khair

2017-01-01

Reactor TRIGA PUSPATI (RTP) Mark II type undergoes safe operation for more than 30 years and the only research reactor exists in Malaysia. The main safety feature of Instrumentation and Control (I&C) system design is such that any failure in the electronic, or its associated components, does not lead to an uncontrolled rate of reactivity. The existed controller using feedback approach to control the reactor power. This paper introduces proposed controllers such as Model Reference Adaptive Control (MRAC) and Proportional Integral Derivatives (PID) controller for the RTP simulation. In RTP, the most important considered parameter is the reactor power and act as nervous system. To design a controller for complex plant like RTP is quite difficult due to high cost and safety factors cause by the failure of the controller. Furthermore, to overcome these problems, a simulator can be used to replace functions the hardware and test could then be simulated using this simulator. In order to find the best controller, several controllers were proposed and the result will be analysed for study the performances of the controller. The output result will be used to find out the best RTP power controller using MATLAB/Simulink and gives result as close as the real RTP performances. Currently, the structures of RTP was design using MATLAB/Simulink tool that consist of fission chamber, controller, control rod position, height-to-worth of control rods and a RTP model. The controller will control the control rod position to make sure that the reactivity still under the limitation parameter. The results given from each controller will be analysed and validated through experiment data collected from RTP.

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

Code of Federal Regulations, 2014 CFR

2014-01-01

... 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... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

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

Code of Federal Regulations, 2013 CFR

2013-01-01

... 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... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

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

Code of Federal Regulations, 2012 CFR

2012-01-01

... 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... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

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

NASA Astrophysics Data System (ADS)

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

2016-09-01

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

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

NASA Technical Reports Server (NTRS)

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

2017-01-01

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

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

NASA Technical Reports Server (NTRS)

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

2017-01-01

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

Temperature Resistant Fiber Bragg Gratings for On-Line and Structural Health Monitoring of the Next-Generation of Nuclear Reactors.

PubMed

Laffont, Guillaume; Cotillard, Romain; Roussel, Nicolas; Desmarchelier, Rudy; Rougeault, Stéphane

2018-06-02

The harsh environment associated with the next generation of nuclear reactors is a great challenge facing all new sensing technologies to be deployed for on-line monitoring purposes and for the implantation of SHM methods. Sensors able to resist sustained periods at very high temperatures continuously as is the case within sodium-cooled fast reactors require specific developments and evaluations. Among the diversity of optical fiber sensing technologies, temperature resistant fiber Bragg gratings are increasingly being considered for the instrumentation of future nuclear power plants, especially for components exposed to high temperature and high radiation levels. Research programs are supporting the developments of optical fiber sensors under mixed high temperature and radiative environments leading to significant increase in term of maturity. This paper details the development of temperature-resistant wavelength-multiplexed fiber Bragg gratings for temperature and strain measurements and their characterization for on-line monitoring into the liquid sodium used as a coolant for the next generation of fast reactors.

Progress in space nuclear reactor power systems technology development - The SP-100 program

NASA Technical Reports Server (NTRS)

Davis, H. S.

1984-01-01

Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

Electrical Power Conversion of River and Tidal Power Generator

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

Muljadi, Eduard; Gevorgian, Vahan; Wright, Alan

As renewable generation has become less expensive during recent decades, and it becomes more accepted by the global population, the focus on renewable generation has expanded to include new types with promising future applications, such as river and tidal generation. Although the utilization of power electronics and electric machines in industry is phenomenal, the emphasis on system design is different for various sectors of industry. In precision control, robotics, and weaponry, the design emphasis is on accuracy and reliability with less concern for the cost of the final product. In energy generation, the cost of energy is the prime concern;more » thus, capital expenditures (CAPEX) and operations and maintenance expenditures (OPEX) are the major design objectives. This paper describes the electrical power conversion aspects of river and tidal generation. Although modern power converter control is available to control the generation side, the design was chosen on the bases of minimizing the CAPEX and OPEX; thus, the architecture is simple and modular for ease of replacement and maintenance. The power conversion is simplified by considering a simple diode bridge and a DC-DC power converter to take advantage of abundant and low-cost photovoltaic inverters that have well-proven grid integration characteristics (i.e., the capability to produce energy with good power quality and control real power and voltage on the grid side).« less

Method of locating a leaking fuel element in a fast breeder power reactor

DOEpatents

Honekamp, John R.; Fryer, Richard M.

1978-01-01

Leaking fuel elements in a fast reactor are identified by measuring the ratio of .sup.134 Xe to .sup.133 Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

75 FR 70042 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-16

... NUCLEAR REGULATORY COMMISSION [Docket Nos (Redacted), License Nos (Redacted), EA (Redacted); NRC- 2010-0351] In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order Modifying License (Effective Immediately) I. The licensees identified in...

75 FR 79423 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-20

... NUCLEAR REGULATORY COMMISSION [Docket Nos. (Redacted), License Nos.: (Redacted), EA (Redacted); NRC- 2010-0351] In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order Modifying License (Effective Immediately) I The licensees identified in...

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.

Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

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

Forsberg, C.W.; Reich, W.J.

1991-09-01

The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactormore » concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs.« less

Thermoelectric power generator with intermediate loop

DOEpatents

Bell, Lon E; Crane, Douglas Todd

2013-05-21

A thermoelectric power generator is disclosed for use to generate electrical power from heat, typically waste heat. An intermediate heat transfer loop forms a part of the system to permit added control and adjustability in the system. This allows the thermoelectric power generator to more effectively and efficiently generate power in the face of dynamically varying temperatures and heat flux conditions, such as where the heat source is the exhaust of an automobile, or any other heat source with dynamic temperature and heat flux conditions.

Thermoelectric power generator with intermediate loop

DOEpatents

Bel,; Lon, E [Altadena, CA; Crane, Douglas Todd [Pasadena, CA

2009-10-27

A thermoelectric power generator is disclosed for use to generate electrical power from heat, typically waste heat. An intermediate heat transfer loop forms a part of the system to permit added control and adjustability in the system. This allows the thermoelectric power generator to more effectively and efficiently generate power in the face of dynamically varying temperatures and heat flux conditions, such as where the heat source is the exhaust of an automobile, or any other heat source with dynamic temperature and heat flux conditions.

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.

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.

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

Preliminary plan for testing a thermionic reactor in the Plum Brook Space Power Facility

NASA Technical Reports Server (NTRS)

Haley, F. A.

1972-01-01

A preliminary plan is presented for testing a thermionic reactor in the Plum Brook Space Power Facility (SPF). A technical approach, cost estimate, manpower estimate, and schedule are presented to cover a 2 year full power reactor test.

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Technical specifications on effluents from nuclear power reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors...

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Technical specifications on effluents from nuclear power reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors...

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

NASA Technical Reports Server (NTRS)

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

1985-01-01

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

77 FR 40092 - License Amendment To Increase the Maximum Reactor Power Level, Florida Power & Light Company, St...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-06

... Increase the Maximum Reactor Power Level, Florida Power & Light Company, St. Lucie, Units 1 and 2 AGENCY... amendment for Renewed Facility Operating License Nos. DPR-67 and NPF-16, issued to Florida Power & Light... St. Lucie County, Florida. The proposed license amendment would increase the maximum thermal power...

Method and system to directly produce electrical power within the lithium blanket region of a magnetically confined, deuterium-tritium (DT) fueled, thermonuclear fusion reactor

DOEpatents

Woolley, Robert D.

1999-01-01

A method for integrating liquid metal magnetohydrodynamic power generation with fusion blanket technology to produce electrical power from a thermonuclear fusion reactor located within a confining magnetic field and within a toroidal structure. A hot liquid metal flows from a liquid metal blanket region into a pump duct of an electromagnetic pump which moves the liquid metal to a mixer where a gas of predetermined pressure is mixed with the pressurized liquid metal to form a Froth mixture. Electrical power is generated by flowing the Froth mixture between electrodes in a generator duct. When the Froth mixture exits the generator the gas is separated from the liquid metal and both are recycled.

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

NASA Astrophysics Data System (ADS)

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

2017-10-01

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

Foundations for the Fourth Generation of Nuclear Power

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

Lake, James Alan

2000-11-01

Plentiful, affordable electrical energy is a critically important commodity to nations wishing to grow their economy. Energy, and more specifically electricity, is the fuel of economic growth. More than one-third of the world’s population (more than 2 billion people), however, live today without access to any electricity. Further, another 2 billion people in the world exist on less than 100 watts of electricity per capita. By comparison, the large economies of Japan and France use more than 800 watts of electricity per capita, and the United States uses nearly 1500 watts of electricity per capita. As the governments of developingmore » nations strive to improve their economies, and hence the standard of living of their people, electricity use is increasing. Several forecasts of electrical generation growth have concluded that world electricity demand will roughly double in the next 20–25 years, and possibly triple by 2050. This electrical generation growth will occur primarily in the rapidly developing and growing economies in Asia and Latin America. This net growth is in addition to the need for replacement generating capacity in the United States and Europe as aging power plants (primarily fossil-fueled) are replaced. This very substantial worldwide electricity demand growth places the issue of where this new electricity generation capacity is to come from squarely in front of the developed countries. They have a fundamental desire (if not a moral obligation) to help these developing countries sustain their economic growth and improve their standard of living, while at the same time protecting the energy (and economic) security of their own countries. There are currently 435 power reactors generating about 16 percent of the world’s electricity. We know full well that nuclear power shows great promise as an economical, safe, and emissions-free source of electrical energy, but it also carries at least the perception of great problems, from public safety to

Design and Control of Integrated Systems for Hydrogen Production and Power Generation

NASA Astrophysics Data System (ADS)

Georgis, Dimitrios

Growing concerns on CO2 emissions have led to the development of highly efficient power plants. Options for increased energy efficiencies include alternative energy conversion pathways, energy integration and process intensification. Solid oxide fuel cells (SOFC) constitute a promising alternative for power generation since they convert the chemical energy electrochemically directly to electricity. Their high operating temperature shows potential for energy integration with energy intensive units (e.g. steam reforming reactors). Although energy integration is an essential tool for increased efficiencies, it leads to highly complex process schemes with rich dynamic behavior, which are challenging to control. Furthermore, the use of process intensification for increased energy efficiency imposes an additional control challenge. This dissertation identifies and proposes solutions on design, operational and control challenges of integrated systems for hydrogen production and power generation. Initially, a study on energy integrated SOFC systems is presented. Design alternatives are identified, control strategies are proposed for each alternative and their validity is evaluated under different operational scenarios. The operational range of the proposed control strategies is also analyzed. Next, thermal management of water gas shift membrane reactors, which are a typical application of process intensification, is considered. Design and operational objectives are identified and a control strategy is proposed employing advanced control algorithms. The performance of the proposed control strategy is evaluated and compared with classical control strategies. Finally SOFC systems for combined heat and power applications are considered. Multiple recycle loops are placed to increase design flexibility. Different operational objectives are identified and a nonlinear optimization problem is formulated. Optimal designs are obtained and their features are discussed and compared

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...

Enhancing power generation of floating wave power generators by utilization of nonlinear roll-pitch coupling

NASA Astrophysics Data System (ADS)

Yerrapragada, Karthik; Ansari, M. H.; Karami, M. Amin

2017-09-01

We propose utilization of the nonlinear coupling between the roll and pitch motions of wave energy harvesting vessels to increase their power generation by orders of magnitude. Unlike linear vessels that exhibit unidirectional motion, our vessel undergoes both pitch and roll motions in response to frontal waves. This significantly magnifies the motion of the vessel and thus improves the power production by several orders of magnitude. The ocean waves result in roll and pitch motions of the vessel, which in turn causes rotation of an onboard pendulum. The pendulum is connected to an electric generator to produce power. The coupled electro-mechanical system is modeled using energy methods. This paper investigates the power generation of the vessel when the ratio between pitch and roll natural frequencies is about 2 to 1. In that case, a nonlinear energy transfer occurs between the roll and pitch motions, causing the vessel to perform coupled pitch and roll motion even though it is only excited in the pitch direction. It is shown that co-existence of pitch and roll motions significantly enhances the pendulum rotation and power generation. A method for tuning the natural frequencies of the vessel is proposed to make the energy generator robust to variations of the frequency of the incident waves. It is shown that the proposed method enhances the power output of the floating wave power generators by multiple orders of magnitude. A small-scale prototype is developed for the proof of concept. The nonlinear energy transfer and the full rotation of the pendulum in the prototype are observed in the experimental tests.

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.

Reliability and safety of the electrical power supply complex of the Hanford production reactors

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

Robbins, F.D.

Safety has been and must continue to be the inviolable modulus by which the operation of a nuclear reactor must be judged. A malfunction in any reactor may well result in a release of fission products which may dissipate over a wide geographical area. Such dissipation may place the health, happiness and even the lives of the people in the region in serious jeopardy. As a result, the property damage and liability cost may reach astronomical values in the order of magnitude of billions of dollars. Reliability of the electrical network is an indispensable factor in attaining a high ordermore » of safety assurance. Progress in the peaceful use of atomic energy may take the form of electrical power generation using the nuclear reactor as a source of thermal energy. In view of these factors it seems appropriate and profitable that a critical engineering study be made of the safety and reliability of the Hanford reactors without regard to cost economics. This individual and independent technical engineering analysis was made without regard to Hanford traditional engineering and administration assignments. The main objective has been to focus attention on areas which seem to merit further detailed study on conditions which seem to need adjustment but most of all on those changes which will improve reactor safety. This report is the result of such a study.« less

10 CFR 50.83 - Release of part of a power reactor facility or site for unrestricted use.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Release of part of a power reactor facility or site for... of a power reactor facility or site for unrestricted use. (a) Prior written NRC approval is required... release. Nuclear power reactor licensees seeking NRC approval shall— (1) Evaluate the effect of releasing...

10 CFR 50.83 - Release of part of a power reactor facility or site for unrestricted use.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Release of part of a power reactor facility or site for... of a power reactor facility or site for unrestricted use. (a) Prior written NRC approval is required... release. Nuclear power reactor licensees seeking NRC approval shall— (1) Evaluate the effect of releasing...

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.

A small, 1400 deg Kelvin, reactor for Brayton space power systems

NASA Technical Reports Server (NTRS)

Lantz, E.; Mayo, W.

1972-01-01

A preliminary cost estimate for a small reactor in Brayton space power systems with (u-233)n or (pu-239)n as the fuel in the T-111 fuel elements totaled to about four million dollars; considered is a 22.8 in. diameter reactor with 247 fuel elements.

10 CFR 50.44 - Combustible gas control for nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 1 2014-01-01 2014-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control... capability for ensuring a mixed atmosphere. (2) Combustible gas control. (i) All boiling water reactors with...

10 CFR 50.44 - Combustible gas control for nuclear power reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control... capability for ensuring a mixed atmosphere. (2) Combustible gas control. (i) All boiling water reactors with...

10 CFR 50.44 - Combustible gas control for nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control... capability for ensuring a mixed atmosphere. (2) Combustible gas control. (i) All boiling water reactors with...

Tokamak power reactor ignition and time dependent fractional power operation

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

Vold, E.L.; Mau, T.K.; Conn, R.W.

1986-06-01

A flexible time-dependent and zero-dimensional plasma burn code with radial profiles was developed and employed to study the fractional power operation and the thermal burn control options for an INTOR-sized tokamak reactor. The code includes alpha thermalization and a time-dependent transport loss which can be represented by any one of several currently popular scaling laws for energy confinement time. Ignition parameters were found to vary widely in density-temperature (n-T) space for the range of scaling laws examined. Critical ignition issues were found to include the extent of confinement time degradation by alpha heating, the ratio of ion to electron transportmore » power loss, and effect of auxiliary heating on confinement. Feedback control of the auxiliary power and ion fuel sources are shown to provide thermal stability near the ignition curve.« 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.

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

NASA Astrophysics Data System (ADS)

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

2018-03-01

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

A citation-based assessment of the performance of U.S. boiling water reactors following extended power up-rates

NASA Astrophysics Data System (ADS)

Heidrich, Brenden J.

Nuclear power plants produce 20 percent of the electricity generated in the U.S. Nuclear generated electricity is increasingly valuable to a utility because it can be produced at a low marginal cost and it does not release any carbon dioxide. It can also be a hedge against uncertain fossil fuel prices. The construction of new nuclear power plants in the U.S. is cautiously moving forward, restrained by high capital costs. Since 1998, nuclear utilities have been increasing the power output of their reactors by implementing extended power up-rates. Power increases of up to 20 percent are allowed under this process. The equivalent of nine large power plants has been added via extended power up-rates. These up-rates require the replacement of large capital equipment and are often performed in concert with other plant life extension activities such as license renewals. This dissertation examines the effect of these extended power up-rates on the safety performance of U.S. boiling water reactors. Licensing event reports are submitted by the utilities to the Nuclear Regulatory Commission, the federal nuclear regulator, for a wide range of abnormal events. Two methods are used to examine the effect of extended power up-rates on the frequency of abnormal events at the reactors. The Crow/AMSAA model, a univariate technique is used to determine if the implementation of an extended power up-rate affects the rate of abnormal events. The method has a long history in the aerospace industry and in the military. At a 95-percent confidence level, the rate of events requiring the submission of a licensing event report decreases following the implementation of an extended power up-rate. It is hypothesized that the improvement in performance is tied to the equipment replacement and refurbishment that is performed as part of the up-rate process. The reactor performance is also analyzed using the proportional hazards model. This technique allows for the estimation of the effects of

Space Power

NASA Technical Reports Server (NTRS)

1984-01-01

Appropriate directions for the applied research and technology programs that will develop space power systems for U.S. future space missions beyond 1995 are explored. Spacecraft power supplies; space stations, space power reactors, solar arrays, thermoelectric generators, energy storage, and communication satellites are among the topics discussed.

Reactor Power for Large Displacement Autonomous Underwater Vehicles

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

McClure, Patrick Ray; Reid, Robert Stowers; Poston, David Irvin

This is a PentaChart on reactor power for large displacement autonomous underwater vehicles. Currently AUVs use batteries or combinations of batteries and fuel cells for power. Battery/fuel cell technology is limited by duration. Batteries and cell fuels are a good match for some missions, but other missions could benefit greatly by a longer duration. The goal is the following: to design nuclear systems to power an AUV and meet design constraints including non-proliferation issues, power level, size constraints, and power conversion limitations. The action plan is to continue development of a range of systems for terrestrial systems and focus onmore » a system for Titan Moon as alternative to Pu-238 for NASA.« less

Advanced Power Conversion Efficiency in Inventive Plasma for Hybrid Toroidal Reactor

NASA Astrophysics Data System (ADS)

Hançerlioğullari, Aybaba; Cini, Mesut; Güdal, Murat

2013-08-01

Apex hybrid reactor has a good potential to utilize uranium and thorium fuels in the future. This toroidal reactor is a type of system that facilitates the occurrence of the nuclear fusion and fission events together. The most important feature of hybrid reactor is that the first wall surrounding the plasma is liquid. The advantages of utilizing a liquid wall are high power density capacity good power transformation productivity, the magnitude of the reactor's operational duration, low failure percentage, short maintenance time and the inclusion of the system's simple technology and material. The analysis has been made using the MCNP Monte Carlo code and ENDF/B-V-VI nuclear data. Around the fusion chamber, molten salts Flibe (LI2BeF4), lead-lithium (PbLi), Li-Sn, thin-lityum (Li20Sn80) have used as cooling materials. APEX reactor has modeled in the torus form by adding nuclear materials of low significance in the specified percentages between 0 and 12 % to the molten salts. In this study, the neutronic performance of the APEX fusion reactor using various molten salts has been investigated. The nuclear parameters of Apex reactor has been searched for Flibe (LI2BeF4) and Li-Sn, for blanket layers. In case of usage of the Flibe (LI2BeF4), PbLi, and thin-lityum (Li20Sn80) salt solutions at APEX toroidal reactors, fissile material production per source neutron, tritium production speed, total fission rate, energy reproduction factor has been calculated, the results obtained for both salt solutions are compared.

Power generation systems and methods

NASA Technical Reports Server (NTRS)

Jones, Jack A. (Inventor); Chao, Yi (Inventor)

2011-01-01

A power generation system includes a plurality of submerged mechanical devices. Each device includes a pump that can be powered, in operation, by mechanical energy to output a pressurized output liquid flow in a conduit. Main output conduits are connected with the device conduits to combine pressurized output flows output from the submerged mechanical devices into a lower number of pressurized flows. These flows are delivered to a location remote of the submerged mechanical devices for power generation.

An appealing photo-powered multi-functional energy system for the poly-generation of hydrogen and electricity

NASA Astrophysics Data System (ADS)

Tang, Tiantian; Li, Kan; Shen, Zhemin; Sun, Tonghua; Wang, Yalin; Jia, Jinping

2015-10-01

This paper focuses on a photo-powered poly-generation system (PPS) that is powered by the photocatalytic oxidation of organic substrate to produce hydrogen energy and electrical energy synchronously. This particular device runs entirely on light energy and chemical energy of substrate without external voltage. The performance measurements and optimization experiments are all investigated by using the low concentration of pure ethanol (EtOH) solution. Compared with the conventional submerged reactor for the photogeneration of hydrogen, the hydrogen and the electric current obtained in the constructed PPS are all relatively stable in experimental period and the numerical values detected are many times higher than that of the former by using various simulated ethanol waste liquid. When using Chinese rice wine as substrate at the same ethanol content level (i.e., 0.1 mol L-1), the production of hydrogen is close to that of the pure ethanol solution in the constructed PPS, but no hydrogen is detected in the conventional submerged reactor. These results demonstrate that the constructed PPS could effectively utilize light energy and perform good capability in poly-generation of hydrogen and electricity.

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.

High power ring methods and accelerator driven subcritical reactor application

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

Tahar, Malek Haj

2016-08-07

High power proton accelerators allow providing, by spallation reaction, the neutron fluxes necessary in the synthesis of fissile material, starting from Uranium 238 or Thorium 232. This is the basis of the concept of sub-critical operation of a reactor, for energy production or nuclear waste transmutation, with the objective of achieving cleaner, safer and more efficient process than today’s technologies allow. Designing, building and operating a proton accelerator in the 500-1000 MeV energy range, CW regime, MW power class still remains a challenge nowadays. There is a limited number of installations at present achieving beam characteristics in that class, e.g.,more » PSI in Villigen, 590 MeV CW beam from a cyclotron, SNS in Oakland, 1 GeV pulsed beam from a linear accelerator, in addition to projects as the ESS in Europe, a 5 MW beam from a linear accelerator. Furthermore, coupling an accelerator to a sub-critical nuclear reactor is a challenging proposition: some of the key issues/requirements are the design of a spallation target to withstand high power densities as well as ensure the safety of the installation. These two domains are the grounds of the PhD work: the focus is on the high power ring methods in the frame of the KURRI FFAG collaboration in Japan: upgrade of the installation towards high intensity is crucial to demonstrate the high beam power capability of FFAG. Thus, modeling of the beam dynamics and benchmarking of different codes was undertaken to validate the simulation results. Experimental results revealed some major losses that need to be understood and eventually overcome. By developing analytical models that account for the field defects, one identified major sources of imperfection in the design of scaling FFAG that explain the important tune variations resulting in the crossing of several betatron resonances. A new formula is derived to compute the tunes and properties established that characterize the effect of the field

The Simulator Development for RDE Reactor

NASA Astrophysics Data System (ADS)

Subekti, Muhammad; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.

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.

Threshold self-powered gamma detector for use as a monitor of power in a nuclear reactor

DOEpatents

LeVert, Francis E.; Cox, Samson A.

1978-01-01

A self-powered gamma monitor for placement near the core of a nuclear reactor comprises a lead prism surrounded by a coaxial thin nickel sheet, the combination forming a collector. A coaxial polyethylene electron barrier encloses the collector and is separated from the nickel sheet by a vacuum region. The electron barrier is enclosed by a coaxial stainless steel emitter which, in turn, is enclosed within a lead casing. When the detector is placed in a flux of gamma rays, a measure of the current flow in an external circuit between emitter and collector provides a measure of the power level of the reactor.

Microfabricated thermoelectric power-generation devices

NASA Technical Reports Server (NTRS)

Fleurial, Jean-Pierre (Inventor); Ryan, Margaret A. (Inventor); Borshchevsky, Alex (Inventor); Phillips, Wayne (Inventor); Kolawa, Elizabeth A. (Inventor); Snyder, G. Jeffrey (Inventor); Caillat, Thierry (Inventor); Kascich, Thorsten (Inventor); Mueller, Peter (Inventor)

2004-01-01

A device for generating power to run an electronic component. The device includes a heat-conducting substrate (composed, e.g., of diamond or another high thermal conductivity material) disposed in thermal contact with a high temperature region. During operation, heat flows from the high temperature region into the heat-conducting substrate, from which the heat flows into the electrical power generator. A thermoelectric material (e.g., a BiTe alloy-based film or other thermoelectric material) is placed in thermal contact with the heat-conducting substrate. A low temperature region is located on the side of the thermoelectric material opposite that of the high temperature region. The thermal gradient generates electrical power and drives an electrical component.

Microfabricated thermoelectric power-generation devices

NASA Technical Reports Server (NTRS)

Fleurial, Jean-Pierre (Inventor); Phillips, Wayne (Inventor); Borshchevsky, Alex (Inventor); Kolawa, Elizabeth A. (Inventor); Ryan, Margaret A. (Inventor); Caillat, Thierry (Inventor); Mueller, Peter (Inventor); Snyder, G. Jeffrey (Inventor); Kascich, Thorsten (Inventor)

2002-01-01

A device for generating power to run an electronic component. The device includes a heat-conducting substrate (composed, e.g., of diamond or another high thermal conductivity material) disposed in thermal contact with a high temperature region. During operation, heat flows from the high temperature region into the heat-conducting substrate, from which the heat flows into the electrical power generator. A thermoelectric material (e.g., a BiTe alloy-based film or other thermoelectric material) is placed in thermal contact with the heat-conducting substrate. A low temperature region is located on the side of the thermoelectric material opposite that of the high temperature region. The thermal gradient generates electrical power and drives an electrical component.

Prediction and modeling of the two-dimensional separation characteristic of a steam generator at a nuclear power station with VVER-1000 reactors

NASA Astrophysics Data System (ADS)

Parchevsky, V. M.; Guryanova, V. V.

2017-01-01

A computational and experimental procedure for construction of the two-dimensional separation curve (TDSC) for a horizontal steam generator (SG) at a nuclear power station (NPS) with VVER-reactors. In contrast to the conventional one-dimensional curve describing the wetness of saturated steam generated in SG as a function of the boiler water level at one, usually rated, load, TDSC is a function of two variables, which are the level and the load of SGB that enables TDSC to be used for wetness control in a wide load range. The procedure is based on two types of experimental data obtained during rated load operation: the nonuniformity factor of the steam load at the outlet from the submerged perforated sheet (SPS) and the dependence of the mass water level in the vicinity of the "hot" header on the water level the "cold" end of SG. The TDSC prediction procedure is presented in the form of an algorithm using SG characteristics, such as steam load and water level as the input and giving the calculated steam wetness at the output. The zoneby-zone calculation method is used. The result is presented in an analytical form (as an empirical correlation) suitable for uploading into controllers or other controls. The predicted TDSC can be used during real-time operation for implementation of different wetness control scenarios (for example, if the effectiveness is a priority, then the minimum water level, minimum wetness, and maximum turbine efficiency should be maintained; if safety is a priority, then the maximum level at the allowable wetness and the maximum water inventory should be kept), for operation of NPS in controlling the frequency and power in a power system, at the design phase (as a part of the simulation complex for verification of design solutions), during construction and erection (in developing software for personnel training simulators), during commissioning tests (to reduce the duration and labor-intensity of experimental activities), and for training.

An Implanted, Stimulated Muscle Powered Piezoelectric Generator

NASA Technical Reports Server (NTRS)

Lewandowski, Beth; Gustafson, Kenneth; Kilgore, Kevin

2007-01-01

A totally implantable piezoelectric generator system able to harness power from electrically activated muscle could be used to augment the power systems of implanted medical devices, such as neural prostheses, by reducing the number of battery replacement surgeries or by allowing periods of untethered functionality. The features of our generator design are no moving parts and the use of a portion of the generated power for system operation and regulation. A software model of the system has been developed and simulations have been performed to predict the output power as the system parameters were varied within their constraints. Mechanical forces that mimic muscle forces have been experimentally applied to a piezoelectric generator to verify the accuracy of the simulations and to explore losses due to mechanical coupling. Depending on the selection of system parameters, software simulations predict that this generator concept can generate up to approximately 700 W of power, which is greater than the power necessary to drive the generator, conservatively estimated to be 50 W. These results suggest that this concept has the potential to be an implantable, self-replenishing power source and further investigation is underway.

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.

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

NASA Astrophysics Data System (ADS)

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

2013-09-01

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

10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... domestic non-power reactors. 50.64 Section 50.64 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... Permits § 50.64 Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors. (a) Applicability. The requirements of this section apply to all non-power reactors. (b) Requirements. (1) The...

10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... domestic non-power reactors. 50.64 Section 50.64 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... Permits § 50.64 Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors. (a) Applicability. The requirements of this section apply to all non-power reactors. (b) Requirements. (1) The...

Fission-powered in-core thermoacoustic sensor

DOE PAGES

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

2016-04-07

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

Fission-powered in-core thermoacoustic sensor

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

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

2016-04-04

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

Methods and strategies for future reactor safety goals

NASA Astrophysics Data System (ADS)

Arndt, Steven Andrew

There have been significant discussions over the past few years by the United States Nuclear Regulatory Commission (NRC), the Advisory Committee on Reactor Safeguards (ACRS), and others as to the adequacy of the NRC safety goals for use with the next generation of nuclear power reactors to be built in the United States. The NRC, in its safety goals policy statement, has provided general qualitative safety goals and basic quantitative health objectives (QHOs) for nuclear reactors in the United States. Risk metrics such as core damage frequency (CDF) and large early release frequency (LERF) have been used as surrogates for the QHOs. In its review of the new plant licensing policy the ACRS has looked at the safety goals, as has the NRC. A number of issues have been raised including what the Commission had in mind when it drafted the safety goals and QHOs, how risk from multiple reactors at a site should be combined for evaluation, how the combination of a new and old reactor at the same site should be evaluated, what the criteria for evaluating new reactors should be, and whether new reactors should be required to be safer than current generation reactors. As part of the development and application of the NRC safety goal policy statement the Commissioners laid out the expectations for the safety of a nuclear power plant but did not address the risk associated with current multi-unit sites, potential modular reactor sites, and hybrid sites that could contain current generation reactors, new passive reactors, and/or modular reactors. The NRC safety goals and the QHOs refer to a "nuclear power plant," but do not discuss whether a "plant" refers to only a single unit or all of the units on a site. There has been much discussion on this issue recently due to the development of modular reactors. Additionally, the risk of multiple reactor accidents on the same site has been largely ignored in the probabilistic risk assessments (PRAs) done to date, and in most risk

75 FR 3946 - License Nos. DPR-42 and DPR-60; Northern States Power Company; Prairie Island Nuclear Generating...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-25

..., Office of Nuclear Reactor Regulation. [FR Doc. 2010-1309 Filed 1-22-10; 8:45 am] BILLING CODE 7590-01-P ... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-282 and 50-306; NRC-2010-0022] License Nos. DPR-42 and DPR-60; Northern States Power Company; Prairie Island Nuclear Generating Plant, Units 1 and 2...

Power conditioning for space nuclear reactor systems

NASA Technical Reports Server (NTRS)

Berman, Baruch

1987-01-01

This paper addresses the power conditioning subsystem for both Stirling and Brayton conversion of space nuclear reactor systems. Included are the requirements summary, trade results related to subsystem implementation, subsystem description, voltage level versus weight, efficiency and operational integrity, components selection, and shielding considerations. The discussion is supported by pertinent circuit and block diagrams. Summary conclusions and recommendations derived from the above studies are included.

Assessment of the Neutronic and Fuel Cycle Performance of the Transatomic Power Molten Salt Reactor Design

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

Robertson, Sean; Dewan, Leslie; Massie, Mark

This report presents results from a collaboration between Transatomic Power Corporation (TAP) and Oak Ridge National Laboratory (ORNL) to provide neutronic and fuel cycle analysis of the TAP core design through the Department of Energy Gateway for Accelerated Innovation in Nuclear (GAIN) Nuclear Energy Voucher program. The TAP concept is a molten salt reactor using configurable zirconium hydride moderator rod assemblies to shift the neutron spectrum in the core from mostly epithermal at beginning of life to thermal at end of life. Additional developments in the ChemTriton modeling and simulation tool provide the critical moderator-to-fuel ratio searches and time-dependent parametersmore » necessary to simulate the continuously changing physics in this complex system. The implementation of continuous-energy Monte Carlo transport and depletion tools in ChemTriton provide for full-core three-dimensional modeling and simulation. Results from simulations with these tools show agreement with TAP-calculated performance metrics for core lifetime, discharge burnup, and salt volume fraction, verifying the viability of reducing actinide waste production with this concept. Additional analyses of mass feed rates and enrichments, isotopic removals, tritium generation, core power distribution, core vessel helium generation, moderator rod heat deposition, and reactivity coeffcients provide additional information to make informed design decisions. This work demonstrates capabilities of ORNL modeling and simulation tools for neutronic and fuel cycle analysis of molten salt reactor concepts.« less

An RF-Powered Micro-Reactor for Efficient Extraction and Hydrolysis

NASA Astrophysics Data System (ADS)

Scott, V.

2014-12-01

An RF sample-processing micro-reactor that was developed as part of potential in situ Exploration Missions to inner- and outer-planetary bodies was designed to utilize aqueous solutions subjected to 60 GHz radiation at 730 mW of input power to extract target organic compounds and molecular and inorganic ions as well as to hydrolyze complex polymeric materials. Successful identification and characterization of these molecules relies on the sample-processing techniques utilized alongside state-of-the-art detection and analysis. For mass and power restrictions put on space exploration missions, smaller and more efficient instruments are highly desirable. The RF micro-reactor potentially offers a simplified alternative to the typical gold-standard extractions that often use solvents, chemicals, and conditions that can vary wildly and depend on the targeted molecules. Instead, this instrument uses a single solvent ­— water — that can be "tuned" under the different experimental conditions, leveraging the operating principles of the Sub-Critical Water Extractor. Proof-of-concept experiments examining the hydrolysis of glycosidic and peptide bonds were successful in demonstrating the RF micro-reactor's capabilities. Progress toward coupling the reactor with a micro-scale sample-handling system enabling slurry delivery has been made and preliminary results on heterogeneous reactions and extractions will be presented.

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.

Oscillating fluid power generator

DOEpatents

Morris, David C

2014-02-25

A system and method for harvesting the kinetic energy of a fluid flow for power generation with a vertically oriented, aerodynamic wing structure comprising one or more airfoil elements pivotably attached to a mast. When activated by the moving fluid stream, the wing structure oscillates back and forth, generating lift first in one direction then in the opposite direction. This oscillating movement is converted to unidirectional rotational movement in order to provide motive power to an electricity generator. Unlike other oscillating devices, this device is designed to harvest the maximum aerodynamic lift forces available for a given oscillation cycle. Because the system is not subjected to the same intense forces and stresses as turbine systems, it can be constructed less expensively, reducing the cost of electricity generation. The system can be grouped in more compact clusters, be less evident in the landscape, and present reduced risk to avian species.

Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

NASA Astrophysics Data System (ADS)

Ilham, Muhammad; Su'ud, Zaki

2017-01-01

Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

Investigation of the possibility of using residual heat reactor energy

NASA Astrophysics Data System (ADS)

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

2017-11-01

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

High density operation for reactor-relevant power exhaust

NASA Astrophysics Data System (ADS)

Wischmeier, M.; ASDEX Upgrade Team; Jet Efda Contributors

2015-08-01

With increasing size of a tokamak device and associated fusion power gain an increasing power flux density towards the divertor needs to be handled. A solution for handling this power flux is crucial for a safe and economic operation. Using purely geometric arguments in an ITER-like divertor this power flux can be reduced by approximately a factor 100. Based on a conservative extrapolation of current technology for an integrated engineering approach to remove power deposited on plasma facing components a further reduction of the power flux density via volumetric processes in the plasma by up to a factor of 50 is required. Our current ability to interpret existing power exhaust scenarios using numerical transport codes is analyzed and an operational scenario as a potential solution for ITER like divertors under high density and highly radiating reactor-relevant conditions is presented. Alternative concepts for risk mitigation as well as strategies for moving forward are outlined.

Improving proliferation resistance of high breeding gain generation 4 reactors using blankets composed of light water reactor waste

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

Hellesen, C.; Grape, S.; Haakanson, A.

2013-07-01

Fertile blankets can be used in fast reactors to enhance the breeding gain as well as the passive safety characteristics. However, such blankets typically result in the production of weapons grade plutonium. For this reason they are often excluded from Generation IV reactor designs. In this paper we demonstrate that using blankets manufactured directly from spent light water (LWR) reactor fuel it is possible to produce a plutonium product with non-proliferation characteristics on a par with spent LWR fuel of 30-50 MWd/kg burnup. The beneficial breeding and safety characteristics are retained. (authors)

Nuclear reactor descriptions for space power systems analysis

NASA Technical Reports Server (NTRS)

Mccauley, E. W.; Brown, N. J.

1972-01-01

For the small, high performance reactors required for space electric applications, adequate neutronic analysis is of crucial importance, but in terms of computational time consumed, nuclear calculations probably yield the least amount of detail for mission analysis study. It has been found possible, after generation of only a few designs of a reactor family in elaborate thermomechanical and nuclear detail to use simple curve fitting techniques to assure desired neutronic performance while still performing the thermomechanical analysis in explicit detail. The resulting speed-up in computation time permits a broad detailed examination of constraints by the mission analyst.

Fast reactor power plant design having heat pipe heat exchanger

DOEpatents

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

1984-08-30

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

Fast reactor power plant design having heat pipe heat exchanger

DOEpatents

Huebotter, Paul R.; McLennan, George A.

1985-01-01

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

77 FR 74697 - Meeting of the ACRS, Subcommittee on U.S. Evolutionary Power Reactor; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2012-12-17

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS, Subcommittee on U.S. Evolutionary Power Reactor; Notice of Meeting The ACRS Subcommittee on U.S. Evolutionary Power Reactor (U.S. EPR) will hold a meeting on January 17, 2013, Room T-2B1, 11545 Rockville Pike...

Electrical Power Conversion of a River and Tidal Power Generator: Preprint

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

Muljadi, Eduard; Gevorgian, Vahan; Wright, Alan

As renewable generation has become less expensive during recent decades, and it becomes more accepted by the global population, the focus on renewable generation has expanded to include new types with promising future applications, such as river and tidal generation. Although the utilization of power electronics and electric machines in industry is phenomenal, the emphasis on system design is different for various sectors of industry. In precision control, robotics, and weaponry, the design emphasis is on accuracy and reliability with less concern for the cost of the final product. In energy generation, the cost of energy is the prime concern;more » thus, capital expenditures (CAPEX) and operations and maintenance expenditures (OPEX) are the major design objectives. This paper describes the electrical power conversion aspects of river and tidal generation. Although modern power converter control is available to control the generation side, the design was chosen on the bases of minimizing the CAPEX and OPEX; thus, the architecture is simple and modular for ease of replacement and maintenance. The power conversion is simplified by considering a simple diode bridge and a DC-DC power converter to take advantage of abundant and low-cost photovoltaic inverters that have well-proven grid integration characteristics (i.e., the capability to produce energy with good power quality and control real power and voltage on the grid side).« less

Computer simulation of magnetization-controlled shunt reactors for calculating electromagnetic transients in power systems

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

Karpov, A. S.

2013-01-15

A computer procedure for simulating magnetization-controlled dc shunt reactors is described, which enables the electromagnetic transients in electric power systems to be calculated. It is shown that, by taking technically simple measures in the control system, one can obtain high-speed reactors sufficient for many purposes, and dispense with the use of high-power devices for compensating higher harmonic components.

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.

Integrated engine generator for aircraft secondary power

NASA Technical Reports Server (NTRS)

Secunde, R. R.

1972-01-01

An integrated engine-generator for aircraft secondary power generation is described. The concept consists of an electric generator located inside a turbojet or turbofan engine and both concentric with and driven by one of the main engine shafts. The electric power conversion equipment and generator controls are located in the aircraft. When properly rated, the generator serves as an engine starter as well as a source of electric power. This configuration reduces or eliminates the need for an external gear box on the engine and permits reduction in the nacelle diameter.

Prospects for the use of SMR and IGCC technologies for power generation in Poland

NASA Astrophysics Data System (ADS)

Wyrwa, Artur; Suwała, Wojciech

2017-11-01

This study is a preliminary assessment of prospects for new power generation technologies that are of particular interest in Poland. We analysed the economic competitiveness of small size integrated gasification combined cycle units (IGCC) and small modular reactors (SMR). For comparison we used one of the most widely applied and universal metric i.e. Levelized Cost of Electricity (LCOE). The LCOE results were complemented with the results of energy-economic model TIMES-PL in order to analyse the economic viability of these technologies under operation regime of the entire power system. The results show that with techno-economic assumptions presented in the paper SMRs are more competitive option as compared to small IGCC units.

Developing the European Center of Competence on VVER-type nuclear power reactors

NASA Astrophysics Data System (ADS)

Geraskin, Nikolay; Pironkov, Lyubomir; Kulikov, Evgeny; Glebov, Vasily

2017-09-01

This paper presents the results of the European educational projects CORONA and CORONA-II which are dedicated to preserving and further developing nuclear knowledge and competencies in the area of VVER-type nuclear power reactors technologies (Water-Water Energetic Reactor, WWER or VVER). The development of the European Center of Competence for VVER-technology is focused on master's degree programmes. The specifics of a systematic approach to training in the area of VVER-type nuclear power reactors technologies are analysed. This paper discusses enhancement of the training opportunities of the European Center that have arisen from advances in methodology and distance education. With a special attention paid to the European Nuclear Education Network (ENEN), the possibilities of further development of the international cooperation between European countries and educational institutions are examined.

Assessment and mitigation of power quality problems for PUSPATI TRIGA Reactor (RTP)

NASA Astrophysics Data System (ADS)

Zakaria, Mohd Fazli; Ramachandaramurthy, Vigna K.

2017-01-01

An electrical power systems are exposed to different types of power quality disturbances. Investigation and monitoring of power quality are necessary to maintain accurate operation of sensitive equipment especially for nuclear installations. This paper will discuss the power quality problems observed at the electrical sources of PUSPATI TRIGA Reactor (RTP). Assessment of power quality requires the identification of any anomalous behavior on a power system, which adversely affects the normal operation of electrical or electronic equipment. A power quality assessment involves gathering data resources; analyzing the data (with reference to power quality standards) then, if problems exist, recommendation of mitigation techniques must be considered. Field power quality data is collected by power quality recorder and analyzed with reference to power quality standards. Normally the electrical power is supplied to the RTP via two sources in order to keep a good reliability where each of them is designed to carry the full load. The assessment of power quality during reactor operation was performed for both electrical sources. There were several disturbances such as voltage harmonics and flicker that exceeded the thresholds. To reduce these disturbances, mitigation techniques have been proposed, such as to install passive harmonic filters to reduce harmonic distortion, dynamic voltage restorer (DVR) to reduce voltage disturbances and isolate all sensitive and critical loads.

Performance Analyses of 38 kWe Turbo-Machine Unit for Space Reactor Power Systems

NASA Astrophysics Data System (ADS)

Gallo, Bruno M.; El-Genk, Mohamed S.

2008-01-01

This paper developed a design and investigated the performance of 38 kWe turbo-machine unit for space nuclear reactor power systems with Closed Brayton Cycle (CBC) energy conversion. The compressor and turbine of this unit are scaled versions of the NASA's BRU developed in the sixties and seventies. The performance results of turbo-machine unit are calculated for rotational speed up to 45 krpm, variable reactor thermal power and system pressure, and fixed turbine and compressor inlet temperatures of 1144 K and 400 K. The analyses used a detailed turbo-machine model developed at the University of New Mexico that accounts for the various energy losses in the compressor and turbine and the effect of compressibility of the He-Xe (40 mole/g) working fluid with increased flow rate. The model also accounts for the changes in the physical and transport properties of the working fluid with temperature and pressure. Results show that a unit efficiency of 24.5% is achievable at rotation speed of 45 krpm and system pressure of 0.75 MPa, assuming shaft and electrical generator efficiencies of 86.7% and 90%. The corresponding net electric power output of the unit is 38.5 kWe, the flow rate of the working fluid is 1.667 kg/s, the pressure ratio and polytropic efficiency for the compressor are 1.60 and 83.1%, and 1.51 and 88.3% for the turbine.

18 CFR 801.12 - Electric power generation.

Code of Federal Regulations, 2012 CFR

2012-04-01

... 18 Conservation of Power and Water Resources 2 2012-04-01 2012-04-01 false Electric power... COMMISSION GENERAL POLICIES § 801.12 Electric power generation. (a) Significant uses are presently being made of the waters of the basin for the generation of electric power at hydro, pumped storage, and...

18 CFR 801.12 - Electric power generation.

Code of Federal Regulations, 2010 CFR

2010-04-01

... 18 Conservation of Power and Water Resources 2 2010-04-01 2010-04-01 false Electric power... COMMISSION GENERAL POLICIES § 801.12 Electric power generation. (a) Significant uses are presently being made of the waters of the basin for the generation of electric power at hydro, pumped storage, and...

18 CFR 801.12 - Electric power generation.

Code of Federal Regulations, 2011 CFR

2011-04-01

... 18 Conservation of Power and Water Resources 2 2011-04-01 2011-04-01 false Electric power... COMMISSION GENERAL POLICIES § 801.12 Electric power generation. (a) Significant uses are presently being made of the waters of the basin for the generation of electric power at hydro, pumped storage, and...

18 CFR 801.12 - Electric power generation.

Code of Federal Regulations, 2014 CFR

2014-04-01

... 18 Conservation of Power and Water Resources 2 2014-04-01 2014-04-01 false Electric power... COMMISSION GENERAL POLICIES § 801.12 Electric power generation. (a) Significant uses are presently being made of the waters of the basin for the generation of electric power at hydro, pumped storage, and...

18 CFR 801.12 - Electric power generation.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 18 Conservation of Power and Water Resources 2 2013-04-01 2012-04-01 true Electric power... COMMISSION GENERAL POLICIES § 801.12 Electric power generation. (a) Significant uses are presently being made of the waters of the basin for the generation of electric power at hydro, pumped storage, and...

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.

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

On the possibility of connecting a non-operating main circulation pump with three pumps in operation without preliminary coast-down of power-generating unit No. 5 in the Novovoronezh nuclear power plant

NASA Astrophysics Data System (ADS)

Vitkovskii, I. L.; Nikonov, S. P.; Ryasnyi, S. I.

2014-02-01

The subject of this paper is a transient caused by connection of a standby loop to three operating circulation pumps at the initial reactor heat rate equal to 70% of the rated value without preliminarily reducing it to 30% of the rated level as required by the safe operation regulations. Failure of the following normal operation systems is supposed: the first- and the second-type warning protection systems, all quick-acting reducing devices releasing steam into the auxiliary manifold, the electric heaters of the pressurizer, the pressurizer injection system, the primary cooling circuit fluid makeup/blow-through systems, and the blocking systems to shut down the main circulation pump after the level in the steam generator is exceeded. In addition, it is supposed that, under transient conditions, the valves of the turbine regulation system will be in the position in which they were at the moment of the initial event until generation of the signal for positive closing of the turbine stop valves. The first signal to actuate the reactor emergency protection system (EPS) is skipped. The failure of all quick-acting reducing devices releasing steam into the atmosphere is assumed. In addition to equipment failure, at the moment when the main circulation pump is connected, the operator erroneously puts in a new setting to maintain the power allowable for four pumps in operation-in the calculations it was taken equal to 104% of the rated level at most considering the accuracy of evaluating and maintaining the reactor heat rate-and the working group of the reactor protection and control system (P&CS) starts moving upward. On reaching the set power level, the automatic reactor power regulator stops operating and the P&CS elements remain in the position in which they are at the moment. Compliance with the design safety criteria for the adopted scenario of the transient is demonstrated.

Gas core reactors for actinide transmutation and breeder applications

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.

1978-01-01

This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions.

Magnetohydrodynamic power generation

NASA Technical Reports Server (NTRS)

Smith, J. L.

1984-01-01

Magnetohydrodynamic (MHD) Power Generation is a concise summary of MHD theory, history, and future trends. Results of the major international MHD research projects are discussed. Data from MHD research is included. Economics of initial and operating costs are considered.

The IRIS Spool-Type Reactor Coolant Pump

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

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

2002-07-01

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

Developing the European Center of Competence on VVER-Type Nuclear Power Reactors

ERIC Educational Resources Information Center

Geraskin, Nikolay; Pironkov, Lyubomir; Kulikov, Evgeny; Glebov, Vasily

2017-01-01

This paper presents the results of the European educational projects CORONA and CORONA-II which are dedicated to preserving and further developing nuclear knowledge and competencies in the area of VVER-type nuclear power reactors technologies (Water-Water Energetic Reactor, WWER or VVER). The development of the European Center of Competence for…

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

NASA Astrophysics Data System (ADS)

Koshelev, A. S.; Kovshov, K. N.; Ovchinnikov, M. A.; Pikulina, G. N.; Sokolov, A. B.

2016-12-01

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

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

Koshelev, A. S., E-mail: alexsander.coshelev@yandex.ru; Kovshov, K. N.; Ovchinnikov, M. A.

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

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

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.

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.

Innovative self-powered submersible microbial electrolysis cell (SMEC) for biohydrogen production from anaerobic reactors.

PubMed

Zhang, Yifeng; Angelidaki, Irini

2012-05-15

A self-powered submersible microbial electrolysis cell (SMEC), in which a specially designed anode chamber and external electricity supply were not needed, was developed for in situ biohydrogen production from anaerobic reactors. In batch experiments, the hydrogen production rate reached 17.8 mL/L/d at the initial acetate concentration of 410 mg/L (5 mM), while the cathodic hydrogen recovery ( [Formula: see text] ) and overall systemic coulombic efficiency (CE(os)) were 93% and 28%, respectively, and the systemic hydrogen yield ( [Formula: see text] ) peaked at 1.27 mol-H(2)/mol-acetate. The hydrogen production increased along with acetate and buffer concentration. The highest hydrogen production rate of 32.2 mL/L/d and [Formula: see text] of 1.43 mol-H(2)/mol-acetate were achieved at 1640 mg/L (20 mM) acetate and 100 mM phosphate buffer. Further evaluation of the reactor under single electricity-generating or hydrogen-producing mode indicated that further improvement of voltage output and reduction of electron losses were essential for efficient hydrogen generation. In addition, alternate exchanging the electricity-assisting and hydrogen-producing function between the two cell units of the SMEC was found to be an effective approach to inhibit methanogens. Furthermore, 16S rRNA genes analysis showed that this special operation strategy resulted same microbial community structures in the anodic biofilms of the two cell units. The simple, compact and in situ applicable SMEC offers new opportunities for reactor design for a microbial electricity-assisted biohydrogen production system. Copyright © 2012 Elsevier Ltd. All rights reserved.

Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors

NASA Astrophysics Data System (ADS)

Guidez, Joel; Saturnin, Anne

2017-11-01

During the operation of a nuclear reactor, the external individual doses received by the personnel are measured and recorded, in conformity with the regulations in force. The sum of these measurements enables an evaluation of the annual collective dose expressed in man·Sv/year. This information is a useful tool when comparing the different design types and reactors. This article discusses the evolution of the collective dose for several types of reactors, mainly based on publications from the NEA and the IAEA. The spread of good practices (optimization of working conditions and of the organization, sharing of lessons learned, etc.) and ongoing improvements in reactor design have meant that over time, the doses of various origins received by the personnel have decreased. In the case of sodium-cooled fast reactors (SFRs), the compilation and summarizing of various documentary resources has enabled them to be situated and compared to other types of reactors of the second and third generations (respectively pressurized water reactors in operation and EPR under construction). From these results, it can be seen that the doses received during the operation of SFR are significantly lower for this type of reactor.

SNAP (Space Nuclear Auxiliary Power) Reactor Overview

DTIC Science & Technology

1984-08-01

so that emphasis could be placed on the development of the space shuttle and the national space station . During 1969 NASA came up with a requirement...which would need the Zr-H reactor system which was the semipermanent orbiting space station . This helped the Zr-H system weather through the major FY 71...provide power for advanced space missions, such as lunar stations or orbiting space platforms, and for interplanetary com- munications. In addition

76 FR 78173 - Options for Developing the Regulatory Basis for Streamlining Non-Power Reactor License Renewal...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-16

... Emergency Preparedness AGENCY: Nuclear Regulatory Commission. ACTION: Notice of public meeting. SUMMARY: The... non-power reactor license renewal and non-power reactor emergency preparedness. This meeting is a... potential enhancements to emergency preparedness requirements. This meeting is open to the public. DATES...

The effectiveness of power-generating complexes constructed on the basis of nuclear power plants combined with additional sources of energy determined taking risk factors into account

NASA Astrophysics Data System (ADS)

Aminov, R. Z.; Khrustalev, V. A.; Portyankin, A. V.

2015-02-01

The effectiveness of combining nuclear power plants equipped with water-cooled water-moderated power-generating reactors (VVER) with other sources of energy within unified power-generating complexes is analyzed. The use of such power-generating complexes makes it possible to achieve the necessary load pickup capability and flexibility in performing the mandatory selective primary and emergency control of load, as well as participation in passing the night minimums of electric load curves while retaining high values of the capacity utilization factor of the entire power-generating complex at higher levels of the steam-turbine part efficiency. Versions involving combined use of nuclear power plants with hydrogen toppings and gas turbine units for generating electricity are considered. In view of the fact that hydrogen is an unsafe energy carrier, the use of which introduces additional elements of risk, a procedure for evaluating these risks under different conditions of implementing the fuel-and-hydrogen cycle at nuclear power plants is proposed. Risk accounting technique with the use of statistical data is considered, including the characteristics of hydrogen and gas pipelines, and the process pipelines equipment tightness loss occurrence rate. The expected intensities of fires and explosions at nuclear power plants fitted with hydrogen toppings and gas turbine units are calculated. In estimating the damage inflicted by events (fires and explosions) occurred in nuclear power plant turbine buildings, the US statistical data were used. Conservative scenarios of fires and explosions of hydrogen-air mixtures in nuclear power plant turbine buildings are presented. Results from calculations of the introduced annual risk to the attained net annual profit ratio in commensurable versions are given. This ratio can be used in selecting projects characterized by the most technically attainable and socially acceptable safety.

Application of Molten Salt Reactor Technology to MMW In-Space NEP and Surface Power Missions

NASA Technical Reports Server (NTRS)

Patton, Bruce; Sorensen, Kirk; Rodgers, Stephen (Technical Monitor)

2002-01-01

Anticipated manned nuclear electric propulsion (NEP) and planetary surface power missions will require multimegawatt nuclear reactors that are lightweight, operationally robust, and scalable in power for widely varying scientific mission objectives. Molten salt reactor technology meets all of these requirements and offers an interesting alternative to traditional multimegawatt gas-cooled and liquid metal concepts.

Coal and Coal/Biomass-Based Power Generation

EPA Science Inventory

For Frank Princiotta's book, Global Climate Change--The Technology Challenge Coal is a key, growing component in power generation globally. It generates 50% of U.S. electricity, and criteria emissions from coal-based power generation are being reduced. However, CO2 emissions m...

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

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

Yehia, Ashraf; Mizuno, Akira

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

Solar energy thermally powered electrical generating system

NASA Technical Reports Server (NTRS)

Owens, William R. (Inventor)

1989-01-01

A thermally powered electrical generating system for use in a space vehicle is disclosed. The rate of storage in a thermal energy storage medium is controlled by varying the rate of generation and dissipation of electrical energy in a thermally powered electrical generating system which is powered from heat stored in the thermal energy storage medium without exceeding a maximum quantity of heat. A control system (10) varies the rate at which electrical energy is generated by the electrical generating system and the rate at which electrical energy is consumed by a variable parasitic electrical load to cause storage of an amount of thermal energy in the thermal energy storage system at the end of a period of insolation which is sufficient to satisfy the scheduled demand for electrical power to be generated during the next period of eclipse. The control system is based upon Kalman filter theory.

POWER-BURST FACILITY (PBF) CONCEPTUAL DESIGN

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

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

1963-06-21

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

Radio-frequency power-assisted performance improvement of a magnetohydrodynamic power generator

NASA Astrophysics Data System (ADS)

Murakami, Tomoyuki; Okuno, Yoshihiro; Yamasaki, Hiroyuki

2005-12-01

We describe a radio-frequency (rf) electromagnetic-field-assisted magnetohydrodynamic power generation experiment, where an inductively coupled rf field (13.56MHz, 5.2kW) is continuously supplied to the disk generator. The rf power assists the precise plasma ignition, by which the otherwise irregular plasma behavior was stabilized. The rf heating suppresses the ionization instability in the plasma behavior and homogenizes the nonuniformity of the plasma structures. The power-generating performance is significantly improved with the aid of the rf power under wide seeding conditions: insufficient, optimum, and excessive seed fractions. The increment of the enthalpy extraction ratio of around 2% is significantly greater than the fraction of the net rf power, that is, 0.16%, to the thermal input.

Power generator driven by Maxwell's demon

NASA Astrophysics Data System (ADS)

Chida, Kensaku; Desai, Samarth; Nishiguchi, Katsuhiko; Fujiwara, Akira

2017-05-01

Maxwell's demon is an imaginary entity that reduces the entropy of a system and generates free energy in the system. About 150 years after its proposal, theoretical studies explained the physical validity of Maxwell's demon in the context of information thermodynamics, and there have been successful experimental demonstrations of energy generation by the demon. The demon's next task is to convert the generated free energy to work that acts on the surroundings. Here, we demonstrate that Maxwell's demon can generate and output electric current and power with individual randomly moving electrons in small transistors. Real-time monitoring of electron motion shows that two transistors functioning as gates that control an electron's trajectory so that an electron moves directionally. A numerical calculation reveals that power generation is increased by miniaturizing the room in which the electrons are partitioned. These results suggest that evolving transistor-miniaturization technology can increase the demon's power output.

Preliminary Comparison of Radioactive Waste Disposal Cost for Fusion and Fission Reactors

NASA Astrophysics Data System (ADS)

Seki, Yasushi; Aoki, Isao; Yamano, Naoki; Tabara, Takashi

1997-09-01

The environmental and economic impact of radioactive waste (radwaste) generated from fusion power reactors using five types of structural materials and a fission reactor has been evaluated and compared. Possible radwaste disposal scenario of fusion radwaste in Japan is considered. The exposure doses were evaluated for the skyshine of gamma-ray during the disposal operation, groundwater migration scenario during the institutional control period of 300 years and future site use scenario after the institutional period. The radwaste generated from a typical light water fission reactor was evaluated using the same methodology as for the fusion reactors. It is found that radwaste from the fusion reactors using F82H and SiC/SiC composites without impurities could be disposed by the shallow land disposal presently applied to the low level waste in Japan. The disposal cost of radwaste from five fusion power reactors and a typical light water reactor were roughly evaluated and compared.

Peak power ratio generator

DOEpatents

Moyer, Robert D.

1985-01-01

A peak power ratio generator is described for measuring, in combination with a conventional power meter, the peak power level of extremely narrow pulses in the gigahertz radio frequency bands. The present invention in a preferred embodiment utilizes a tunnel diode and a back diode combination in a detector circuit as the only high speed elements. The high speed tunnel diode provides a bistable signal and serves as a memory device of the input pulses for the remaining, slower components. A hybrid digital and analog loop maintains the peak power level of a reference channel at a known amount. Thus, by measuring the average power levels of the reference signal and the source signal, the peak power level of the source signal can be determined.

Peak power ratio generator

DOEpatents

Moyer, R.D.

A peak power ratio generator is described for measuring, in combination with a conventional power meter, the peak power level of extremely narrow pulses in the gigahertz radio frequency bands. The present invention in a preferred embodiment utilizes a tunnel diode and a back diode combination in a detector circuit as the only high speed elements. The high speed tunnel diode provides a bistable signal and serves as a memory device of the input pulses for the remaining, slower components. A hybrid digital and analog loop maintains the peak power level of a reference channel at a known amount. Thus, by measuring the average power levels of the reference signal and the source signal, the peak power level of the source signal can be determined.

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

NASA Technical Reports Server (NTRS)

Anghaie, S.; Chen, G.

1996-01-01

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

Magnetic Flux Compression Reactor Concepts for Spacecraft Propulsion and Power (MSFC Center Director's Discretionary Fund; Project No. 99-24). Part 1

NASA Technical Reports Server (NTRS)

Litchford, R. J.; Robertson, G. A.; Hawk, C. W.; Turner, M. W.; Koelfgen, S.; Litchford, Ron J. (Technical Monitor)

2001-01-01

This technical publication (TP) examines performance and design issues associated with magnetic flux compression reactor concepts for nuclear/chemical pulse propulsion and power. Assuming that low-yield microfusion detonations or chemical detonations using high-energy density matter can eventually be realized in practice, various magnetic flux compression concepts are conceivable. In particular, reactors in which a magnetic field would be compressed between an expanding detonation-driven plasma cloud and a stationary structure formed from a high-temperature superconductor are envisioned. Primary interest is accomplishing two important functions: (1) Collimation and reflection of a hot diamagnetic plasma for direct thrust production, and (2) electric power generation for fusion standoff drivers and/or dense plasma formation. In this TP, performance potential is examined, major technical uncertainties related to this concept accessed, and a simple performance model for a radial-mode reactor developed. Flux trapping effectiveness is analyzed using a skin layer methodology, which accounts for magnetic diffusion losses into the plasma armature and the stationary stator. The results of laboratory-scale experiments on magnetic diffusion in bulk-processed type II superconductors are also presented.

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...

BOILER-SUPERHEATED REACTOR

DOEpatents

Heckman, T.P.

1961-05-01

A nuclear power reactor of the type in which a liquid moderator-coolant is transformed by nuclear heating into a vapor that may be used to drive a turbo- generator is described. The core of this reactor comprises a plurality of freely suspended tubular fuel elements, called fuel element trains, within which nonboiling pressurized liquid moderator-coolant is preheated and sprayed through orifices in the walls of the trains against the outer walls thereof to be converted into vapor. Passage of the vapor ovcr other unwetted portions of the outside of the fuel elements causes the steam to be superheated. The moderatorcoolant within the fuel elements remains in the liqUid state, and that between the fuel elements remains substantiaily in the vapor state. A unique liquid neutron-absorber control system is used. Advantages expected from the reactor design include reduced fuel element failure, increased stability of operation, direct response to power demand, and circulation of a minimum amount of liquid moderatorcoolant. (A.G.W.)

Forecasting Electric Power Generation of Photovoltaic Power System for Energy Network

NASA Astrophysics Data System (ADS)

Kudo, Mitsuru; Takeuchi, Akira; Nozaki, Yousuke; Endo, Hisahito; Sumita, Jiro

Recently, there has been an increase in concern about the global environment. Interest is growing in developing an energy network by which new energy systems such as photovoltaic and fuel cells generate power locally and electric power and heat are controlled with a communications network. We developed the power generation forecast method for photovoltaic power systems in an energy network. The method makes use of weather information and regression analysis. We carried out forecasting power output of the photovoltaic power system installed in Expo 2005, Aichi Japan. As a result of comparing measurements with a prediction values, the average prediction error per day was about 26% of the measured power.

Solar power generation system for reducing leakage current

NASA Astrophysics Data System (ADS)

Wu, Jinn-Chang; Jou, Hurng-Liahng; Hung, Chih-Yi

2018-04-01

This paper proposes a transformer-less multi-level solar power generation system. This solar power generation system is composed of a solar cell array, a boost power converter, an isolation switch set and a full-bridge inverter. A unipolar pulse-width modulation (PWM) strategy is used in the full-bridge inverter to attenuate the output ripple current. Circuit isolation is accomplished by integrating the isolation switch set between the solar cell array and the utility, to suppress the leakage current. The isolation switch set also determines the DC bus voltage for the full-bridge inverter connecting to the solar cell array or the output of the boost power converter. Accordingly, the proposed transformer-less multi-level solar power generation system generates a five-level voltage, and the partial power of the solar cell array is also converted to AC power using only the full-bridge inverter, so the power efficiency is increased. A prototype is developed to validate the performance of the proposed transformer-less multi-level solar power generation system.

Documentation for MeshKit - Reactor Geometry (&mesh) Generator

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

Jain, Rajeev; Mahadevan, Vijay

2015-09-30

This report gives documentation for using MeshKit’s Reactor Geometry (and mesh) Generator (RGG) GUI and also briefly documents other algorithms and tools available in MeshKit. RGG is a program designed to aid in modeling and meshing of complex/large hexagonal and rectilinear reactor cores. RGG uses Argonne’s SIGMA interfaces, Qt and VTK to produce an intuitive user interface. By integrating a 3D view of the reactor with the meshing tools and combining them into one user interface, RGG streamlines the task of preparing a simulation mesh and enables real-time feedback that reduces accidental scripting mistakes that could waste hours of meshing.more » RGG interfaces with MeshKit tools to consolidate the meshing process, meaning that going from model to mesh is as easy as a button click. This report is designed to explain RGG v 2.0 interface and provide users with the knowledge and skills to pilot RGG successfully. Brief documentation of MeshKit source code, tools and other algorithms available are also presented for developers to extend and add new algorithms to MeshKit. RGG tools work in serial and parallel and have been used to model complex reactor core models consisting of conical pins, load pads, several thousands of axially varying material properties of instrumentation pins and other interstices meshes.« less

Generation-IV Nuclear Energy Systems

NASA Astrophysics Data System (ADS)

McFarlane, Harold

2008-05-01

Nuclear power technology has evolved through roughly three generations of system designs: a first generation of prototypes and first-of-a-kind units implemented during the period 1950 to 1970; a second generation of industrial power plants built from 1970 to the turn of the century, most of which are still in operation today; and a third generation of evolutionary advanced reactors which began being built by the turn of the 20^th century, usually called Generation III or III+, which incorporate technical lessons learned through more than 12,000 reactor-years of operation. The Generation IV International Forum (GIF) is a cooperative international endeavor to develop advanced nuclear energy systems in response to the social, environmental and economic requirements of the 21^st century. Six Generation IV systems under development by GIF promise to enhance the future contribution and benefits of nuclear energy. All Generation IV systems aim at performance improvement, new applications of nuclear energy, and/or more sustainable approaches to the management of nuclear materials. High-temperature systems offer the possibility of efficient process heat applications and eventually hydrogen production. Enhanced sustainability is achieved primarily through adoption of a closed fuel cycle with reprocessing and recycling of plutonium, uranium and minor actinides using fast reactors. This approach provides significant reduction in waste generation and uranium resource requirements.

Nanostructured Bulk Thermoelectric Generator for Efficient Power Harvesting for Self-powered Sensor Networks

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

Zhang, Yanliang; Butt, Darryl; Agarwal, Vivek

2015-07-01

The objective of this Nuclear Energy Enabling Technology research project is to develop high-efficiency and reliable thermoelectric generators for self-powered wireless sensors nodes utilizing thermal energy from nuclear plant or fuel cycle. The power harvesting technology has crosscutting significance to address critical technology gaps in monitoring nuclear plants and fuel cycle. The outcomes of the project will lead to significant advancement in sensors and instrumentation technology, reducing cost, improving monitoring reliability and therefore enhancing safety. The self-powered wireless sensor networks could support the long-term safe and economical operation of all the reactor designs and fuel cycle concepts, as well asmore » spent fuel storage and many other nuclear science and engineering applications. The research is based on recent breakthroughs in high-performance nanostructured bulk (nanobulk) thermoelectric materials that enable high-efficiency direct heat-to-electricity conversion over a wide temperature range. The nanobulk thermoelectric materials that the research team at Boise State University and University of Houston has developed yield up to a 50% increase in the thermoelectric figure of merit, ZT, compared with state-of-the-art bulk counterparts. This report focuses on the selection of optimal thermoelectric materials for this project. The team has performed extensive study on two thermoelectric materials systems, i.e. the half-Heusler materials, and the Bismuth-Telluride materials. The report contains our recent research results on the fabrication, characterization and thermoelectric property measurements of these two materials.« less

Potential Application of a Thermoelectric Generator in Passive Cooling System of Nuclear Power Plants

NASA Astrophysics Data System (ADS)

Wang, Dongqing; Liu, Yu; Jiang, Jin; Pang, Wei; Lau, Woon Ming; Mei, Jun

2017-05-01

In the design of nuclear power plants, various natural circulation passive cooling systems are considered to remove residual heat from the reactor core in the event of a power loss and maintain the plant's safety. These passive systems rely on gravity differences of fluids, resulting from density differentials, rather than using an external power-driven system. Unfortunately, a major drawback of such systems is their weak driving force, which can negatively impact safety. In such systems, there is a temperature difference between the heat source and the heat sink, which potentially offers a natural platform for thermoelectric generator (TEG) applications. While a previous study designed and analyzed a TEG-based passive core cooling system, this paper considers TEG applications in other passive cooling systems of nuclear power plants, after which the concept of a TEG-based passive cooling system is proposed. In such a system, electricity is produced using the system's temperature differences through the TEG, and this electricity is used to further enhance the cooling process.

Hydrogen Generation by Koh-Ethanol Plasma Electrolysis Using Double Compartement Reactor

NASA Astrophysics Data System (ADS)

Saksono, Nelson; Sasiang, Johannes; Dewi Rosalina, Chandra; Budikania, Trisutanti

2018-03-01

This study has successfully investigated the generation of hydrogen using double compartment reactor with plasma electrolysis process. Double compartment reactor is designed to achieve high discharged voltage, high concentration, and also reduce the energy consumption. The experimental results showed the use of double compartment reactor increased the productivity ratio 90 times higher compared to Faraday electrolysis process. The highest hydrogen production obtained is 26.50 mmol/min while the energy consumption can reach up 1.71 kJ/mmol H2 at 0.01 M KOH solution. It was shown that KOH concentration, addition of ethanol, cathode depth, and temperature have important effects on hydrogen production, energy consumption, and process efficiency.

Novel, Integrated Reactor/Power Conversion System (LMR-AMTEC)

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

Dmitry V. Paramonov, Lead Collaborator

2001-07-31

The overall objective of NERI Project Number 99-0198 is to assess the technical and economic feasibility, develop engineering solutions and determine a range of potential applications for ''Novel Integrated Reactor/Energy conversion Systems''. The near term goal is the design of a power supply for developing countries in remote locations in a proliferation resistant, reliable and economical way. The heart of the concept is the use of a single loop liquid metal fast reactor (LMR) with conversion of the heat directly into electricity in a Alkali Metal Thermal to Electric Converter (AMTEC). The first year of the project focused on themore » feasibility issues with a long life, high temperature liquid metal-cooled core; selection of the working fluid, core-to-AMTEC coupling scheme and interface parameters; and, energy conversion systems design and performance. Report Number STD-ES-01-0028, Revision 0, dated July 31, 2001, summarizes the work performed by Westinghouse personnel in Year One and report number UNM-ISNPS-3-2000, dated October 2000, summarizes the work performed by the Institute for Space and Nuclear Power Studies at the University of New Mexico in Year One.« less

Evaluation Of Different Power Conditioning Options For Stirling Generators

NASA Astrophysics Data System (ADS)

Garrigos, A.; Blanes, J. M.; Carrasco, J. A.; Maset, E.; Montalban, G.; Ejea, J.; Ferreres, A.; Sanchis, E.

2011-10-01

Free-piston Stirling engines are an interesting alternative for electrical power systems, especially in deep space missions where photovoltaic systems are not feasible. This kind of power generators contains two main parts, the Stirling machine and the linear alternator that converts the mechanical energy from the piston movement to electrical energy. Since the generated power is in AC form, several aspects should be assessed to use such kind of generators in a spacecraft power system: AC/DC topologies, power factor correction, power regulation techniques, integration into the power system, etc. This paper details power generator operation and explores different power conversion approaches.

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.-

Piezoelectric power generation using friction-induced vibration

NASA Astrophysics Data System (ADS)

Tadokoro, Chiharu; Matsumoto, Aya; Nagamine, Takuo; Sasaki, Shinya

2017-06-01

In order to examine the feasibility of power generation by using friction-induced vibration with a piezoelectric element, we performed experiments and numerical analysis. In the experiments, the generated power in the piezoelectric element and the displacement of an oscillator were measured by a newly developed apparatus that embodied a single-degree-of-freedom (1-DOF) system with friction. In the numerical analysis, an analytical model of a 1-DOF system with friction and piezoelectric element was proposed to simulate the experiments. The experimental results demonstrated that the power of a few microwatts was generated by sliding between a steel ball and a steel plate lubricated with glycerol. In this study, a maximum power of approximately 10 μW was generated at a driving velocity of 40 mm s-1 and a normal load of 15 N. The numerical results demonstrated good qualitative agreement with the experimental results. This implies that this analytical model can be applied to optimize the oscillator design in piezoelectric power generation using friction-induced vibration.

Engine-integrated solid oxide fuel cells for efficient electrical power generation on aircraft

NASA Astrophysics Data System (ADS)

Waters, Daniel F.; Cadou, Christopher P.

2015-06-01

This work investigates the use of engine-integrated catalytic partial oxidation (CPOx) reactors and solid oxide fuel cells (SOFCs) to reduce fuel burn in vehicles with large electrical loads like sensor-laden unmanned air vehicles. Thermodynamic models of SOFCs, CPOx reactors, and three gas turbine (GT) engine types (turbojet, combined exhaust turbofan, separate exhaust turbofan) are developed and checked against relevant data and source material. Fuel efficiency is increased by 4% and 8% in the 50 kW and 90 kW separate exhaust turbofan systems respectively at only modest cost in specific power (8% and 13% reductions respectively). Similar results are achieved in other engine types. An additional benefit of hybridization is the ability to provide more electric power (factors of 3 or more in some cases) than generator-based systems before encountering turbine inlet temperature limits. A sensitivity analysis shows that the most important parameters affecting the system's performance are operating voltage, percent fuel oxidation, and SOFC assembly air flows. Taken together, this study shows that it is possible to create a GT-SOFC hybrid where the GT mitigates balance of plant losses and the SOFC raises overall system efficiency. The result is a synergistic system with better overall performance than stand-alone components.

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

NASA Astrophysics Data System (ADS)

Fic, Adam; Składzień, Jan; Gabriel, Michał

2015-03-01

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

Power Maximization Control of Variable Speed Wind Generation System Using Permanent Magnet Synchronous Generator

NASA Astrophysics Data System (ADS)

Morimoto, Shigeo; Nakamura, Tomohiko; Takeda, Yoji

This paper proposes the sensorless output power maximization control of the wind generation system. A permanent magnet synchronous generator (PMSG) is used as a variable speed generator in the proposed system. The generator torque is suitably controlled according to the generator speed and thus the power from a wind turbine settles down on the maximum power point by the proposed MPPT control method, where the information of wind velocity is not required. Moreover, the maximum available generated power is obtained by the optimum current vector control. The current vector of PMSG is optimally controlled according to the generator speed and the required torque in order to minimize the losses of PMSG considering the voltage and current constraints. The proposed wind power generation system can be achieved without mechanical sensors such as a wind velocity detector and a position sensor. Several experimental results show the effectiveness of the proposed control method.

Experimental power density distribution benchmark in the TRIGA Mark II reactor

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

Snoj, L.; Stancar, Z.; Radulovic, V.

2012-07-01

In order to improve the power calibration process and to benchmark the existing computational model of the TRIGA Mark II reactor at the Josef Stefan Inst. (JSI), a bilateral project was started as part of the agreement between the French Commissariat a l'energie atomique et aux energies alternatives (CEA) and the Ministry of higher education, science and technology of Slovenia. One of the objectives of the project was to analyze and improve the power calibration process of the JSI TRIGA reactor (procedural improvement and uncertainty reduction) by using absolutely calibrated CEA fission chambers (FCs). This is one of the fewmore » available power density distribution benchmarks for testing not only the fission rate distribution but also the absolute values of the fission rates. Our preliminary calculations indicate that the total experimental uncertainty of the measured reaction rate is sufficiently low that the experiments could be considered as benchmark experiments. (authors)« less

Hydrogen-based power generation from bioethanol steam reforming

NASA Astrophysics Data System (ADS)

Tasnadi-Asztalos, Zs.; Cormos, C. C.; Agachi, P. S.

2015-12-01

This paper is evaluating two power generation concepts based on hydrogen produced from bioethanol steam reforming at industrial scale without and with carbon capture. The power generation from bioethanol conversion is based on two important steps: hydrogen production from bioethanol catalytic steam reforming and electricity generation using a hydrogen-fuelled gas turbine. As carbon capture method to be assessed in hydrogen-based power generation from bioethanol steam reforming, the gas-liquid absorption using methyl-di-ethanol-amine (MDEA) was used. Bioethanol is a renewable energy carrier mainly produced from biomass fermentation. Steam reforming of bioethanol (SRE) provides a promising method for hydrogen and power production from renewable resources. SRE is performed at high temperatures (e.g. 800-900°C) to reduce the reforming by-products (e.g. ethane, ethene). The power generation from hydrogen was done with M701G2 gas turbine (334 MW net power output). Hydrogen was obtained through catalytic steam reforming of bioethanol without and with carbon capture. For the evaluated plant concepts the following key performance indicators were assessed: fuel consumption, gross and net power outputs, net electrical efficiency, ancillary consumptions, carbon capture rate, specific CO2 emission etc. As the results show, the power generation based on bioethanol conversion has high energy efficiency and low carbon footprint.

Ocean Current Power Generator. Final Report

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

O'Sullivan, G. A.

2002-07-26

The Ocean Power Generator is both technically and economically suitable for deployment in the Gulf Stream from the US Navy facility in Dania, Florida. Yet to be completed is the calibration test in the Chesapeake Bay with the prototype dual hydroturbine Underwater Electric Kite. For the production units a revised design includes two ballast tanks mounted as pontoons to provide buoyancy and depth control. The power rating of the Ocean Power Generator has been doubled to 200 kW ready for insertion into the utility grid. The projected cost for a 10 MW installation is $3.38 per watt, a cost thatmore » is consistent with wind power pricing when it was in its deployment infancy, and a cost that is far better than photovoltaics after 25 years of research and development. The Gulf Stream flows 24 hours per day, and water flow is both environmentally and ecologically perfect as a renewable energy source. No real estate purchases are necessary, and you cannot see, hear, smell, or touch an Ocean Power Generator.« less

Group Constants Generation of the Pseudo Fission Products for Fast Reactor Burnup Calculations

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

Gil, Choong-Sup; Kim, Do Heon; Chang, Jonghwa

The pseudo fission products for the burnup calculations of the liquid metal fast reactor were generated. The cross-section data and fission product yield data of ENDF/B-VI were used for the pseudo fission product data of U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242. The pseudo fission product data can be used with the KAFAX-F22 or -E66, which are the MATXS-format libraries for analyses of the liquid metal fast reactor at KAERI and were distributed through the OECD/NEA. The 80-group MATXS-format libraries of the 172 fission products were generated and the burnup chains for generation of the pseudo fission products were prepared.

Inherently Safe Fission Power System for Lunar Outposts

NASA Astrophysics Data System (ADS)

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

2013-09-01

This paper presents the Solid Core-Sectored Compact Reactor (SC-SCoRe) and power system for future lunar outposts. The power system nominally provides 38 kWe continuously for 21 years, employs static components and has no single point failures in reactor cooling or power generation. The reactor core has six sectors, each has a separate pair of primary and secondary loops with liquid NaK-56 working fluid, thermoelectric (TE) power conversion and heat-pipes radiator panels. The electromagnetic (EM) pumps in the primary and secondary loops, powered with separate TE power units, ensure operation reliability and passive decay heat removal from the reactor after shutdown. The reactor poses no radiological concerns during launch, and remains sufficiently subcritical, with the radial reflector dissembled, when submerged in wet sand and the core flooded with seawater, following a launch abort accident. After 300 years of storage below grade on the Moon, the total radioactivity in the post-operation reactor drops below 164 Ci, a low enough radioactivity for a recovery and safe handling of the reactor.

Combined fuel and air staged power generation system

DOEpatents

Rabovitser, Iosif K; Pratapas, John M; Boulanov, Dmitri

2014-05-27

A method and apparatus for generation of electric power employing fuel and air staging in which a first stage gas turbine and a second stage partial oxidation gas turbine power operated in parallel. A first portion of fuel and oxidant are provided to the first stage gas turbine which generates a first portion of electric power and a hot oxidant. A second portion of fuel and oxidant are provided to the second stage partial oxidation gas turbine which generates a second portion of electric power and a hot syngas. The hot oxidant and the hot syngas are provided to a bottoming cycle employing a fuel-fired boiler by which a third portion of electric power is generated.

Performance simulation of a plasma magnetohydrodynamic power generator

NASA Astrophysics Data System (ADS)

Huang, Hulin; Li, Linyong; Zhu, Guiping

2018-05-01

The performance of magnetohydrodynamic (MHD) power generator is affected by many issues, among which the load coefficient k is of great importance. This paper reveals the relationship between the k and the performance of MHD generator by numerical simulation on Faraday-type MHD power generator using He/Xe as working plasma. The results demonstrate that the power generation efficiency increases with an increment of the load factor. However, the enthalpy extraction firstly increases then decreases with the load factor increasing. The enthalpy extraction rate reaches the maximum when the load coefficient k equals to 0.625, which infers the best performance of the power generator channel with the maximum electricity production.

Light Water Reactor Sustainability Program Integrated Program Plan

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

McCarthy, Kathryn A.; Busby, Jeremy; Hallbert, Bruce

2014-04-01

Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution tomore » the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.« less

Electricity generation and transmission planning in deregulated power markets

NASA Astrophysics Data System (ADS)

He, Yang

This dissertation addresses the long-term planning of power generation and transmission facilities in a deregulated power market. Three models with increasing complexities are developed, primarily for investment decisions in generation and transmission capacity. The models are presented in a two-stage decision context where generation and transmission capacity expansion decisions are made in the first stage, while power generation and transmission service fees are decided in the second stage. Uncertainties that exist in the second stage affect the capacity expansion decisions in the first stage. The first model assumes that the electric power market is not constrained by transmission capacity limit. The second model, which includes transmission constraints, considers the interactions between generation firms and the transmission network operator. The third model assumes that the generation and transmission sectors make capacity investment decisions separately. These models result in Nash-Cournot equilibrium among the unregulated generation firms, while the regulated transmission network operator supports the competition among generation firms. Several issues in the deregulated electric power market can be studied with these models such as market powers of generation firms and transmission network operator, uncertainties of the future market, and interactions between the generation and transmission sectors. Results deduced from the developed models include (a) regulated transmission network operator will not reserve transmission capacity to gain extra profits; instead, it will make capacity expansion decisions to support the competition in the generation sector; (b) generation firms will provide more power supplies when there is more demand; (c) in the presence of future uncertainties, the generation firms will add more generation capacity if the demand in the future power market is expected to be higher; and (d) the transmission capacity invested by the

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.

The Birth of Nuclear-Generated Electricity

DOE R&D Accomplishments Database

1999-09-01

The Experimental Breeder Reactor-I (EBR-I), built in Idaho in 1949, generated the first usable electricity from nuclear power on December 20, 1951. More importantly, the reactor was used to prove that it was possible to create more nuclear fuel in the reactor than it consumed during operation -- fuel breeding. The EBR-I facility is now a National Historic Landmark open to the public.

Compensation for Harmonic Currents and Reactive Power in Wind Power Generation System using PWM Inverter

NASA Astrophysics Data System (ADS)

Shinohara, Katsuji; Shinhatsubo, Kurato; Iimori, Kenichi; Yamamoto, Kichiro; Saruban, Takamichi; Yamaemori, Takahiro

In recent year, consciousness of environmental problems is enhancing, and the price of the electric power purchased by an electric power company is established expensive for the power plant utilizing the natural energy. So, the introduction of the wind power generation is promoted in Japan. Generally, squirrel-cage induction machines are widely used as a generator in wind power generation system because of its small size, lightweight and low-cost. However, the induction machines do not have a source of excitation. Thus, it causes the inrush currents and the instantaneous voltage drop when the generator is directly connected to a power grid. To reduce the inrush currents, an AC power regulator is used. Wind power generations are frequently connected to and disconnected from the power grid. However, when the inrush currents are reduced, harmonic currents are caused by phase control of the AC power regulator. And the phase control of AC power regulator cannot control the power factor. Therefore, we propose the use of the AC power regulator to compensate for the harmonic currents and reactive power in the wind power generation system, and demonstrate the validity of its system by simulated and experimental results.

Silver Recovery and Power Generation from Ammonia Chelated Silver Solution in a Bio-Electrochemical Reactor

NASA Astrophysics Data System (ADS)

Ho, N. A. D.; Babel, S.

2017-06-01

Silver has valuable features and limited availability, and thus recovery from wastewater or aqueous solutions plays an important role in environmental protection and economic profits. In this study, silver recovery along with power generation and COD removal were investigated in a bio-electrochemical system (BES). The BES comprised of an anode and a cathode chamber which were separated by a cation exchange membrane to prevent the cross-over of electrolytes. During the biological oxidation of acetate as an electron donor in the anode chamber, the reduction of ammonia chelated silver ions as electron acceptors in the cathode side occurred spontaneously. Results showed that a silver recovery of 99% and COD removal efficiency of 60% were achieved at the initial silver concentration of 1,000 mg/L after 48 hours of operation. The power generation improved 4.66%, from 3,618 to 3,795 mW/m3, by adding NaNO3 of 850 mg/L to the catholyte containing 2,000 mg/L of silver ions. Deposits on the cathode surface were characterized using scanning electron microscope (SEM) and energy dispersive X-ray (EDX). Metallic silver with dendritic structures and high purity were detected. This study demonstrated that BES technology can be employed to recover silver from complex chelating solution, produce electricity, and treat wastewater.

46 CFR 111.10-4 - Power requirements, generating sources.

Code of Federal Regulations, 2010 CFR

2010-10-01

... ELECTRIC SYSTEMS-GENERAL REQUIREMENTS Power Supply § 111.10-4 Power requirements, generating sources. (a... generators which supply both ship's service and propulsion power do not need additional ship's service... 46 Shipping 4 2010-10-01 2010-10-01 false Power requirements, generating sources. 111.10-4 Section...

46 CFR 111.10-4 - Power requirements, generating sources.

Code of Federal Regulations, 2011 CFR

2011-10-01

... ELECTRIC SYSTEMS-GENERAL REQUIREMENTS Power Supply § 111.10-4 Power requirements, generating sources. (a... generators which supply both ship's service and propulsion power do not need additional ship's service... 46 Shipping 4 2011-10-01 2011-10-01 false Power requirements, generating sources. 111.10-4 Section...

Nuclear power: levels of safety.

PubMed

Lidsky, L M

1988-02-01

The rise and fall of the nuclear power industry in the United States is a well-documented story with enough socio-technological conflict to fill dozens of scholarly, and not so scholarly, books. Whatever the reasons for the situation we are now in, and no matter how we apportion the blame, the ultimate choice of whether to use nuclear power in this country is made by the utilities and by the public. Their choices are, finally, based on some form of risk-benefit analysis. Such analysis is done in well-documented and apparently logical form by the utilities and in a rather more inchoate but not necessarily less accurate form by the public. Nuclear power has failed in the United States because both the real and perceived risks outweigh the potential benefits. The national decision not to rely upon nuclear power in its present form is not an irrational one. A wide ranging public balancing of risk and benefit requires a classification of risk which is clear and believable for the public to be able to assess the risks associated with given technological structures. The qualitative four-level safety ladder provides such a framework. Nuclear reactors have been designed which fit clearly and demonstrably into each of the possible qualitative safety levels. Surprisingly, it appears that safer may also mean cheaper. The intellectual and technical prerequisites are in hand for an important national decision. Deployment of a qualitatively different second generation of nuclear reactors can have important benefits for the United States. Surprisingly, it may well be the "nuclear establishment" itself, with enormous investments of money and pride in the existing nuclear systems, that rejects second generation reactors. It may be that we will not have a second generation of reactors until the first generation of nuclear engineers and nuclear power advocates has retired.

Hydrogen-based power generation from bioethanol steam reforming

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

Tasnadi-Asztalos, Zs., E-mail: tazsolt@chem.ubbcluj.ro; Cormos, C. C., E-mail: cormos@chem.ubbcluj.ro; Agachi, P. S.

This paper is evaluating two power generation concepts based on hydrogen produced from bioethanol steam reforming at industrial scale without and with carbon capture. The power generation from bioethanol conversion is based on two important steps: hydrogen production from bioethanol catalytic steam reforming and electricity generation using a hydrogen-fuelled gas turbine. As carbon capture method to be assessed in hydrogen-based power generation from bioethanol steam reforming, the gas-liquid absorption using methyl-di-ethanol-amine (MDEA) was used. Bioethanol is a renewable energy carrier mainly produced from biomass fermentation. Steam reforming of bioethanol (SRE) provides a promising method for hydrogen and power production frommore » renewable resources. SRE is performed at high temperatures (e.g. 800-900°C) to reduce the reforming by-products (e.g. ethane, ethene). The power generation from hydrogen was done with M701G2 gas turbine (334 MW net power output). Hydrogen was obtained through catalytic steam reforming of bioethanol without and with carbon capture. For the evaluated plant concepts the following key performance indicators were assessed: fuel consumption, gross and net power outputs, net electrical efficiency, ancillary consumptions, carbon capture rate, specific CO{sub 2} emission etc. As the results show, the power generation based on bioethanol conversion has high energy efficiency and low carbon footprint.« less

Power generation method including membrane separation

DOEpatents

Lokhandwala, Kaaeid A.

2000-01-01

A method for generating electric power, such as at, or close to, natural gas fields. The method includes conditioning natural gas containing C.sub.3+ hydrocarbons and/or acid gas by means of a membrane separation step. This step creates a leaner, sweeter, drier gas, which is then used as combustion fuel to run a turbine, which is in turn used for power generation.

Apparatus and method for closed-loop control of reactor power in minimum time

DOEpatents

Bernard, Jr., John A.

1988-11-01

Closed-loop control law for altering the power level of nuclear reactors in a safe manner and without overshoot and in minimum time. Apparatus is provided for moving a fast-acting control element such as a control rod or a control drum for altering the nuclear reactor power level. A computer computes at short time intervals either the function: .rho.=(.beta.-.rho.).omega.-.lambda..sub.e '.rho.-.SIGMA..beta..sub.i (.lambda..sub.i -.lambda..sub.e ')+l* .omega.+l* [.omega..sup.2 +.lambda..sub.e '.omega.] or the function: .rho.=(.beta.-.rho.).omega.-.lambda..sub.e .rho.-(.lambda..sub.e /.lambda..sub.e)(.beta.-.rho.)+l* .omega.+l* [.omega..sup.2 +.lambda..sub.e .omega.-(.lambda..sub.e /.lambda..sub.e).omega.] These functions each specify the rate of change of reactivity that is necessary to achieve a specified rate of change of reactor power. The direction and speed of motion of the control element is altered so as to provide the rate of reactivity change calculated using either or both of these functions thereby resulting in the attainment of a new power level without overshoot and in minimum time. These functions are computed at intervals of approximately 0.01-1.0 seconds depending on the specific application.

A NEUTRONIC REACTOR

DOEpatents

Luebke, E.A.; Vandenberg, L.B.

1959-09-01

A nuclear reactor for producing thermoelectric power is described. The reactor core comprises a series of thermoelectric assemblies, each assembly including fissionable fuel as an active element to form a hot junction and a thermocouple. The assemblies are disposed parallel to each other to form spaces and means are included for Introducing an electrically conductive coolant between the assemblies to form cold junctions of the thermocouples. An electromotive force is developed across the entire series of the thermoelectric assemblies due to fission heat generated in the fuel causing a current to flow perpendicular to the flow of coolant and is distributed to a load outside of the reactor by means of bus bars electrically connected to the outermost thermoelectric assembly.

Thermoelectric fabrics: toward power generating clothing.

PubMed

Du, Yong; Cai, Kefeng; Chen, Song; Wang, Hongxia; Shen, Shirley Z; Donelson, Richard; Lin, Tong

2015-03-23

Herein, we demonstrate that a flexible, air-permeable, thermoelectric (TE) power generator can be prepared by applying a TE polymer (e.g. poly(3,4-ethylenedioxythiophene):poly(4-styrenesulfonate)) coated commercial fabric and subsequently by linking the coated strips with a conductive connection (e.g. using fine metal wires). The poly(3,4-ethylenedioxythiophene):poly(4-styrenesulfonate) coated fabric shows very stable TE properties from 300 K to 390 K. The fabric device can generate a TE voltage output (V) of 4.3 mV at a temperature difference (ΔT) of 75.2 K. The potential for using fabric TE devices to harvest body temperature energy has been discussed. Fabric-based TE devices may be useful for the development of new power generating clothing and self-powered wearable electronics.

Space reactor power 1986 - A year of choices and transition

NASA Technical Reports Server (NTRS)

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

1986-01-01

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

Nuclear power generation and fuel cycle report 1997

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

NONE

1997-09-01

Nuclear power is an important source of electric energy and the amount of nuclear-generated electricity continued to grow as the performance of nuclear power plants improved. In 1996, nuclear power plants supplied 23 percent of the electricity production for countries with nuclear units, and 17 percent of the total electricity generated worldwide. However, the likelihood of nuclear power assuming a much larger role or even retaining its current share of electricity generation production is uncertain. The industry faces a complex set of issues including economic competitiveness, social acceptance, and the handling of nuclear waste, all of which contribute to themore » uncertain future of nuclear power. Nevertheless, for some countries the installed nuclear generating capacity is projected to continue to grow. Insufficient indigenous energy resources and concerns over energy independence make nuclear electric generation a viable option, especially for the countries of the Far East.« less

Innovative open air brayton combined cycle systems for the next generation nuclear power plants

NASA Astrophysics Data System (ADS)

Zohuri, Bahman

The purpose of this research was to model and analyze a nuclear heated multi-turbine power conversion system operating with atmospheric air as the working fluid. The air is heated by a molten salt, or liquid metal, to gas heat exchanger reaching a peak temperature of 660 0C. The effects of adding a recuperator or a bottoming steam cycle have been addressed. The calculated results are intended to identify paths for future work on the next generation nuclear power plant (GEN-IV). This document describes the proposed system in sufficient detail to communicate a good understanding of the overall system, its components, and intended uses. The architecture is described at the conceptual level, and does not replace a detailed design document. The main part of the study focused on a Brayton --- Rankine Combined Cycle system and a Recuperated Brayton Cycle since they offer the highest overall efficiencies. Open Air Brayton power cycles also require low cooling water flows relative to other power cycles. Although the Recuperated Brayton Cycle achieves an overall efficiency slightly less that the Brayton --- Rankine Combined Cycle, it is completely free of a circulating water system and can be used in a desert climate. Detailed results of modeling a combined cycle Brayton-Rankine power conversion system are presented. The Rankine bottoming cycle appears to offer a slight efficiency advantage over the recuperated Brayton cycle. Both offer very significant advantages over current generation Light Water Reactor steam cycles. The combined cycle was optimized as a unit and lower pressure Rankine systems seem to be more efficient. The combined cycle requires a lot less circulating water than current power plants. The open-air Brayton systems appear to be worth investigating, if the higher temperatures predicted for the Next Generation Nuclear Plant do materialize.

Split-core heat-pipe reactors for out-of-pile thermionic power systems.

NASA Technical Reports Server (NTRS)

Niederauer, G.; Lantz, E.; Breitweiser, R.

1971-01-01

Description of the concept of splitting a heat-pipe reactor for out-of-core thermionics into two identical halves and using the resulting center gap for reactivity control. Short Li-W reactor heat pipes penetrate the axial reflectors and form a heat exchanger with long heat pipes which wind through the shield to the thermionic diodes. With one reactor half anchored to the shield, the other is attached to a long arm with a pivot behind the shield and swings through a small arc for reactivity control. A safety shim prevents large reactivity inputs, and a fueled control arm drive shaft acts as a power stabilizer. Reactors fueled with U-235C and with U-233C have been studied.-

Fuel element concept for long life high power nuclear reactors

NASA Technical Reports Server (NTRS)

Mcdonald, G. E.; Rom, F. E.

1969-01-01

Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

Dual-loop self-optimizing robust control of wind power generation with Doubly-Fed Induction Generator.

PubMed

Chen, Quan; Li, Yaoyu; Seem, John E

2015-09-01

This paper presents a self-optimizing robust control scheme that can maximize the power generation for a variable speed wind turbine with Doubly-Fed Induction Generator (DFIG) operated in Region 2. A dual-loop control structure is proposed to synergize the conversion from aerodynamic power to rotor power and the conversion from rotor power to the electrical power. The outer loop is an Extremum Seeking Control (ESC) based generator torque regulation via the electric power feedback. The ESC can search for the optimal generator torque constant to maximize the rotor power without wind measurement or accurate knowledge of power map. The inner loop is a vector-control based scheme that can both regulate the generator torque requested by the ESC and also maximize the conversion from the rotor power to grid power. An ℋ(∞) controller is synthesized for maximizing, with performance specifications defined based upon the spectrum of the rotor power obtained by the ESC. Also, the controller is designed to be robust against the variations of some generator parameters. The proposed control strategy is validated via simulation study based on the synergy of several software packages including the TurbSim and FAST developed by NREL, Simulink and SimPowerSystems. Copyright © 2015 ISA. Published by Elsevier Ltd. All rights reserved.

Power Generation for River and Tidal Generators

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

Muljadi, Eduard; Wright, Alan; Gevorgian, Vahan

Renewable energy sources are the second largest contributor to global electricity production, after fossil fuels. The integration of renewable energy continued to grow in 2014 against a backdrop of increasing global energy consumption and a dramatic decline in oil prices during the second half of the year. As renewable generation has become less expensive during recent decades, and it becomes more accepted by the global population, the focus on renewable generation has expanded from primarily wind and solar to include new types with promising future applications, such as hydropower generation, including river and tidal generation. Today, hydropower is considered onemore » of the most important renewable energy sources. In river and tidal generation, the input resource flow is slower but also steadier than it is in wind or solar generation, yet the level of water turbulent flow may vary from one place to another. This report focuses on hydrokinetic power conversion.« less

Aircraft photovoltaic power-generating system

NASA Astrophysics Data System (ADS)

Doellner, Oscar Leonard

Photovoltaic cells, appropriately cooled and operating in the combustion-created high radiant-intensity environment of gas-turbine and jet engines, may replace the conventional (gearbox-driven) electrical power generators aboard jet aircraft. This study projects significant improvements not only in aircraft electrical power-generating-system performance, but also in overall aircraft performance. Jet-engine design modifications incorporating this concept not only save weight (and thus fuel), but are - in themselves - favorable to jet-engine performance. The dissertation concentrates on operational, constructional, structural, thermal, optical, radiometrical, thin-film, and solid-state theoretical aspects of the overall project.

Concentrated solar power generation using solar receivers

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

Anderson, Bruce N.; Treece, William Dean; Brown, Dan

Inventive concentrated solar power systems using solar receivers, and related devices and methods, are generally described. Low pressure solar receivers are provided that function to convert solar radiation energy to thermal energy of a working fluid, e.g., a working fluid of a power generation or thermal storage system. In some embodiments, low pressure solar receivers are provided herein that are useful in conjunction with gas turbine based power generation systems.

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

Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

NASA Technical Reports Server (NTRS)

Roth, R. J.

1976-01-01

The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

Piezoelectric devices for generating low power

NASA Astrophysics Data System (ADS)

Chilibon, Irinela

2016-12-01

This paper reviews concepts and applications in low-power electronics and energy harvesting technologies. Various piezoelectric materials and devices for small power generators useful in renewable electricity are presented. The vibrating piezoelectric device differs from the typical electrical power source in that it has capacitive rather than inductive source impedance, and may be driven by mechanical vibrations of varying amplitude. In general, vibration energy could be converted into electrical energy using one of three techniques: electrostatic charge, magnetic fields and piezoelectric. A low power piezoelectric generator, having a PZT element was realised in order to supply small electronic elements, such as optoelectronic small devices, LEDs, electronic watches, small sensors, interferometry with lasers or Micro-electro-mechanical System (MEMS) array with multi-cantilevers.

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.

Enhanced electricity generation by triclosan and iron anodes in the three-chambered membrane bio-chemical reactor (TC-MBCR).

PubMed

Song, Jing; Liu, Lifen; Yang, Fenglin; Ren, Nanqi; Crittenden, John

2013-11-01

A three-chambered membrane bio-chemical reactor (TC-MBCR) was developed. The stainless steel membrane modules were used as cathodes and iron plates in the middle chamber served as the anode. The TC-MBCR was able to reduce fouling, remove triclosan (TCS) from a synthetic wastewater treatment and enhance electricity generation by ~60% compared with the cell voltage before TCS addition. The TC-MBCR system generated a relatively stable power output (cell voltage ~0.2V) and the corrosion of iron plates contributed to electricity generation together with microbes on iron anode. The permeation flow from anode to cathode chamber was considered important in electricity generation. In addition, the negatively charged cathode membrane and Fe(2+)/Fe(3+) released by iron plates mitigated membrane fouling by approximately 30%, as compared with the control. The removal of COD and total phosphorus was approximately 99% and 90%. The highest triclosan removal rate reached 97.9%. Copyright © 2013. Published by Elsevier Ltd.

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.

Calculated power distribution of a thermionic, beryllium oxide reflected, fast-spectrum reactor

NASA Technical Reports Server (NTRS)

Mayo, W.; Lantz, E.

1973-01-01

A procedure is developed and used to calculate the detailed power distribution in the fuel elements next to a beryllium oxide reflector of a fast-spectrum, thermionic reactor. The results of the calculations show that, although the average power density in these outer fuel elements is not far from the core average, the power density at the very edge of the fuel closest to the beryllium oxide is about 1.8 times the core avearge.

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.

Materials technology for an advanced space power nuclear reactor concept: Program summary

NASA Technical Reports Server (NTRS)

Gluyas, R. E.; Watson, G. K.

1975-01-01

The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

Modular Analysis of Automobile Exhaust Thermoelectric Power Generation System

NASA Astrophysics Data System (ADS)

Deng, Y. D.; Zhang, Y.; Su, C. Q.

2015-06-01

In this paper, an automobile exhaust thermoelectric power generation system is packaged into a model with its own operating principles. The inputs are the engine speed and power, and the output is the power generated by the system. The model is divided into two submodels. One is the inlet temperature submodel, and the other is the power generation submodel. An experimental data modeling method is adopted to construct the inlet temperature submodel, and a theoretical modeling method is adopted to construct the power generation submodel. After modeling, simulation is conducted under various engine operating conditions to determine the variation of the power generated by the system. Finally, the model is embedded into a Honda Insight vehicle model to explore the energy-saving effect of the system on the vehicle under Economic Commission for Europe and cyc-constant_60 driving cycles.

Space power reactor in-core thermionic multicell evolutionary (S-prime) design

NASA Astrophysics Data System (ADS)

Determan, William R.; Van Hagan, Tom H.

1993-01-01

A 5- to 40-kWe moderated in-core thermionic space nuclear power system (TI-SNPS) concept was developed to address the TI-SNPS program requirements. The 40-kWe baseline design uses multicell Thermionic Fuel Elements (TFEs) in a zirconium hydride moderated reactor to achieve a specific mass of 18.2 We/kg and a net end-of-mission (EOM) efficiency of 8.2%. The reactor is cooled with a single NaK-78 pumped loop, which rejects the heat through a 24 m2 heat pipe space radiator.

Nuclear power in the 21st century: Challenges and possibilities.

PubMed

Horvath, Akos; Rachlew, Elisabeth

2016-01-01

The current situation and possible future developments for nuclear power--including fission and fusion processes--is presented. The fission nuclear power continues to be an essential part of the low-carbon electricity generation in the world for decades to come. There are breakthrough possibilities in the development of new generation nuclear reactors where the life-time of the nuclear waste can be reduced to some hundreds of years instead of the present time-scales of hundred thousand of years. Research on the fourth generation reactors is needed for the realisation of this development. For the fast nuclear reactors, a substantial research and development effort is required in many fields--from material sciences to safety demonstration--to attain the envisaged goals. Fusion provides a long-term vision for an efficient energy production. The fusion option for a nuclear reactor for efficient production of electricity has been set out in a focussed European programme including the international project of ITER after which a fusion electricity DEMO reactor is envisaged.

Fast-spectrum space-power-reactor concepts using boron control devices

NASA Technical Reports Server (NTRS)

Mayo, W.

1973-01-01

Several fast-spectrum space power reactor concepts that use boron carbide control devices were examined to determine the neutronic feasibility of the designs. The designs considered were (1) a 199-fuel-pin, 12-poison-reflector-control-drum reactor; (2) a 232-fuel-pin reactor with 12 reflector drums and three in-core control rods; (3) a 337-fuel-pin design with 12 incore control rods; and a 181-fuel-pin design with six drums closely coupled to the core to increase reactivity per drum. Adequate reactivity control and excess reactivity could be obtained for each concept, and the goals of 50,000 hours at 2.17 thermal megawatts with a lithium-7 coolant outlet temperature of 1222 K could be met without exceeding the 1-percent-clad-creep criterion. Heating rates in the boron carbide were calculated, but a heat transfer analysis was not done.

The role of nuclear reactors in space exploration and development

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

Lipinski, R.J.

2000-07-01

The United States has launched more than 20 radioisotopic thermoelectric generators (RTGs) into space over the past 30 yr but has launched only one nuclear reactor, and that was in 1965. Russia has launched more than 30 reactors. The RTGs use the heat of alpha decay of {sup 238}Pu for power and typically generate <1 kW of electricity. Apollo, Pioneer, Voyager, Viking, Galileo, Ulysses, and Cassini all used RTGs. Space reactors use the fission energy of {sup 235}U; typical designs are for 100 to 1000 kW of electricity. The only US space reactor launch (SNAP-10A) was a demonstration mission. Onemore » reason for the lack of space reactor use by the United States was the lack of space missions that required high power. But, another was the assumed negative publicity that would accompany a reactor launch. The net result is that all space reactor programs after 1970 were terminated before an operating space reactor could be developed, and they are now many years from recovering the ability to build them. Two major near-term needs for space reactors are the human exploration of Mars and advanced missions to and beyond the orbit of Jupiter. To help obtain public acceptance of space reactors, one must correct some of the misconceptions concerning space reactors and convey the following facts to the public and to decision makers: Space reactors are 1000 times smaller in power and size than a commercial power reactor. A space reactor at launch is only as radioactive as a pile of dirt 60 m (200 ft) across. A space reactor contains no plutonium at launch. It does not become significantly radioactive until it is turned on, and it will be engineered so that no launch accident can turn it on, even if that means fueling it after launch. The reactor will not be turned on until it is in a high stable orbit or even on an earth-escape trajectory for some missions. The benefits of space reactors are that they give humanity a stairway to the planets and perhaps the stars. They open a

Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

NASA Astrophysics Data System (ADS)

Ternovykh, Mikhail; Tikhomirov, Georgy; Saldikov, Ivan; Gerasimov, Alexander

2017-09-01

Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing.

Optimal Output of Distributed Generation Based On Complex Power Increment

NASA Astrophysics Data System (ADS)

Wu, D.; Bao, H.

2017-12-01

In order to meet the growing demand for electricity and improve the cleanliness of power generation, new energy generation, represented by wind power generation, photovoltaic power generation, etc has been widely used. The new energy power generation access to distribution network in the form of distributed generation, consumed by local load. However, with the increase of the scale of distribution generation access to the network, the optimization of its power output is becoming more and more prominent, which needs further study. Classical optimization methods often use extended sensitivity method to obtain the relationship between different power generators, but ignore the coupling parameter between nodes makes the results are not accurate; heuristic algorithm also has defects such as slow calculation speed, uncertain outcomes. This article proposes a method called complex power increment, the essence of this method is the analysis of the power grid under steady power flow. After analyzing the results we can obtain the complex scaling function equation between the power supplies, the coefficient of the equation is based on the impedance parameter of the network, so the description of the relation of variables to the coefficients is more precise Thus, the method can accurately describe the power increment relationship, and can obtain the power optimization scheme more accurately and quickly than the extended sensitivity method and heuristic method.

Reactor/Brayton power systems for nuclear electric spacecraft

NASA Technical Reports Server (NTRS)

Layton, J. P.

1980-01-01

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

A Burst Mode, Ultrahigh Temperature UF4 Vapor Core Reactor Rankine Cycle Space Power System Concept

NASA Technical Reports Server (NTRS)

Dugan, E. T.; Kahook, S. D.; Diaz, N. J.

1996-01-01

Static and dynamic neutronic analyses have been performed on an innovative burst mode (100's of MW output for a few thousand seconds) Ulvahigh Temperature Vapor Core Reactor (UTVR) space nuclear power system. The NVTR employs multiple, neutronically-coupled fissioning cores and operates on a direct, closed Rankine cycle using a disk Magnetohydrodynamic (MHD) generater for energy conversion. The UTVR includes two types of fissioning core regions: (1) the central Ultrahigh Temperature Vapor Core (UTVC) which contains a vapor mixture of highly enriched UF4 fuel and a metal fluoride working fluid and (2) the UF4 boiler column cores located in the BeO moderator/reflector region. The gaseous nature of the fuel the fact that the fuel is circulating, the multiple coupled fissioning cores, and the use of a two phase fissioning fuel lead to unique static and dynamic neutronic characteristics. Static neutronic analysis was conducted using two-dimensional S sub n, transport theory calculations and three-dimensional Monte Carlo transport theory calculations. Circulating-fuel, coupled-core point reactor kinetics equations were used for analyzing the dynamic behavior of the UTVR. In addition to including reactivity feedback phenomena associated with the individual fissioning cores, the effects of core-to-core neutronic and mass flow coupling between the UTVC and the surrounding boiler cores were also included in the dynamic model The dynamic analysis of the UTVR reveals the existence of some very effectlve inherent reactivity feedback effects that are capable of quickly stabilizing this system, within a few seconds, even when large positive reactivity insertions are imposed. If the UTVC vapor fuel density feedback is suppressed, the UTVR is still inherently stable because of the boiler core liquid-fuel volume feedback; in contrast, suppression of the vapor fuel density feedback in 'conventional" gas core cavity reactors causes them to become inherently unstable. Due to the

Eugene P. Wigner's Visionary Contributions to Generations-I through IV Fission Reactors

NASA Astrophysics Data System (ADS)

Carré, Frank

2014-09-01

Among Europe's greatest scientists who fled to Britain and America in the 1930s, Eugene P. Wigner made instrumental advances in reactor physics, reactor design and technology, and spent nuclear fuel processing for both purposes of developing atomic weapons during world-war II and nuclear power afterwards. Wigner who had training in chemical engineering and self-education in physics first gained recognition for his remarkable articles and books on applications of Group theory to Quantum mechanics, Solid state physics and other topics that opened new branches of Physics.

Integrated engine-generator concept for aircraft electric secondary power

NASA Technical Reports Server (NTRS)

Secunde, R. R.; Macosko, R. P.; Repas, D. S.

1972-01-01

The integrated engine-generator concept of locating an electric generator inside an aircraft turbojet or turbofan engine concentric with, and driven by, one of the main engine shafts is discussed. When properly rated, the generator can serve as an engine starter as well as a generator of electric power. The electric power conversion equipment and generator controls are conveniently located in the aircraft. Preliminary layouts of generators in a large engine together with their physical sizes and weights indicate that this concept is a technically feasible approach to aircraft secondary power.

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.

Extension of the supercritical carbon dioxide brayton cycle to low reactor power operation: investigations using the coupled anl plant dynamics code-SAS4A/SASSYS-1 liquid metal reactor code system.

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

Moisseytsev, A.; Sienicki, J. J.

2012-05-10

Significant progress has been made on the development of a control strategy for the supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle enabling removal of power from an autonomous load following Sodium-Cooled Fast Reactor (SFR) down to decay heat levels such that the S-CO{sub 2} cycle can be used to cool the reactor until decay heat can be removed by the normal shutdown heat removal system or a passive decay heat removal system such as Direct Reactor Auxiliary Cooling System (DRACS) loops with DRACS in-vessel heat exchangers. This capability of the new control strategy eliminates the need for use of amore » separate shutdown heat removal system which might also use supercritical CO{sub 2}. It has been found that this capability can be achieved by introducing a new control mechanism involving shaft speed control for the common shaft joining the turbine and two compressors following reduction of the load demand from the electrical grid to zero. Following disconnection of the generator from the electrical grid, heat is removed from the intermediate sodium circuit through the sodium-to-CO{sub 2} heat exchanger, the turbine solely drives the two compressors, and heat is rejected from the cycle through the CO{sub 2}-to-water cooler. To investigate the effectiveness of shaft speed control, calculations are carried out using the coupled Plant Dynamics Code-SAS4A/SASSYS-1 code for a linear load reduction transient for a 1000 MWt metallic-fueled SFR with autonomous load following. No deliberate motion of control rods or adjustment of sodium pump speeds is assumed to take place. It is assumed that the S-CO{sub 2} turbomachinery shaft speed linearly decreases from 100 to 20% nominal following reduction of grid load to zero. The reactor power is calculated to autonomously decrease down to 3% nominal providing a lengthy window in time for the switchover to the normal shutdown heat removal system or for a passive decay heat removal system to become effective. However, the

Erosion-corrosion entrainment of iron-containing compounds as a source of deposits in steam generators used at nuclear power plants equipped with VVER reactors

NASA Astrophysics Data System (ADS)

Tomarov, G. V.; Shipkov, A. A.

2011-03-01

The main stages and processes through which deposits are generated, migrate, and precipitate in the metal-secondary coolant system of power units at nuclear power plants are analyzed and determined. It is shown that substances produced by the mechanism of general erosion-corrosion are the main source of the ionic-colloid form of iron, which is the main component of deposits in a steam generator. Ways for controlling the formation of deposits in a nuclear power plant's steam generator are proposed together with methods for estimating their efficiency.

Method and apparatus for improving the performance of a nuclear power electrical generation system

DOEpatents

Tsiklauri, Georgi V.; Durst, Bruce M.

1995-01-01

A method and apparatus for improving the efficiency and performance a of nuclear electrical generation system that comprises the addition of steam handling equipment to an existing plant that results in a surprising increase in plant performance. More particularly, a gas turbine electrical generation system with heat recovery boiler is installed along with a high pressure and a low pressure mixer superheater. Depending upon plant characteristics, the existing moisture separator reheater (MSR) can be either augmented or done away with. The instant invention enables a reduction in T.sub.hot without a derating of the reactor unit, and improves efficiency of the plant's electrical conversion cycle. Coupled with this advantage is a possible extension of the plant's fuel cycle length due to an increased electrical conversion efficiency. The reduction in T.sub.hot further allows for a surprising extension of steam generator life. An additional advantage is the reduction in erosion/corrosion of secondary system components including turbine blades and diaphragms. The gas turbine generator used in the instant invention can also replace or augment existing peak or emergency power needs.

Robust Power Management Control for Stand-Alone Hybrid Power Generation System

NASA Astrophysics Data System (ADS)

Kamal, Elkhatib; Adouane, Lounis; Aitouche, Abdel; Mohammed, Walaa

2017-01-01

This paper presents a new robust fuzzy control of energy management strategy for the stand-alone hybrid power systems. It consists of two levels named centralized fuzzy supervisory control which generates the power references for each decentralized robust fuzzy control. Hybrid power systems comprises: a photovoltaic panel and wind turbine as renewable sources, a micro turbine generator and a battery storage system. The proposed control strategy is able to satisfy the load requirements based on a fuzzy supervisor controller and manage power flows between the different energy sources and the storage unit by respecting the state of charge and the variation of wind speed and irradiance. Centralized controller is designed based on If-Then fuzzy rules to manage and optimize the hybrid power system production by generating the reference power for photovoltaic panel and wind turbine. Decentralized controller is based on the Takagi-Sugeno fuzzy model and permits us to stabilize each photovoltaic panel and wind turbine in presence of disturbances and parametric uncertainties and to optimize the tracking reference which is given by the centralized controller level. The sufficient conditions stability are formulated in the format of linear matrix inequalities using the Lyapunov stability theory. The effectiveness of the proposed Strategy is finally demonstrated through a SAHPS (stand-alone hybrid power systems) to illustrate the effectiveness of the overall proposed method.

Apollo experience report: Power generation system

NASA Technical Reports Server (NTRS)

Bell, D., III; Plauche, F. M.

1973-01-01

A comprehensive review of the design philosophy and experience of the Apollo electrical power generation system is presented. The review of the system covers a period of 8 years, from conception through the Apollo 12 lunar-landing mission. The program progressed from the definition phase to hardware design, system development and qualification, and, ultimately, to the flight phase. Several problems were encountered; however, a technology evolved that enabled resolution of the problems and resulted in a fully manrated power generation system. These problems are defined and examined, and the corrective action taken is discussed. Several recommendations are made to preclude similar occurrences and to provide a more reliable fuel-cell power system.

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.

PARTIAL ECONOMIC STUDY OF STEAM COOLED HEAVY WATER MODERATED REACTORS

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

None

1960-04-01

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

Green Power Partnership Top 30 On-site Generation

EPA Pesticide Factsheets

EPA's Green Power Partnership is a voluntary program designed to reduce the environmental impact of electricity generation by promoting renewable energy. These partners are generating and consuming the most green power on-site within the GPP.

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.

COMPREHENSIVE STANDARDS: THE POWER GENERATION CASE

EPA Science Inventory

This study presents an illustrative data base of material quantities and environmental effluents in the fuel cycles for alternative technologies of thermally generated power. The entire fuel cycle for each of the alternative ten technologies is outlined for a representative power...

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

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

Capacity expansion model of wind power generation based on ELCC

NASA Astrophysics Data System (ADS)

Yuan, Bo; Zong, Jin; Wu, Shengyu

2018-02-01

Capacity expansion is an indispensable prerequisite for power system planning and construction. A reasonable, efficient and accurate capacity expansion model (CEM) is crucial to power system planning. In most current CEMs, the capacity of wind power generation is considered as boundary conditions instead of decision variables, which may lead to curtailment or over construction of flexible resource, especially at a high renewable energy penetration scenario. This paper proposed a wind power generation capacity value(CV) calculation method based on effective load-carrying capability, and a CEM that co-optimizes wind power generation and conventional power sources. Wind power generation is considered as decision variable in this model, and the model can accurately reflect the uncertainty nature of wind power.

Physics and nuclear power

NASA Astrophysics Data System (ADS)

Buttery, N. E.

2008-03-01

Nuclear power owes its origin to physicists. Fission was demonstrated by physicists and chemists and the first nuclear reactor project was led by physicists. However as nuclear power was harnessed to produce electricity the role of the engineer became stronger. Modern nuclear power reactors bring together the skills of physicists, chemists, chemical engineers, electrical engineers, mechanical engineers and civil engineers. The paper illustrates this by considering the Sizewell B project and the role played by physicists in this. This covers not only the roles in design and analysis but in problem solving during the commissioning of first of a kind plant. Looking forward to the challenges to provide sustainable and environmentally acceptable energy sources for the future illustrates the need for a continuing synergy between physics and engineering. This will be discussed in the context of the challenges posed by Generation IV reactors.

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

NASA Astrophysics Data System (ADS)