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.

Megawatt Class Nuclear Space Power Systems (MCNSPS) conceptual design and evaluation report. Volume 2, technologies 1: Reactors, heat transport, integration issues

NASA Technical Reports Server (NTRS)

Wetch, J. R.

1988-01-01

The objectives of the Megawatt Class Nuclear Space Power System (MCNSPS) study are summarized and candidate systems and subsystems are described. Particular emphasis is given to the heat rejection system and the space reactor subsystem.

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.

NASA Growth Space Station missions and candidate nuclear/solar power systems

NASA Technical Reports Server (NTRS)

Heller, Jack A.; Nainiger, Joseph J.

1987-01-01

A brief summary is presented of a NASA study contract and in-house investigation on Growth Space Station missions and appropriate nuclear and solar space electric power systems. By the year 2000 some 300 kWe will be needed for missions and housekeeping power for a 12 to 18 person Station crew. Several Space Station configurations employing nuclear reactor power systems are discussed, including shielding requirements and power transmission schemes. Advantages of reactor power include a greatly simplified Station orientation procedure, greatly reduced occultation of views of the earth and deep space, near elimination of energy storage requirements, and significantly reduced station-keeping propellant mass due to very low drag of the reactor power system. The in-house studies of viable alternative Growth Space Station power systems showed that at 300 kWe a rigid silicon solar cell array with NiCd batteries had the highest specific mass at 275 kg/kWe, with solar Stirling the lowest at 40 kg/kWe. However, when 10 year propellant mass requirements are factored in, the 300 kWe nuclear Stirling exhibits the lowest total mass.

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.

Soviet space nuclear reactor incidents - Perception versus reality

NASA Technical Reports Server (NTRS)

Bennett, Gary L.

1992-01-01

Since the Soviet Union reportedly began flying nuclear power sources in 1965 it has had four publicly known accidents involving space reactors, two publicly known accidents involving radioisotope power sources and one close call with a space reactor (Cosmos 1900). The reactor accidents, particularly Cosmos 954 and Cosmos 1402, indicated that the Soviets had adopted burnup as their reentry philosophy which is consistent with the U.S. philosophy from the 1960s and 1970s. While quantitative risk analyses have shown that the Soviet accidents have not posed a serious risk to the world's population, concerns still remain about Soviet space nuclear safety practices.

Perspectives of The Interagency Nuclear Safety Review Panel (INSRP) on future nuclear powered space missions

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

Gray, L.B.; Pyatt, D.W.; Sholtis, J.A.

1993-01-10

The Interagency Nuclear Safety Review Panel (INSRP) has provided reviews of all nuclear powered spacecraft launched by the United States. The two most recent launches were Ulysses in 1990 and Galileo in 1989. One reactor was launched in 1965 (SNAP-10A). All other U.S. space missions have utilized radioisotopic thermoelectric generators (RTGs). There are several missions in the next few years that are to be nuclear powered, including one that would utilize the Topaz II reactor purchased from Russia. INSRP must realign itself to perform parallel safety assessments of a reactor powered space mission, which has not been done in aboutmore » thirty years, and RTG powered missions.« less

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

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.

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.

A Comparison of Fast-Spectrum and Moderated Space Fission Reactors

NASA Astrophysics Data System (ADS)

Poston, David I.

2005-02-01

The reactor neutron spectrum is one of the fundamental design choices for any fission reactor, but the implications of using a moderated spectrum are vastly different for space reactors as opposed to terrestrial reactors. In addition, the pros and cons of neutron spectra are significantly different among many of the envisioned space power applications. This paper begins with a discussion of the neutronic differences between fast-spectrum and moderated space reactors. This is followed by a discussion of the pros and cons of fast-spectrum and moderated space reactors separated into three areas—technical risk, performance, and safety/safeguards. A mix of quantitative and qualitative arguments is presented, and some conclusions generally can be made regarding neutron spectrum and space power application. In most cases, a fast-spectrum system appears to be the better alternative (mostly because of simplicity and higher potential operating temperatures); however, in some cases, such as a low-power (<100-kWt) surface reactor, a moderated spectrum could provide a better approach. In all cases, the determination of which spectrum is preferred is a strong function of the metrics provided by the "customer"— i.e., if a certain level of performance is required, it could provide a different solution than if a certain level of safeguards is required (which in some cases could produce a null solution). The views expressed in this document are those of the author and do not necessarily reflect agreement by the Government.

Kilopower: Small and Affordable Fission Power Systems for Space

NASA Technical Reports Server (NTRS)

Mason, Lee; Palac, Don; Gibson, Marc

2017-01-01

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

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

NASA Technical Reports Server (NTRS)

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

1988-01-01

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

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

NASA Astrophysics Data System (ADS)

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

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

Goals of thermionic program for space power

NASA Technical Reports Server (NTRS)

English, R. E.

1981-01-01

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

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.

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.

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.

Application of Molten Salt Reactor Technology to Nuclear Electric Propulsion Mission

NASA Technical Reports Server (NTRS)

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

2002-01-01

Nuclear electric propulsion (NEP) and planetary surface power missions require 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 gas cooled, liquid metal, and heat pipe space reactors.

Movable-molybdenum-reflector reactivity experiments for control studies of compact space power reactor concepts

NASA Technical Reports Server (NTRS)

Fox, T. A.

1973-01-01

An experimental reflector reactivity study was made with a compact cylindrical reactor using a uranyl fluoride - water fuel solution. The reactor was axially unreflected and radially reflected with segments of molybdenum. The reflector segments were displaced incrementally in both the axial and radial dimensions, and the shutdown of each configuration was measured by using the pulsed-neutron source technique. The reactivity effects for axial and radial displacement of reflector segments are tabulated separately and compared. The experiments provide data for control-system studies of compact-space-power-reactor concepts.

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.

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.

Analysis of space reactor system components: Investigation through simulation and non-nuclear testing

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.

The use of fission energy in space power and propulsion systems offers considerable advantages over chemical propulsion. Fission provides over six orders of magnitude higher energy density, which translates to higher vehicle specific impulse and lower specific mass. These characteristics enable ambitious space exploration missions. The natural space radiation environment provides an external source of protons and high energy, high Z particles that can result in the production of secondary neutrons through interactions in reactor structures. Applying the approximate proton source in geosynchronous orbit during a solar particle event, investigation using MCNPX 2.5.b for proton transport through the SAFE-400 heat pipe cooled reactor indicates an incoming secondary neutron current of (1.16 +/- 0.03) x 107 n/s at the core-reflector interface. This neutron current may affect reactor operation during low power maneuvers (e.g., start-up) and may provide a sufficient reactor start-up source. It is important that a reactor control system be designed to automatically adjust to changes in reactor power levels, maintaining nominal operation without user intervention. A robust, autonomous control system is developed and analyzed for application during reactor start-up, accounting for fluctuations in the radiation environment that result from changes in vehicle location or to temporal variations in the radiation field. Development of a nuclear reactor for space applications requires a significant amount of testing prior to deployment of a flight unit. High confidence in fission system performance can be obtained through relatively inexpensive non-nuclear tests performed in relevant environments, with the heat from nuclear fission simulated using electric resistance heaters. A series of non-nuclear experiments was performed to characterize various aspects of reactor operation. This work includes measurement of reactor core deformation due to material thermal expansion and implementation of a virtual reactivity feedback control loop; testing and thermal hydraulic characterization of the coolant flow paths for two space reactor concepts; and analysis of heat pipe operation during start-up and steady state operation.

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

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

Wright, Steven A.; Sanchez, Travis

2005-02-06

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

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.

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.

Development and Testing of Space Fission Technology at NASA-MSFC

NASA Technical Reports Server (NTRS)

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

2008-01-01

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

Power-Conversion Concept Designed for the Jupiter Icy Moons Orbiter

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2004-01-01

The Jupiter Icy Moons Orbiter (JIMO) is a bold new mission being developed by NASA's Office of Space Science under Project Prometheus. JIMO is examining the potential of nuclear electric propulsion (NEP) technology to efficiently deliver scientific payloads to three of Jupiter's moons: Callisto, Ganymede, and Europa. A critical element of the NEP spacecraft is the space reactor power system (SRPS), consisting of the nuclear reactor, power conversion, heat rejection, and power management and distribution (PMAD).

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.

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.

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.

Tethered nuclear power for the Space Station

NASA Technical Reports Server (NTRS)

Bents, D. J.

1985-01-01

A nuclear space power system the SP-100 is being developed for future missions where large amounts of electrical power will be required. Although it is primarily intended for unmanned spacecraft, it can be adapted to a manned space platform by tethering it above the station through an electrical transmission line which isolates the reactor far away from the inhabited platform and conveys its power back to where it is needed. The transmission line, used in conjunction with an instrument rate shield, attenuates reactor radiation in the vicinity of the space station to less than one-one hundredth of the natural background which is already there. This combination of shielding and distance attenuation is less than one-tenth the mass of boom-mounted or onboard man-rated shields that are required when the reactor is mounted nearby. This paper describes how connection is made to the platform (configuration, operational requirements) and introduces a new element the coaxial transmission tube which enables efficient transmission of electrical power through long tethers in space. Design methodology for transmission tubes and tube arrays is discussed. An example conceptual design is presented that shows SP-100 at three power levels 100 kWe, 300 kWe, and 1000 kWe connected to space station via a 2 km HVDC transmission line/tether. Power system performance, mass, and radiation hazard are estimated with impacts on space station architecture and operation.

Tethered nuclear power for the space station

NASA Technical Reports Server (NTRS)

Bents, D. J.

1985-01-01

A nuclear space power system the SP-100 is being developed for future missions where large amounts of electrical power will be required. Although it is primarily intended for unmanned spacecraft, it can be adapted to a manned space platform by tethering it above the station through an electrical transmission line which isolates the reactor far away from the inhabited platform and conveys its power back to where it is needed. The transmission line, used in conjunction with an instrument rate shield, attenuates reactor radiation in the vicinity of the space station to less than one-one hundredth of the natural background which is already there. This combination of shielding and distance attenuation is less than one-tenth the mass of boom-mounted or onboard man-rated shields that are required when the reactor is mounted nearby. This paper describes how connection is made to the platform (configuration, operational requirements) and introduces a new element the coaxial transmission tube which enables efficient transmission of electrical power through long tethers in space. Design methodology for transmission tubes and tube arrays is discussed. An example conceptual design is presented that shows SP-100 at three power levels 100 kWe, 300 kWe, and 1000 kWe connected to space station via a 2 km HVDC transmission line/tether. Power system performance, mass, and radiation hazard are estimated with impacts on space station architecture and operation.

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

NASA Astrophysics Data System (ADS)

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

2004-02-01

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

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.

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.

SP-100 program developments

NASA Technical Reports Server (NTRS)

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

1985-01-01

An update is provided on the status of the Sp-100 Space Reactor Power Program. The historical background that led to the program is reviewed and the overall program objectives and development approach are discussed. The results of the mission studies identify applications for which space nuclear power is desirable and even essential. Results of a series of technology feasibility experiments are expected to significantly improve the earlier technology data base for engineering development. The conclusion is reached that a nuclear reactor space power system can be developed by the early 1990s to meet emerging mission performance requirements.

Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

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

D. Kokkinos

2005-04-28

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophymore » on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.« less

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

NASA Astrophysics Data System (ADS)

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

Reactor design and integration into a nuclear electric spacecraft

NASA Technical Reports Server (NTRS)

Phillips, W. M.; Koenig, D. R.

1978-01-01

One of the well-defined applications for nuclear power in space is nuclear electric propulsion (NEP). Mission studies have identified the optimum power level (400 kWe). A single Shuttle launch requirement and science-package integration have added additional constraints to the design. A reactor design which will meet these constraints has been studied. The reactor employs 90 fuel elements, each heat pipe cooled. Reactor control is obtained with BeO/B4C drums in a BeO reflector. The balance of the spacecraft is shielded from the reactor with LiH. Power conditioning and reactor control drum drives are located behind the LiH with the power conditioning. Launch safety, mechanical design and integration with the power conversion subsystem are discussed.

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

Fitzpatrick, F.C.; Gray, D.D.; Hyndman, J.R.

The thermal, ecological, and social impacts of a 40-reactor NEC are compared to impacts from four 10-reactor NECs and ten 4-reactor power plants. The comparison was made for surrogate sites in western Tennessee. The surrogate site for the 40-reactor NEC is located on Kentucky Lake. A layout is postulated for ten clusters of four reactors each with 2.5-mile spacing between clusters. The plants use natural-draft cooling towers. A transmission system is proposed for delivering the power (48,000 MW) to five load centers. Comparable transmission systems are proposed for the 10-reactor NECs and the 4-reactor dispersed sites delivering power to themore » same load centers. (auth)« less

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

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

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

1981-01-01

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

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

NASA Technical Reports Server (NTRS)

Wetch, J. R.

1988-01-01

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

High Neutron Fluence Survivability Testing of Advanced Fiber Bragg Grating Sensors

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

Fielder, Robert S.; Klemer, Daniel; Stinson-Bagby, Kelly L.

2004-02-04

The motivation for the reported research was to support NASA space nuclear power initiatives through the development of advanced fiber optic sensors for space-based nuclear power applications. The purpose of the high-neutron fluence testing was to demonstrate the survivability of fiber Bragg grating (FBG) sensors in a fission reactor environment. 520 FBGs were installed in the Ford reactor at the University of Michigan. The reactor was operated for 1012 effective full power hours resulting in a maximum neutron fluence of approximately 5x1019 n/cm2, and a maximum gamma dose of 2x103 MGy gamma. This work is significant in that, to themore » knowledge of the authors, the exposure levels obtained are approximately 1000 times higher than for any previously published experiment. Four different fiber compositions were evaluated. An 87% survival rate was observed for fiber Bragg gratings located at the fuel centerline. Optical Frequency Domain Reflectometry (OFDR), originally developed at the NASA Langley Research Center, can be used to interrogate several thousand low-reflectivity FBG strain and/or temperature sensors along a single optical fiber. A key advantage of the OFDR sensor technology for space nuclear power is the extremely low mass of the sensor, which consists of only a silica fiber 125{mu}m in diameter. The sensors produced using this technology will fill applications in nuclear power for current reactor plants, emerging Generation-IV reactors, and for space nuclear power. The reported research was conducted by Luna Innovations and was funded through a Small Business Innovative Research (SBIR) contract with the NASA Glenn Research Center.« less

High Neutron Fluence Survivability Testing of Advanced Fiber Bragg Grating Sensors

NASA Astrophysics Data System (ADS)

Fielder, Robert S.; Klemer, Daniel; Stinson-Bagby, Kelly L.

2004-02-01

The motivation for the reported research was to support NASA space nuclear power initiatives through the development of advanced fiber optic sensors for space-based nuclear power applications. The purpose of the high-neutron fluence testing was to demonstrate the survivability of fiber Bragg grating (FBG) sensors in a fission reactor environment. 520 FBGs were installed in the Ford reactor at the University of Michigan. The reactor was operated for 1012 effective full power hours resulting in a maximum neutron fluence of approximately 5×1019 n/cm2, and a maximum gamma dose of 2×103 MGy gamma. This work is significant in that, to the knowledge of the authors, the exposure levels obtained are approximately 1000 times higher than for any previously published experiment. Four different fiber compositions were evaluated. An 87% survival rate was observed for fiber Bragg gratings located at the fuel centerline. Optical Frequency Domain Reflectometry (OFDR), originally developed at the NASA Langley Research Center, can be used to interrogate several thousand low-reflectivity FBG strain and/or temperature sensors along a single optical fiber. A key advantage of the OFDR sensor technology for space nuclear power is the extremely low mass of the sensor, which consists of only a silica fiber 125μm in diameter. The sensors produced using this technology will fill applications in nuclear power for current reactor plants, emerging Generation-IV reactors, and for space nuclear power. The reported research was conducted by Luna Innovations and was funded through a Small Business Innovative Research (SBIR) contract with the NASA Glenn Research Center.

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).

A Spherical Torus Nuclear Fusion Reactor Space Propulsion Vehicle Concept for Fast Interplanetary Travel

NASA Technical Reports Server (NTRS)

Williams, Craig H.; Borowski, Stanley K.; Dudzinski, Leonard A.; Juhasz, Albert J.

1998-01-01

A conceptual vehicle design enabling fast outer solar system travel was produced predicated on a small aspect ratio spherical torus nuclear fusion reactor. Initial requirements were for a human mission to Saturn with a greater than 5% payload mass fraction and a one way trip time of less than one year. Analysis revealed that the vehicle could deliver a 108 mt crew habitat payload to Saturn rendezvous in 235 days, with an initial mass in low Earth orbit of 2,941 mt. Engineering conceptual design, analysis, and assessment was performed on all ma or systems including payload, central truss, nuclear reactor (including divertor and fuel injector), power conversion (including turbine, compressor, alternator, radiator, recuperator, and conditioning), magnetic nozzle, neutral beam injector, tankage, start/re-start reactor and battery, refrigeration, communications, reaction control, and in-space operations. Detailed assessment was done on reactor operations, including plasma characteristics, power balance, power utilization, and component design.

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

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

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

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

2004-02-04

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

ETR, TRA642. BASEMENT SPACE ALLOCATION FOR EXPERIMENTERS CA. 1966, SOUTHEAST ...

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

ETR, TRA-642. BASEMENT SPACE ALLOCATION FOR EXPERIMENTERS CA. 1966, SOUTHEAST QUADRANT OF FLOOR. WESTINGHOUSE ATOMIC POWER DIVISION (WAPD) AND BETTIS ATOMIC POWER LABORATORY (BAPL) CONSUME MOST OF THE QUADRANT. PHILLIPS PETROLEUM COMPANY ETR-E-2256, 12/1966. INL INDEX NO. 532-0642-00-706-021256, REV. F. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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.

Nuclear Reactors for Space Power, Understanding the Atom Series.

ERIC Educational Resources Information Center

Corliss, William R.

The historical development of rocketry and nuclear technology includes a specific description of Systems for Nuclear Auxiliary Power (SNAP) programs. Solar cells and fuel cells are considered as alternative power supplies for space use. Construction and operation of space power plants must include considerations of the transfer of heat energy to…

Propulsion and Power Technologies for the NASA Exploration Vision: A Research Perspective

NASA Technical Reports Server (NTRS)

Litchford, Ron J.

2004-01-01

Future propulsion and power technologies for deep space missions are profiled in this viewgraph presentation. The presentation includes diagrams illustrating possible future travel times to other planets in the solar system. The propulsion technologies researched at Marshall Space Flight Center (MSFC) include: 1) Chemical Propulsion; 2) Nuclear Propulsion; 3) Electric and Plasma Propulsion; 4) Energetics. The presentation contains additional information about these technologies, as well as space reactors, reactor simulation, and the Propulsion Research Laboratory (PRL) at MSFC.

Advanced Thermal Simulator Testing: Thermal Analysis and Test Results

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Dickens, Ricky; Dixon, David; Reid, Robert; Adams, Mike; Davis, Joe

2008-01-01

Work at the NASA Marshall Space Flight Center seeks to develop high fidelity, electrically heated thermal simulators that represent fuel elements in a nuclear reactor design to support non-nuclear testing applicable to the development of a space nuclear power or propulsion system. Comparison between the fuel pins and thermal simulators is made at the outer fuel clad surface, which corresponds to the outer sheath surface in the thermal simulator. The thermal simulators that are currently being tested correspond to a SNAP derivative reactor design that could be applied for Lunar surface power. These simulators are designed to meet the geometric and power requirements of a proposed surface power reactor design, accommodate testing of various axial power profiles, and incorporate imbedded instrumentation. This paper reports the results of thermal simulator analysis and testing in a bare element configuration, which does not incorporate active heat removal, and testing in a water-cooled calorimeter designed to mimic the heat removal that would be experienced in a reactor core.

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 infrastructures. Additionally, many Generation IV (Gen IV) reactor concepts have goals for optimizing investment recovery and economic efficiency that promote significant reductions in plant operations and maintenance staff over current-generation nuclear power plants. To accomplish these Gen IV goals and also address the SRPS remote-siting challenges, higher levels of automation, fault tolerance, and advanced diagnostic capabilities are needed to provide nearly autonomous operations with anticipatory maintenance. Essentially, the SRPS control system for several anticipated terrestrial applications can benefit from the kind of operational autonomy that is necessary for deep space and planetary SRPS-enabled missions. Investigation of the state of the technology for autonomous control confirmed that control systems with varying levels of autonomy have been employed in robotic, transportation, spacecraft, and manufacturing applications. As an example, NASA has pursued autonomy for spacecraft and surface exploration vehicles (e.g., rovers) to reduce mission costs, increase efficiency for communications between ground control and the vehicle, and enable independent operation of the vehicle during times of communications blackout. However, autonomous control has not been implemented for an operating terrestrial nuclear power plant nor has there been any experience beyond automating simple control loops for space reactors. Current automated control technologies for nuclear power plants are reasonably mature, and fully automated control of normal SRPS operations is clearly feasible. However, the space-based and remote terrestrial applications of SRPS modules require autonomous capabilities that can accommodate nonoptimum operations when degradation, failure, and other off-normal events challenge the performance of the reactor while immediate human intervention is not possible. The independent action provided by autonomous control, which is distinct from the more limited self action of automated control, can satisfy these conditions. Key characteristics that distinguish autonomous control include: (1) intelligence to confirm system performance and detect degraded or failed conditions, (2) optimization to minimize stress on SRPS components and efficiently react to operational events without compromising system integrity, (3) robustness to accommodate uncertainties and changing conditions, and (4) flexibility and adaptability to accommodate failures through reconfiguration among available control system elements or adjustment of control system strategies, algorithms, or parameters.« less

A Power Conversion Concept for the Jupiter Icy Moons Orbiter

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2003-01-01

The Jupiter Icy Moons Orbiter (JIMO) is a bold new mission under development by the Office of Space Science at NASA Headquarters. ITMO is examining the potential of Nuclear Electric Propulsion (NEP) technology to efficiently deliver scientific payloads to three Jovian moons: Callisto, Ganymede, and Europa. A critical element of the NEP vehicle is the reactor power system, consisting of the nuclear reactor, power conversion, heat rejection, and power management and distribution (PMAD). The emphasis of this paper is on the non-nuclear elements of the reactor power system.

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

NASA Technical Reports Server (NTRS)

Bowles, K. J.

1974-01-01

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

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.

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

NASA Technical Reports Server (NTRS)

Wetch, J. R.

1988-01-01

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

MTR,TRA603. EXPERIMENTERS' SPACE ALLOCATIONS IN BASEMENT AS OF 1963. SHIELDED ...

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

MTR,TRA-603. EXPERIMENTERS' SPACE ALLOCATIONS IN BASEMENT AS OF 1963. SHIELDED CUBICLES WERE IDENTIFIED BY SPONSORING LABORATORY AND ITS TEST HOLE NUMBER IN THE REACTOR, IE, "KAPL HB-1" SIGNIFIED KNOLLS ATOMIC POWER LABORATORY, HORIZONTAL BEAM NO. 1. "WAPD" WAS WESTINGHOUSE ATOMIC POWER DIVISION. CATCH TANKS AND SAMPLE STATIONS FOR TEST LOOPS WERE ASSOCIATED WITH THESE CUBICLES. NOTE DESKS, STORAGE CABINETS, SWITCH GEAR, INSTRUMENT PANELS. PHILLIPS PETROLEUM COMPANY MTR-E-5205, 4/1963. INL INDEX NO. 531-0603-00-706-009757, REV. 5. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

NASA Technical Reports Server (NTRS)

Springborn, R. H.

1971-01-01

Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

Legal and Regulatroy Obstacles to Nuclear Fission Technology in Space

NASA Astrophysics Data System (ADS)

Force, Melissa K.

2013-09-01

In forecasting the prospective use of small nuclear reactors for spacecraft and space-based power stations, the U.S. Air Force describes space as "the ultimate high ground," providing access to every part of the globe. But is it? A report titled "Energy Horizons: United States Air Force Energy Science &Technology Vision 2011-2026," focuses on core Air Force missions in space energy generation, operations and propulsion and recognizes that investments into small modular nuclear fission reactors can be leveraged for space-based systems. However, the report mentions, as an aside, that "potential catastrophic outcomes" are an element to be weighed and provides no insight into the monumental political and legal will required to overcome the mere stigma of nuclear energy, even when referring only to the most benign nuclear power generation systems - RTGs. On the heels of that report, a joint Department of Energy and NASA team published positive results from the demonstration of a uranium- powered fission reactor. The experiment was perhaps most notable for exemplifying just how effective the powerful anti-nuclear lobby has been in the United States: It was the first such demonstration of its kind in nearly fifty years. Space visionaries must anticipate a difficult war, consisting of multiple battles that must be waged in order to obtain a license to fly any but the feeblest of nuclear power sources in space. This paper aims to guide the reader through the obstacles to be overcome before nuclear fission technology can be put to use in space.

Nuclear energy technology

NASA Technical Reports Server (NTRS)

Buden, David

1992-01-01

An overview of space nuclear energy technologies is presented. The development and characteristics of radioisotope thermoelectric generators (RTG's) and space nuclear power reactors are discussed. In addition, the policy and issues related to public safety and the use of nuclear power sources in space are addressed.

Application of the Enabler to nuclear electric propulsion

NASA Astrophysics Data System (ADS)

Pierce, Bill L.

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

Studies of PuF sub 6 and transplutonic materials' critical properties for space high power nuclear pumped lasers

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

Gu, A.G.; Miller, M.S.

1991-01-01

All space missions require a reliable, compact source of energy. This paper describes preliminary neutronics studies of pocket'' reactor concepts employing PuF{sub 6} and transplutonic materials as fuels for space high power/energy Nuclear Pumped Lasers (NPLs). Previous research has studied NPL reactor concepts with thin fuel layers, aerosol fuels and gaseous UF{sub 6}. The total reactor volumes for compact reactors with these types of fuels typically range from 3 m{sup 3} to 50 m{sup 3}. By employing PuF{sub 6} and transplutonic fuels at the same low densities, a calculated value for Keff of 1.2 has been achieved for conditions ofmore » 900 K and 5 atm, with total reactor volumes of 1.5 m{sup 3} for PuF{sub 6}, 0.51 m{sup 3} for Am-242m, 0.58 m{sup 3} for Cm-245 and 0.63 m{sup 3} for Cf-249.« less

Preliminary survey of 21st century civil mission applications of space nuclear power

NASA Technical Reports Server (NTRS)

Mankins, John C.; Olivieri, J.; Hepenstal, A.

1987-01-01

The purpose was to collect and categorize a forecast of civilian space missions and their power requirements, and to assess the suitability of an SP-100 class space reactor power system to those missions. A wide variety of missions were selected for examination. The applicability of an SP-100 type of nuclear power system was assessed for each of the selected missions; a strawman nuclear power system configuration was drawn up for each mission. The main conclusions are as follows: (1) Space nuclear power in the 50 kW sub e plus range can enhance or enable a wide variety of ambitious civil space mission; (2) Safety issues require additional analyses for some applications; (3) Safe space nuclear reactor disposal is an issue for some applications; (4) The current baseline SP-100 conical radiator configuration is not applicable in all cases; (5) Several applications will require shielding greater than that provided by the baseline shadow-shield; and (6) Long duration, continuous operation, high reliability missions may exceed the currently designed SP-100 lifetime capabilities.

Space rocket engine on the base of the reactor-pumped laser for the interplanetary flights and earth orbital applications

NASA Astrophysics Data System (ADS)

Gulevich, Andrey V.; Dyachenko, Peter P.; Kukharchuk, Oleg F.; Zrodnikov, Anatoly V.

2000-01-01

In this report the concept of vehicle-based reactor-laser engine for long time interplanetary and interorbital (LEO to GEO) flights is proposed. Reactor-pumped lasers offer the perspective way to create on the base of modern nuclear and lasers technologies the low mass and high energy density, repetitively pulsed vehicle-based laser of average power 100 kW. Nowadays the efficiency of nuclear-to-optical energy conversion reached the value of 2-3%. The demo model of reactor-pumped laser facility is under construction in Institute for Physics and Power Engineering (Obninsk, Russia). It enable us to hope that using high power laser on board of the vehicle could make the effective space laser engine possible. Such engine may provide the high specific impulse ~1000-2000 s with the thrust up to 10-100 n. Some calculation results of the characteristics of vehicle-based reactor-laser thermal engine concept are also presented. .

Satellite nuclear power station: An engineering analysis

NASA Technical Reports Server (NTRS)

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

1973-01-01

A nuclear-MHD power plant system which uses a compact non-breeder reactor to produce power in the multimegawatt range is analyzed. It is shown that, operated in synchronous orbit, the plant would transmit power safely to the ground by a microwave beam. Fuel reprocessing would take place in space, and no radioactive material would be returned to earth. Even the effect of a disastrous accident would have negligible effect on earth. A hydrogen moderated gas core reactor, or a colloid-core, or NERVA type reactor could also be used. The system is shown to approach closely the ideal of economical power without pollution.

Realizing "2001: A Space Odyssey": Piloted Spherical Torus Nuclear Fusion Propulsion

NASA Technical Reports Server (NTRS)

Williams, Craig H.; Dudzinski, Leonard A.; Borowski, Stanley K.; Juhasz, Albert J.

2005-01-01

A conceptual vehicle design enabling fast, piloted outer solar system travel was created predicated on a small aspect ratio spherical torus nuclear fusion reactor. The initial requirements were satisfied by the vehicle concept, which could deliver a 172 mt crew payload from Earth to Jupiter rendezvous in 118 days, with an initial mass in low Earth orbit of 1,690 mt. Engineering conceptual design, analysis, and assessment was performed on all major systems including artificial gravity payload, central truss, nuclear fusion reactor, power conversion, magnetic nozzle, fast wave plasma heating, tankage, fuel pellet injector, startup/re-start fission reactor and battery bank, refrigeration, reaction control, communications, mission design, and space operations. Detailed fusion reactor design included analysis of plasma characteristics, power balance/utilization, first wall, toroidal field coils, heat transfer, and neutron/x-ray radiation. Technical comparisons are made between the vehicle concept and the interplanetary spacecraft depicted in the motion picture 2001: A Space Odyssey.

SP-100 Program: space reactor system and subsystem investigations

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

Harty, R.B.

1983-09-30

For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs.

SP-100 program: Space reactor system and subsystem investigations

NASA Astrophysics Data System (ADS)

Harty, R. B.

1983-09-01

For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. The nuclear safety review/approval process that is required for a space reactor system is summarized. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that is expected and to provide information that could be usable in future programs.

OAST Space Theme Workshop. Volume 3: Working group summary. 6: Power (P-2). A. Statement. B. Technology needs (form 1). C. Priority assessment (form 2)

NASA Technical Reports Server (NTRS)

1976-01-01

Power requirements for the multipurpose space power platform, for space industrialization, SETI, the solar system exploration facility, and for global services are assessed for various launch dates. Priorities and initiatives for the development of elements of space power systems are described for systems using light power input (solar energy source) or thermal power input, (solar, chemical, nuclear, radioisotopes, reactors). Systems for power conversion, power processing, distribution and control are likewise examined.

SP-100 design, safety, and testing

NASA Technical Reports Server (NTRS)

Cox, Carl. M.; Mahaffey, Michael M.; Smith, Gary L.

1991-01-01

The SP-100 Program is developing a nuclear reactor power system that can enhance and/or enable future civilian and military space missions. The program is directed to develop space reactor technology to provide electrical power in the range of tens to hundreds of kilowatts. The major nuclear assembly test is to be conducted at the Hanford Site near Richland, Washington, and is designed to validate the performance of the 2.4-MWt nuclear and heat transport assembly.

Autonomous Control of Space Reactor Systems

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

Belle R. Upadhyaya; K. Zhao; S.R.P. Perillo

2007-11-30

Autonomous and semi-autonomous control is a key element of space reactor design in order to meet the mission requirements of safety, reliability, survivability, and life expectancy. Interrestrial nuclear power plants, human operators are avilable to perform intelligent control functions that are necessary for both normal and abnormal operational conditions.

Space nuclear reactors — A post-operational disposal strategy

NASA Astrophysics Data System (ADS)

Angelo, Joseph A.; Buden, David

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

Analysis and Down Select of Flow Passages for Thermal Hydraulic Testing of a SNAP Derived Reactor

NASA Technical Reports Server (NTRS)

Godfroy, T. J.; Sadasivan, P.; Masterson, S.

2007-01-01

As past of the Vision for Space Exploration, man will return to the moon. To enable safe and productive time on the lunar surface will require adequate power resources. To provide the needed power and to give mission planners all landing site possibilities, including a permanently dark crater, a nuclear reactor provides the most options. Designed to be l00kWt providing approx. 25kWe this power plants would be very effective in delivering dependable, site non-specific power to crews or robotic missions on the lunar surface. An affordable reference reactor based upon the successful SNAP program of the 1960's and early 1970's has been designed by Los Alamos National Laboratory that will meet such a requirement. Considering current funding, environmental, and schedule limitations this lunar surface power reactor will be tested using non-nuclear simulators to simulate the heat from fission reactions. Currently a 25kWe surface power SNAP derivative reactor is in the early process of design and testing with collaboration between Los Alamos National Laboratory, Idaho National Laboratory, Glenn Research Center, Marshall Space Flight Center, and Sandia National Laboratory to ensure that this new design is affordable and can be tested using non-nuclear methods as have proven so effective in the past. This paper will discuss the study and down selection of a flow passage concept for a approx. 25kWe lunar surface power reactor. Several different flow passages designs were evaluated using computational fluid dynamics to determine pressure drop and a structural assessment to consider thermal and stress of the passage walls. The reactor design basis conditions are discussed followed by passage problem setup and results for each concept. A recommendation for passage design is made with rationale for selection.

Review of the Tri-Agency Space Nuclear Reactor Power System Technology Program

NASA Technical Reports Server (NTRS)

Ambrus, J. H.; Wright, W. E.; Bunch, D. F.

1984-01-01

The Space Nuclear Reactor Power System Technology Program designated SP-100 was created in 1983 by NASA, the U.S. Department of Defense, and the Defense Advanced Research Projects Agency. Attention is presently given to the development history of SP-100 over the course of its first year, in which it has been engaged in program objectives' definition, the analysis of civil and military missions, nuclear power system functional requirements' definition, concept definition studies, the selection of primary concepts for technology feasibility validation, and the acquisition of initial experimental and analytical results.

Design and optimization of the heat rejection system for a liquid cooled thermionic space nuclear reactor power system

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

Moriarty, M.P.

1993-01-15

The heat transport subsystem for a liquid metal cooled thermionic space nuclear power system was modelled using algorithms developed in support of previous nuclear power system study programs, which date back to the SNAP-10A flight system. The model was used to define the optimum dimensions of the various components in the heat transport subsystem subjected to the constraints of minimizing mass and achieving a launchable package that did not require radiator deployment. The resulting design provides for the safe and reliable cooling of the nuclear reactor in a proven lightweight design.

Design and optimization of the heat rejection system for a liquid cooled thermionic space nuclear reactor power system

NASA Astrophysics Data System (ADS)

Moriarty, Michael P.

1993-01-01

The heat transport subsystem for a liquid metal cooled thermionic space nuclear power system was modelled using algorithms developed in support of previous nuclear power system study programs, which date back to the SNAP-10A flight system. The model was used to define the optimum dimensions of the various components in the heat transport subsystem subjected to the constraints of minimizing mass and achieving a launchable package that did not require radiator deployment. The resulting design provides for the safe and reliable cooling of the nuclear reactor in a proven lightweight design.

Preliminary assessment of high power, NERVA-class dual-mode space nuclear propulsion and power systems

NASA Astrophysics Data System (ADS)

Buksa, John J.; Kirk, William L.; Cappiello, Michael W.

A preliminary assessment of the technical feasibility and mass competitiveness of a dual-mode nuclear propulsion and power system based on the NERVA rocket engine has been completed. Results indicate that the coupling of the Rover reactor to a direct Brayton power conversion system can be accomplished through a number of design features. Furthermore, based on previously published and independently calculated component masses, the dual-mode system was found to have the potential to be mass competitive with propulsion/power systems that use separate reactors. The uncertainties of reactor design modification and shielding requirements were identified as important issues requiring future investigation.

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).

Physics and potentials of fissioning plasmas for space power and propulsion

NASA Technical Reports Server (NTRS)

Thom, K.; Schwenk, F. C.; Schneider, R. T.

1976-01-01

Fissioning uranium plasmas are the nuclear fuel in conceptual high-temperature gaseous-core reactors for advanced rocket propulsion in space. A gaseous-core nuclear rocket would be a thermal reactor in which an enriched uranium plasma at about 10,000 K is confined in a reflector-moderator cavity where it is nuclear critical and transfers its fission power to a confining propellant flow for the production of thrust at a specific impulse up to 5000 sec. With a thrust-to-engine weight ratio approaching unity, the gaseous-core nuclear rocket could provide for propulsion capabilities needed for manned missions to the nearby planets and for economical cislunar ferry services. Fueled with enriched uranium hexafluoride and operated at temperatures lower than needed for propulsion, the gaseous-core reactor scheme also offers significant benefits in applications for space and terrestrial power. They include high-efficiency power generation at low specific mass, the burnup of certain fission products and actinides, the breeding of U-233 from thorium with short doubling times, and improved convenience of fuel handling and processing in the gaseous phase.

Multi-Megawatt Space Nuclear Power Generation

DTIC Science & Technology

1993-06-28

electric generation, both for open- and closed-cycle opera- tion. These reactors use the particulate fuel of the type developed for HTGR reactors. What...commercial HTGR power reactors, the particles are held in place and directly cooled. Figure 2.7 shows the two types of fuel particles developed for...of MW(e), for pulsed energy devices. The FBR would use HTGR -type particle fuel , contained in a annular bed be- tween two porous frits. Helium would

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

DTIC Science & Technology

1980-03-01

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

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 systems for space power and propulsion

NASA Technical Reports Server (NTRS)

Klein, M.

1971-01-01

As exploration and utilization of space proceeds through the 1970s, 1980s, and beyond, spacecraft in earth orbit will become increasingly larger, spacecraft will travel deeper into space, and space activities will involve more complex operations. These trends require increasing amounts of energy for power and propulsion. The role to be played by nuclear energy is presented, including plans for deep space missions using radioisotope generators, the reactor power systems for earth orbiting stations and satellites, and the role of nuclear propulsion in space transportation.

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.

Development and Utilization of Space Fission Power Systems

NASA Technical Reports Server (NTRS)

Houts, Michael G.; Mason, Lee S.; Palac, Donald T.; Harlow, Scott E.

2009-01-01

Space fission power systems could enable advanced civilian space missions. Terrestrially, thousands of fission systems have been operated since 1942. In addition, the US flew a space fission system in 1965, and the former Soviet Union flew 33 such systems prior to the end of the Cold War. Modern design and development practices, coupled with 65 years of experience with terrestrial reactors, could enable the affordable development of space fission power systems for near-term planetary surface applications.

Development and Utilization of Space Fission Power Systems

NASA Technical Reports Server (NTRS)

Houts, Michael; Mason, Lee S.; Palac, Donald T.; Harlow, Scott E.

2008-01-01

Space fission power systems could enable advanced civilian space missions. Terrestrially, thousands of fission systems have been operated since 1942. In addition, the US flew a space fission system in 1965, and the former Soviet Union flew 33 such systems prior to the end of the Cold War. Modern design and development practices, coupled with 65 years of experience with terrestrial reactors, could enable the affordable development of space fission power systems for near-term planetary surface applications.

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

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2000-01-01

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

Cermet-fueled reactors for advanced space applications

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

Cowan, C.L.; Palmer, R.S.; Taylor, I.N.

Cermet-fueled nuclear reactors are attractive candidates for high-performance advanced space power systems. The cermet consists of a hexagonal matrix of a refractory metal and a ceramic fuel, with multiple tubular flow channels. The high performance characteristics of the fuel matrix come from its high strength at elevated temperatures and its high thermal conductivity. The cermet fuel concept evolved in the 1960s with the objective of developing a reactor design that could be used for a wide range of mobile power generating sytems, including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests for the cermet fuel weremore » carried out by General Electric as part of the 710 Project (General Electric 1966), and by Argonne National Laboratory in the Direct Nuclear Rocket Program (1965). Development programs for cermet fuel are currently under way at Argonne National Laboratory and Pacific Northwest Laboratory. The high temperature qualification tests from the 1960s have provided a base for the incorporation of cermet fuel in advanced space applications. The status of the cermet fuel development activities and descriptions of the key features of the cermet-fueled reactor design are summarized in this paper.« less

The NASA CSTI high capacity power project

NASA Technical Reports Server (NTRS)

Winter, J.; Dudenhoefer, J.; Juhasz, A.; Schwarze, G.; Patterson, R.; Ferguson, D.; Titran, R.; Schmitz, P.; Vandersande, J.

1992-01-01

The SP-100 Space Nuclear Power Program was established in 1983 by DOD, DOE, and NASA as a joint program to develop technology for military and civil applications. Starting in 1986, NASA has funded a technology program to maintain the momentum of promising aerospace technology advancement started during Phase 1 of SP-100 and to strengthen, in key areas, the chances for successful development and growth capability of space nuclear reactor power systems for a wide range of future space applications. The elements of the Civilian Space Technology Initiative (CSTI) High Capacity Power Project include Systems Analysis, Stirling Power Conversion, Thermoelectric Power Conversion, Thermal Management, Power Management, Systems Diagnostics, Environmental Interactions, and Material/Structural Development. Technology advancement in all elements is required to provide the growth capability, high reliability and 7 to 10 year lifetime demanded for future space nuclear power systems. The overall project will develop and demonstrate the technology base required to provide a wide range of modular power systems compatible with the SP-100 reactor which facilitates operation during lunar and planetary day/night cycles as well as allowing spacecraft operation at any attitude or distance from the sun. Significant accomplishments in all of the project elements will be presented, along with revised goals and project timelines recently developed.

The NASA CSTI high capacity power project

NASA Astrophysics Data System (ADS)

Winter, J.; Dudenhoefer, J.; Juhasz, A.; Schwarze, G.; Patterson, R.; Ferguson, D.; Titran, R.; Schmitz, P.; Vandersande, J.

1992-08-01

The SP-100 Space Nuclear Power Program was established in 1983 by DOD, DOE, and NASA as a joint program to develop technology for military and civil applications. Starting in 1986, NASA has funded a technology program to maintain the momentum of promising aerospace technology advancement started during Phase 1 of SP-100 and to strengthen, in key areas, the chances for successful development and growth capability of space nuclear reactor power systems for a wide range of future space applications. The elements of the Civilian Space Technology Initiative (CSTI) High Capacity Power Project include Systems Analysis, Stirling Power Conversion, Thermoelectric Power Conversion, Thermal Management, Power Management, Systems Diagnostics, Environmental Interactions, and Material/Structural Development. Technology advancement in all elements is required to provide the growth capability, high reliability and 7 to 10 year lifetime demanded for future space nuclear power systems. The overall project will develop and demonstrate the technology base required to provide a wide range of modular power systems compatible with the SP-100 reactor which facilitates operation during lunar and planetary day/night cycles as well as allowing spacecraft operation at any attitude or distance from the sun. Significant accomplishments in all of the project elements will be presented, along with revised goals and project timelines recently developed.

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.

Improved Nuclear Reactor and Shield Mass Model for Space Applications

NASA Technical Reports Server (NTRS)

Robb, Kevin

2004-01-01

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

CALANDRIA TYPE SODIUM GRAPHITE REACTOR

DOEpatents

Peterson, R.M.; Mahlmeister, J.E.; Vaughn, N.E.; Sanders, W.J.; Williams, A.C.

1964-02-11

A sodium graphite power reactor in which the unclad graphite moderator and fuel elements are contained within a core tank is described. The core tank is submersed in sodium within the reactor vessel. Extending longitudinally through the core thnk are process tubes with fuel elements positioned therein. A bellows sealing means allows axial expansion and construction of the tubes. Within the core tank, a leakage plenum is located below the graphite, and above the graphite is a gas space. A vent line regulates the gas pressure in the space, and another line removes sodium from the plenum. The sodium coolant flows from the lower reactor vessel through the annular space between the fuel elements and process tubes and out into the reactor vessel space above the core tank. From there, the heated coolant is drawn off through an outlet line and sent to the heat exchange. (AEC)

SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

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

None

2013-09-25

U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in amore » remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.« less

SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

ScienceCinema

None

2018-01-16

U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in a remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.

Space Nuclear Facility test capability at the Baikal-1 and IGR sites Semipalatinsk-21, Kazakhstan

NASA Astrophysics Data System (ADS)

Hill, T. J.; Stanley, M. L.; Martinell, J. S.

1993-01-01

The International Space Technology Assessment Program was established 1/19/92 to take advantage of the availability of Russian space technology and hardware. DOE had two delegations visit CIS and assess its space nuclear power and propulsion technologies. The visit coincided with the Conference on Nuclear Power Engineering in Space Nuclear Rocket Engines at Semipalatinsk-21 (Kurchatov, Kazakhstan) on Sept. 22-25, 1992. Reactor facilities assessed in Semipalatinski-21 included the IVG-1 reactor (a nuclear furnace, which has been modified and now called IVG-1M), the RA reactor, and the Impulse Graphite Reactor (IGR), the CIS version of TREAT. Although the reactor facilities are being maintained satisfactorily, the support infrastructure appears to be degrading. The group assessment is based on two half-day tours of the Baikals-1 test facility and a brief (2 hr) tour of IGR; because of limited time and the large size of the tour group, it was impossible to obtain answers to all prepared questions. Potential benefit is that CIS fuels and facilities may permit USA to conduct a lower priced space nuclear propulsion program while achieving higher performance capability faster, and immediate access to test facilities that cannot be available in this country for 5 years. Information needs to be obtained about available data acquisition capability, accuracy, frequency response, and number of channels. Potential areas of interest with broad application in the U.S. nuclear industry are listed.

Space and energy. [space systems for energy generation, distribution and control

NASA Technical Reports Server (NTRS)

Bekey, I.

1976-01-01

Potential contributions of space to energy-related activities are discussed. Advanced concepts presented include worldwide energy distribution to substation-sized users using low-altitude space reflectors; powering large numbers of large aircraft worldwide using laser beams reflected from space mirror complexes; providing night illumination via sunlight-reflecting space mirrors; fine-scale power programming and monitoring in transmission networks by monitoring millions of network points from space; prevention of undetected hijacking of nuclear reactor fuels by space tracking of signals from tagging transmitters on all such materials; and disposal of nuclear power plant radioactive wastes in space.

Engineering design aspects of the heat-pipe power system

NASA Technical Reports Server (NTRS)

Capell, B. M.; Houts, M. G.; Poston, D. I.; Berte, M.

1997-01-01

The Heat-pipe Power System (HPS) is a near-term, low-cost space power system designed at Los Alamos that can provide up to 1,000 kWt for many space nuclear applications. The design of the reactor is simple, modular, and adaptable. The basic design allows for the use of a variety of power conversion systems and reactor materials (including the fuel, clad, and heat pipes). This paper describes a project that was undertaken to develop a database supporting many engineering aspects of the HPS design. The specific tasks discussed in this paper are: the development of an HPS materials database, the creation of finite element models that will allow a wide variety of investigations, and the verification of past calculations.

Engineering and Fabrication Considerations for Cost-Effective Space Reactor Shield Development

NASA Astrophysics Data System (ADS)

Berg, Thomas A.; Disney, Richard K.

2004-02-01

Investment in developing nuclear power for space missions cannot be made on the basis of a single mission. Current efforts in the design and fabrication of the reactor module, including the reactor shield, must be cost-effective and take into account scalability and fabricability for planned and future missions. Engineering considerations for the shield need to accommodate passive thermal management, varying radiation levels and effects, and structural/mechanical issues. Considering these challenges, design principles and cost drivers specific to the engineering and fabrication of the reactor shield are presented that contribute to lower recurring mission costs.

Engineering and Fabrication Considerations for Cost-Effective Space Reactor Shield Development

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

Berg, Thomas A.; Disney, Richard K.

Investment in developing nuclear power for space missions cannot be made on the basis of a single mission. Current efforts in the design and fabrication of the reactor module, including the reactor shield, must be cost-effective and take into account scalability and fabricability for planned and future missions. Engineering considerations for the shield need to accommodate passive thermal management, varying radiation levels and effects, and structural/mechanical issues. Considering these challenges, design principles and cost drivers specific to the engineering and fabrication of the reactor shield are presented that contribute to lower recurring mission costs.

Reliability and mass analysis of dynamic power conversion systems with parallel of standby redundancy

NASA Technical Reports Server (NTRS)

Juhasz, A. J.; Bloomfield, H. S.

1985-01-01

A combinatorial reliability approach is used to identify potential dynamic power conversion systems for space mission applications. A reliability and mass analysis is also performed, specifically for a 100 kWe nuclear Brayton power conversion system with parallel redundancy. Although this study is done for a reactor outlet temperature of 1100K, preliminary system mass estimates are also included for reactor outlet temperatures ranging up to 1500 K.

Potential civil mission applications for space nuclear power systems

NASA Technical Reports Server (NTRS)

Ambrus, J. H.; Beatty, R. G. G.

1985-01-01

It is pointed out that the energy needs of spacecraft over the last 25 years have been met by photovoltaic arrays with batteries, primary fuel cells, and radioisotope thermoelectric generators (RTG). However, it might be difficult to satisfy energy requirements for the next generation of space missions with the currently used energy sources. Applications studies have emphasized the need for a lighter, cheaper, and more compact high-energy source than the scaling up of current technologies would permit. These requirements could be satisfied by a nuclear reactor power system. The joint NASA/DOD/DOE SP-100 program is to explore and evaluate this option. Critical elements of the technology are also to be developed, taking into account space reactor systems of the 100 kW class. The present paper is concerned with some of the civil mission categories and concepts which are enabled or significantly enhanced by the performance characteristics of a nuclear reactor energy system.

A nuclear driven metallic vapor MHD coupled with MPD thrusters

NASA Technical Reports Server (NTRS)

Anghaie, Samim; Kumar, Ratan

1991-01-01

Nuclear energy as a source of power for space missions, represents an enabling technology for advanced and ambitious space applications. Nuclear fuel in a gaseous or liquid form has been configured as a promising and practical candidate in this regard. The present study investigates and performs a feasibility analysis of an innovative concept for space power generation and propulsion. The system embodies a conceptual nuclear reactor with an MHD generator and coupled to MPD thrusters. The reactor utilizes liquid uranium in droplet form as fuel and superheated metallic vapor as the working fluid. This ultrahigh temperature vapor core reactor brings forward varied and challenging technical issues, and it has been addressed to in this paper. A parametric study of the conceived system has been performed in a qualitative and quantitative manner. Preliminary results show enough promise for further indepth analysis of this novel system.

Thermally Simulated 32kW Direct-Drive Gas-Cooled Reactor: Design, Assembly, and Test

NASA Astrophysics Data System (ADS)

Godfroy, Thomas J.; Kapernick, Richard J.; Bragg-Sitton, Shannon M.

2004-02-01

One of the power systems under consideration for nuclear electric propulsion is a direct-drive gas-cooled reactor coupled to a Brayton cycle. In this system, power is transferred from the reactor to the Brayton system via a circulated closed loop gas. To allow early utilization, system designs must be relatively simple, easy to fabricate, and easy to test using non-nuclear heaters to closely mimic heat from fission. This combination of attributes will allow pre-prototypic systems to be designed, fabricated, and tested quickly and affordably. The ability to build and test units is key to the success of a nuclear program, especially if an early flight is desired. The ability to perform very realistic non-nuclear testing increases the success probability of the system. In addition, the technologies required by a concept will substantially impact the cost, time, and resources required to develop a successful space reactor power system. This paper describes design features, assembly, and test matrix for the testing of a thermally simulated 32kW direct-drive gas-cooled reactor in the Early Flight Fission - Test Facility (EFF-TF) at Marshall Space Flight Center. The reactor design and test matrix are provided by Los Alamos National Laboratories.

Summary of aerospace and nuclear engineering activities

NASA Technical Reports Server (NTRS)

1988-01-01

The Texas A&M Nuclear and Aerospace engineering departments have worked on five different projects for the NASA/USRA Advanced Design Program during the 1987/88 year. The aerospace department worked on two types of lunar tunnelers that would create habitable space. The first design used a heated cone to melt the lunar regolith, and the second used a conventional drill to bore its way through the crust. Both used a dump truck to get rid of waste heat from the reactor as well as excess regolith from the tunneling operation. The nuclear engineering department worked on three separate projects. The NEPTUNE system is a manned, outer-planetary explorer designed with Jupiter exploration as the baseline mission. The lifetime requirement for both reactor and power-conversion systems was twenty years. The second project undertaken for the power supply was a Mars Sample Return Mission power supply. This was designed to produce 2 kW of electrical power for seven years. The design consisted of a General Purpose Heat Source (GPHS) utilizing a Stirling engine as the power conversion unit. A mass optimization was performed to aid in overall design. The last design was a reactor to provide power for propulsion to Mars and power on the surface. The requirements of 300 kW of electrical power output and a mass of less than 10,000 Rg were set. This allowed the reactor and power conversion unit to fit within the Space Shuttle cargo bay.

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

NASA Technical Reports Server (NTRS)

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

1991-01-01

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

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

NASA Technical Reports Server (NTRS)

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

1991-01-01

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

In Space Nuclear Power as an Enabling Technology for Deep Space Exploration

NASA Technical Reports Server (NTRS)

Sackheim, Robert L.; Houts, Michael

2000-01-01

Deep Space Exploration missions, both for scientific and Human Exploration and Development (HEDS), appear to be as weight limited today as they would have been 35 years ago. Right behind the weight constraints is the nearly equally important mission limitation of cost. Launch vehicles, upper stages and in-space propulsion systems also cost about the same today with the same efficiency as they have had for many years (excluding impact of inflation). Both these dual mission constraints combine to force either very expensive, mega systems missions or very light weight, but high risk/low margin planetary spacecraft designs, such as the recent unsuccessful attempts for an extremely low cost mission to Mars during the 1998-99 opportunity (i.e., Mars Climate Orbiter and the Mars Polar Lander). When one considers spacecraft missions to the outer heliopause or even the outer planets, the enormous weight and cost constraints will impose even more daunting concerns for mission cost, risk and the ability to establish adequate mission margins for success. This paper will discuss the benefits of using a safe in-space nuclear reactor as the basis for providing both sufficient electric power and high performance space propulsion that will greatly reduce mission risk and significantly increase weight (IMLEO) and cost margins. Weight and cost margins are increased by enabling much higher payload fractions and redundant design features for a given launch vehicle (higher payload fraction of IMLEO). The paper will also discuss and summarize the recent advances in nuclear reactor technology and safety of modern reactor designs and operating practice and experience, as well as advances in reactor coupled power generation and high performance nuclear thermal and electric propulsion technologies. It will be shown that these nuclear power and propulsion technologies are major enabling capabilities for higher reliability, higher margin and lower cost deep space missions design to reliably reach the outer planets for scientific exploration.

Nuclear power for space based systems

NASA Astrophysics Data System (ADS)

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

1991-09-01

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

Nuclear safety for the space exploration initiative

NASA Technical Reports Server (NTRS)

Dix, Terry E.

1991-01-01

The results of a study to identify potential hazards arising from nuclear reactor power systems for use on the lunar and Martian surfaces, related safety issues, and resolutions of such issues by system design changes, operating procedures, and other means are presented. All safety aspects of nuclear reactor power systems from prelaunch ground handling to eventual disposal were examined consistent with the level of detail for SP-100 reactor design at the 1988 System Design Review and for launch vehicle and space transport vehicle designs and mission descriptions as defined in the 90-day Space Exploration Initiative (SEI) study. Information from previous aerospace nuclear safety studies was used where appropriate. Safety requirements for the SP-100 space nuclear reactor system were compiled. Mission profiles were defined with emphasis on activities after low earth orbit insertion. Accident scenarios were then qualitatively defined for each mission phase. Safety issues were identified for all mission phases with the aid of simplified event trees. Safety issue resolution approaches of the SP-100 program were compiled. Resolution approaches for those safety issues not covered by the SP-100 program were identified. Additionally, the resolution approaches of the SP-100 program were examined in light of the moon and Mars missions.

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.

Introduction to Reactor Statics Modules, RS-1. Nuclear Engineering Computer Modules.

ERIC Educational Resources Information Center

Edlund, Milton C.

The nine Reactor Statics Modules are designed to introduce students to the use of numerical methods and digital computers for calculation of neutron flux distributions in space and energy which are needed to calculate criticality, power distribution, and fuel burn-up for both slow neutron and fast neutron fission reactors. The diffusion…

Early Program Development

NASA Image and Video Library

1963-01-01

This artist's concept from 1963 shows a proposed NERVA (Nuclear Engine for Rocket Vehicle Application) incorporating the NRX-A1, the first NERVA-type cold flow reactor. The NERVA engine, based on Kiwi nuclear reactor technology, was intended to power a RIFT (Reactor-In-Flight-Test) nuclear stage, for which Marshall Space Flight Center had development responsibility.

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.-

Liquid Metal Pump Technologies for Nuclear Surface Power

NASA Technical Reports Server (NTRS)

Polzin, Kurt A.

2007-01-01

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

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

NASA Astrophysics Data System (ADS)

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

2005-02-01

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

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

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

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

2005-02-06

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

Thermionic fast spectrum reactor-converter on the basis of multi-cell TFE

NASA Astrophysics Data System (ADS)

Ponomarev-Stepnoi, N. N.; Kompaniets, G. V.; Poliakov, D. N.; Stepennov, B. S.; Andreev, P. V.; Zhabotinsky, E. E.; Nikolaev, Yu. V.; Lapochkin, N. V.

2001-02-01

Today Russian experts have technological experience in development of in-core thermionic converters for reactors of space nuclear power plants. Such a converter contains nuclear fuel inside and really represents a fuel element of a reactor. Two types of reactors can be considered on the basis of these thermionic fuel elements: with thermal or intermediate neutron spectrum, and with fast neutron spectrum. The first type is characterized by the presence of moderator in core that ensures most economical usage of nuclear fuel. The estimation shows that moderated system is the most effective in the power range of about 5 ... 100 kWe. The power systems of higher level are characterized by larger dimensions due to the presence of moderator. The second type of reactor is considered for higher power levels. This power range is about hundreds kWe. Dimensions of the fast reactor and core configuration are determined by the necessity to ensure the required net output power, on the one hand, and the necessity to ensure critical state on the other hand. In the case of using in-core thermionic fuel elements of the specified design, minimal reactor output power is determined by reactor criticality condition, and maximum reactor power output is determined by specifications and launcher capabilities. In the present paper the effective multiplication factor of a fast spectrum reactor on the basis of a multi-cell TFE developed by ``Lutch'' is considered a function of the total number of TFEs in the reactor. The MCU Monte-Carlo code, developed in Russia (Alekseev, et al., 1991), was used for computations. TFE computational models are placed in the nodes of a uniform triangular lattice and surrounded with pressure vessel and a side reflector. Ordinary fuel pins without thermionic converters were used instead of some TFEs to optimize criticality parameters, dimensions and output power of the reactor. General weight parameters of the reactor are presented in the paper. .

Operators in the Plum Brook Reactor Facility Control Room

NASA Image and Video Library

1970-03-21

Donald Rhodes, left, and Clyde Greer, right, monitor the operation of the National Aeronautics and Space Administration’s (NASA) Plum Brook Reactor Facility from the control room. The 60-megawatt test reactor, NASA’s only reactor, was the eighth largest test reactor in the world. The facility was built by the Lewis Research Center in the late 1950s to study the effects of radiation on different materials that could be used to construct nuclear propulsion systems for aircraft or rockets. The reactor went critical for the first time in 1961. For the next two years, two operators were on duty 24 hours per day working on the fission process until the reactor reached its full-power level in 1963. Reactor Operators were responsible for monitoring and controlling the reactor systems. Once the reactor was running under normal operating conditions, the work was relatively uneventful. Normally the reactor was kept at a designated power level within certain limits. Occasionally the operators had to increase the power for a certain test. The shift supervisor and several different people would get together and discuss the change before boosting the power. All operators were required to maintain a Reactor Operator License from the Atomic Energy Commission. The license included six months of training, an eight-hour written exam, a four-hour walkaround, and testing on the reactor controls.

Simulated nuclear reactor fuel assembly

DOEpatents

Berta, V.T.

1993-04-06

An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

A feasibility assessment of installation, operation and disposal options for nuclear reactor power system concepts for a NASA growth space station

NASA Technical Reports Server (NTRS)

Bloomfield, Harvey S.; Heller, Jack A.

1987-01-01

A preliminary feasibility assessment of the integration of reactor power system concepts with a projected growth space station architecture was conducted to address a variety of installation, operational disposition, and safety issues. A previous NASA sponsored study, which showed the advantages of space station - attached concepts, served as the basis for this study. A study methodology was defined and implemented to assess compatible combinations of reactor power installation concepts, disposal destinations, and propulsion methods. Three installation concepts that met a set of integration criteria were characterized from a configuration and operational viewpoint, with end-of-life disposal mass identified. Disposal destinations that met current aerospace nuclear safety criteria were identified and characterized from an operational and energy requirements viewpoint, with delta-V energy requirement as a key parameter. Chemical propulsion methods that met current and near-term application criteria were identified and payload mass and delta-V capabilities were characterized. These capabilities were matched against concept disposal mass and destination delta-V requirements to provide the feasibility of each combination.

A feasibility assessment of nuclear reactor power system concepts for the NASA Growth Space Station

NASA Technical Reports Server (NTRS)

Bloomfield, H. S.; Heller, J. A.

1986-01-01

A preliminary feasibility assessment of the integration of reactor power system concepts with a projected growth Space Station architecture was conducted to address a variety of installation, operational, disposition and safety issues. A previous NASA sponsored study, which showed the advantages of Space Station - attached concepts, served as the basis for this study. A study methodology was defined and implemented to assess compatible combinations of reactor power installation concepts, disposal destinations, and propulsion methods. Three installation concepts that met a set of integration criteria were characterized from a configuration and operational viewpoint, with end-of-life disposal mass identified. Disposal destinations that met current aerospace nuclear safety criteria were identified and characterized from an operational and energy requirements viewpoint, with delta-V energy requirement as a key parameter. Chemical propulsion methods that met current and near-term application criteria were identified and payload mass and delta-V capabilities were characterized. These capabilities were matched against concept disposal mass and destination delta-V requirements to provide a feasibility of each combination.

Simulated nuclear reactor fuel assembly

DOEpatents

Berta, Victor T.

1993-01-01

An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

TECHNICAL AND MEDICAL ASPECTS OF NUCLAR-POWERED AIRPLANES: THE ROLE OF NUCLEAR POWER IN THE ATMOSPHERE AND IN SPACE (in Swedish)

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

von Ublisch, H.

1961-01-01

An evaluation is given of the merits of potential nuclear propulsion systems for aircraft and rockets as well as of small nuclear power plants for auxiliary use in space exploration. The protection of personnel and passengers against gamma and high-velocity radiation is discussed, and the shielding properties of various materials are analysed. Mobile shielding would be required for airplanes even when landed and the reactor is shut down. No significant pollutton of the atmosphere is expected from leaking reactors, but accidents constitute a real danger. The prospects of realizing the ion motor and the photon motor are speculated upon. (auth)

The use of dual mode thermionic reactors in supporting Earth orbital and space exploration missions

NASA Astrophysics Data System (ADS)

Zubrin, Robert M.; Sulmeisters, Tal K.

1993-01-01

Missions requiring large amounts of electric power to support their payload functions can be enabled through the employment of nuclear electric power reactors, which in some cases can also assist the mission by making possible the employment of high specific impulse electric propulsion. However it is found that the practicality and versality of using a power reactor to provide advanced propulsion is enormously enhanced if the reactor is configured in such a way to allow it to generate a certain amount of direct thrust as well. The use of such a system allows the creation of a common bus upper stage that can provide both high power and high impulse (with short orbit transfer times). It is shown that such a system, termed an Integral Power and Propulsion Stage (IPAPS), is optimal for supporting many Earth, Lunar, planetary and asteroidal observation, exploration, and communication support missions, and it is therefore recommended that the nuclear power reactor ultimately selected by the government for development and production be one that can be configured for such a function.

Reactor Testing and Qualification: Prioritized High-level Criticality Testing Needs

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

S. Bragg-Sitton; J. Bess; J. Werner

2011-09-01

Researchers at the Idaho National Laboratory (INL) were tasked with reviewing possible criticality testing needs to support development of the fission surface power system reactor design. Reactor physics testing can provide significant information to aid in development of technologies associated with small, fast spectrum reactors that could be applied for non-terrestrial power systems, leading to eventual system qualification. Several studies have been conducted in recent years to assess the data and analyses required to design and build a space fission power system with high confidence that the system will perform as designed [Marcille, 2004a, 2004b; Weaver, 2007; Parry et al.,more » 2008]. This report will provide a summary of previous critical tests and physics measurements that are potentially applicable to the current reactor design (both those that have been benchmarked and those not yet benchmarked), summarize recent studies of potential nuclear testing needs for space reactor development and their applicability to the current baseline fission surface power (FSP) system design, and provide an overview of a suite of tests (separate effects, sub-critical or critical) that could fill in the information database to improve the accuracy of physics modeling efforts as the FSP design is refined. Some recommendations for tasks that could be completed in the near term are also included. Specific recommendations on critical test configurations will be reserved until after the sensitivity analyses being conducted by Los Alamos National Laboratory (LANL) are completed (due August 2011).« less

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.

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.

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

NASA Astrophysics Data System (ADS)

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

2004-02-01

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

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

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

Buden, D.

1993-08-11

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

Summary of space nuclear reactor power systems, 1983 - 1992

NASA Astrophysics Data System (ADS)

Buden, D.

1993-08-01

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

Startup thaw concept for the SP-100 space reactor power system

NASA Technical Reports Server (NTRS)

Kirpich, A.; Das, A.; Choe, H.; Mcnamara, E.; Switick, D.; Bhandari, P.

1990-01-01

A thaw concept for a space reactor power system which employs lithium as a circulant for both the heat-transport and the heat-rejection fluid loops is presented. An exemplary thermal analysis for a 100-kWe (i.e., SP-100) system is performed. It is shown that the design of the thaw system requires a thorough knowledge of the various physical states of the circulant throughout the system, both spatially and temporally, and that the design has to provide adequate margins for the system to avoid a structural or thermally induced damage.

Development costs for a nuclear electric propulsion stage.

NASA Technical Reports Server (NTRS)

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

1973-01-01

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

Compatibility tests of materials for a lithium-cooled space power reactor concept

NASA Technical Reports Server (NTRS)

Sinclair, J. H.

1973-01-01

Materials for a lithium-cooled space power reactor concept must be chemically compatible for up to 50,000 hr at high temperature. Capsule tests at 1040 C (1900 F) were made of material combinations of prime interest: T-111 in direct contact with uranium mononitride (UN), Un in vacuum separated from T-111 by tungsten wire, UN with various oxygen impurity levels enclosed in tungsten wire lithium-filled T-111 capsules, and TZM and lithium together in T-111 capsules. All combinations were compatible for over 2800 hr except for T-111 in direct contact with UN.

Space-Based Solar Power System Architecture

DTIC Science & Technology

2012-12-01

alternatives to fossil fuels: nuclear fission reactors, hydroelectric power, wind turbines and solar power to name just a few. Each has advantages...powered laser could give people or wildlife serious burns or cause blindness and possibly even death depending on exposure proximity and time. 11

Accident analysis of heavy water cooled thorium breeder reactor

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

Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

2015-04-16

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k,more » and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition. The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.« less

Accident analysis of heavy water cooled thorium breeder reactor

NASA Astrophysics Data System (ADS)

Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

2015-04-01

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition. The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.

Analysis of a boron-carbide-drum-controlled critical reactor experiment

NASA Technical Reports Server (NTRS)

Mayo, W. T.

1972-01-01

In order to validate methods and cross sections used in the neutronic design of compact fast-spectrum reactors for generating electric power in space, an analysis of a boron-carbide-drum-controlled critical reactor was made. For this reactor the transport analysis gave generally satisfactory results. The calculated multiplication factor for the most detailed calculation was only 0.7-percent Delta k too high. Calculated reactivity worth of the control drums was $11.61 compared to measurements of $11.58 by the inverse kinetics methods and $11.98 by the inverse counting method. Calculated radial and axial power distributions were in good agreement with experiment.

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

None

Progress is reported on fundamental research in: crystal physics, reactions at metal surfaces, spectroscopy of ionic media, structure of metals, theory of alloying, physical properties, sintering, deformation of crystalline solids, x ray diffraction, metallurgy of superconducting materials, and electron microscope studies. Long-randge applied research studies were conducted for: zirconium metallurgy, materials compatibility, solid reactions, fuel element development, mechanical properties, non-destructive testing, and high-temperature materials. Reactor development support work was carried out for: gas-cooled reactor program, molten-salt reactor, high-flux isotope reactor, space-power program, thorium-utilization program, advanced-test reactor, Army Package Power Reactor, Enrico Fermi fast-breeder reactor, and water desalination program. Other programmore » activities, for which research was conducted, included: thermonuclear project, transuraniunn program, and post-irradiation examination laboratory. Separate abstracts were prepared for 30 sections of the report. (B.O.G.)« less

Observation of nuclear reactors on satellites with a balloon-borne gamma-ray telescope

NASA Technical Reports Server (NTRS)

O'Neill, Terrence J.; Kerrick, Alan D.; Ait-Ouamer, Farid; Tumer, O. Tumay; Zych, Allen D.

1989-01-01

Four Soviet nuclear-powered satellites flying over a double Compton gamma-ray telescope resulted in the detection of gamma rays with 0.3-8.0 MeV energies on April 15, 1988, as the balloonborne telescope searched, from a 35-km altitude, for celestial gamma-ray sources. The satellites included Cosmos 1900 and 1932. The USSR is the only nation currently employing moderated nuclear reactors for satellite power; reactors in space may cause significant problems for gamma-ray astronomy by increasing backgrounds, especially in the case of gamma-ray bursts.

Design of a heatpipe-cooled Mars-surface fission reactor

NASA Astrophysics Data System (ADS)

Poston, David I.; Kapernick, Richard J.; Guffee, Ray M.; Reid, Robert S.; Lipinski, Ronald J.; Wright, Steven A.; Talandis, Regina A.

2002-01-01

The next generation of robotic missions to Mars will most likely require robust power sources in the range of 3 to 20 kWe. Fission systems are well suited to provide safe, reliable, and economic power within this range. The goal of this study is to design a compact, low-mass fission system that meets Mars-surface power requirements, while maintaining a high level of safety and reliability at a relatively low cost. The Heatpipe Power System (HPS) is one possible approach for producing near-term, low-cost, space fission power. The goal of the HPS project is to devise an attractive space fission system that can be developed quickly and affordably. The primary ways of doing this are by using existing technology and by designing the system for inexpensive testing. If the system can be designed to allow highly prototypic testing with electrical heating, then an exhaustive test program can be carried out quickly and inexpensively, and thorough testing of the actual flight unit can be performed-which is a major benefit to reliability. Over the past 4 years, three small HPS proof-of-concept technology demonstrations have been conducted, and each has been highly successful. The Heatpipe-Operated Mars Exploration Reactor (HOMER) is a derivative of the HPS designed especially for producing power on the surface of Mars. The HOMER-15 is a 15-kWt reactor that couples with a 3-kWe Stirling engine power system. The reactor contains stainless-steel (SS)-clad uranium nitride (UN) fuel pins that are structurally and thermally bonded to SS/sodium heatpipes. Fission energy is conducted from the fuel pins to the heatpipes, which then carry the heat to the Stirling engine. This paper describes the attributes, specifications, and performance of a 15-kWt HOMER reactor. .

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

NASA Technical Reports Server (NTRS)

Martin, James J.; Reid, Robert S.

2004-01-01

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

System Design for a Nuclear Electric Spacecraft Utilizing Out-of-core Thermionic Conversion

NASA Technical Reports Server (NTRS)

Estabrook, W. C.; Phillips, W. M.; Hsieh, T.

1976-01-01

Basic guidelines are presented for a nuclear space power system which utilizes heat pipes to transport thermal power from a fast nuclear reactor to an out of core thermionic converter array. Design parameters are discussed for the nuclear reactor, heat pipes, thermionic converters, shields (neutron and gamma), waste heat rejection systems, and the electrical bus bar-cable system required to transport the high current/low voltage power to the processing equipment. Dimensions are compatible with shuttle payload bay constraints.

Institute of Electrical and Electronics Engineers, Nuclear Science Symposium, 18th, and Nuclear Power Systems Symposium, 3rd, San Francisco, Calif., November 3-5, 1971, Proceedings.

NASA Technical Reports Server (NTRS)

1972-01-01

Potential advantages of fusion power reactors are discussed together with the protection of the public from radioactivity produced in nuclear power reactors, and the significance of tritium releases to the environment. Other subjects considered are biomedical instrumentation, radiation damage problems, low level environmental radionuclide analysis systems, nuclear techniques in environmental research, nuclear instrumentation, and space and plasma instrumentation. Individual items are abstracted in this issue.

Support vector machines for nuclear reactor state estimation

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

Zavaljevski, N.; Gross, K. C.

2000-02-14

Validation of nuclear power reactor signals is often performed by comparing signal prototypes with the actual reactor signals. The signal prototypes are often computed based on empirical data. The implementation of an estimation algorithm which can make predictions on limited data is an important issue. A new machine learning algorithm called support vector machines (SVMS) recently developed by Vladimir Vapnik and his coworkers enables a high level of generalization with finite high-dimensional data. The improved generalization in comparison with standard methods like neural networks is due mainly to the following characteristics of the method. The input data space is transformedmore » into a high-dimensional feature space using a kernel function, and the learning problem is formulated as a convex quadratic programming problem with a unique solution. In this paper the authors have applied the SVM method for data-based state estimation in nuclear power reactors. In particular, they implemented and tested kernels developed at Argonne National Laboratory for the Multivariate State Estimation Technique (MSET), a nonlinear, nonparametric estimation technique with a wide range of applications in nuclear reactors. The methodology has been applied to three data sets from experimental and commercial nuclear power reactor applications. The results are promising. The combination of MSET kernels with the SVM method has better noise reduction and generalization properties than the standard MSET algorithm.« less

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

NASA Technical Reports Server (NTRS)

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

1992-01-01

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

The past as prologue - A look at historical flight qualifications for space nuclear systems

NASA Technical Reports Server (NTRS)

Bennett, Gary L.

1992-01-01

Currently the U.S. is sponsoring production of radioisotope thermoelectric generators (RTGs) for the Cassini mission to Saturn; the SP-100 space nuclear reactor power system for NASA applications; a thermionic space reactor program for DoD applications as well as early work on nuclear propulsion. In an era of heightened public concern about having successful space ventures it is important that a full understanding be developed of what it means to 'flight qualify' a space nuclear system. As a contribution to the ongoing work this paper reviews several qualification programs, including the general-purpose heat source radioisotope thermoelectric generators (GPHS-RTGs) as developed for the Galileo and Ulysses missions, the SNAP-10A space reactor, the Nuclear Engine for Rocket Vehicle Applications (NERVA), the F-1 chemical engine used on the Saturn-V, and the Space Shuttle Main Engines (SSMEs). Similarities and contrasts are noted.

The past as prologue - A look at historical flight qualifications for space nuclear systems

NASA Astrophysics Data System (ADS)

Bennett, Gary L.

Currently the U.S. is sponsoring production of radioisotope thermoelectric generators (RTGs) for the Cassini mission to Saturn; the SP-100 space nuclear reactor power system for NASA applications; a thermionic space reactor program for DoD applications as well as early work on nuclear propulsion. In an era of heightened public concern about having successful space ventures it is important that a full understanding be developed of what it means to 'flight qualify' a space nuclear system. As a contribution to the ongoing work this paper reviews several qualification programs, including the general-purpose heat source radioisotope thermoelectric generators (GPHS-RTGs) as developed for the Galileo and Ulysses missions, the SNAP-10A space reactor, the Nuclear Engine for Rocket Vehicle Applications (NERVA), the F-1 chemical engine used on the Saturn-V, and the Space Shuttle Main Engines (SSMEs). Similarities and contrasts are noted.

Trade studies for nuclear space power systems

NASA Technical Reports Server (NTRS)

Smith, John M.; Bents, David J.; Bloomfield, Harvey S.

1991-01-01

As human visions of space applications expand and as we probe further out into the universe, our needs for power will also expand, and missions will evolve which are enabled by nuclear power. A broad spectrum of missions which are enhanced or enabled by nuclear power sources have been defined. These include Earth orbital platforms, deep space platforms, planetary exploration, and terrestrial resource exploration. The recently proposed Space Exploration Initiative (SEI) to the Moon and Mars has more clearly defined these missions and their power requirements. Presented here are results of recent studies of radioisotope and nuclear reactor energy sources, combined with various energy conversion devices for Earth orbital applications, SEI lunar/Mars rovers, surface power, and planetary exploration.

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

NASA Technical Reports Server (NTRS)

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

2005-01-01

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

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).

Reactor Statics Module, RS-9: Multigroup Diffusion Program Using an Exponential Acceleration Technique.

ERIC Educational Resources Information Center

Macek, Victor C.

The nine Reactor Statics Modules are designed to introduce students to the use of numerical methods and digital computers for calculation of neutron flux distributions in space and energy which are needed to calculate criticality, power distribution, and fuel burnup for both slow neutron and fast neutron fission reactors. The last module, RS-9,…

SNAP (Space Nuclear Auxiliary Power) reactor overview. Final report, June 1982-December 1983

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

Voss, S.S.

1984-08-01

The SNAP reactor programs are outlined in this report. A summary of the program is included along with a technical outline of the SER, S2DR, SNAP 10A/SNAPSHOT, S8ER, and S8DR reactor systems. Specifications of the designs, the design logic and a conclusion outlining some of the program weaknesses are given.

Silane-Pyrolysis Reactor With Nonuniform Heating

NASA Technical Reports Server (NTRS)

Iya, Sridhar K.

1991-01-01

Improved reactor serves as last stage in system processing metallurgical-grade silicon feedstock into silicon powder of ultrahigh purity. Silane pyrolized to silicon powder and hydrogen gas via homogeneous decomposition reaction in free space. Features set of individually adjustable electrical heaters and purge flow of hydrogen to improve control of pyrolysis conditions. Power supplied to each heater set in conjunction with flow in reactor to obtain desired distribution of temperature as function of position along reactor.

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

NASA Technical Reports Server (NTRS)

1972-01-01

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

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

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.

Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

NASA Astrophysics Data System (ADS)

Dixon, David D.; Hiatt, Matthew T.; Poston, David I.; Kapernick, Richard J.

2006-01-01

A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

Structural Materials and Fuels for Space Power Plants

NASA Technical Reports Server (NTRS)

Bowman, Cheryl; Busby, Jeremy; Porter, Douglas

2008-01-01

A fission reactor combined with Stirling convertor power generation is one promising candidate in on-going Fission Surface Power (FSP) studies for future lunar and Martian bases. There are many challenges for designing and qualifying space-rated nuclear power plants. In order to have an affordable and sustainable program, NASA and DOE designers want to build upon the extensive foundation in nuclear fuels and structural materials. This talk will outline the current Fission Surface Power program and outline baseline design options for a lunar power plant with an emphasis on materials challenges. NASA first organized an Affordable Fission Surface Power System Study Team to establish a reference design that could be scrutinized for technical and fiscal feasibility. Previous papers and presentations have discussed this study process in detail. Considerations for the reference design included that no significant nuclear technology, fuels, or material development were required for near term use. The desire was to build upon terrestrial-derived reactor technology including conventional fuels and materials. Here we will present an overview of the reference design, Figure 1, and examine the materials choices. The system definition included analysis and recommendations for power level and life, plant configuration, shielding approach, reactor type, and power conversion type. It is important to note that this is just one concept undergoing refinement. The design team, however, understands that materials selection and improvement must be an integral part of the system development.

A critical review of the state of foreign space technology

NASA Technical Reports Server (NTRS)

Grey, J.; Gerard, M.

1978-01-01

A conference was held to exchange technical information in the area of space technology. Soviet system capability and technology both in Intersputnik and in the domestic Ekran system was discussed in detail. The thermonic power conversion system used in the Soviet Topaz nuclear power reactor was described in detail. Other areas of examination included: (1) Bioastronautics; (2) Space based industry; (3) Propulsion; (4) Astrodynamics; (5) Contact with extraterrestrial intelligence; and (6) Space rescue and safety.

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 expected within the next 30 to 50 years, as predicted by the Hubbert model and confirmed by other global energy consumption prognoses. Having invested national resources into the development of NGNP, the technology and experience accumulated during the project needs to be documented clearly and in sufficient detail for young engineers coming on-board at both DOE and NASA to acquire it. Hands on training on reactor operation, test rigs of turbomachinery, and heat exchanger components, as well as computational tools will be needed. Senior scientist/engineers involved with the development of NGNP should also be encouraged to participate as lecturers, instructors, or adjunct professors at local universities having engineering (mechanical, electrical, nuclear/chemical, and/or materials) as one of their fields of study.

Orbital Space Solar Power Option for a Lunar Village

NASA Technical Reports Server (NTRS)

Johnson, Les

2017-01-01

One of the most significant challenges to the implementation of a continuously manned lunar base is power. During the lunar day (14 Earth days), it is conceptually simple to deploy solar arrays to generate the estimated 35 kilowatts of continuous power required. However, generating this level of power during the lunar night (also 14 Earth days) has been an extremely difficult problem to solve. Conventional solutions range from the requirement that the base be located at the lunar south pole so as to take advantage of the continuous sunshine available there to developing a space-qualified nuclear reactor and power plant to generate the needed energy. There is a third option: Use the soon-to-be-available Space Launch System to place a space based solar power station in lunar orbit that would beam the needed energy to the lunar base. Several detailed studies have been performed by NASA, universities and others looking at the lunar south pole for locating the base. The results are encouraging: by taking advantage of the moon's orbital tilt, large solar arrays can be deployed there to track the sun continuously and generate the power needed to sustain the base. The problem with this approach is inherent to its design: it will only work at the lunar south pole. There is no other site on the Moon with geometry favorable to generating continuous solar power. NASA has also considered the development of a compact fission reactor and power plant to generate the needed power, allowing the base to be sited anywhere on the Moon. The problem with this approach is that there are no space fission reactors available, none are being planned and the cost of developing one is prohibitively expensive. Using an orbiting space based solar power station to generate electrical power and beam it to a base sited anywhere on the moon should therefore be considered. The technology to collect sunlight, generate greater than the estimated 35 kilowatts of power, and beam it to the surface using microwaves is available today. The problem with this concept in the past would have been the mass and packaging volume (for launch) required to put such a system in place in lunar orbit. This problem is potentially solved with the advent of the Space Launch System (SLS). The SLS, with its 70 mT launch capacity, it more than capable of placing such a system into lunar orbit in a single launch. This paper will examine the potential use of an SLS-launched, space solar power system in lunar orbit as the primary power source for a first-generation, continuously-occupied lunar base and compare it with the other power generation and storage options previously considered.

Magnetic Materials Suitable for Fission Power Conversion in Space Missions

NASA Technical Reports Server (NTRS)

Bowman, Cheryl L.

2012-01-01

Terrestrial fission reactors use combinations of shielding and distance to protect power conversion components from elevated temperature and radiation. Space mission systems are necessarily compact and must minimize shielding and distance to enhance system level efficiencies. Technology development efforts to support fission power generation scenarios for future space missions include studying the radiation tolerance of component materials. The fundamental principles of material magnetism are reviewed and used to interpret existing material radiation effects data for expected fission power conversion components for target space missions. Suitable materials for the Fission Power System (FPS) Project are available and guidelines are presented for bounding the elevated temperature/radiation tolerance envelope for candidate magnetic materials.

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.

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.

Fission Surface Power Technology Development Update

NASA Technical Reports Server (NTRS)

Palac, Donald T.; Mason, Lee S.; Houts, Michael G.; Harlow, Scott

2011-01-01

Power is a critical consideration in planning exploration of the surfaces of the Moon, Mars, and places beyond. Nuclear power is an important option, especially for locations in the solar system where sunlight is limited or environmental conditions are challenging (e.g., extreme cold, dust storms). NASA and the Department of Energy are maintaining the option for fission surface power for the Moon and Mars by developing and demonstrating technology for a fission surface power system. The Fission Surface Power Systems project has focused on subscale component and subsystem demonstrations to address the feasibility of a low-risk, low-cost approach to space nuclear power for surface missions. Laboratory demonstrations of the liquid metal pump, reactor control drum drive, power conversion, heat rejection, and power management and distribution technologies have validated that the fundamental characteristics and performance of these components and subsystems are consistent with a Fission Surface Power preliminary reference concept. In addition, subscale versions of a non-nuclear reactor simulator, using electric resistance heating in place of the reactor fuel, have been built and operated with liquid metal sodium-potassium and helium/xenon gas heat transfer loops, demonstrating the viability of establishing system-level performance and characteristics of fission surface power technologies without requiring a nuclear reactor. While some component and subsystem testing will continue through 2011 and beyond, the results to date provide sufficient confidence to proceed with system level technology readiness demonstration. To demonstrate the system level readiness of fission surface power in an operationally relevant environment (the primary goal of the Fission Surface Power Systems project), a full scale, 1/4 power Technology Demonstration Unit (TDU) is under development. The TDU will consist of a non-nuclear reactor simulator, a sodium-potassium heat transfer loop, a power conversion unit with electrical controls, and a heat rejection system with a multi-panel radiator assembly. Testing is planned at the Glenn Research Center Vacuum Facility 6 starting in 2012, with vacuum and liquid-nitrogen cold walls to provide simulation of operationally relevant environments. A nominal two-year test campaign is planned including a Phase 1 reactor simulator and power conversion test followed by a Phase 2 integrated system test with radiator panel heat rejection. The testing is expected to demonstrate the readiness and availability of fission surface power as a viable power system option for NASA's exploration needs. In addition to surface power, technology development work within this project is also directly applicable to in-space fission power and propulsion systems.

Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

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

Sheryl Morton; Carl Baily; Tom Hill

Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. Itmore » provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a lowtemperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.« less

Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

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

Morton, Sheryl L.; Baily, Carl E.; Hill, Thomas J.

Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. Itmore » provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a low-temperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.« less

Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

NASA Astrophysics Data System (ADS)

Morton, Sheryl L.; Baily, Carl E.; Hill, Thomas J.; Werner, James E.

2006-01-01

Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. It provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a low-temperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.

NEUTRON DENSITY CONTROL IN A NEUTRONIC REACTOR

DOEpatents

Young, G.J.

1959-06-30

The method and means for controlling the neutron density in a nuclear reactor is described. It describes the method and means for flattening the neutron density distribution curve across the reactor by spacing the absorbing control members to varying depths in the central region closer to the center than to the periphery of the active portion of the reactor to provide a smaller neutron reproduction ratio in the region wherein the members are inserted, than in the remainder of the reactor thereby increasing the over-all potential power output.

Visible spectral power emitted from a laser produced uranium plasma

NASA Technical Reports Server (NTRS)

Williams, M. D.; Jalufka, N. W.

1975-01-01

The development of plasma-core nuclear reactors for advanced terrestrial and space-power sources is researched. Experimental measurements of the intensity and the spectral distribution of radiation from a nonfissioning uranium plasma are reported.

Megawatt Class Nuclear Space Power Systems (MCNSPS) conceptual design and evaluation report. Volume 3, technologies 2: Power conversion

NASA Technical Reports Server (NTRS)

Wetch, J. R.

1988-01-01

The major power conversion concepts considered for the Megawatt Class Nuclear Space Power System (MCNSPS) are discussed. These concepts include: (1) Rankine alkali-metal-vapor turbine alternators; (2) in-core thermionic conversion; (3) Brayton gas turbine alternators; and (4) free piston Stirling engine linear alternators. Considerations important to the coupling of these four conversion alternatives to an appropriate nuclear reactor heat source are examined along with the comparative performance characteristics of the combined systems meeting MCNSPS requirements.

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.

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

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.; Sawicki, Jerzy T.

2003-01-01

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

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

NASA Astrophysics Data System (ADS)

Juhasz, Albert J.; Sawicki, Jerzy T.

2004-02-01

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

A Compact Nuclear Fusion Reactor for Space Flights

NASA Astrophysics Data System (ADS)

Nastoyashchiy, Anatoly F.

2006-05-01

A small-scale nuclear fusion reactor is suggested based on the concepts of plasma confinement (with a high pressure gas) which have been patented by the author. The reactor considered can be used as a power setup in space flights. Among the advantages of this reactor is the use of a D3He fuel mixture which at burning gives main reactor products — charged particles. The energy balance considerably improves, as synchrotron radiation turn out "captured" in the plasma volume, and dangerous, in the case of classical magnetic confinement, instabilities in the direct current magnetic field configuration proposed do not exist. As a result, the reactor sizes are quite suitable (of the order of several meters). A possibility of making reactive thrust due to employment of ejection of multiply charged ions formed at injection of pellets from some adequate substance into the hot plasma center is considered.

Space station program phase B definition: Nuclear reactor-powered space station cost and schedules

NASA Technical Reports Server (NTRS)

1971-01-01

Tabulated data are presented on the costs, schedules, and technical characteristics for the space station phases C and D program. The work breakdown structure, schedule data, program ground rules, program costs, cost-estimating rationale, funding schedules, and supporting data are included.

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

NASA Astrophysics Data System (ADS)

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

2006-01-01

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

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.

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.

Radiation Shielding Design and Orientation Considerations for a 1 kWe Heat Pipe Cooled Reactor Utilized to Bore Through the Ice Caps of Mars

NASA Astrophysics Data System (ADS)

Fensin, Michael L.; Elliott, John O.; Lipinski, Ronald J.; Poston, David I.

2006-01-01

The goal in designing any space power system is to develop a system able to meet the mission requirements for success while minimizing the overall costs. The mission requirements for the this study was to develop a reactor (with Stirling engine power conversion) and shielding configuration able to fit, along with all the other necessary science equipment, in a Cryobot 3 m high with ~0.5 m diameter hull, produce 1 kWe for 5yrs, and not adversely affect the mission science by keeping the total integrated dose to the science equipment below 150 krad. Since in most space power missions the overall system mass dictates the mission cost, the shielding designs in this study incorporated Martian water extracted at the startup site in order to minimize the tungsten and LiH mass loading at launch. Different reliability and mass minimization concerns led to three design configuration evolutions. With the help of implementing Martian water and configuring the reactor as far from the science equipment as possible, the needed tungsten and LiH shield mass was minimized. This study further characterizes the startup dose and the necessary mission requirements in order to ensure integrity of the surface equipment during reactor startup phase.

Marshall Space Flight Center and the Reactor-in-Flight Stage: A Look Back at Using Nuclear Propulsion to Power Space Vehicles in the 1960's

NASA Technical Reports Server (NTRS)

Wright, Mike

2003-01-01

This paper examines the Marshall Space Flight Center s role in the Reactor-In-Flight (RIlT) project that NASA was involved with in the early 1960 s. The paper outlines the project s relation to the joint NASA-Atomic Energy Commission nuclear initiative known as Project Rover. It describes the justification for the RIFT project, its scope, and the difficulties that were encountered during the project. It also provides as assessment of NASA s overall capabilities related to nuclear propulsion in the early 1960 s.

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

NASA Astrophysics Data System (ADS)

Bess, John Darrell

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

SP-100 flight qualification testing assessment

NASA Technical Reports Server (NTRS)

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

1988-01-01

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

The Satellite Nuclear Power Station - An option for future power generation.

NASA Technical Reports Server (NTRS)

Williams, J. R.; Clement, J. D.

1973-01-01

A new concept in nuclear power generation is being explored which essentially eliminates major objections to nuclear power. The Satellite Nuclear Power Station, remotely operated in synchronous orbit, would transmit power safely to the ground by a microwave beam. Fuel reprocessing would take place in space and no radioactive materials would ever be returned to earth. Even the worst possible accident to such a plant should have negligible effect on the earth. An exploratory study of a satellite nuclear power station to provide 10,000 MWe to the earth has shown that the system could weigh about 20 million pounds and cost less than $1000/KWe. An advanced breeder reactor operating with an MHD power cycle could achieve an efficiency of about 50% with a 1100 K radiator temperature. If a hydrogen moderated gas core reactor is used, its breeding ratio of 1.10 would result in a fuel doubling time of a few years. A rotating fluidized bed or NERVA type reactor might also be used. The efficiency of power transmission from synchronous orbit would range from 70% to 80%.

A Small Fission Power System with Stirling Power Conversion for NASA Science Missions

NASA Technical Reports Server (NTRS)

Mason, Lee; Carmichael, Chad

2011-01-01

In early 2010, a joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) study team developed a concept for a 1 kWe Fission Power System with a 15-year design life that could be available for a 2020 launch to support future NASA science missions. The baseline concept included a solid block uranium-molybdenum reactor core with embedded heat pipes and distributed thermoelectric converters directly coupled to aluminum radiator fins. A short follow-on study was conducted at NASA Glenn Research Center (GRC) to evaluate an alternative power conversion approach. The GRC study considered the use of free-piston Stirling power conversion as a substitution to the thermoelectric converters. The resulting concept enables a power increase to 3 kWe with the same reactor design and scalability to 10 kW without changing the reactor technology. This paper presents the configuration layout, system performance, mass summary, and heat transfer analysis resulting from the study.

A Summary of Closed Brayton Cycle Development Activities at NASA

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2009-01-01

NASA has been involved in the development of Closed Brayton Cycle (CBC) power conversion technology since the 1960's. CBC systems can be coupled to reactor, isotope, or solar heat sources and offer the potential for high efficiency, long life, and scalability to high power. In the 1960's and 1970's, NASA and industry developed the 10 kW Brayton Rotating Unit (BRU) and the 2 kW mini-BRU demonstrating technical feasibility and performance, In the 1980's, a 25 kW CBC Solar Dynamic (SD) power system option was developed for Space Station Freedom and the technology was demonstrated in the 1990's as part of the 2 kW SO Ground Test Demonstration (GTD). Since the early 2000's, NASA has been pursuing CBC technology for space reactor applications. Before it was cancelled, the Jupiter Icy Moons Orbiter (HMO) mission was considering a 100 kWclass CBC system coupled to a gas-cooled fission reactor. Currently, CBC technology is being explored for Fission Surface Power (FSP) systems to provide base power on the moon and Mars. These recent activities have resulted in several CBC-related technology development projects including a 50 kW Alternator Test Unit, a 20 kW Dual Brayton Test Loop, a 2 kW Direct Drive Gas Brayton Test Loop, and a 12 kW FSP Power Conversion Unit design.

Development concept for a small, split-core, heat-pipe-cooled nuclear reactor

NASA Technical Reports Server (NTRS)

Lantz, E.; Breitwieser, R.; Niederauer, G. F.

1974-01-01

There have been two main deterrents to the development of semiportable nuclear reactors. One is the high development costs; the other is the inability to satisfy with assurance the questions of operational safety. This report shows how a split-core, heat-pipe cooled reactor could conceptually eliminate these deterrents, and examines and summarizes recent work on split-core, heat-pipe reactors. A concept for a small reactor that could be developed at a comparatively low cost is presented. The concept would extend the technology of subcritical radioisotope thermoelectric generators using 238 PuO2 to the evolution of critical space power reactors using 239 PuO2.

Comparison of evolving photovoltaic and nuclear power systems for earth orbital applications

NASA Technical Reports Server (NTRS)

Rockey, D. E.; Jones, R. M.; Schulman, I.

1982-01-01

Photovoltaic and fission reactor orbital power systems are compared in terms of the end-to-end system power-to-mass ratios. Three PV systems are examined, i.e., a solid substrate with a cell array and a NiCd battery, a modified SEP array and an NiH2 battery, and a 62-micron Si cell array and a fuel cell. All arrays were modeled to be 13.5% efficient and to produce 25 kW dc. The SP-100 reactor consists of the heat source, radiation shield, heat pipes to transfer thermal energy from the reactor to thermoelectric elements, and a waste heat radiator. Consideration is given to system applications in orbits ranging from LEO to GEO, and to mission durations of 1, 5, and 10 yr. PV systems are concluded to be flight-proven, useful out of radiation belts, and best for low to moderate power levels. Limitations exist for operations where atmospheric drag may become a factor and due to the size of a large PV power supply. Space nuclear reactors will continue under development and uses at high power levels and in low altitude orbits are foreseen.

A Closed Brayton Power Conversion Unit Concept for Nuclear Electric Propulsion for Deep Space Missions

NASA Astrophysics Data System (ADS)

Joyner, Claude Russell; Fowler, Bruce; Matthews, John

2003-01-01

In space, whether in a stable satellite orbit around a planetary body or traveling as a deep space exploration craft, power is just as important as the propulsion. The need for power is especially important for in-space vehicles that use Electric Propulsion. Using nuclear power with electric propulsion has the potential to provide increased payload fractions and reduced mission times to the outer planets. One of the critical engineering and design aspects of nuclear electric propulsion at required mission optimized power levels is the mechanism that is used to convert the thermal energy of the reactor to electrical power. The use of closed Brayton cycles has been studied over the past 30 or years and shown to be the optimum approach for power requirements that range from ten to hundreds of kilowatts of power. It also has been found to be scalable to higher power levels. The Closed Brayton Cycle (CBC) engine power conversion unit (PCU) is the most flexible for a wide range of power conversion needs and uses state-of-the-art, demonstrated engineering approaches. It also is in use with many commercial power plants today. The long life requirements and need for uninterrupted operation for nuclear electric propulsion demands high reliability from a CBC engine. A CBC engine design for use with a Nuclear Electric Propulsion (NEP) system has been defined based on Pratt & Whitney's data from designing long-life turbo-machines such as the Space Shuttle turbopumps and military gas turbines and the use of proven integrated control/health management systems (EHMS). An integrated CBC and EHMS design that is focused on using low-risk and proven technologies will over come many of the life-related design issues. This paper will discuss the use of a CBC engine as the power conversion unit coupled to a gas-cooled nuclear reactor and the design trends relative to its use for powering electric thrusters in the 25 kWe to 100kWe power level.

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

NASA Astrophysics Data System (ADS)

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

2005-02-01

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

Status of NASA's Stirling Space Power Converter Program

NASA Technical Reports Server (NTRS)

Dudenhoefer, James E.; Winter, Jerry M.

1991-01-01

An overview is presented of the NASA-Lewis Free-Piston Stirling Space Power Convertor Technology Program. The goal is to develop the technology base needed to meet the long duration, high capacity power requirements for future NASA space initiatives. Efforts are focused upon increasing system power output and system thermal and electric energy conversion efficiency at least fivefold over current SP-100 technology, and on achieving systems that are compatible with space nuclear reactors. Stirling experience in space and progress toward 1050 and 1300 K Stirling Space Power Converters is discussed. Fabrication is nearly completed for the 1050 K Component Test Power Converters (CTPC); results of motoring tests of cold end (525 K), are presented. The success of these and future designs is dependent upon supporting research and technology efforts including heat pipes, bearings, superalloy joining technologies, high efficiency alternators, life and reliability testing and predictive methodologies. An update is provided of progress in some of these technologies leading off with a discussion of free-piston Stirling experience in space.

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.

Nuclear Propulsion through Direct Conversion of Fusion Energy: The Fusion Driven Rocket

NASA Technical Reports Server (NTRS)

Slough, John; Pancotti, Anthony; Kirtley, David; Pihl, Christopher; Pfaff, Michael

2012-01-01

The future of manned space exploration and development of space depends critically on the creation of a dramatically more proficient propulsion architecture for in-space transportation. A very persuasive reason for investigating the applicability of nuclear power in rockets is the vast energy density gain of nuclear fuel when compared to chemical combustion energy. Current nuclear fusion efforts have focused on the generation of electric grid power and are wholly inappropriate for space transportation as the application of a reactor based fusion-electric system creates a colossal mass and heat rejection problem for space application.

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.

Development of NASA's Small Fission Power System for Science and Human Exploration

NASA Technical Reports Server (NTRS)

Gibson, Marc A.; Mason, Lee; Bowman, Cheryl; Poston, David I.; McClure, Patrick R.; Creasy, John; Robinson, Chris

2014-01-01

Exploration of our solar system has brought great knowledge to our nation's scientific and engineering community over the past several decades. As we expand our visions to explore new, more challenging destinations, we must also expand our technology base to support these new missions. NASA's Space Technology Mission Directorate is tasked with developing these technologies for future mission infusion and continues to seek answers to many existing technology gaps. One such technology gap is related to compact power systems (greater than 1 kWe) that provide abundant power for several years where solar energy is unavailable or inadequate. Below 1 kWe, Radioisotope Power Systems have been the workhorse for NASA and will continue, assuming its availability, to be used for lower power applications similar to the successful missions of Voyager, Ulysses, New Horizons, Cassini, and Curiosity. Above 1 kWe, fission power systems become an attractive technology offering a scalable modular design of the reactor, shield, power conversion, and heat transport subsystems. Near term emphasis has been placed in the 1-10kWe range that lies outside realistic radioisotope power levels and fills a promising technology gap capable of enabling both science and human exploration missions. History has shown that development of space reactors is technically, politically, and financially challenging and requires a new approach to their design and development. A small team of NASA and DOE experts are providing a solution to these enabling FPS technologies starting with the lowest power and most cost effective reactor series named "Kilopower" that is scalable from approximately 1-10 kWe.

Development of NASA's Small Fission Power System for Science and Human Exploration

NASA Technical Reports Server (NTRS)

Gibson, Marc A.; Mason, Lee S.; Bowman, Cheryl L.; Poston, David I.; McClure, Patrick R.; Creasy, John; Robinson, Chris

2015-01-01

Exploration of our solar system has brought many exciting challenges to our nations scientific and engineering community over the past several decades. As we expand our visions to explore new, more challenging destinations, we must also expand our technology base to support these new missions. NASAs Space Technology Mission Directorate is tasked with developing these technologies for future mission infusion and continues to seek answers to many existing technology gaps. One such technology gap is related to compact power systems (1 kWe) that provide abundant power for several years where solar energy is unavailable or inadequate. Below 1 kWe, Radioisotope Power Systems have been the workhorse for NASA and will continue to be used for lower power applications similar to the successful missions of Voyager, Ulysses, New Horizons, Cassini, and Curiosity. Above 1 kWe, fission power systems become an attractive technology offering a scalable modular design of the reactor, shield, power conversion, and heat transport subsystems. Near term emphasis has been placed in the 1-10kWe range that lies outside realistic radioisotope power levels and fills a promising technology gap capable of enabling both science and human exploration missions. History has shown that development of space reactors is technically, politically, and financially challenging and requires a new approach to their design and development. A small team of NASA and DOE experts are providing a solution to these enabling FPS technologies starting with the lowest power and most cost effective reactor series named Kilopower that is scalable from approximately 1-10 kWe.

Coaxial tube array space transmission line characterization

NASA Technical Reports Server (NTRS)

Switzer, Colleen A.; Bents, David J.

1987-01-01

The coaxial tube array tether/transmission line used to connect an SP-100 nuclear power system to the space station was characterized over the range of reactor-to-platform separation distances of 1 to 10 km. Characterization was done with respect to array performance, physical dimensions and masses. Using a fixed design procedure, a family of designs was generated for the same power level (300 kWe), power loss (1.5 percent), and meteoroid survival probability (99.5 percent over 10 yr). To differentiate between vacuum insulated and gas insulated lines, two different maximum values of the E field were considered: 20 kV/cm (appropriate to vacuum insulation) and 50 kV/cm (compressed SF6). Core conductor, tube, bumper, standoff, spacer and bumper support dimensions, and masses were also calculated. The results of the characterization show mainly how transmission line size and mass scale with reactor-to-platform separation distance.

Coaxial tube array space transmission line characterization

NASA Astrophysics Data System (ADS)

Switzer, Colleen A.; Bents, David J.

The coaxial tube array tether/transmission line used to connect an SP-100 nuclear power system to the space station was characterized over the range of reactor-to-platform separation distances of 1 to 10 km. Characterization was done with respect to array performance, physical dimensions and masses. Using a fixed design procedure, a family of designs was generated for the same power level (300 kWe), power loss (1.5 percent), and meteoroid survival probability (99.5 percent over 10 yr). To differentiate between vacuum insulated and gas insulated lines, two different maximum values of the E field were considered: 20 kV/cm (appropriate to vacuum insulation) and 50 kV/cm (compressed SF6). Core conductor, tube, bumper, standoff, spacer and bumper support dimensions, and masses were also calculated. The results of the characterization show mainly how transmission line size and mass scale with reactor-to-platform separation distance.

Advanced shield development for a fission surface power system for the lunar surface

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

A. E. Craft; I. J. Silver; C. M. Clark

A nuclear reactor power system such as the affordable fission surface power system enables a potential outpostonthemoon.Aradiation shieldmustbe included in the reactor system to reduce the otherwise excessive dose to the astronauts and other vital system components. The radiation shield is typically the most massive component of a space reactor system, and thus must be optimized to reduce mass asmuchas possible while still providing the required protection.Various shield options for an on-lander reactor system are examined for outpost distances of 400m and 1 kmfromthe reactor. Also investigated is the resulting mass savings from the use of a high performance cermetmore » fuel. A thermal analysis is performed to determine the thermal behaviours of radiation shields using borated water. For an outpost located 1000m from the core, a tetramethylammonium borohydride shield is the lightest (5148.4 kg), followed by a trilayer shield (boron carbide–tungsten–borated water; 5832.3 kg), and finally a borated water shield (6020.7 kg). In all of the final design cases, the temperature of the borated water remains below 400 K.« less

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.

Systems definition space based power conversion systems: Executive summary

NASA Technical Reports Server (NTRS)

1977-01-01

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

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.

Space fusion energy conversion using a field reversed configuration reactor: A new technical approach for space propulsion and power

NASA Technical Reports Server (NTRS)

Schulze, Norman R.; Miley, George H.; Santarius, John F.

1991-01-01

The fusion energy conversion design approach, the Field Reversed Configuration (FRC) - when burning deuterium and helium-3, offers a new method and concept for space transportation with high energy demanding programs, like the Manned Mars Mission and planetary science outpost missions require. FRC's will increase safety, reduce costs, and enable new missions by providing a high specific power propulsion system from a high performance fusion engine system that can be optimally designed. By using spacecraft powered by FRC's the space program can fulfill High Energy Space Missions (HESM) in a manner not otherwise possible. FRC's can potentially enable the attainment of high payload mass fractions while doing so within shorter flight times.

Nuclear safety considerations in the conceptual design of a fast reactor for space electric power and propulsion

NASA Technical Reports Server (NTRS)

Hsieh, T.-M.; Koenig, D. R.

1977-01-01

Some nuclear safety aspects of a 3.2 mWt heat pipe cooled fast reactor with out-of-core thermionic converters are discussed. Safety related characteristics of the design including a thin layer of B4C surrounding the core, the use of heat pipes and BeO reflector assembly, the elimination of fuel element bowing, etc., are highlighted. Potential supercriticality hazards and countermeasures are considered. Impacts of some safety guidelines of space transportation system are also briefly discussed, since the currently developing space shuttle would be used as the primary launch vehicle for the nuclear electric propulsion spacecraft.

Nuclear power--key to man's extraterrestrial civilization

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

Angelo, J.A.; Buden, D.

1982-08-01

The start of the Third Millennium will be highlighted by the establishment of man's extraterrestrial civilization with three technical cornerstones leading to the off-planet expansion of the human resource base. These are the availability of compact energy sources for power and propulsion, the creation of permanent manned habitats in space, and the ability to process materials anywhere in the Solar System. In the 1990s and beyond, nuclear reactors could represent the prime source of both space power and propulsion. The manned and unmanned space missions of tomorrow will demand first kilowatt and then megawatt levels of power. Various nuclear powermore » plant technologies are discussed, with emphasis on derivatives from the nuclear rocket technology.« less

Free-piston Stirling Engine system considerations for various space power applications

NASA Technical Reports Server (NTRS)

Dochat, George R.; Dhar, Manmohan

1991-01-01

Free-Piston Stirling Engines (FPSE) have the potential to provide high reliability, long life, and efficient operation. Therefore, they are excellent candidates for the dynamic power conversion module of a space-based, power-generating system. FPSE can be coupled with many potential heat sources (radioisotope, solar, or nuclear reactor), various heat input systems (pumped loop, heat pipe), heat rejection (pumped loop or heat pipe), and various power management and distribution systems (ac, dc, high or low voltage, and fixed or variable load). This paper reviews potential space missions that can be met using free-piston Stirling engines and discusses options of various system integration approaches. This paper briefly outlines the program and recent progress.

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

NASA Technical Reports Server (NTRS)

1972-01-01

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

Programmatic status of NASA's CSTI high capacity power Stirling space power converter program

NASA Technical Reports Server (NTRS)

Dudenhoefer, James E.

1990-01-01

An overview is presented of the NASA Lewis Research Center Free-Piston Stirling Space Power Converter Technology Development Program. This work is being conducted under NASA's Civil Space Technology Initiative (CSTI). The goal of the CSTI High Capacity Power element is to develop the technology base needed to meet the long duration, high capacity power requirements for future NASA space initiatives. Efforts are focused upon increasing system thermal and electric energy conversion efficiency at least fivefold over current SP-100 technology, and on achieving systems that are compatible with space nuclear reactors. The status of test activities with the Space Power Research Engine (SPRE) is discussed. Design deficiencies are gradually being corrected and the power converter is now outputting 11.5 kWe at a temperature ratio of 2 (design output is 12.5 kWe). Detail designs were completed for the 1050 K Component Test Power Converter (CTPC). The success of these and future designs is dependent upon supporting research and technology efforts including heat pipes, gas bearings, superalloy joining technologies and high efficiency alternators. An update of progress in these technologies is provided.

High Power LaB6 Plasma Source Performance for the Lockheed Martin Compact Fusion Reactor Experiment

NASA Astrophysics Data System (ADS)

Heinrich, Jonathon

2016-10-01

Lockheed Martin's Compact Fusion Reactor (CFR) concept is a linear encapsulated ring cusp. Due to the complex field geometry, plasma injection into the device requires careful consideration. A high power thermionic plasma source (>0.25MW; >10A/cm2) has been developed with consideration to phase space for optimal coupling. We present the performance of the plasma source, comparison with alternative plasma sources, and plasma coupling with the CFR field configuration. ©2016 Lockheed Martin Corporation. All Rights Reserved.

Integrated head package cable carrier for a nuclear power plant

DOEpatents

Meuschke, Robert E.; Trombola, Daniel M.

1995-01-01

A cabling arrangement is provided for a nuclear reactor located within a containment. Structure inside the containment is characterized by a wall having a near side surrounding the reactor vessel defining a cavity, an operating deck outside the cavity, a sub-space below the deck and on a far side of the wall spaced from the near side, and an operating area above the deck. The arrangement includes a movable frame supporting a plurality of cables extending through the frame, each connectable at a first end to a head package on the reactor vessel and each having a second end located in the sub-space. The frame is movable, with the cables, between a first position during normal operation of the reactor when the cables are connected to the head package, located outside the sub-space proximate the head package, and a second position during refueling when the cables are disconnected from the head package, located in the sub-space. In a preferred embodiment, the frame straddles the top of the wall in a substantially horizontal orientation in the first position, pivots about an end distal from the head package to a substantially vertically oriented intermediate position, and is guided, while remaining about vertically oriented, along a track in the sub-space to the second position.

Material Requirements, Selection And Development for the Proposed JIMO SpacePower System

NASA Astrophysics Data System (ADS)

Ring, P. J.; Sayre, E. D.

2004-02-01

NASA is proposing a major new nuclear Space initiative-The Jupiter Icy Moons Orbiter (JIMO). A mission such as this inevitably requires a significant power source both for propulsion and for on-board power. Three reactor concepts, liquid metal cooled, heat pipe cooled and gas cooled are being considered together with three power conversion systems Brayton (cycle), Thermoelectric and Stirling cycles, and possibly Photo voltaics for future systems. Regardless of the reactor system selected it is almost certain that high temperature (materials), refractory alloys, will be required. This paper revisits the material selection options, reviewing the rationale behind the SP-100 selection of Nb-1Zr as the major cladding and structural material and considers the alternatives and developments needed for the longer duty cycle of the JIMO power supply. A side glance is also taken at the basis behind the selection of Uranium nitride fuel over UO2 or UC and a brief discussion of the reason for the selection of Lithium as the liquid metal coolant for SP-100 over other liquid metals.

Power Generation from Nuclear Reactors in Aerospace Applications

NASA Technical Reports Server (NTRS)

English, Robert E.

1982-01-01

Power generation in nuclear powerplants in space is addressed. In particular, the states of technology of the principal competitive concepts for power generation are assessed. The possible impact of power conditioning on power generation is also discussed. For aircraft nuclear propulsion, the suitability of various technologies is cursorily assessed for flight in the Earth's atmosphere; a program path is suggested to ease the conditions of first use of aircraft nuclear propulsion.

Assessment of Space Power Related Measurement Requirements of the Strategic Defense Initiative

DTIC Science & Technology

1989-04-01

calibration techniques are available and estimated uncertainties vary between 5 and 10%. At low rf power levels (~ 10mW ), NIST maintains standard calibration... bands single or dual six-port automatic network analyzers [24] are used as transfer systems with detectors calibrated using the NIST micro calorimeter...Probable designs for the multimegawatt space reactor program indicate the need to measure neutron fluxes up to 1016 neutrons/cm2- s (1019 neutrons

Analysis of loss-of-coolant accident for a fast-spectrum lithium-cooled nuclear reactor for space-power applications

NASA Technical Reports Server (NTRS)

Turney, G. E.; Petrik, E. J.; Kieffer, A. W.

1972-01-01

A two-dimensional, transient, heat-transfer analysis was made to determine the temperature response in the core of a conceptual space-power nuclear reactor following a total loss of reactor coolant. With loss of coolant from the reactor, the controlling mode of heat transfer is thermal radiation. In one of the schemes considered for removing decay heat from the core, it was assumed that the 4 pi shield which surrounds the core acts as a constant-temperature sink (temperature, 700 K) for absorption of thermal radiation from the core. Results based on this scheme of heat removal show that melting of fuel in the core is possible only when the emissivity of the heat-radiating surfaces in the core is less than about 0.40. In another scheme for removing the afterheat, the core centerline fuel pin was replaced by a redundant, constant temperature, coolant channel. Based on an emissivity of 0.20 for all material surfaces in the core, the calculated maximum fuel temperature for this scheme of heat removal was 2840 K, or about 90 K less than the melting temperature of the UN fuel.

Compatibility of Space Nuclear Power Plant Materials in an Inert He/Xe Working Gas Containing Reactive Impurities

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

MM Hall

2006-01-31

A major materials selection and qualification issue identified in the Space Materials Plan is the potential for creating materials compatibility problems by combining dissimilar reactor core, Brayton Unit and other power conversion plant materials in a recirculating, inert He/Xe gas loop containing reactive impurity gases. Reported here are results of equilibrium thermochemical analyses that address the compatibility of space nuclear power plant (SNPP) materials in high temperature impure He gas environments. These studies provide early information regarding the constraints that exist for SNPP materials selection and provide guidance for establishing test objectives and environments for SNPP materials qualification testing.

Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

NASA Technical Reports Server (NTRS)

Garber, Anne E.; Dickens, Ricky E.

2011-01-01

The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

NASA Astrophysics Data System (ADS)

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

2017-08-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Finally, manual reactor power oscillations were observed in real-time.

Experiences in utilization of research reactors in Yugoslavia

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

Copic, M.; Gabrovsek, Z.; Pop-Jordanov, J.

1971-06-15

The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied bymore » means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro envisage the role of research reactors in the promotion of nuclear power programs in relation to the size of the program, the competence of domestic industries and the degree of independence where fuel supply is concerned. (author)« less

Space nuclear power applied to electric propulsion

NASA Technical Reports Server (NTRS)

Vicente, F. A.; Karras, T.; Darooka, D.; Isenberg, L.

1989-01-01

Space reactor power systems with characteristics ideal for advanced spacecraft systems applications are discussed. These characteristics are: high power-to-weight ratio (15 to 33 W/kg); high volume density (high ballistic coefficient); no preferential orientation in orbit; long operational life; high reliability; and total launch and operational safety. These characteristics allow the use of electric propulsion to raise spacecraft from low earth parking orbits to operational orbits, greatly increasing the useful orbit payload for a given launch vehicle by eliminating the need for a separation injection stage. A proposed demonstration mission is described.

Report of NASA Lunar Energy Enterprise Case Study Task Force

NASA Technical Reports Server (NTRS)

Kearney, John J.

1989-01-01

The Lunar Energy Enterprise Case Study Task Force was asked to determine the economic viability and commercial potential of mining and extracting He-3 from the lunar soil, and transporting the material to Earth for use in a power-generating fusion reactor. Two other space energy projects, the Space Power Station (SPS) and the Lunar Power Station (LPS), were also reviewed because of several interrelated aspects of these projects. The specific findings of the Task Force are presented. Appendices contain related papers generated by individual Task Force Members.

Parameter study of dual-mode space nuclear fission solid core power and propulsion systems, NUROC3A. AMS report No. 1239c

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

Smith, W.W.; Layton, J.P.

1976-09-13

The three-volume report describes a dual-mode nuclear space power and propulsion system concept that employs an advanced solid-core nuclear fission reactor coupled via heat pipes to one of several electric power conversion systems. The NUROC3A systems analysis code was designed to provide the user with performance characteristics of the dual-mode system. Volume 3 describes utilization of the NUROC3A code to produce a detailed parameter study of the system.

Technology for Bayton-cycle powerplants using solar and nuclear energy

NASA Technical Reports Server (NTRS)

English, R. E.

1986-01-01

Brayton cycle gas turbines have the potential to use either solar heat or nuclear reactors for generating from tens of kilowatts to tens of megawatts of power in space, all this from a single technology for the power generating system. Their development for solar energy dynamic power generation for the space station could be the first step in an evolution of such powerplants for a very wide range of applications. At the low power level of only 10 kWe, a power generating system has already demonstrated overall efficiency of 0.29 and operated 38 000 hr. Tests of improved components show that these components would raise that efficiency to 0.32, a value twice that demonstrated by any alternate concept. Because of this high efficiency, solar Brayton cycle power generators offer the potential to increase power per unit of solar collector area to levels exceeding four times that from photovoltaic powerplants using present technology for silicon solar cells. The technologies for solar mirrors and heat receivers are reviewed and assessed. This Brayton technology for solar powerplants is equally suitable for use with the nuclear reactors. The available long time creep data on the tantalum alloy ASTAR-811C show that such Brayton cycles can evolve to cycle peak temperatures of 1500 K (2240 F). And this same technology can be extended to generate 10 to 100 MW in space by exploiting existing technology for terrestrial gas turbines in the fields of both aircraft propulsion and stationary power generation.

Reactivity change in a fast-spectrum space power reactor due to a 328-meter-per-second (1075-ft/sec) impact

NASA Technical Reports Server (NTRS)

Peoples, J. A., Jr.; Puthoff, R. L.

1973-01-01

Application of nuclear reactors in space will present operational problems. One such problem is the possibility of an earth impact at velocities in excess of 305 m/sec (1000 ft/sec). This report shows the results of an impact against concrete at 328 m/sec (1075 ft/sec) and examines the deformed core to estimate the range of activity inserted as a result of the impact. The results of this examination are that the deformation of the reactor core within the containment vessel left only an estimated 2.7 percent void in the core and that the reactivity inserted due to this impact deformation could be from 4.0 to 10.25 dollars.

Progress update of NASA's free-piston Stirling space power converter technology project

NASA Technical Reports Server (NTRS)

Dudenhoefer, James E.; Winter, Jerry M.; Alger, Donald

1992-01-01

A progress update is presented of the NASA LeRC Free-Piston Stirling Space Power Converter Technology Project. This work is being conducted under NASA's Civil Space Technology Initiative (CSTI). The goal of the CSTI High Capacity Power Element is to develop the technology base needed to meet the long duration, high capacity power requirements for future NASA space initiatives. Efforts are focused upon increasing system power output and system thermal and electric energy conversion efficiency at least five fold over current SP-100 technology, and on achieving systems that are compatible with space nuclear reactors. This paper will discuss progress toward 1050 K Stirling Space Power Converters. Fabrication is nearly completed for the 1050 K Component Test Power Converter (CTPC); results of motoring tests of the cold end (525 K), are presented. The success of these and future designs is dependent upon supporting research and technology efforts including heat pipes, bearings, superalloy joining technologies, high efficiency alternators, life and reliability testing, and predictive methodologies. This paper will compare progress in significant areas of component development from the start of the program with the Space Power Development Engine (SPDE) to the present work on CTPC.

Scale Effects on Magnet Systems of Heliotron-Type Reactors

NASA Astrophysics Data System (ADS)

S, Imagawa; A, Sagara

2005-02-01

For power plants heliotron-type reactors have attractive advantages, such as no current-disruptions, no current-drive, and wide space between helical coils for the maintenance of in-vessel components. However, one disadvantage is that a major radius has to be large enough to obtain large Q-value or to produce sufficient space for blankets. Although the larger radius is considered to increase the construction cost, the influence has not been understood clearly, yet. Scale effects on superconducting magnet systems have been estimated under the conditions of a constant energy confinement time and similar geometrical parameters. Since the necessary magnetic field with a larger radius becomes lower, the increase rate of the weight of the coil support to the major radius is less than the square root. The necessary major radius will be determined mainly by the blanket space. The appropriate major radius will be around 13 m for a reactor similar to the Large Helical Device (LHD).

Overview of NASA Lewis Research Center free-piston Stirling engine technology activities applicable to space power systems

NASA Technical Reports Server (NTRS)

Slaby, Jack G.

1987-01-01

A brief overview is presented of the development and technological activities of the free-piston Stirling engine. The engine started as a small scale fractional horsepower engine which demonstrated basic engine operating principles and the advantages of being hermetically sealed, highly efficient, and simple. It eventually developed into the free piston Stirling engine driven heat pump, and then into the SP-100 Space Reactor Power Program from which came the Space Power Demonstrator Engine (SPDE). The SPDE successfully operated for over 300 hr and delivered 20 kW of PV power to an alternator plunger. The SPDE demonstrated that a dynamic power conversion system can, with proper design, be balanced; and the engine performed well with externally pumped hydrostatic gas bearings.

Nuclear design of a vapor core reactor for space nuclear propulsion

NASA Astrophysics Data System (ADS)

Dugan, Edward T.; Watanabe, Yoichi; Kuras, Stephen A.; Maya, Isaac; Diaz, Nils J.

1993-01-01

Neutronic analysis methodology and results are presented for the nuclear design of a vapor core reactor for space nuclear propulsion. The Nuclear Vapor Thermal Reactor (NVTR) Rocket Engine uses modified NERVA geometry and systems which the solid fuel replaced by uranium tetrafluoride vapor. The NVTR is an intermediate term gas core thermal rocket engine with specific impulse in the range of 1000-1200 seconds; a thrust of 75,000 lbs for a hydrogen flow rate of 30 kg/s; average core exit temperatures of 3100 K to 3400 K; and reactor thermal powers of 1400 to 1800 MW. Initial calculations were performed on epithermal NVTRs using ZrC fuel elements. Studies are now directed at thermal NVTRs that use fuel elements made of C-C composite. The large ZrC-moderated reactors resulted in thrust-to-weight ratios of only 1 to 2; the compact C-C composite systems yield thrust-to-weight ratios of 3 to 5.

Nuclear power supplies: Their potential and the practical problems to their achievement for space missions

NASA Technical Reports Server (NTRS)

Colston, B. W.

1986-01-01

Various issues associated with getting technology development of nuclear power systems moving at a pace which will support the anticipated need for such systems in later years is discussed. The projected power needs of such advanced space elements as growth space stations and lunar and planetary vehicles and bases are addressed briefly, and the relevance of nuclear power systems is discussed. A brief history and status of the U.S. nuclear reactor systems is provided, and some of the problems (real and/or perceived) are dealt with briefly. Key areas on which development attention should be focused in the near future are identified, and a suggested approach is recommended to help accelerate the process.

Early Program Development

NASA Image and Video Library

2004-04-15

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

SNPSAM - Space Nuclear Power System Analysis Model

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.; Seo, Jong T.

The current version of SNPSAM is described, and the results of the integrated thermoeletric SP-100 system performance studies using SNPSAM are reported. The electric power output, conversion efficiency, coolant temperatures, and specific pumping power of the system are calculated as functions of the reactor thermal power and the liquid metal coolant type (Li or NaK-78) during steady state operation. The transient behavior of the system is also discussed.

Investigation and Feasibility Assessment of TOPAZ-2 Derivations for Space Power Applications

NASA Technical Reports Server (NTRS)

Parlos, Alexander G.; Peddicord, Kenneth L.

1998-01-01

The ability to provide continuous power at significant levels is of utmost importance for many space missions, from simple satellite operations to manned Mars missions. One of the main problems faced in delivering solar or chemical space power in the tens of kW range, is the increasingly massive nature of the power source and the costs associated with its launch, operation and maintenance. A national program had been initiated to study the feasibility of using certain advanced technologies in developing an efficient lightweight space power source. The starting point for these studies has been the Russian TOPAZ-2 space reactor system, with the ultimate goal to aid in the development of a TOPAZ-2 derivative which will be ready for flight by the year 2000. The main objective of this project has been to perform feasibility assessment and trade studies which would allow the development of an advanced space nuclear power system based on the in-core thermionic fuel element technology currently used in the Russian TOPAZ-2 reactor. Two of the important considerations in developing the concept are: (1) compliance of the current TOPAZ-2 and of any advanced designs with U.S. nuclear safety expectations, and (2) compliance of the design with the seven years lifetime requirement. The project was composed of two major phases. The initial phase of the project has concentrated on understanding the TOPAZ-2 thermionic reactor in sufficient detail to allow several follow-on tasks. The primary interest during this first phase has been given on identifying the potential of the TOPAZ-2 design for further improvements. The second phase of the project has focused on the feasibility of a TOPAZ-2 system capable of delivering 30-50 kWe. Towards the elimination of single-point failures in the load voltage regulation system an active voltage regulator has been designed to be used in conjunction with the available shunt load voltage regulator. The possible use of a dual-loop, model-based adaptive control system for load-following in the TOPAZ-2 has also been investigated. The objective of this fault-tolerant, autonomous control system is to deliver the demanded electric power at the desired voltage level, by appropriately manipulating the neutron power through the control drums. As a result, sufficient thermal power is produced to meet the required demand in the presence of dynamically changing system operating conditions and potential sensor failures. The designed controller is proposed for use in combination with the currently available shunt regulators, or as a back-up controller when other means of power system control, including some of the sensors, fail.

Static Converter for High Energy Utilization, Modular, Small Nuclear Power Plants

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

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

2002-07-01

This paper presents and analyzes the performance of high efficiency, high total energy utilization, static converters, which could be used in conjunction with small nuclear reactor plants in remote locations and in undersea applications, requiring little or no maintenance. The converters consist of a top cycle of Alkali Metal Thermal-to-Electric Conversion (AMTEC) units and PbTe thermoelectric (TE) bottom cycle. In addition to converting the reactor thermal power to electricity at 1150 K or less, at a thermodynamic efficiency in the low to mid thirties, the heat rejection from the TE bottom cycle could be used for space heating, industrial processing,more » or sea water desalination. The results indicated that for space heating applications, where the rejected thermal power from the TE bottom cycle is removed by natural convection of ambient air, a total utilization of the reactor thermal power of > 80% is possible. When operated at 1030 K, potassium AMTEC/TE converters are not only more efficient than the sodium AMTEC/TE converters but produce more electrical power. The present analysis showed that a single converter could be sized to produce up to 100 kWe and 70 kWe, for the Na-AMTEC/TE units when operating at 1150 K and the K-AMTEC/TE units when operating at 1030 K, respectively. Such modularity is an added advantage to the high-energy utilization of the present AMTEC/TE converters. (authors)« less

Shock and vibration tests of uranium mononitride fuel pellets for a space power nuclear reactor

NASA Technical Reports Server (NTRS)

Adams, D. W.

1972-01-01

Shock and vibration tests were conducted on cylindrically shaped, depleted, uranium mononitride (UN) fuel pellets. The structural capabilities of the pellets were determined under exposure to shock and vibration loading which a nuclear reactor may encounter during launching into space. Various combinations of diametral and axial clearances between the pellets and their enclosing structures were tested. The results of these tests indicate that for present fabrication of UN pellets, a diametral clearance of 0.254 millimeter and an axial clearance of 0.025 millimeter are tolerable when subjected to launch-induced loads.

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

NASA Technical Reports Server (NTRS)

1976-01-01

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

Liquid-Metal Pump Technologies for Nuclear Surface Power

NASA Technical Reports Server (NTRS)

Polzin, K. A.

2007-01-01

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

Design of a Mechanical NaK Pump for Fission Space Power Systems

NASA Technical Reports Server (NTRS)

Mireles, Omar R.; Bradley, David; Godfroy, Thomas

2010-01-01

Alkali liquid metal cooled fission reactor concepts are under development for mid-range spaceflight power requirements. One such concept utilizes a sodium-potassium eutectic (NaK) as the primary loop working fluid. Traditionally, linear induction pumps have been used to provide the required flow and head conditions for liquid metal systems but can be limited in performance. This paper details the design, build, and check-out test of a mechanical NaK pump. The pump was designed to meet reactor cooling requirements using commercially available components modified for high temperature NaK service.

Design of a Mechanical NaK Pump for Fission Space Power

NASA Technical Reports Server (NTRS)

Mireles, Omar R.; Bradley, David E.; Godfroy, Thomas

2011-01-01

Alkali liquid metal cooled fission reactor concepts are under development for spaceflight power requirements. One such concept utilizes a sodium-potassium eutectic (NaK) as the primary loop working fluid, which has specific pumping requirements. Traditionally, electromagnetic linear induction pumps have been used to provide the required flow and pressure head conditions for NaK systems but they can be limited in performance, efficiency, and number of available vendors. The objective of the project was to develop a mechanical NaK centrifugal pump that takes advantages of technology advances not available in previous liquid metal mechanical pump designs. This paper details the design, build, and performance test of a mechanical NaK pump developed at NASA Marshall Space Flight Center. The pump was designed to meet reactor cooling requirements using commercially available components modified for high temperature NaK service.

Disposal of radioactive iodine in space

NASA Technical Reports Server (NTRS)

Burns, R. E.; Defield, J. G.

1978-01-01

The possibility of space disposal of iodine waste from nuclear power reactors is investigated. The space transportation system utilized relies upon the space shuttle, a liquid hydrogen/liquid oxygen orbit transfer vehicle, and a solid propellant final stage. The iodine is assumed to be in the form of either an iodide or an iodate, and calculations assume that the final destination is either solar orbit or solar system escape. It is concluded that space disposal of iodine is feasible.

Pratt & Whitney ESCORT derivative for mars surface power

NASA Astrophysics Data System (ADS)

Feller, Gerald J.; Joyner, Russell

1999-01-01

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

Orbital transfer of large space structures with nuclear electric rockets

NASA Technical Reports Server (NTRS)

Silva, T. H.; Byers, D. C.

1980-01-01

This paper discusses the potential application of electric propulsion for orbit transfer of a large spacecraft structure from low earth orbit to geosynchronous altitude in a deployed configuration. The electric power was provided by the spacecraft nuclear reactor space power system on a shared basis during transfer operations. Factors considered with respect to system effectiveness included nuclear power source sizing, electric propulsion thruster concept, spacecraft deployment constraints, and orbital operations and safety. It is shown that the favorable total impulse capability inherent in electric propulsion provides a potential economic advantage over chemical propulsion orbit transfer vehicles by reducing the number of Space Shuttle flights in ground-to-orbit transportation requirements.

Central electrical utility power for a satellite ring city in low earth orbit space

NASA Technical Reports Server (NTRS)

Myers, Ira T.; Faymon, Karl A.; Patton, A. D.

1989-01-01

Information is given in viewgraph form on central electrical power for a satellite ring city, defined as a group of large free flyers of 10 to 20 units with perhaps 100 people in each unit, and organized in a circle so that power can be fed from a central location. The free flyers would be located at 300 to 700 miles in altitude, and spaced about a kilometer apart. Potential activities of a ring city are listed as well as the electrical power needs. Information is given on costs and individual and centralized solar arrays and nuclear reactor systems.

Poisoning of Heat Pipes

NASA Technical Reports Server (NTRS)

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

2007-01-01

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

Multiple Restart Testing of a Stainless Steel Sodium Heat Pipe Module

NASA Astrophysics Data System (ADS)

Martin, James; Mireles, Omar; Reid, Robert

2005-02-01

A heat pipe cooled reactor is one of several candidate reactor concepts being considered for space power and propulsion systems to support future space exploration activities. Long life heat pipe modules, with concepts verified through a combination of theoretical analysis and experimental evaluations, would be necessary to establish the viability of this option. A number of stainless steel/sodium heat pipe modules have been designed and fabricated to support experimental testing of a Safe Affordable Fission Engine (SAFE) project, a 100-kWt core design pursued jointly by the Marshall Space Flight Center and the Los Alamos National Laboratory. One of the SAFE heat pipe modules was successfully subjected to over 200 restarts, examining the behavior of multiple passive freeze/thaw operations. Typical operation included a 1-hour startup to an average evaporator temperature of 1000 K followed by a 15-minute hold at temperature. Nominal maximum input power to the evaporator (measured at the power supply) during the hold period was 1.9 kW, with approximately 1.6 kW calculated as the axial power transfer to the condenser (the 300W difference was lost to environment at the evaporator surface). Between heating cycles the module was cooled to less than 325 K, returning the sodium to a frozen state in preparation for the next startup cycle.

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

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.

Multi-MW Closed Cycle MHD Nuclear Space Power Via Nonequilibrium He/Xe Working Plasma

NASA Technical Reports Server (NTRS)

Litchford, Ron J.; Harada, Nobuhiro

2011-01-01

Prospects for a low specific mass multi-megawatt nuclear space power plant were examined assuming closed cycle coupling of a high-temperature fission reactor with magnetohydrodynamic (MHD) energy conversion and utilization of a nonequilibrium helium/xenon frozen inert plasma (FIP). Critical evaluation of performance attributes and specific mass characteristics was based on a comprehensive systems analysis assuming a reactor operating temperature of 1800 K for a range of subsystem mass properties. Total plant efficiency was expected to be 55.2% including plasma pre-ionization power, and the effects of compressor stage number, regenerator efficiency and radiation cooler temperature on plant efficiency were assessed. Optimal specific mass characteristics were found to be dependent on overall power plant scale with 3 kg/kWe being potentially achievable at a net electrical power output of 1-MWe. This figure drops to less than 2 kg/kWe when power output exceeds 3 MWe. Key technical issues include identification of effective methods for non-equilibrium pre-ionization and achievement of frozen inert plasma conditions within the MHD generator channel. A three-phase research and development strategy is proposed encompassing Phase-I Proof of Principle Experiments, a Phase-II Subscale Power Generation Experiment, and a Phase-III Closed-Loop Prototypical Laboratory Demonstration Test.

Sealed head access area enclosure

DOEpatents

Golden, Martin P.; Govi, Aldo R.

1978-01-01

A liquid-metal-cooled fast breeder power reactor is provided with a sealed head access area enclosure disposed above the reactor vessel head consisting of a plurality of prefabricated structural panels including a center panel removably sealed into position with inflatable seals, and outer panels sealed into position with semipermanent sealant joints. The sealant joints are located in the joint between the edge of the panels and the reactor containment structure and include from bottom to top an inverted U-shaped strip, a lower layer of a room temperature vulcanizing material, a separator strip defining a test space therewithin, and an upper layer of a room temperature vulcanizing material. The test space is tapped by a normally plugged passage extending to the top of the enclosure for testing the seal or introducing a buffer gas thereinto.

Nuclear power propulsion system for spacecraft

NASA Astrophysics Data System (ADS)

Koroteev, A. S.; Oshev, Yu. A.; Popov, S. A.; Karevsky, A. V.; Solodukhin, A. Ye.; Zakharenkov, L. E.; Semenkin, A. V.

2015-12-01

The proposed designs of high-power space tugs that utilize solar or nuclear energy to power an electric jet engine are reviewed. The conceptual design of a nuclear power propulsion system (NPPS) is described; its structural diagram, gas circuit, and electric diagram are discussed. The NPPS incorporates a nuclear reactor, a thermal-to-electric energy conversion system, a system for the conversion and distribution of electric energy, and an electric propulsion system. Two criterion parameters were chosen in the considered NPPS design: the temperature of gaseous working medium at the nuclear reactor outlet and the rotor speed of turboalternators. The maintenance of these parameters at a given level guarantees that the needed electric voltage is generated and allows for power mode control. The processes of startup/shutdown and increasing/reducing the power, the principles of distribution of electric energy over loads, and the probable emergencies for the proposed NPPS design are discussed.

Evolution of systems concepts for a 100 kWe class Space Nuclear Power System

NASA Technical Reports Server (NTRS)

Katucki, R.; Josloff, A.; Kirpich, A.; Florio, F.

1985-01-01

Conceptual designs for the SP-100 Space Nuclear Power System have been prepared that meet baseline, backup and growth program scenarios. Near-term advancement in technology was considered in the design of the Baseline Concept. An improved silicon-germanium thermoelectric technique is used to convert the heat from a fast-spectrum, liquid lithium cooled reactor. This system produces a net power of 100 kWe with a 10-year end of life, under the specific constraints of area and volume. Output of the Backup Concept is estimated to be 60 kWe for a 10-year end of life. This system differs from the Baseline Concept because currently available thermoelectric conversion is used from energy supplied by a liquid sodium cooled reactor. The Growth Concept uses Stirling engine conversion to produce 100 kWe within the constraints of mass and volume. The Growth Concept can be scaled up to produce a 1 MWe output that uses the same type reactor developed for the Baseline Concept. Assessments made for each of the program scenarios indicate the key development efforts needed to initiate detailed design and hardware program phases. Development plans were prepared for each scenario that detail the work elements and show the program activities leading to a state of flight readiness.

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

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

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

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

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

1993-01-01

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

A search for neutrino oscillations using the CHOOZ 1 km baseline reactor neutrino experiment

NASA Astrophysics Data System (ADS)

George, Jean

1999-10-01

Neutrino oscillation searches are an active field of research due to the implications their discovery may have for the solar neutrino anomaly as well as for the atmospheric neutrino anomaly. Their discovery may also have broad ramifications for the Standard Model of Particle Physics as a whole. Results from an oscillation search using the CHOOZ long baseline reactor neutrino experiment are presented in this thesis. These results are based on the data taken from June 1997 through April 1998 when the two reactors ran at combined thermal power levels ranging from zero power to their full power level of 8.5 GW. Electron flavored antineutrinos emanating from the reactors were detected through the inverse beta decay channel using a liquid scintillating calorimeter located at a distance of approximately 1 km from the reactor sources. The underground experimental site (300 MWE) provided natural shielding from the background of cosmic ray muons-leading to a background rate more than an order of magnitude lower than the full power signal rate. From the agreement between the detected and expected neutrino event rates no evidence for neutrino oscillations was found (at the 90% C.L.) for the oscillation parameter space governed by Δm 2 > 0.8 × 10-3 eV2 for maximal mixing and by sin2 2Θ > 0.18 for large values of Δm2.

Multi-megawatt power system trade study

NASA Astrophysics Data System (ADS)

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

2002-01-01

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

Experimental Evaluation of a Water Shield for a Surface Power Reactor

NASA Technical Reports Server (NTRS)

Pearson, J. B.; Reid, R.; Sadasivan, P.; Stewart, E.

2007-01-01

A water based shielding system is being investigated for use on initial lunar surface power systems. The use of water may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. A representative lunar surface reactor design is evaluated at various power levels in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The evaluation compares the experimental data from the WST to CFD models. Performance of a water shield on the lunar surface is predicted by CFD models anchored to test data, and by matching relevant dimensionless parameters.

An assessment and validation study of nuclear reactors for low power space applications

NASA Technical Reports Server (NTRS)

Klein, A. C.; Gedeon, S. R.; Morey, D. C.

1987-01-01

The feasibility and safety of six conceptual small, low power nuclear reactor designs was evaluated. Feasibility evaluations included the determination of sufficient reactivity margins for seven years of full power operation and safe shutdown as well as handling during pre-launch assembly phases. Safety evaluations were concerned with the potential for maintaining subcritical conditions in the event of launch or transportation accidents. These included water immersion accident scenarios both with and without water flooding the core. Results show that most of the concepts can potentially meet the feasibility and safety requirements; however, due to the preliminary nature of the designs considered, more detailed designs will be necessary to enable these concepts to fully meet the safety requirements.

Nuclear thermionic power plants in the 50-300 kWe range.

NASA Technical Reports Server (NTRS)

Van Hoomissen, J. E.; Sawyer, C. D.; Prickett, W. Z.

1972-01-01

This paper reviews the results of recent studies performed by General Electric on in-core thermionic reactor power plants in the 50-300 kWe range. In particular, a 100 kWe manned Space Base mission and a 240 kWe unmanned electric propulsion mission are singled out as representative design points for this concept.

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.

Nuclear reactor system study for NASA/JPL

NASA Technical Reports Server (NTRS)

Palmer, R. G.; Lundberg, L. B.; Keddy, E. S.; Koenig, D. R.

1982-01-01

Reactor shielding, safety studies, and heat pipe development work are described. Monte Carlo calculations of gamma and neutron shield configurations show that substantial weight penalties are incurred if exposure at 25 m to neutrons and gammas must be limited to 10 to the 12th power nvt and 10 to the 6th power rad, instead of the 10 to the 13th power nvt and 10 to the 7th power rad values used earlier. For a 1.6 MW sub t reactor, the required shield weight increases from 400 to 815 kg. Water immersion critically calculations were extended to study the effect of water in fuel void spaces as well as in the core heat pipes. These show that the insertion into the core of eight blades of B4C with a mass totaling 2.5 kg will guarantee subcriticality. The design, fabrication procedure, and testing of a 4m long molybdenum/lithium heat pipe are described. It appears that an excess of oxygen in the wick prevented the attainment of expected performance capability.

A potassium Rankine multimegawatt nuclear electric propulsion concept

NASA Technical Reports Server (NTRS)

Baumeister, E.; Rovang, R.; Mills, J.; Sercel, J.; Frisbee, R.

1990-01-01

Multimegawatt nuclear electric propulsion (NEP) has been identified as a potentially attractive option for future space exploratory missions. A liquid-metal-cooled reactor, potassium Rankine power system that is being developed is suited to fulfill this application. The key features of the nuclear power system are described, and system characteristics are provided for various potential NEP power ranges and operational lifetimes. The results of recent mission studies are presented to illustrate some of the potential benefits to future space exploration to be gained from high-power NEP. Specifically, mission analyses have been performed to assess the mass and trip time performance of advanced NEP for both cargo and piloted missions to Mars.

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.

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.

Application of the monolithic solid oxide fuel cell to space power systems

NASA Astrophysics Data System (ADS)

Myles, Kevin M.; Bhattacharyya, Samit K.

1991-01-01

The monolithic solid-oxide fuel cell (MSOFC) is a promising electrochemical power generation device that is currently under development at Argonne National Laboratory. The extremely high power density of the MSOFC leads to MSOFC systems that have sufficiently high energy densities that they are excellent candidates for a number of space missions. The fuel cell can also be operated in reverse, if it can be coupled to an external power source, to regenerate the fuel and oxidant from the water product. This feature further enhances the potential mission applications of the MSOFC. In this paper, the current status of the fuel cell development is presented—the focus being on fabrication and currently achievable performance. In addition, a specific example of a space power system, featuring a liquid metal cooled fast spectrum nuclear reactor and a monolithic solid oxide fuel cell, is presented to demonstrate the features of an integrated system.

Safe Affordable Fission Engine-(SAFE-) 100a Heat Exchanger Thermal and Structural Analysis

NASA Technical Reports Server (NTRS)

Steeve, B. E.

2005-01-01

A potential fission power system for in-space missions is a heat pipe-cooled reactor coupled to a Brayton cycle. In this system, a heat exchanger (HX) transfers the heat of the reactor core to the Brayton gas. The Safe Affordable Fission Engine- (SAFE-) 100a is a test program designed to thermally and hydraulically simulate a 95 Btu/s prototypic heat pipe-cooled reactor using electrical resistance heaters on the ground. This Technical Memorandum documents the thermal and structural assessment of the HX used in the SAFE-100a program.

Space reactor fuel element testing in upgraded TREAT

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

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

1993-01-14

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

Space reactor fuel element testing in upgraded TREAT

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

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

1993-05-01

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

Gas Foil Bearing Misalignment and Unbalance Effects

NASA Technical Reports Server (NTRS)

Howard, Samuel A.

2008-01-01

The effects of misalignment and unbalance on gas foil bearings are presented. The future of U.S. space exploration includes plans to conduct science missions aboard space vehicles, return humans to the Moon, and place humans on Mars. All of these endeavors are of long duration, and require high amounts of electrical power for propulsion, life support, mission operations, etc. One potential source of electrical power of sufficient magnitude and duration is a nuclear-fission-based system. The system architecture would consist of a nuclear reactor heat source with the resulting thermal energy converted to electrical energy through a dynamic power conversion and heat rejection system. Various types of power conversion systems can be utilized, but the Closed Brayton Cycle (CBC) turboalternator is one of the leading candidates. In the CBC, an inert gas heated by the reactor drives a turboalternator, rejects excess heat to space through a heat exchanger, and returns to the reactor in a closed loop configuration. The use of the CBC for space power and propulsion is described in more detail in the literature (Mason, 2003). In the CBC system just described, the process fluid is a high pressure inert gas such as argon, krypton, or a helium-xenon mixture. Due to the closed loop nature of the system and the associated potential for damage to components in the system, contamination of the working fluid is intolerable. Since a potential source of contamination is the lubricant used in conventional turbomachinery bearings, Gas Foil Bearings (GFB) have high potential for the rotor support system. GFBs are compliant, hydrodynamic journal and thrust bearings that use a gas, such as the CBC working fluid, as their lubricant. Thus, GFBs eliminate the possibility of contamination due to lubricant leaks into the closed loop system. Gas foil bearings are currently used in many commercial applications, both terrestrial and aerospace. Aircraft Air Cycle Machines (ACMs) and ground-based microturbines have demonstrated histories of successful long-term operation using GFBs (Heshmat et al., 2000). Small aircraft propulsion engines, helicopter gas turbines, and high-speed electric motors are potential future applications.

Transient Approximation of SAFE-100 Heat Pipe Operation

NASA Technical Reports Server (NTRS)

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

2005-01-01

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

[Anthropogenic sources of radiation hazard in the near-Earth space].

PubMed

Fedoseev, G A

2004-01-01

All plausible artificial radioactive sources entering the near-Earth space (NES) were systematized and consequences of various large radiation accidents and catastrophes to Earth and NES were analyzed. Aggressive "population" of near-Earth orbits by space stations with rotating crews, unmanned research platforms and observatories extends "borderlines" of the noosphere raising at the same time concerns about the noosphere radiation safety and global radioecology. Specifically, consideration is given to the facts of negative effects of space power reactor facilities on results of orbital astrophysical investigations.

Survey of Current and Next Generation Space Power Technologies

DTIC Science & Technology

2006-06-26

different thermodynamic cycles, such as the Brayton, Rankine, and Stirling cycles, alkali metal thermal electric converters ( AMTEC ) and thermionic...efficiencies @ 1700K. The primary issue with this system is the integration of the converter technology into the nuclear reactor core. AMTEC (static...Alkali metal thermal to electric converters ( AMTECs ) are thermally powered electrochemical concentration cells that convert heat energy directly to DC

1986 Nuclear Science Symposium, 33rd, and 1986 Symposium on Nuclear Power Systems, 18th, Washington, DC, Oct. 29-31, 1986, Proceedings

NASA Technical Reports Server (NTRS)

Stubblefield, F. W. (Editor)

1987-01-01

Papers are presented on space, low-energy physics, and general nuclear science instrumentations. Topics discussed include data acquisition systems and circuits, nuclear medicine imaging and tomography, and nuclear radiation detectors. Consideration is given to high-energy physics instrumentation, reactor systems and safeguards, health physics instrumentation, and nuclear power systems.

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

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

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

2007-01-16

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

Lunar He-3, fusion propulsion, and space development

NASA Technical Reports Server (NTRS)

Santarius, John F.

1992-01-01

The recent identification of a substantial lunar resource of the fusion energy fuel He-3 may provide the first terrestrial market for a lunar commodity and, therefore, a major impetus to lunar development. The impact of this resource-when burned in D-He-3 fusion reactors for space power and propulsion-may be even more significant as an enabling technology for safe, efficient exploration and development of space. One possible reactor configuration among several options, the tandem mirror, illustrates the potential advantages of fusion propulsion. The most important advantage is the ability to provide either fast, piloted vessels or high-payload-fraction cargo vessels due to a range of specific impulses from 50 sec to 1,000,000 sec at thrust-to-weight ratios from 0.1 to 5x10(exp -5). Fusion power research has made steady, impressive progress. It is plausible, and even probable, that fusion rockets similar to the designs presented here will be available in the early part of the twenty-first century, enabling a major expansion of human presence into the solar system.

Development of High Fidelity, Fuel-Like Thermal Simulators for Non-Nuclear Testing

NASA Technical Reports Server (NTRS)

Bragg-Sitton, S. M.; Farmer, J.; Dixon, D.; Kapernick, R.; Dickens, R.; Adams, M.

2007-01-01

Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Work at the NASA Marshall Space Flight Center seeks to develop high fidelity thermal simulators that not only match the static power profile that would be observed in an operating, fueled nuclear reactor, but to also match the dynamic fuel pin performance during feasible transients. Comparison between the fuel pins and thermal simulators is made at the fuel clad surface, which corresponds to the sheath surface in the thermal simulator. Static and dynamic fuel pin performance was determined using SINDA-FLUINT analysis, and the performance of conceptual thermal simulator designs was compared to the expected nuclear performance. Through a series of iterative analysis, a conceptual high fidelity design will be developed, followed by engineering design, fabrication, and testing to validate the overall design process. Although the resulting thermal simulator will be designed for a specific reactor concept, establishing this rigorous design process will assist in streamlining the thermal simulator development for other reactor concepts.

Radiation and temperature effects on electronic components investigated under the CSTI high capacity power project

NASA Technical Reports Server (NTRS)

Schwarze, Gene E.; Niedra, Janis M.; Frasca, Albert J.; Wieserman, William R.

1993-01-01

The effects of nuclear radiation and high temperature environments must be fully known and understood for the electronic components and materials used in both the Power Conditioning and Control subsystem and the reactor Instrumentation and Control subsystem of future high capacity nuclear space power systems. This knowledge is required by the designer of these subsystems in order to develop highly reliable, long-life power systems for future NASA missions. A review and summary of the experimental results obtained for the electronic components and materials investigated under the power management element of the Civilian Space Technology Initiative (CSTI) high capacity power project are presented: (1) neutron, gamma ray, and temperature effects on power semiconductor switches, (2) temperature and frequency effects on soft magnetic materials; and (3) temperature effects on rare earth permanent magnets.

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 nuclear reactors and radiation protection, thermal physics, physical chemistry and technology of liquid metal coolants, and physics of radiation-induced defects, and radiation materials science. The activity of the institute is aimed at solving matters concerned with technological development of large-scale nuclear power engineering on the basis of a closed nuclear fuel cycle with the use of fast-neutron reactors (referred to henceforth as fast reactors), development of innovative nuclear and conventional technologies, and extension of their application fields.

Lightweight Damage Tolerant Radiators for In-Space Nuclear Electric Power and Propulsion

NASA Technical Reports Server (NTRS)

Craven, Paul; SanSoucie, Michael P.; Tomboulian, Briana; Rogers, Jan; Hyers, Robert

2014-01-01

Nuclear electric propulsion (NEP) is a promising option for high-speed in-space travel due to the high energy density of nuclear power sources and efficient electric thrusters. Advanced power conversion technologies for converting thermal energy from the reactor to electrical energy at high operating temperatures would benefit from lightweight, high temperature radiator materials. Radiator performance dictates power output for nuclear electric propulsion systems. Pitch-based carbon fiber materials have the potential to offer significant improvements in operating temperature and mass. An effort at the NASA Marshall Space Flight Center to show that woven high thermal conductivity carbon fiber mats can be used to replace standard metal and composite radiator fins to dissipate waste heat from NEP systems is ongoing. The goals of this effort are to demonstrate a proof of concept, to show that a significant improvement of specific power (power/mass) can be achieved, and to develop a thermal model with predictive capabilities. A description of this effort is presented.

Development of crawler type device using new measuring system

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

Maruyama, T.; Sasaki, T.; Yagi, T.

1995-08-01

This paper reports the development and field application of a new device which examine shell to shell weld joints of RPV. In a BWR type nuclear power plant, there is narrow space around the Reactor Pressure Vessel (RPV) because RPV is enclosed by the Reactor Shield Wall (RSW) and thermal insulations. The developed device is characterized by a new position measuring system and magnet wheels for driving. The new position measuring system uses laser beam and ultrasonic wave. The magnet wheels make the device travel freely in the narrow space between RPV and insulation. This device is tested on mock-upsmore » and applied examination of RPVs to verify field applicability.« less

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.

Advanced Space Nuclear Reactors from Fiction to Reality

NASA Astrophysics Data System (ADS)

Popa-Simil, L.

The advanced nuclear power sources are used in a large variety of science fiction movies and novels, but their practical development is, still, in its early conceptual stages, some of the ideas being confirmed by collateral experiments. The novel reactor concept uses the direct conversion of nuclear energy into electricity, has electronic control of reactivity, being surrounded by a transmutation blanket and very thin shielding being small and light that at its very limit may be suitable to power an autonomously flying car. It also provides an improved fuel cycle producing minimal negative impact to environment. The key elements started to lose the fiction attributes, becoming viable actual concepts and goals for the developments to come, and on the possibility to achieve these objectives started to become more real because the theory shows that using the novel nano-technologies this novel reactor might be achievable in less than a century.

Investigation of Natural and Man-Made Radiation Effects on Crews on Long Duration Space Missions

NASA Technical Reports Server (NTRS)

Bolch, Wesley E.; Parlos, Alexander

1996-01-01

Over the past several years, NASA has studied a variety of mission scenarios designed to establish a permanent human presence on the surface of Mars. Nuclear electric propulsion (NEP) is one of the possible elements in this program. During the initial stages of vehicle design work, careful consideration must be given to not only the shielding requirements of natural space radiation, but to the shielding and configuration requirements of the on-board reactors. In this work, the radiation transport code MCNP has been used to make initial estimates of crew exposures to reactor radiation fields for a specific manned NEP vehicle design. In this design, three 25 MW(sub th), scaled SP-100-class reactors are shielded by three identical shields. Each shield has layers of beryllium, tungsten, and lithium hydride between the reactor and the crew compartment. Separate calculations are made of both the exiting neutron and gamma fluxes from the reactors during beginning-of-life, full-power operation. This data is then used as the source terms for particle transport in MCNP. The total gamma and neutron fluxes exiting the reactor shields are recorded and separate transport calculations are then performed for a 10 g/sq cm crew compartment aluminum thickness. Estimates of crew exposures have been assessed for various thicknesses of the shield tungsten and lithium hydride layers. A minimal tungsten thickness of 20 cm is required to shield the reactor photons below the 0.05 Sv/y man-made radiation limit. In addition to a 20-cm thick tungsten layer, a 40-cm thick lithium hydride layer is required to shield the reactor neutrons below the annual limit. If the tungsten layer is 30-cm thick, the lithium hydride layer should be at least 30-cm thick. These estimates do not take into account the photons generated by neutron interactions inside the shield because the MCNP neutron cross sections did not allow reliable estimates of photon production in these materials. These results, along with natural space radiation shielding estimates calculated by NASA Langley Research Center, have been used to provide preliminary input data into a new Macintosh-based software tool. A skeletal version of this tool being developed will allow rapid radiation exposure and risk analyses to be performed on a variety of Lunar and Mars missions utilizing nuclear-powered vehicles.

Technicians Manufacture a Nozzle for the Kiwi B-1-B Engine

NASA Image and Video Library

1964-05-21

Technicians manufacture a nozzle for the Kiwi B-1-B nuclear rocket engine in the Fabrication Shop’s vacuum oven at the National Aeronautics and Space Administration (NASA) Lewis Research Center. The Nuclear Engine for Rocket Vehicle Applications (NERVA) was a joint NASA and Atomic Energy Commission (AEC) endeavor to develop a nuclear-powered rocket for both long-range missions to Mars and as a possible upper-stage for the Apollo Program. The early portion of the program consisted of basic reactor and fuel system research. This was followed by a series of Kiwi reactors built to test basic nuclear rocket principles in a non-flying nuclear engine. The next phase, NERVA, would create an entire flyable engine. The final phase of the program, called Reactor-In-Flight-Test, would be an actual launch test. The AEC was responsible for designing the nuclear reactor and overall engine. NASA Lewis was responsible for developing the liquid-hydrogen fuel system. The turbopump, which pumped the fuels from the storage tanks to the engine, was the primary tool for restarting the engine. The NERVA had to be able to restart in space on its own using a safe preprogrammed startup system. Lewis researchers endeavored to design and test this system. This non-nuclear Kiwi engine, seen here, was being prepared for tests at Lewis’ High Energy Rocket Engine Research Facility (B-1) located at Plum Brook Station. The tests were designed to start an unfueled Kiwi B-1-B reactor and its Aerojet Mark IX turbopump without any external power.

Impact of Radiation Hardness and Operating Temperatures of Silicon Carbide Electronics on Space Power System Mass

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.; Tew, Roy C.; Schwarze, Gene E.

1998-01-01

The effect of silicon carbide (SiC) electronics operating temperatures on Power Management and Distribution (PMAD), or Power Conditioning (PC), subsystem radiator size and mass requirements was evaluated for three power output levels (100 kW(e) , 1 MW(e), and 10 MW(e)) for near term technology ( i.e. 1500 K turbine inlet temperature) Closed Cycle Gas Turbine (CCGT) power systems with a High Temperature Gas Reactor (HTGR) heat source. The study was conducted for assumed PC radiator temperatures ranging from 370 to 845 K and for three scenarios of electrical energy to heat conversion levels which needed to be rejected to space by means of the PC radiator. In addition, during part of the study the radiation hardness of the PC electronics was varied at a fixed separation distance to estimate its effect on the mass of the instrument rated reactor shadow shield. With both the PC radiator and the conical shadow shield representing major components of the overall power system the influence of the above on total power system mass was also determined. As expected, results show that the greatest actual mass savings achieved by the use of SiC electronics occur with high capacity power systems. Moreover, raising the PC radiator temperature above 600 K yields only small additional system mass savings. The effect of increased radiation hardness on total system mass is to reduce system mass by virtue of lowering the shield mass.

Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Demonstration. CEDT Phase 1 Preliminary Design Documentation

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

Sanchez, Rene Gerardo; Hutchinson, Jesson D.; Mcclure, Patrick Ray

2015-08-20

The intent of the integral experiment request IER 299 (called KiloPower by NASA) is to assemble and evaluate the operational performance of a compact reactor configuration that closely resembles the flight unit to be used by NASA to execute a deep space exploration mission. The reactor design will include heat pipes coupled to Stirling engines to demonstrate how one can generate electricity when extracting energy from a “nuclear generated” heat source. This series of experiments is a larger scale follow up to the DUFF series of experiments1,2 that were performed using the Flat-Top assembly.

HIGH TEMPERATURE, HIGH POWER HETEROGENEOUS NUCLEAR REACTOR

DOEpatents

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

1960-06-14

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

Pellet bed reactor for nuclear propelled vehicles: Part 2: Missions and vehicle integration trades

NASA Technical Reports Server (NTRS)

Haloulakos, V. E.

1991-01-01

Mission and vehicle integration tradeoffs involving the use of the pellet bed reactor (PBR) for nuclear powered vehicles is discussed, with much of the information being given in viewgraph form. Information is given on propellant tank geometries, shield weight requirements for conventional tank configurations, effective specific impulse, radiation mapping, radiation dose rate after shutdown, space transfer vehicle design data, a Mars mission summary, sample pellet bed nuclear orbit transfer vehicle mass breakdown, and payload fraction vs. velocity increment.

Emergency Cooling of Nuclear Power Plant Reactors With Heat Removal By a Forced-Draft Cooling Tower

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

Murav’ev, V. P., E-mail: murval1@mail.ru

The feasibility of heat removal during emergency cooling of a reactor by a forced-draft cooling tower with accumulation of the peak heat release in a volume of precooled water is evaluated. The advantages of a cooling tower over a spray cooling pond are demonstrated: it requires less space, consumes less material, employs shorter lines in the heat removal system, and provides considerably better protection of the environment from wetting by entrained moisture.

Optimization of a heat-pipe-cooled space radiator for use with a reactor-powered Stirling engine

NASA Technical Reports Server (NTRS)

Moriarty, Michael P.; French, Edward P.

1987-01-01

The design optimization of a reactor-Stirling heat-pipe-cooled radiator is presented. The radiator is a self-deploying concept that uses individual finned heat pipe 'petals' to reject waste heat from a Stirling engine. Radiator optimization methodology is presented, and the results of a parametric analysis of the radiator design variables for a 100-kW(e) system are given. The additional steps of optiminzing the radiator resulted in a net system mass savings of 3 percent.

Heat pipe nuclear reactor for space power

NASA Technical Reports Server (NTRS)

Koening, D. R.

1976-01-01

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

Pulsed Magnetic Field Driven Gas Core Reactors for Space Power & Propulsion Applications

NASA Technical Reports Server (NTRS)

Anghaie, Samim; Smith, Blair; Knight, Travis; Butler, Carey

2003-01-01

The present results indicated that: 1. A pulsed magnetic driven fission power concept, PMD-GCR is developed for closed (NER) and semi-open (NTR) operations. 2. In power mode, power is generated at alpha less than 1 for power levels of hundreds of KW or higher 3. IN semi open NTR mode, PMD-GCR generates thrust at I(sub sp) approx. 5,000 s and jet power approx. 5KW/Kg. 4. PMD-GCR is highly subcritical and is actively driven to critically. 5. Parallel path with fusion R&D needs in many areas including magnet and plasma.

Fuel processing in integrated micro-structured heat-exchanger reactors

NASA Astrophysics Data System (ADS)

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

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

Conceptual design studies of the Electron Cyclotron launcher for DEMO reactor

NASA Astrophysics Data System (ADS)

Moro, Alessandro; Bruschi, Alex; Franke, Thomas; Garavaglia, Saul; Granucci, Gustavo; Grossetti, Giovanni; Hizanidis, Kyriakos; Tigelis, Ioannis; Tran, Minh-Quang; Tsironis, Christos

2017-10-01

A demonstration fusion power plant (DEMO) producing electricity for the grid at the level of a few hundred megawatts is included in the European Roadmap [1]. The engineering design and R&D for the electron cyclotron (EC), ion cyclotron and neutral beam systems for the DEMO reactor is being performed by Work Package Heating and Current Drive (WPHCD) in the framework of EUROfusion Consortium activities. The EC target power to the plasma is about 50 MW, in which the required power for NTM control and burn control is included. EC launcher conceptual design studies are here presented, showing how the main design drivers of the system have been taken into account (physics requirements, reactor relevant operations, issues related to its integration as in-vessel components). Different options for the antenna are studied in a parameters space including a selection of frequencies, injection angles and launch points to get the best performances for the antenna configuration, using beam tracing calculations to evaluate plasma accessibility and deposited power. This conceptual design studies comes up with the identification of possible limits, constraints and critical issues, essential in the selection process of launcher setup solution.

Nuclear Design of the HOMER-15 Mars Surface Fission Reactor

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

Poston, David I.

2002-07-01

The next generation of robotic missions to Mars will most likely require robust power sources in the range of 3 to 20 kWe. Fission systems are well suited to provide safe, reliable, and economic power within this range. The goal of this study is to design a compact, low-mass fission system that meets Mars surface power requirements, while maintaining a high level of safety and reliability at a relatively low cost. The Heat pipe Power System (HPS) is one possible approach for producing near-term, low-cost, space fission power. The goal of the HPS project is to devise an attractive spacemore » fission system that can be developed quickly and affordably. The primary ways of doing this are by using existing technology and by designing the system for inexpensive testing. If the system can be designed to allow highly prototypic testing with electrical heating, then an exhaustive test program can be carried out quickly and inexpensively, and thorough testing of the actual flight unit can be performed - which is a major benefit to reliability. Over the past 4 years, three small HPS proof-of-concept technology demonstrations have been conducted, and each has been highly successful. The Heat pipe-Operated Mars Exploration Reactor (HOMER) is a derivative of the HPS designed especially for producing power on the surface of Mars. The HOMER-15 is a 15-kWt reactor that couples with a 3-kWe Stirling engine power system. The reactor contains stainless-steel (SS)-clad uranium nitride (UN) fuel pins that are structurally and thermally bonded to SS/sodium heat pipes. Fission energy is conducted from the fuel pins to the heat pipes, which then carry the heat to the Stirling engine. This paper describes conceptual design and nuclear performance the HOMER-15 reactor. (author)« less

High Power MPD Nuclear Electric Propulsion (NEP) for Artificial Gravity HOPE Missions to Callisto

NASA Technical Reports Server (NTRS)

McGuire, Melissa L.; Borowski, Stanley K.; Mason, Lee M.; Gilland, James

2003-01-01

This documents the results of a one-year multi-center NASA study on the prospect of sending humans to Jupiter's moon, Callisto, using an all Nuclear Electric Propulsion (NEP) space transportation system architecture with magnetoplasmadynamic (MPD) thrusters. The fission reactor system utilizes high temperature uranium dioxide (UO2) in tungsten (W) metal matrix cermet fuel and electricity is generated using advanced dynamic Brayton power conversion technology. The mission timeframe assumes on-going human Moon and Mars missions and existing space infrastructure to support launch of cargo and crewed spacecraft to Jupiter in 2041 and 2045, respectively.

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

NASA Astrophysics Data System (ADS)

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

2006-01-01

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

Comparisons of selected laser beam power missions to conventionally powered missions

NASA Technical Reports Server (NTRS)

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

1993-01-01

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

Evaluation of refractory-metal-clad uranium nitride and uranium dioxide fuel pins after irradiation for times up to 10 450 hours at 990 C

NASA Technical Reports Server (NTRS)

Bowles, K. J.; Gluyas, R. E.

1975-01-01

The effects of some materials variables on the irradiation performance of fuel pins for a lithium-cooled space power reactor design concept were examined. The variables studied were UN fuel density, fuel composition, and cladding alloy. All pins were irradiated at about 990 C in a thermal neutron environment to the design fuel burnup. An 85-percent dense UN fuel gave the best overall results in meeting the operational goals. The T-111 cladding on all specimens was embrittled, possibly by hydrogen in the case of the UN fuel and by uranium and oxygen in the case of the UO2 fuel. Tests with Cb-1Zr cladding indicate potential use of this cladding material. The UO2 fueled specimens met the operational goals of less than 1 percent cladding strain, but other factors make UO2 less attractive than low-density UN for the contemplated space power reactor use.

A 20,000-Kilowatt Nuclear Turboelectric Power Supply for Manned Space Vehicles

NASA Technical Reports Server (NTRS)

English, Robert E.; Slone, Henry O.; Bernatowicz, Daniel T.; Davison, Elmer H.; Lieblein, Seymour

1959-01-01

A conceptual design of a nuclear turboelectric powerplant, producing 20,000 kilowatts of power suitable for manned space vehicles is presented. The study indicates that the radiator necessary for rejecting cycle waste heat is the dominant weight, and emphasis is placed on the selection of cycle operating conditions in order to reduce this weight. A thermodynamic cycle using sodium vapor as the working fluid and operating at a turbine-inlet temperature of 2500 R was selected. The total powerplant weight was calculated to be approximately 6 pounds per kilowatt. The radiator contributes approximately 2.1 pounds per kilowatt to the total weight and the reactor and reactor shield contribute approximately 0.24 and 1.2 pounds per kilowatt, respectively. The generator, turbine, and piping add significantly to the total weight (between 0.5 and 0.6 lb/kw), but the heat exchanger, pumps, and so on are less important. Several important research areas associated with the development of a reliable nuclear turboelectric powerplant of the type analyzed are discussed.

Safety and environmental constraints on space applications of fusion energy

NASA Technical Reports Server (NTRS)

Roth, J. Reece

1990-01-01

Some of the constraints are examined on fusion reactions, plasma confinement systems, and fusion reactors that are intended for such space related missions as manned or unmanned operations in near earth orbit, interplanetary missions, or requirements of the SDI program. Of the many constraints on space power and propulsion systems, those arising from safety and environmental considerations are emphasized since these considerations place severe constraints on some fusion systems and have not been adequately treated in previous studies.

High-temperature, high-power-density thermionic energy conversion for space

NASA Technical Reports Server (NTRS)

Morris, J. F.

1977-01-01

Theoretic converter outputs and efficiencies indicate the need to consider thermionic energy conversion (TEC) with greater power densities and higher temperatures within reasonable limits for space missions. Converter-output power density, voltage, and efficiency as functions of current density were determined for 1400-to-2000 K emitters with 725-to-1000 K collectors. The results encourage utilization of TEC with hotter-than-1650 K emitters and greater-than-6W sq cm outputs to attain better efficiencies, greater voltages, and higher waste-heat-rejection temperatures for multihundred-kilowatt space-power applications. For example, 1800 K, 30 A sq cm TEC operation for NEP compared with the 1650 K, 5 A/sq cm case should allow much lower radiation weights, substantially fewer and/or smaller emitter heat pipes, significantly reduced reactor and shield-related weights, many fewer converters and associated current-collecting bus bars, less power conditioning, and lower transmission losses. Integration of these effects should yield considerably reduced NEP specific weights.

Application of a Virtual Reactivity Feedback Control Loop in Non-Nuclear Testing of a Fast Spectrum Reactor

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Forsbacka, Matthew

2004-01-01

For a compact, fast-spectrum reactor, reactivity feedback is dominated by core deformation at elevated temperature. Given the use of accurate deformation measurement techniques, it is possible to simulate nuclear feedback in non-nuclear electrically heated reactor tests. Implementation of simulated reactivity feedback in response to measured deflection is being tested at the NASA Marshall Space Flight Center Early Flight Fission Test Facility (EFF-TF). During tests of the SAFE-100 reactor prototype, core deflection was monitored using a high resolution camera. "virtual" reactivity feedback was accomplished by applying the results of Monte Carlo calculations (MCNPX) to core deflection measurements; the computational analysis was used to establish the reactivity worth of van'ous core deformations. The power delivered to the SAFE-100 prototype was then dusted accordingly via kinetics calculations, The work presented in this paper will demonstrate virtual reactivity feedback as core power was increased from 1 kilowatt(sub t), to 10 kilowatts(sub t), held approximately constant at 10 kilowatts (sub t), and then allowed to decrease based on the negative thermal reactivity coefficient.

Interior of the Plum Brook Reactor Facility

NASA Image and Video Library

1961-02-21

A view inside the 55-foot high containment vessel of the National Aeronautics and Space Administration (NASA) Plum Brook Reactor Facility in Sandusky, Ohio. The 60-megawatt test reactor went critical for the first time in 1961 and began its full-power research operations in 1963. From 1961 to 1973, this reactor performed some of the nation’s most advanced nuclear research. The reactor was designed to determine the behavior of metals and other materials after long durations of irradiation. The materials would be used to construct a nuclear-powered rocket. The reactor core, where the chain reaction occurred, sat at the bottom of the tubular pressure vessel, seen here at the center of the shielding pool. The core contained fuel rods with uranium isotopes. A cooling system was needed to reduce the heat levels during the reaction. A neutron-impervious reflector was also employed to send many of the neutrons back to the core. The Plum Brook Reactor Facility was constructed from high-density concrete and steel to prevent the excess neutrons from escaping the facility, but the water in the pool shielded most of the radiation. The water, found in three of the four quadrants served as a reflector, moderator, and coolant. In this photograph, the three 20-ton protective shrapnel shields and hatch have been removed from the top of the pressure tank revealing the reactor tank. An overhead crane could be manipulated to reach any section of this room. It was used to remove the shrapnel shields and transfer equipment.

Nuclear power technology requirements for NASA exploration missions

NASA Technical Reports Server (NTRS)

Bloomfield, Harvey S.

1990-01-01

It is pointed out that future exploration of the moon and Mars will mandate developments in many areas of technology. In particular, major advances will be required in planet surface power systems. Critical nuclear technology challenges that can enable strategic self-sufficiency, acceptable operational costs, and cost-effective space transportation goals for NASA exploration missions have been identified. Critical technologies for surface power systems include stationary and mobile nuclear reactor and radioisotope heat sources coupled to static and dynamic power conversion devices. These technologies can provide dramatic reductions in mass, leading to operational and transportation cost savings. Critical technologies for space transportation systems include nuclear thermal rocket and nuclear electric propulsion options, which present compelling concepts for significantly reducing mass, cost, or travel time required for Earth-Mars transport.

Rotating Fluidized Bed Reactor for Space Nuclear Propulsion. Annual Report; Design Studies and Experimental Results

NASA Technical Reports Server (NTRS)

1971-01-01

The rotating fluidized bed reactor concept is being investigated for possible application in nuclear propulsion systems. Physics calculations show U-233 to be superior to U-235 as a fuel for a cavity reactor of this type. Preliminary estimates of the effect of hydrogen in the reactor, reflector material, and power peaking are given. A preliminary engineering analysis was made for U-235 and U-233 fueled systems. An evaluation of the parameters affecting the design of the system is given, along with the thrust-to-weight ratios. The experimental equipment is described, as are the special photographic techniques and procedures. Characteristics of the fluidized bed and experimental results are given, including photographic evidence of bed fluidization at high rotational velocities.

Submersion criticality safety of tungsten-rhenium urania cermet fuel for space propulsion and power applications

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

A.E. Craft; R. C. O'Brien; S. D. Howe

Nuclear thermal rockets are the preferred propulsion technology for a manned mission to Mars, and tungsten–uranium oxide cermet fuels could provide significant performance and cost advantages for nuclear thermal rockets. A nuclear reactor intended for use in space must remain subcritical before and during launch, and must remain subcritical in launch abort scenarios where the reactor falls back to Earth and becomes submerged in terrestrial materials (including seawater, wet sand, or dry sand). Submersion increases reflection of neutrons and also thermalizes the neutron spectrum, which typically increases the reactivity of the core. This effect is typically very significant for compact,more » fast-spectrum reactors. This paper provides a submersion criticality safety analysis for a representative tungsten/uranium oxide fueled reactor with a range of fuel compositions. Each submersion case considers both the rhenium content in the matrix alloy and the uranium oxide volume fraction in the cermet. The inclusion of rhenium significantly improves the submersion criticality safety of the reactor. While increased uranium oxide content increases the reactivity of the core, it does not significantly affect the submersion behavior of the reactor. There is no significant difference in submersion behavior between reactors with rhenium distributed within the cermet matrix and reactors with a rhenium clad in the coolant channels. The combination of the flooding of the coolant channels in submersion scenarios and the presence of a significant amount of spectral shift absorbers (i.e. high rhenium concentration) further decreases reactivity for short reactor cores compared to longer cores.« less

Thermoelectric converter for SP-100 space reactor power system

NASA Technical Reports Server (NTRS)

Terrill, W. R.; Haley, V. F.

1986-01-01

Conductively coupling the thermoelectric converter to the heat source and the radiator maximizes the utilization of the reactor and radiator temperatures and thereby minimizes the power system weight. This paper presents the design for the converter and the individual thermoelectric cells that are the building block modules for the converter. It also summarizes progress on the fabrication of initial cells and the results obtained from the preparation of a manufacturing plan. The design developed for the SP-100 system utilizes thermally conductive compliant pads that can absorb the displacement and distortion caused by the combinations of temperatures and thermal expansion coefficients. The converter and cell designs provided a 100 kWe system which met the system requirements. Initial cells were fabricated and tested.

Reactor-pumped laser facility at DOE's Nevada Test Site

NASA Astrophysics Data System (ADS)

Lipinski, Ronald J.

1994-05-01

The Nevada Test Site (NTS) is one excellent possibility for a laser power beaming site. It is in the low latitudes of the U.S., is in an exceptionally cloud-free area of the southwest, is already an area of restricted access (which enhances safety considerations), and possesses a highly skilled technical team with extensive engineering and research capabilities from underground testing of our nation's nuclear deterrence. The average availability of cloud-free clear line of site to a given point in space is about 84%. With a beaming angle of +/- 60 degree(s) from the zenith, about 52 geostationary-orbit (GEO) satellites could be accessed continuously from NTS. In addition, the site would provide an average view factor of about 10% for orbital transfer from low earth orbit to GEO. One of the major candidates for a long-duration, high- power laser is a reactor-pumped laser being developed by DOE. The extensive nuclear expertise at NTS makes this site a prime candidate for utilizing the capabilities of a rector pumped laser for power beaming. The site then could be used for many dual-use roles such as industrial material processing research, defense testing, and removing space debris.

Apparatus for localizing disturbances in pressurized water reactors (PWR)

DOEpatents

Sykora, Dalibor

1989-01-01

The invention according to CS-PS 177386, entitled ''Apparatus for increasing the efficiency and passivity of the functioning of a bubbling-vacuum system for localizing disturbances in nuclear power plants with a pressurized water reactor'', concerns an important area of nuclear power engineering that is being developed in the RGW member countries. The invention solves the problems of increasing the reliability and intensification during the operation of the above very important system for guaranteeing the safety of the standard nuclear power plants of Soviet design. The essence of the invention consists in the installation of a simple passively operating supplementary apparatus. Consequently, the following can be observed in the system: first an improvement and simultaneous increase in the reliability of its function during the critical transition period, which follows the filling of the second space with air from the first space; secondly, elimination of the hitherto unavoidable initiating role of the active sprinkler-condensation device present; thirdly, a more effective performance and subjection of the elements to disintegration of the water flowing from the bubbling condenser into the first space; and fourthly, an enhanced utilization of the heat-conducting ability of the water reservoir of the bubbling condenser. Representatives of the supplementary apparatus are autonomous and local secondary systems of the sprinkler-sprayer without an insert, which spray the water under the effect of gravity. 1 fig.

Speculations on future opportunities to evolve Brayton powerplants aboard the space station

NASA Technical Reports Server (NTRS)

English, Robert E.

1987-01-01

The Space Station provides a unique, low-risk environment in which to evolve new capabilities. In this way, the Space Station will grow in capacity, in its range of capabilities, and its economy of operation as a laboratory and as a center for space operations. Although both Rankine and Brayton cycles, two concepts for solar dynamic power generation, now compete to power the station, this paper confines its attention to the Brayton cycle using a mixture of He and Xe as its working fluid. Such a Brayton powerplant to supply the station's increasing demands for both electric power and heat has the potential to gradually evolve higher and higher performance by exploiting already-evolved materials (ASTAR-811C and molten-Li heat storage), its peak cycle temperature rising ultimately to 1500 K. Adapting the station to exploit long tethers (200 to 300 km long) could yield increases in payloads to LEO, to GEO, and to distant destinations in the solar system. Such tethering of the Space Station would not only require additional power for electric propulsion but also would so increase nuclear safety that nuclear powerplants might provide this power. From an 8000-kWt SP-100 reactor, thermoelectric power generation could produce 300 kWe, or adapted solar-Brayton cycle, 2400 to 2800 kWe.

Issues relating to spent nuclear fuel storage on the Oak Ridge Reservation

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

Klein, J.A.; Turner, D.W.

1994-12-31

Currently, about 2,800 metric tons of spent nuclear fuel (SNF) is stored in the US, 1,000 kg of SNF (or about 0.03% of the nation`s total) are stored at the US Department of Energy (DOE) complex in Oak Ridge, Tennessee. However small the total quantity of material stored at Oak Ridge, some of the material is quite singular in character and, thus, poses unique management concerns. The various types of SNF stored at Oak Ridge will be discussed including: (1) High-Flux Isotope Reactor (HFIR) and future Advanced Neutron Source (ANS) fuels; (2) Material Testing Reactor (MTR) fuels, including Bulk Shieldingmore » Reactor (BSR) and Oak Ridge Research Reactor (ORR) fuels; (3) Molten Salt Reactor Experiment (MSRE) fuel; (4) Homogeneous Reactor Experiment (HRE) fuel; (5) Miscellaneous SNF stored in Oak Ridge National Laboratory`s (ORNL`s) Solid Waste Storage Areas (SWSAs); (6) SNF stored in the Y-12 Plant 9720-5 Warehouse including Health. Physics Reactor (HPRR), Space Nuclear Auxiliary Power (SNAP-) 10A, and DOE Demonstration Reactor fuels.« less

Predictions of ion energy distributions and radical fluxes in radio frequency biased inductively coupled plasma etching reactors

NASA Astrophysics Data System (ADS)

Hoekstra, Robert J.; Kushner, Mark J.

1996-03-01

Inductively coupled plasma (ICP) reactors are being developed for low gas pressure (<10s mTorr) and high plasma density ([e]≳1011 cm-3) microelectronics fabrication. In these reactors, the plasma is generated by the inductively coupled electric field while an additional radio frequency (rf) bias is applied to the substrate. One of the goals of these systems is to independently control the magnitude of the ion flux by the inductively coupled power deposition, and the acceleration of ions into the substrate by the rf bias. In high plasma density reactors the width of the sheath above the wafer may be sufficiently thin that ions are able to traverse it in approximately 1 rf cycle, even at 13.56 MHz. As a consequence, the ion energy distribution (IED) may have a shape typically associated with lower frequency operation in conventional reactive ion etching tools. In this paper, we present results from a computer model for the IED incident on the wafer in ICP etching reactors. We find that in the parameter space of interest, the shape of the IED depends both on the amplitude of the rf bias and on the ICP power. The former quantity determines the average energy of the IED. The latter quantity controls the width of the sheath, the transit time of ions across the sheath and hence the width of the IED. In general, high ICP powers (thinner sheaths) produce wider IEDs.

Space nuclear power systems 1989; Proceedings of the 6th Symposium, Albuquerque, NM, Jan. 8-12, 1989. Vols. 1 & 2

NASA Technical Reports Server (NTRS)

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

1992-01-01

The present conference discusses such space nuclear power (SNP) issues as current design trends for SDI applications, ultrahigh heat-flux systems with curved surface subcooled nucleate boiling, design and manufacturing alternatives for low cost production of SNPs, a lightweight radioisotope heater for the Galileo mission, compatible materials for uranium fluoride-based gas core SNPs, Johnson noise thermometry for SNPs, and uranium nitride/rhenium compatibility studies for the SP-100 SNP. Also discussed are system issues in antimatter energy conversion, the thermal design of a heat source for a Brayton cycle radioisotope power system, structural and thermal analyses of an isotope heat source, a novel plant protection strategy for transient reactors, and beryllium toxicity.

Fission Surface Power for the Exploration and Colonization of Mars

NASA Technical Reports Server (NTRS)

Houts, Mike; Porter, Ron; Gaddis, Steve; Van Dyke, Melissa; Martin, Jim; Godfroy, Tom; Bragg-Sitton, Shannon; Garber, Anne; Pearson, Boise

2006-01-01

The colonization of Mars will require abundant energy. One potential energy source is nuclear fission. Terrestrial fission systems are highly developed and have the demonstrated ability to safely produce tremendous amounts of energy. In space, fission systems not only have the potential to safely generate tremendous amounts of energy, but could also potentially be used on missions where alternatives are not practical. Programmatic risks such as cost and schedule are potential concerns with fission surface power (FSP) systems. To be mission enabling, FSP systems must be affordable and programmatic risk must be kept acceptably low to avoid jeopardizing exploration efforts that may rely on FSP. Initial FSP systems on Mars could be "workhorse" units sized to enable the establishment of a Mars base and the early growth of a colony. These systems could be nearly identical to FSP systems used on the moon. The systems could be designed to be safe, reliable, and have low development and recurring costs. Systems could also be designed to fit on relatively small landers. One potential option for an early Mars FSP system would be a 100 kWt class, NaK cooled system analogous to space reactors developed and flown under the U.S. "SNAP" program or those developed and flown by the former Soviet Union ("BUK" reactor). The systems could use highly developed fuel and materials. Water and Martian soil could be used to provide shielding. A modern, high-efficiency power conversion subsystem could be used to reduce required reactor thermal power. This, in turn, would reduce fuel burnup and radiation damage .effects by reducing "nuclear" fuels and materials development costs. A realistic, non-nuclear heated and fully integrated technology demonstration unit (TDU) could be used to reduce cost and programmatic uncertainties prior to initiating a flight program.

Experimental detailed power distribution in a fast spectrum thermionic reactor fuel element at the core/BeO reflector interface region

NASA Technical Reports Server (NTRS)

Klann, P. G.; Lantz, E.

1973-01-01

A zero-power critical assembly was designed, constructed, and operated for the prupose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-7-cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power conversion system. The critical assembly was modified to simulate a fast spectrum advanced thermionics reactor by: (1) using BeO as a reflector in place of some of the existing molybdenum, (2) substituting Nb-1Zr tubing for some of the existing Ta tubing, and (3) inserting four full-scale mockups of thermionic type fuel elements near the core and BeO reflector boundary. These mockups were surrounded with a buffer zone having the equivalent thermionic core composition. In addition to measuring the critical mass of this thermionic configuration, a detailed power distribution in one of the thermionic element stages in the mixed spectrum region was measured. A power peak to average ratio of two was observed for this fuel stage at the midplane of the core and adjacent to the reflector. Also, the power on the outer surface adjacent to the BeO was slightly more than a factor of two larger than the power on the inside surface of a 5.08 cm (2.0 in.) high annular fuel segment with a 2.52 cm (0.993 in. ) o.d. and a 1.86 cm (0.731 in.) i.d.

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.

Operation of the WWR-S Reactor in Poland and in the Period 1961-1962; DOKLAD OB ISPOL'ZOVANII I EKSPLUATATSII REAKTORA TIPA VVR-C V POL'SHE ZA 1961-1962 GODA

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

Aleksandrowicz, J.

1963-03-01

The experimental equipment used in the work at the horizontal reactor channels is listed. Diagrams of the utilization of the nominal reactor power and core loading are given, the reactivity fractions of the separate fuel assemblies are detonated, together with the diagram of reactivity versus burnup. Reactor channels and space used for sample irradiation and isotope production are described, and the total number of irradiations is given. Results of the measurements connected with the routine reactor operation are quoted, namely: analysis of water purity in the primary circuit, analysis of the work of the ion exchanger and mechanical filter, andmore » analysis of air activity in the special ventilation system. Data are given concerning radiation protection of personnel, including individual monitoring. Leak testing of the fuel elements is discussed. Damage of the reactor equipment and appearance of alarm signals are described. (auth)« less

New Concepts for Compact Space Reactor Power Systems for Space Based Radar Applications: A Feasibility Study

DTIC Science & Technology

1989-12-01

SPENT FUEL REPROCESSING COULD ALSO BE EMPLOYED IRRADIATION EXPERIENCE - EXTREMELY LIMITED - JOINT US/UK PROGRAM (ONGOING) - TUI/KFK PROGRAM (CANCELED...only the use of off-the-shelf technologies. For example, conventional fuel technology (uranium dioxide), conventional thermionic conversion...advanced fuel (Americium oxide, A1TI2O3) and advanced thermionic conversion. Concept C involves use of an advanced fuel (Americium oxide, Arri203

A pilot level decision analysis of thermionic reactor development strategy for nuclear electric propulsion

NASA Technical Reports Server (NTRS)

Menke, M. M.; Judd, B. R.

1973-01-01

The development policy for thermionic reactors to provide electric propulsion and power for space exploration was analyzed to develop a logical procedure for selecting development alternatives that reflect the technical feasibility, JPL/NASA project objectives, and the economic environment of the project. The partial evolution of a decision model from the underlying philosophy of decision analysis to a deterministic pilot phase is presented, and the general manner in which this decision model can be employed to examine propulsion development alternatives is illustrated.

Heat transfer system

DOEpatents

Not Available

1980-03-07

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

Heat transfer system

DOEpatents

McGuire, Joseph C.

1982-01-01

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

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

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.

2014-01-01

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

A three-dimensional radiation image display on a real space image created via photogrammetry

NASA Astrophysics Data System (ADS)

Sato, Y.; Ozawa, S.; Tanifuji, Y.; Torii, T.

2018-03-01

The Fukushima Daiichi Nuclear Power Station (FDNPS), operated by Tokyo Electric Power Company Holdings, Inc., went into meltdown after the occurrence of a large tsunami caused by the Great East Japan Earthquake of March 11, 2011. The radiation distribution measurements inside the FDNPS buildings are indispensable to execute decommissioning tasks in the reactor buildings. We have developed a three-dimensional (3D) image reconstruction method for radioactive substances using a compact Compton camera. Moreover, we succeeded in visually recognizing the position of radioactive substances in real space by the integration of 3D radiation images and the 3D photo-model created using photogrammetry.

Gas Foil Bearing Technology Advancements for Closed Brayton Cycle Turbines

NASA Technical Reports Server (NTRS)

Howard, Samuel A.; Bruckner, Robert J.; DellaCorte, Christopher; Radil, Kevin C.

2007-01-01

Closed Brayton Cycle (CBC) turbine systems are under consideration for future space electric power generation. CBC turbines convert thermal energy from a nuclear reactor, or other heat source, to electrical power using a closed-loop cycle. The operating fluid in the closed-loop is commonly a high pressure inert gas mixture that cannot tolerate contamination. One source of potential contamination in a system such as this is the lubricant used in the turbomachine bearings. Gas Foil Bearings (GFB) represent a bearing technology that eliminates the possibility of contamination by using the working fluid as the lubricant. Thus, foil bearings are well suited to application in space power CBC turbine systems. NASA Glenn Research Center is actively researching GFB technology for use in these CBC power turbines. A power loss model has been developed, and the effects of a very high ambient pressure, start-up torque, and misalignment, have been observed and are reported here.

Development of a three-dimensional transient code for reactivity-initiated events of BWRs (boiling water reactors) - Models and code verifications

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

Uematsu, Hitoshi; Yamamoto, Toru; Izutsu, Sadayuki

1990-06-01

A reactivity-initiated event is a design-basis accident for the safety analysis of boiling water reactors. It is defined as a rapid transient of reactor power caused by a reactivity insertion of over $1.0 due to a postulated drop or abnormal withdrawal of the control rod from the core. Strong space-dependent feedback effects are associated with the local power increase due to control rod movement. A realistic treatment of the core status in a transient by a code with a detailed core model is recommended in evaluating this event. A three-dimensional transient code, ARIES, has been developed to meet this need.more » The code simulates the event with three-dimensional neutronics, coupled with multichannel thermal hydraulics, based on a nonequilibrium separated flow model. The experimental data obtained in reactivity accident tests performed with the SPERT III-E core are used to verify the entire code, including thermal-hydraulic models.« less

Summary and evaluation of the Strategic Defense Initiative Space Power Architecture Study

NASA Technical Reports Server (NTRS)

Edenburn, M. (Editor); Smith, J. M. (Editor)

1989-01-01

The Space Power Architecture Study (SPAS) identified and evaluated power subsystem options for multimegawatt electric (MMWE) space based weapons and surveillance platforms for the Strategic Defense Initiative (SDI) applications. Steady state requirements of less than 1 MMWE are adequately covered by the SP-100 nuclear space power program and hence were not addressed in the SPAS. Four steady state power systems less than 1 MMWE were investigated with little difference between them on a mass basis. The majority of the burst power systems utilized H(2) from the weapons and were either closed (no effluent), open (effluent release) or steady state with storage (no effluent). Closed systems used nuclear or combustion heat source with thermionic, Rankine, turboalternator, fuel cell and battery conversion devices. Open systems included nuclear or combustion heat sources using turboalternator, magnetohydrodynamic, fuel cell or battery power conversion devices. The steady state systems with storage used the SP-100 or Star-M reactors as energy sources and flywheels, fuel cells or batteries to store energy for burst applications. As with other studies the open systems are by far the lightest, most compact and simplist (most reliable) systems. However, unlike other studies the SPAS studied potential platform operational problems caused by effluents or vibration.

A Stainless-Steel, Uranium-Dioxide, Potassium-Heatpipe-Cooled Surface Reactor

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

Amiri, Benjamin W.; Nuclear and Radiological Engineering Department, University of Florida, Gainesville, FL 32611; Sims, Bryan T.

2006-01-20

One of the primary goals in designing a fission power system is to ensure that the system can be developed at a low cost and on an acceptable schedule without compromising reliability. The Heatpipe Power System (HPS) is one possible approach for producing near-term, low-cost, space fission power. The Heatpipe Operated Moon Exploration Reactor (HOMER-25) is a HPS designed to produce 25-kWe on the lunar surface for 5 full-power years. The HOMER-25 core is made up of 93% enriched UO2 fuel pins and stainless-steel (SS)/potassium (K) heatpipes in a SS monolith. The heatpipes transport heat generated in the core throughmore » the water shield to a potassium boiler, which drives six Stirling engines. The operating heatpipe temperature is 880 K and the peak fast fluence is 1.6e21 n/cm2, which is well within an established database for the selected materials. The HOMER-25 is designed to be buried in 1.5 m of lunar regolith during operation. By using technology and materials which do not require extensive technology development programs, the HOMER-25 could be developed at a relatively low cost. This paper describes the attributes, specifications, and performance of the HOMER-25 reactor system.« less

Neutrino Physics at Kalinin Nuclear Power Plant: 2002 - 2017

NASA Astrophysics Data System (ADS)

Alekseev, I.; Belov, V.; Brudanin, V.; Danilov, M.; Egorov, V.; Filosofov, D.; Fomina, M.; Hons, Z.; Kazartsev, S.; Kobyakin, A.; Kuznetsov, A.; Machikhiliyan, I.; Medvedev, D.; Nesterov, V.; Olshevsky, A.; Pogorelov, N.; Ponomarev, D.; Rozova, I.; Rumyantseva, N.; Rusinov, V.; Salamatin, A.; Shevchik, Ye; Shirchenko, M.; Shitov, Yu; Skrobova, N.; Starostin, A.; Svirida, D.; Tarkovsky, E.; Tikhomirov, I.; Vlášek, J.; Zhitnikov, I.; Zinatulina, D.

2017-12-01

The results of the research in the field of neutrino physics obtained at Kalinin nuclear power plant during 15 years are presented. The investigations were performed in two directions. The first one includes GEMMA I and GEMMA II experiments for the search of the neutrino magnetic moment, where the best result in the world on the value of the upper limit of this quantity was obtained. The second direction is tied with the measurements by a solid scintillator detector DANSS designed for remote on-line diagnostics of nuclear reactor parameters and search for short range neutrino oscillations. DANSS is now installed at the Kalinin Nuclear Power Plant under the 4-th unit on a movable platform. Measurements of the antineutrino flux demonstrated that the detector is capable to reflect the reactor thermal power with an accuracy of about 1.5% in one day. Investigations of the neutrino flux and their energy spectrum at different distances allowed to study a large fraction of a sterile neutrino parameter space indicated by recent experiments and perform the reanalysis of the reactor neutrino fluxes. Status of the short range oscillation experiment is presented together with some preliminary results based on about 170 days of active data taking during the first year of operation.

Thermionic system evaluated test (TSET) facility description

NASA Astrophysics Data System (ADS)

Fairchild, Jerry F.; Koonmen, James P.; Thome, Frank V.

1992-01-01

A consortium of US agencies are involved in the Thermionic System Evaluation Test (TSET) which is being supported by the Strategic Defense Initiative Organization (SDIO). The project is a ground test of an unfueled Soviet TOPAZ-II in-core thermionic space reactor powered by electrical heat. It is part of the United States' national thermionic space nuclear power program. It will be tested in Albuquerque, New Mexico at the New Mexico Engineering Research Institute complex by the Phillips Laboratoty, Sandia National Laboratories, Los Alamos National Laboratory, and the University of New Mexico. One of TSET's many objectives is to demonstrate that the US can operate and test a complete space nuclear power system, in the electrical heater configuration, at a low cost. Great efforts have been made to help reduce facility costs during the first phase of this project. These costs include structural, mechanical, and electrical modifications to the existing facility as well as the installation of additional emergency systems to mitigate the effects of utility power losses and alkali metal fires.

Testing of Liquid Metal Components for Nuclear Surface Power Systems

NASA Technical Reports Server (NTRS)

Polzin, Kurt A.; Godfroy, Thomas J.; Pearson, J. Boise

2010-01-01

The Early Flight Fission Test Facility (EFF-TF) was established by the Marshall Space Flight Center (MSFC) to provide a capability for performing hardware-directed activities to support multiple in-space nuclear reactor concepts by using a non-nuclear test methodology. This includes fabrication and testing at both the module/component level and near prototypic reactor configurations. The EFF-TF is currently supporting an effort to develop an affordable fission surface power (AFSP) system that could be deployed on the Lunar surface. The AFSP system is presently based on a pumped liquid metal-cooled (Sodium-Potassium eutectic, NaK-78) reactor design. This design was derived from the only fission system that the United States has deployed for space operation, the Systems for Nuclear Auxiliary Power (SNAP) 10A reactor, which was launched in 1965. Two prototypical components recently tested at MSFC were a pair of Stirling power conversion units that would be used in a reactor system to convert heat to electricity, and an annular linear induction pump (ALIP) that uses travelling electromagnetic fields to pump the liquid metal coolant through the reactor loop. First ever tests were conducted at MSFC to determine baseline performance of a pair of 1 kW Stirling convertors using NaK as the hot side working fluid. A special test rig was designed and constructed and testing was conducted inside a vacuum chamber at MSFC. This test rig delivered pumped NaK for the hot end temperature to the Stirlings and water as the working fluid on the cold end temperature. These test were conducted through a hot end temperature range between 400 to 550C in increments of 50 C and a cold end temperature range from 30 to 70 C in 20 C increments. Piston amplitudes were varied from 6 to 1 1mm in .5 mm increments. A maximum of 2240 Watts electric was produced at the design point of 550 hot end, 40 C cold end with a piston amplitude of 10.5mm. This power level was reached at a gross thermal efficiency of 28%. A baseline performance map was established for the pair of 1kW Stirling convertors. The performance data will then be used for design modification to the Stirling convertors. The ALIP tested at MSFC has no moving parts and no direct electrical connections to the liquid metal containing components. Pressure is developed by the interaction of the magnetic field produced by the stator and the current which flows as a result of the voltage induced in the liquid metal contained in the pump duct. Flow is controlled by variation of the voltage supplied to the pump windings. Under steady-state conditions, pump performance is measured for flow rates from 0.5-4.3 kg/s. The pressure rise developed by the pump to support these flow rates is roughly 5-65 kPa. The RMS input voltage (phase-to-phase voltage) ranges from 5-120 V, while the frequency can be varied arbitrarily up to 60 Hz. Performance is quantified at different loop temperature levels from 50 C up to 650 C, which is the peak operating temperature of the proposed AFSP reactor. The transient response of the pump is also evaluated to determine its behavior during startup and shut-down procedures.

10 CFR 52.167 - Issuance of manufacturing license.

Code of Federal Regulations, 2010 CFR

2010-01-01

... proposed reactor(s) can be incorporated into a nuclear power plant and operated at sites having... design and manufacture the proposed nuclear power reactor(s); (5) The proposed inspections, tests... the construction of a nuclear power facility using the manufactured reactor(s). (2) A holder of a...

Advanced refractory metals and composites for extraterrestrial power systems

NASA Technical Reports Server (NTRS)

Titran, R. H.; Grobstein, Toni L.

1990-01-01

Concepts for future space power systems include nuclear and focused solar heat sources coupled to static and dynamic power-conversion devices; such systems must be designed for service lives as long as 30 years, despite service temperatures of the order of 1600 K. Materials are a critical technology-development factor in such aspects of these systems as reactor fuel containment, environmental protection, power management, and thermal management. Attention is given to the prospective performance of such refractory metals as Nb, W, and Mo alloys, W fiber-reinforced Nb-matrix composites, and HfC precipitate-strengthened W-Re alloys.

Nuclear Science Symposium, 19th, and Nuclear Power Systems Symposium, 4th, Miami, Fla., December 6-8, 1972, Proceedings.

NASA Technical Reports Server (NTRS)

1973-01-01

Major topics covered include radiation monitoring instrumentation, nuclear circuits and systems, biomedical applications of nuclear radiation in diagnosis and therapy, plasma research for fusion power, reactor control and instrumentation, nuclear power standards, and applications of digital computers in nuclear power plants. Systems and devices for space applications are described, including the Apollo alpha spectrometer, a position sensitive detection system for UV and X-ray photons, a 4500-volt electron multiplier bias supply for satellite use, spark chamber systems, proportional counters, and other devices. Individual items are announced in this issue.

Anticipatory control of xenon in a pressurized water reactor

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

Impink, A.J. Jr.

1987-02-10

A method is described for automatically dampening xenon-135 spatial transients in the core of a pressurized water reactor having control rods which regulate reactor power level, comprising the steps of: measuring the neutron flu in the reactor core at a plurality of axially spaced locations on a real-time, on-line basis; repetitively generating from the neutron flux measurements, on a point-by-point basis, signals representative of the current axial distribution of xenon-135, and signals representative of the current rate of change of the axial distribution of xenon-135; generating from the xenon-135 distribution signals and the rate of change of xenon distribution signals,more » control signals for reducing the xenon transients; and positioning the control rods as a function of the control signals to dampen the xenon-135 spatial transients.« less

Automated startup of the MIT research reactor

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

Kwok, K.S.

1992-01-01

This summary describes the development, implementation, and testing of a generic method for performing automated startups of nuclear reactors described by space-independent kinetics under conditions of closed-loop digital control. The technique entails first obtaining a reliable estimate of the reactor's initial degree of subcriticality and then substituting that estimate into a model-based control law so as to permit a power increase from subcritical on a demanded trajectory. The estimation of subcriticality is accomplished by application of the perturbed reactivity method. The shutdown reactor is perturbed by the insertion of reactivity at a known rate. Observation of the resulting period permitsmore » determination of the initial degree of subcriticality. A major advantage to this method is that repeated estimates are obtained of the same quantity. Hence, statistical methods can be applied to improve the quality of the calculation.« less

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...

Closed Brayton cycle power conversion systems for nuclear reactors :

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

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

2006-04-01

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

Design analysis and risk assessment for a single stage to orbit nuclear thermal rocket

NASA Astrophysics Data System (ADS)

Labib, Satira I.

Recent advances in high power density fuel materials have renewed interest in nuclear thermal rockets (NTRs) as a viable propulsion technology for future space exploration. This thesis describes the design of three NTR reactor engines designed for the single stage to orbit launch of payloads from 1-15 metric tons. Thermal hydraulic and rocket engine analyses indicate that the proposed rocket engines are able to reach specific impulses in excess of 700 seconds. Neutronics analyses performed using MCNP5 demonstrate that the hot excess reactivity, shutdown margin, and submersion criticality requirements are satisfied for each NTR reactor. The reactors each consist of a 40 cm diameter core packed with hexagonal tungsten cermet fuel elements. The core is surrounded by radial and axial beryllium reflectors and eight boron carbide control drums. At the same power level, the 40 cm reactor results in the lowest radiation dose rate of the three reactors. Radiation dose rates decrease to background levels ~3.5 km from the launch site. After a one-year decay time, all of the activated materials produced by an NTR launch would be classified as Class A low-level waste. The activation of air produces significant amounts of argon-41 and nitrogen-16 within 100 m of the launch. The derived air concentration, DAC, from the activation products decays to less than unity within two days, with only argon-41 remaining. After 10 minutes of full power operation the 120 cm core corresponding to a 15 MT payload contains 2.5 x 1013, 1.4 x 1012, 1.5 x 1012, and 7.8 x 10 7 Bq of 131I, 137Cs, 90Sr, and 239Pu respectively. The decay heat after shutdown increases with increasing reactor power with a maximum decay heat of 108 kW immediately after shutdown for the 15 MT payload.

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

Mirror fusion propulsion system: A performance comparison with alternate propulsion systems for the manned Mars Mission

NASA Technical Reports Server (NTRS)

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

1993-01-01

The performance characteristics of several propulsion technologies applied to piloted Mars missions are compared. The characteristics that are compared are Initial Mass in Low Earth Orbit (IMLEO), mission flexibility, and flight times. The propulsion systems being compared are both demonstrated and envisioned: Chemical (or Cryogenic), Nuclear Thermal Rocket (NTR) solid core, NTR gas core, Nuclear Electric Propulsion (NEP), and a mirror fusion space propulsion system. The proposed magnetic mirror fusion reactor, known as the Mirror Fusion Propulsion System (MFPS), is described. The description is an overview of a design study that was conducted to convert a mirror reactor experiment at Lawrence Livermore National Lab (LLNL) into a viable space propulsion system. Design principles geared towards minimizing mass and maximizing power available for thrust are identified and applied to the LLNL reactor design, resulting in the MFPS. The MFPS' design evolution, reactor and fuel choices, and system configuration are described. Results of the performance comparison shows that the MFPS minimizes flight time to 60 to 90 days for flights to Mars while allowing continuous return-home capability while at Mars. Total MFPS IMLEO including propellant and payloads is kept to about 1,000 metric tons.

Mirror fusion propulsion system - A performance comparison with alternate propulsion systems for the manned Mars mission

NASA Technical Reports Server (NTRS)

Deveny, M.; Carpenter, S.; O'Connell, T.; Schulze, N.

1993-01-01

The performance characteristics of several propulsion technologies applied to piloted Mars missions are compared. The characteristics that are compared are Initial Mass in Low Earth Orbit (IMLEO), mission flexibility, and flight times. The propulsion systems being compared are both demonstrated and envisioned: Chemical (or Cryogenic), Nuclear Thermal Rocket (NTR) solid core, NTR gas core, Nuclear Electric Propulsion (NEP), and a mirror fusion space propulsion system. The proposed magnetic mirror fusion reactor, known as the Mirror Fusion Propulsion System (MFPS), is described. The description is an overview of a design study that was conducted to convert a mirror reactor experiment at Lawrence Livermore National Lab (LLNL) into a viable space propulsion system. Design principles geared towards minimizing mass and maximizing power available for thrust are identified and applied to the LLNL reactor design, resulting in the MFPS. The MFPS' design evolution, reactor and fuel choices, and system configuration are described. Results of the performance comparison shows that the MFPS minimizes flight time to 60 to 90 days for flights to Mars while allowing continuous return-home capability while at Mars. Total MFPS IMLEO including propellant and payloads is kept to about 1,000 metric tons.

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.

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...

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 strength of the negative reactivity feedback in the UTVR, it is found that external reactivity insertions alone are inadequate for bringing about significant power level changes during normal reactor operations. Additional methods of reactivity control such as variations in the gaseous fuel mass flow rate, are needed to achieve the desired power level oontrol.

Double Retort System for Materials Compatibility Testing

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

V. Munne; EV Carelli

2006-02-23

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

Flying Reactors: The Political Feasibility of Nuclear Power in Space

DTIC Science & Technology

2004-04-01

anxiety that ground based nuclear power caused in the minds of the public starting in the early 1970s. The risks and rewards of SNP are in the public...the most. Even extremely small amounts lodging in the body, particularly the lungs are thought to certainly cause lung cancer . Fear of a plutonium...Some have argued that significant increases in worldwide cancer can be directly attributed to the additional plutonium that was added to the

New 5 Kilowatt Free-piston Stirling Space Convertor Developments

NASA Technical Reports Server (NTRS)

Brandhorst, Henry W., Jr.; Chapman, Peter A., Jr.

2007-01-01

The NASA Vision for Exploration of the moon may someday require a nuclear reactor coupled with a free-piston Stirling convertor at a power level of 30-40 kW. In the 1990s, Mechanical Technology Inc. s Stirling Engine Systems Division (some of whose Stirling personnel are now at Foster-Miller, Inc.) developed a 25 kW free piston Stirling Space Power Demonstrator Engine under the SP-100 program. This system consisted of two 12.5 kW engines connected at their hot ends and mounted in tandem to cancel vibration. Recently, NASA and DoE have been developing dual 55 W and 80 W Stirling convertor systems for potential use with radioisotope heat sources. Total test times of all convertors in this effort exceed 120,000 hours. Recently, NASA began a new project with Auburn University to develop a 5 kW, single convertor for potential use in a lunar surface reactor power system. Goals of this development program include a specific power in excess of 140 W/kg at the convertor level, lifetime in excess of five years and a control system that will safely manage the convertors in case of an emergency. Auburn University awarded a subcontract to Foster-Miller, Inc. to undertake development of the 5 kW Stirling Convertor Assembly. The characteristics of the design along with progress in developing the system will be described.

New 5 kW free-piston Stirling space convertor developments

NASA Astrophysics Data System (ADS)

Brandhorst, Henry W., Jr.; Chapman, Peter A., Jr.

2008-07-01

The NASA Vision for Exploration of the moon may someday require a nuclear reactor coupled with a free-piston Stirling convertor at a power level of 30-40 kW. In the 1990s, Mechanical Technology Inc.'s Stirling Engine Systems Division (some of whose Stirling personnel are now at Foster-Miller, Inc.) developed a 25 kW free-piston Stirling Space Power Demonstrator Engine under the SP-100 program. This system consisted of two 12.5 kW engines connected at their hot ends and mounted in tandem to cancel vibration. Recently, NASA and DoE have been developing dual 55 and 80 W Stirling convertor systems for potential use with radioisotope heat sources. Total test times of all convertors in this effort exceed 120,000 h. Recently, NASA began a new project with Auburn University to develop a 5 kW, single convertor for potential use in a lunar surface reactor power system. Goals of this development program include a specific power in excess of 140 W/kg at the convertor level, lifetime in excess of five years and a control system that will safely manage the convertors in case of an emergency. Auburn University awarded a subcontract to Foster-Miller, Inc. to undertake development of the 5 kW Stirling convertor assembly. The characteristics of the design along with progress in developing the system will be described.

Nuclear Science Symposium, 27th, and Symposium on Nuclear Power Systems, 12th, Orlando, Fla., November 5-7, 1980, Proceedings

NASA Technical Reports Server (NTRS)

Martini, M.

1981-01-01

Advances in instrumentation for use in nuclear-science studies are described. Consideration is given to medical instrumentation, computerized fluoroscopy, environmental instrumentation, data acquisition techniques, semiconductor detectors, microchannel plates and photomultiplier tubes, reactor instrumentation, neutron detectors and proportional counters, and space instrumentation.

Mass and Performance Estimates for 5 to 1000 kW(e) Nuclear Reactor Power Systems for Space Applications

DTIC Science & Technology

1990-12-01

Albany Street Cambridge, MA 02139 Dave Berwald Grumman Aerospace Corporation MS B20-05 Bethpage, NY 11714 F. Best Assistant Professor Texas A&M... Zielinski U. S. Department of Energy SAN-ACR Division 13333 Broadway Oakland, CA 94612 G. L. Zigler Science & Engineering Associates 6301

A 1 GeV CW FFAG High Intensity Proton Driver

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

Johnstone, C.; Sheehy, S. L.

2012-05-01

The drive for high beam power, high duty cycle, and reliable beams at reasonable cost has focused world attention on fixed-field accelerators, notably Fixed-Field Alternating Gradient accelerators (FFAGs). High-intensity GeV proton drivers are of particular interest, as these encounter duty cycle and space-charge limits in the synchrotron and machine size concerns in the weaker-focusing cyclotron. Recently, the concept of isochronous orbits has been explored and developed for non-scaling FFAGs using powerful new methodologies in FFAG accelerator design. These new breeds of FFAGs have been identified by international collaborations for serious study thanks to their potential applications including Accelerator Driven Subcriticalmore » Reactors (ADS) a nd Accelerator Transmutation of Waste. The extreme reliability requirements for ADS mandate CW operation capability and the FFAG s strong focusing, particularly in the vertical, will serve to mitigate the effect of space charge (as compared with the weak- focusing cyclotron). This paper reports on these new advances in FFAG accelerator technology and presents a stable, 0.25-1GeV isochronous FFAG for an accelerator driven subcritical reactor.« less

Post Irradiation Evaluation of Thermal Control Coatings and Solid Lubricants to Support Fission Surface Power Systems

NASA Astrophysics Data System (ADS)

Bowman, Cheryl L.; Jaworske, Donald A.; Stanford, Malcolm K.; Persinger, Justin A.; Khorsandi, Behrooz; Blue, Thomas E.

2007-01-01

The development of a nuclear power system for space missions, such as the Jupiter Icy Moons Orbiter or a lunar outpost, requires substantially more compact reactor design than conventional terrestrial systems. In order to minimize shielding requirements and hence system weight, the radiation tolerance of component materials within the power conversion and heat rejection systems must be defined. Two classes of coatings, thermal control paints and solid lubricants, were identified as material systems for which limited radiation hardness information was available. Screening studies were designed to explore candidate coatings under a predominately fast neutron spectrum. The Ohio State Research Reactor Facility staff performed irradiation in a well characterized, mixed energy spectrum and performed post irradiation analysis of representative coatings for thermal control and solid lubricant applications. Thermal control paints were evaluated for 1 MeV equivalent fluences from 1013 to 1015 n/cm2. No optical degradation was noted although some adhesive degradation was found at higher fluence levels. Solid lubricant coatings were evaluated for 1 MeV equivalent fluences from 1015 to 1016 n/cm2 with coating adhesion and flexibility used for post irradiation evaluation screening. The exposures studied did not lead to obvious property degradation indicating the coatings would have survived the radiation environment for the previously proposed Jupiter mission. The results are also applicable to space power development programs such as fission surface power for future lunar and Mars missions.

Post Irradiation Evaluation of Thermal Control Coatings and Solid Lubricants to Support Fission Surface Power Systems

NASA Technical Reports Server (NTRS)

Bowman, Cheryl L.; Jaworske, Donald A.; Stanford, Malcolm K.; Persinger, Justin A.; Khorsandi, Behrooz; Blue, Thomas E.

2007-01-01

The development of a nuclear power system for space missions, such as the Jupiter Icy Moons Orbiter or a lunar outpost, requires substantially more compact reactor design than conventional terrestrial systems. In order to minimize shielding requirements and hence system weight, the radiation tolerance of component materials within the power conversion and heat rejection systems must be defined. Two classes of coatings, thermal control paints and solid lubricants, were identified as material systems for which limited radiation hardness information was available. Screening studies were designed to explore candidate coatings under a predominately fast neutron spectrum. The Ohio State Research Reactor Facility staff performed irradiation in a well characterized, mixed energy spectrum and performed post irradiation analysis of representative coatings for thermal control and solid lubricant applications. Thermal control paints were evaluated for 1 MeV equivalent fluences from 10(exp 13) to 10(exp 15) n per square centimeters. No optical degradation was noted although some adhesive degradation was found at higher fluence levels. Solid lubricant coatings were evaluated for 1 MeV equivalent fluences from 10(exp 15) to 10(exp 16) n per square centimeters with coating adhesion and flexibility used for post irradiation evaluation screening. The exposures studied did not lead to obvious property degradation indicating the coatings would have survived the radiation environment for the previously proposed Jupiter mission. The results are also applicable to space power development programs such as fission surface power for future lunar and Mars missions.

NASA's future space power needs and requirements

NASA Technical Reports Server (NTRS)

Schnyer, A. D.; Sovie, Ronald J.

1990-01-01

The National Space Policy of 1988 established the U.S.'s long-range civil space goals, and has served to guide NASA's recent planning for future space mission operations. One of the major goals was to extend the human presence beyond earth's boundaries and to advance the scientific knowledge of the solar system. A broad spectrum of potential civil space mission opportunities and interests are currently being investigated by NASA to meet the espoused goals. Participation in many of these missions requires power systems with capabilities far beyond what exists today. In other mission examples, advanced power systems technology could enhance mission performance significantly. Power system requirements and issues that need resolution to ensure eventual mission accomplishment are addressed, in conjunction with the ongoing NASA technology development efforts and the need for even greater innovative efforts to match the ambitious solar exploration mission goals. Particular attention is given to potential lunar surface operations and technology goals, based on investigations to date. It is suggested that the nuclear reactor power systems can best meet long-life requirements as well as dramatically reduce the earth-surface-to-lunar-surface transportation costs due to the lunar day/night cycle impact on the solar system's energy storage mass requirements. The state of the art of candidate power systems and elements for the lunar application and the respective exploration technology goals for mission life requirements from 10 to 25 years are examined.

Long titanium heat pipes for high-temperature space radiators

NASA Technical Reports Server (NTRS)

Girrens, S. P.; Ernst, D. M.

1982-01-01

Titanium heat pipes are being developed to provide light weight, reliable heat rejection devices as an alternate radiator design for the Space Reactor Power System (SP-100). The radiator design includes 360 heat pipes, each of which is 5.2 m long and dissipates 3 kW of power at 775 K. The radiator heat pipes use potassium as the working fluid, have two screen arteries for fluid return, a roughened surface distributive wicking system, and a D shaped cross section container configuration. A prototype titanium heat pipe, 5.5 m long, was fabricated and tested in space simulating conditions. Results from startup and isothermal operation tests are presented. These results are also compared to theoretical performance predictions that were used to design the heat pipe initially.

Long titanium heat pipes for high-temperature space radiators

NASA Technical Reports Server (NTRS)

Girrens, S. P.; Ernst, D. M.

1982-01-01

Titanium heat pipes are being developed to provide light weight, reliable heat rejection devices as an alternate radiator design for the Space Reactor Power System (SP-100). The radiator design includes 360 heat pipes, each of which is 5.2 m long and dissipates 3 kW of power at 775 K. The radiator heat pipes use potassium as the working fluid, have two screen arteries for fluid return, a roughened surface distributive wicking system, and a D-shaped cross-section container configuration. A prototype titanium heat pipe, 5.5-m long, has been fabricated and tested in space-simulating conditions. Results from startup and isothermal operation tests are presented. These results are also compared to theoretical performance predictions that were used to design the heat pipe initially.

Nuclear Systems Kilopower Overview

NASA Technical Reports Server (NTRS)

Palac, Don; Gibson, Marc; Mason, Lee; Houts, Michael; McClure, Patrick; Robinson, Ross

2016-01-01

The Nuclear Systems Kilopower Project was initiated by NASAs Space Technology Mission Directorate Game Changing Development Program in fiscal year 2015 to demonstrate subsystem-level technology readiness of small space fission power in a relevant environment (Technology Readiness Level 5) for space science and human exploration power needs. The Nuclear Systems Kilopower Project consists of two elements. The primary element is the Kilopower Prototype Test, also called the Kilopower Reactor Using Stirling Technology(KRUSTY) Test. This element consists of the development and testing of a fission ground technology demonstrator of a 1 kWe fission power system. A 1 kWe system matches requirements for some robotic precursor exploration systems and future potential deep space science missions, and also allows a nuclear ground technology demonstration in existing nuclear test facilities at low cost. The second element, the Mars Kilopower Scalability Study, consists of the analysis and design of a scaled-up version of the 1 kWe reference concept to 10 kWe for Mars surface power projected requirements, and validation of the applicability of the KRUSTY experiment to key technology challenges for a 10 kWe system. If successful, these two elements will lead to initiation of planning for a technology demonstration of a 10 kWe fission power capability for Mars surface outpost power.

10 CFR 72.1 - Purpose.

Code of Federal Regulations, 2011 CFR

2011-01-01

... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Provisions § 72.1 Purpose. The... receive, transfer, and possess power reactor spent fuel, power reactor-related Greater than Class C (GTCC... reactor spent fuel, high-level radioactive waste, power reactor-related GTCC waste, and other radioactive...

10 CFR 72.1 - Purpose.

Code of Federal Regulations, 2010 CFR

2010-01-01

... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Provisions § 72.1 Purpose. The... receive, transfer, and possess power reactor spent fuel, power reactor-related Greater than Class C (GTCC... reactor spent fuel, high-level radioactive waste, power reactor-related GTCC waste, and other radioactive...

RELAP-7 Development Updates

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

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

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

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...

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...

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

NASA Astrophysics Data System (ADS)

Schriener, Timothy M.

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

Multiple Restart Testing of a Stainless Steel Sodium Heat Pipe Module

NASA Technical Reports Server (NTRS)

Martin, James; Mireles, Omar; Reid, Robert

2005-01-01

A heat pipe cooled reactor is one of several candidate reactor cores being considered for space power and propulsion systems to support future space exploration activities. Long life heat pipe modules. with designs verified through a combination of theoretical analysis and experimental evaluations. would be necessary to establish the viability of this option. A hardware-based program was initiated to begin experimental testing of components to verify compliance of proposed designs. To this end, a number of stainless steel/sodium heat pipe modules have been designed and fabricated to support experimental testing of a Safe Affordable Fission Engine (SAFE) project, a 100-kWt core design pursued jointly by the Marshall Space Flight Center and the Los Alamos National Laboratory. One of the SAFE heat pipe modules was successfully subjected to over 200 restarts. examining the behavior of multiple passive freeze/thaw operations. Typical operation included a 1-hour startup to an average evaporator temperature of 1000 K followed by a 15 minute hold at temperature. Nominal maximum input power during the hold period was 1.9 kW. Between heating cycles the module was cooled to less than 325 K, returning the sodium to a frozen state in preparation fop the next startup cycle.

Testing of Liquid Metal Components for Nuclear Surface Power Systems

NASA Technical Reports Server (NTRS)

Polzin, K. A.; Pearson, J. B.; Godfroy, T. J.; Schoenfeld, M.; Webster, K.; Briggs, M. H.; Geng, S. M.; Adkins, H. E.; Werner, J. E.

2010-01-01

The capability to perform testing at both the module/component level and in near prototypic reactor configurations using a non-nuclear test methodology allowed for evaluation of two components critical to the development of a potential nuclear fission power system for the lunar surface. A pair of 1 kW Stirling power convertors, similar to the type that would be used in a reactor system to convert heat to electricity, were integrated into a reactor simulator system to determine their performance using pumped NaK as the hot side working fluid. The performance in the pumped-NaK system met or exceed the baseline performance measurements where the converters were electrically heated. At the maximum hot-side temperature of 550 C the maximum output power was 2375 watts. A specially-designed test apparatus was fabricated and used to quantify the performance of an annular linear induction pump that is similar to the type that could be used to circulate liquid metal through the core of a space reactor system. The errors on the measurements were generally much smaller than the magnitude of the measurements, permitting accurate performance evaluation over a wide range of operating conditions. The pump produced flow rates spanning roughly 0.16 to 5.7 l/s (2.5 to 90 GPM), and delta p levels from less than 1 kPa to 90 kPa (greater than 0.145 psi to roughly 13 psi). At the nominal FSP system operating temperature of 525 C the maximum efficiency was just over 4%.

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...

Discharge ignition in the diaphragm configuration supplied by DC non-pulsing voltage

NASA Astrophysics Data System (ADS)

Hlochová, L.; Hlavatá, L.; Kozáková, Z.; Krčma, F.

2016-05-01

This work deals with the ignition of the discharge in the diaphragm configuration generated in water solutions containing supporting NaCl electrolyte. The reactor has volume of 110 ml and it is made of polycarbonate. HV electrodes made of stainless steel are placed in this reactor. Ceramic (Shapal-MTM) diaphragm is placed in the barrier separating the cathode and the anode space. An electric power source supplies the reactor by constant DC voltage up to 4 kV and electric current up to 300 mA. The discharge ignition is compared in the reactor with different sizes of diaphragms. Measurements are carried out in electrolyte solutions with the same conductivity. Images of plasma streamers and bubble formation are taken by an ICCD camera iStar 734. Electrical characteristics are measured by an oscilloscope LeCroy LT 374 L in order to determine breakdown moments at different experimental conditions.

Preliminary Design of a SP-100/Stirling Radiatively Coupled Heat Exchanger

NASA Technical Reports Server (NTRS)

Schmitz, Paul; Tower, Leonard; Dawson, Ronald; Blue, Brian; Dunn, Pat

1995-01-01

Several methods for coupling the SP-100 space nuclear reactor to the NASA Lewis Research Center's Free Piston Stirling Power Convertor (FPSPC) are presented. A 25 kWe, dual opposed Stirling convertor configuration is used in these designs. The concepts use radiative coupling between the SP-100 lithium loop and the sodium heat pipe of the Stirling convertor to transfer the heat from the reactor to the convertor. Four separate configurations are presented. Masses for the four designs vary from 41 to 176 kgs. Each design's structure, heat transfer characteristics, and heat pipe performance are analytically modeled.

Shield materials recommended for space power nuclear reactors

NASA Technical Reports Server (NTRS)

Kaszubinski, L. J.

1973-01-01

Lithium hydride is recommended for neutron attenuation and depleted uranium is recommended for gamma ray attenuation. For minimum shield weights these materials must be arranged in alternate layers to attenuate the secondary gamma rays efficiently. In the regions of the shield near the reactor, where excessive fissioning occurs in the uranium, a tungsten alloy is used instead. Alloys of uranium such as either the U-0.5Ti or U-8Mo are available to accommodate structural requirements. The zone-cooled casting process is recommended for lithium hydride fabrication. Internal honeycomb reinforcement to control cracks in the lithium hydride is recommended.

Dynamic Computer Model of a Stirling Space Nuclear Power System

DTIC Science & Technology

2006-05-04

diagram of electric propulsion…………………………………. 17 Figure 2-1. General NEP structure……………………………………………………….20 Figure 2-2. Fission of uranium -235...Figure 2-1. General NEP structure. [20] 21 Figure 2-2. Fission of uranium -235. In a fast reactor, the average number of neutrons...that is modeled for this project is a 600 kW(t) fast fission reactor consisting of uranium nitride fuel and sodium potassium coolant. Its dynamic

Preliminary design of a SP-100/Stirling radiatively coupled heat exchanger

NASA Astrophysics Data System (ADS)

Schmitz, Paul; Tower, Leonard; Dawson, Ronald; Blue, Brian; Dunn, Pat

1995-10-01

Several methods for coupling the SP-100 space nuclear reactor to the NASA Lewis Research Center's Free Piston Stirling Power Convertor (FPSPC) are presented. A 25 kWe, dual opposed Stirling convertor configuration is used in these designs. The concepts use radiative coupling between the SP-100 lithium loop and the sodium heat pipe of the Stirling convertor to transfer the heat from the reactor to the convertor. Four separate configurations are presented. Masses for the four designs vary from 41 to 176 kgs. Each design's structure, heat transfer characteristics, and heat pipe performance are analytically modeled.

Robotic influence in the conceptual design of mechanical systems in space and vice versa - A survey

NASA Technical Reports Server (NTRS)

Sanger, George F.

1988-01-01

A survey of methods using robotic devices to construct structural elements in space is presented. Two approaches to robotic construction are considered: one in which the structural elements are designed using conventional aerospace techniques which tend to constrain the function aspects of robotics and one in which the structural elements are designed from the conceptual stage with built-in robotic features. Examples are presented of structural building concepts using robotics, including the construction of the SP-100 nuclear reactor power system, a multimirror large aperture IR space telescope concept, retrieval and repair in space, and the Flight Telerobotic Servicer.

Future space transport

NASA Technical Reports Server (NTRS)

Grishin, S. D.; Chekalin, S. V.

1984-01-01

Prospects for the mastery of space and the basic problems which must be solved in developing systems for both manned and cargo spacecraft are examined. The achievements and flaws of rocket boosters are discussed as well as the use of reusable spacecraft. The need for orbiting satellite solar power plants and related astrionics for active control of large space structures for space stations and colonies in an age of space industrialization is demonstrated. Various forms of spacecraft propulsion are described including liquid propellant rocket engines, nuclear reactors, thermonuclear rocket engines, electrorocket engines, electromagnetic engines, magnetic gas dynamic generators, electromagnetic mass accelerators (rail guns), laser rocket engines, pulse nuclear rocket engines, ramjet thermonuclear rocket engines, and photon rockets. The possibilities of interstellar flight are assessed.

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...

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

NASA Technical Reports Server (NTRS)

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

1973-01-01

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

Thermionic fuel element for the S-prime reactor

NASA Astrophysics Data System (ADS)

Van Hagan, Thomas H.; Drees, Elizabeth A.

1993-01-01

Technical aspects of the thermionic fuel element (TFE) design proposed for the S-PRIME space nuclear power system are discussed. Topics covered include the rational for selecting a multicell TFE approach, a technical description of the S-PRIME TFE and its estimated performance, and the technology readiness of the design, which emphasizes techology maturity and low risk.

SP-100 lithium thaw design, analysis, and testing

NASA Astrophysics Data System (ADS)

Choe, Hwang; Schrag, Michael R.; Koonce, David R.; Gamble, Robert E.; Halfen, Frank J.; Kirpich, Aaron S.

1993-01-01

The thaw design has been established for the 100 kWe SP-100 Space Reactor Power System. System thaw/startup analysis has confirmed that all system thaw requirements are met, and that rethaw and restart can be easily accomplished with this design. In addition, a series of lithium thaw characterization tests has been performed, confirming key design assumptions.

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.

Five Lectures on Nuclear Reactors Presented at Cal Tech

DOE R&D Accomplishments Database

Weinberg, Alvin M.

1956-02-10

The basic issues involved in the physics and engineering of nuclear reactors are summarized. Topics discussed include theory of reactor design, technical problems in power reactors, physical problems in nuclear power production, and future developments in nuclear power. (C.H.)

Performance Testing of a Liquid Metal Pump for In-Space Power Systems

NASA Technical Reports Server (NTRS)

Polzin, Kurt

2011-01-01

Fission surface power (FSP) systems could be used to provide power on the surface of the moon, Mars, or other planets and moons of our solar system. Fission power systems could provide excellent performance at any location, including those near the poles or other permanently shaded regions, and offer the capability to provide on demand power at any time, even at large distances from the sun. Fission-based systems also offer the potential for outposts, crew and science instruments to operate in a power-rich environment. NASA has been exploring technologies with the goal of reducing the cost and technical risk of employing FSP systems. A reference 40 kWe option has been devised that is cost-competitive with alternatives while providing more power for less mass anywhere on the lunar surface. The reference FSP system is also readily extensible for use on Mars, where it would be capable of operating through global dust storms and providing year-round power at any Martian latitude. Detailed development of the FSP concept and the reference mission are documented in various other reports. The development discussed in this paper prepares the way for testing of the Technology Demonstration Unit (TDU), which is a 10 kWe end-to-end test of FSP technologies intended to raise the entire FSP system to technology readiness level (TRL) 6. The Early Flight Fission Test Facility (EFF-TF) was established by NASA s Marshall Space Flight Center (MSFC) to provide a capability for performing hardware-directed activities to support multiple in-space nuclear reactor concepts by using a nonnuclear test methodology. This includes fabrication and testing at both the module/component level and at near prototypic reactor components and configurations allowing for realistic thermal-hydraulic evaluations of systems. The liquid-metal pump associated with the FSP system must be compatible with the liquid NaK coolant and have adequate performance to enable a viable flight system. Idaho National Laboratory (INL) was tasked with the modeling, design, and fabrication of an ALIP suitable for the FSP reference mission. A prototypic ALIP was fabricated under the direction of INL and shipped to MSFC for inclusion in the Technology Demonstration Unit (TDU), a quarter-scale end-to-end reactor simulator system that is scheduled for testing at NASA-GRC. Before inclusion in the TDU, the ALIP was tested in the ALIP test circuit (ATC), which is a rig developed and operated at MSFC for the specific purpose of providing accurate quantification of liquid metal pump performance. Data showing the pump performance curves (pressure, flowrate, and pump efficiency) are presented for various operating power levels, demonstrating the full performance envelope of the pump.

Space or terrestrial energy?

NASA Astrophysics Data System (ADS)

Boulet, L.

Consideration is given to the possibility of generating sufficient energy at acceptable costs on earth to offset the need to build solar power satellite systems (SPS). Electricity usage, one of the basic driving forces of developed nations, grows with the population. Currently comprising 33 pct of the total world energy used, electricity is projected to grow to a 50-55 pct share in the 21st century. Future terrestrial electrical energy sources include carbon-based fuels, nuclear (fusion or fission), and the renewable solar technologies. Carbon-based fuel supplies can last until 2030 AD, about the same as fission plants with recycled fuel. Breeder reactors would stretch the nuclear fuels to the year 3000. Solar technologies offer more immediate solutions than fusion reactors and can produce 50 pct of the power available from the construction of the maximum number of nuclear power plants. The addition of SPS would further augment the total. Combinations of all the technologies are recommended, with local research for the most appropriate technology for each nation.

High Efficiency push-pull class E amplifiers for fusion rocket engines

NASA Astrophysics Data System (ADS)

Gaitan, Gabriel; Ham, Eric; Cohen, S. A.; Swanson, Charles; Chen, Minjie; Brunkhorst, Christopher

2017-10-01

In a Field Reversed Configuration fusion reactor, ions in the plasma are heated by an antenna operating at RF frequencies. This paper presents how push-pull class E amplifiers can be used to efficiently drive this antenna in the MHz range, from 0.5MHz to 4 MHz, while maintaining low harmonic content in the output signal. We offer four different configurations that present a trade-off between efficiency and low harmonic content. The paper presents theoretical values and breadboard results from these configurations, which operate at a power of around 100W. For a practical design, multiple amplifiers would be linked in parallel and would power the RF antenna at around 1MW. These designs provide multiple different options for reactor systems that could be used in a variety of applications, from power plants on the ground to rocket engines in space. This work was supported, in part, by DOE Contract Number DE-AC02-09CH11466 and Princeton Environmental Institute.

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

NASA Technical Reports Server (NTRS)

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

2012-01-01

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

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...

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.

Annual Report to Congress of the Atomic Energy Commission for 1965

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

Seaborg, Glenn T.

1966-01-31

The document represents the 1965 Annual Report of the Atomic Energy Commission (AEC) to Congress. The report opens with a Foreword - a letter from President Lyndon B. Johnson. The main portion is divided into 3 major sections for 1965, plus 10 appendices and the index. Section names and chapters are as follows. Part One reports on Developmental and Promotional Activities with the following chapters: (1) The Atomic Energy Program - 1965; (2) The Industrial Base ; (3) Industrial Relations; (4) Operational Safety; (5) Source and Special Nuclear Materials Production; (6) The Nuclear Defense Effort; (7) Civilian Nuclear Power; (8)more » Nuclear Space Applications; (9) Auxiliary Electrical Power for Land and Sea; (10) Military Reactors; (11) Advanced Reactor Technology and Nuclear Safety Research; (12) The Plowshare Program; (13) Isotopes and Radiation Development; (14) Facilities and Projects for Basic Research; (15) International Cooperation; and, (16) Nuclear Education and Information. Part Two reports on Regulatory Activities with the following chapters: (1) Licensing and Regulating the Atom; (2) Reactors and other Nuclear Facilities; and, (3) Control of Radioactive Materials. Part Three reports on Adjudicatory Activities.« less

Scaling of surface-plasma reactors with a significantly increased energy density for NO conversion.

PubMed

Malik, Muhammad Arif; Xiao, Shu; Schoenbach, Karl H

2012-03-30

Comparative studies revealed that surface plasmas developing along a solid-gas interface are significantly more effective and energy efficient for remediation of toxic pollutants in air than conventional plasmas propagating in air. Scaling of the surface plasma reactors to large volumes by operating them in parallel suffers from a serious problem of adverse effects of the space charges generated at the dielectric surfaces of the neighboring discharge chambers. This study revealed that a conductive foil on the cathode potential placed between the dielectric plates as a shield not only decoupled the discharges, but also increased the electrical power deposited in the reactor by a factor of about forty over the electrical power level obtained without shielding and without loss of efficiency for NO removal. The shield had no negative effect on efficiency, which is verified by the fact that the energy costs for 50% NO removal were about 60 eV/molecule and the energy constant, k(E), was about 0.02 L/J in both the shielded and unshielded cases. Copyright © 2012 Elsevier B.V. All rights reserved.

New 5 Kilowatt Free-Piston Stirling Space Converter Developments

NASA Astrophysics Data System (ADS)

Brandhorst, Henry W.

2007-01-01

NASA has recently funded development of a 5 kW (or greater) free-piston Stirling conversion system for reactor power systems. A nominal 5 kW converter allows two of these units to be dynamically balanced. A group of three dual-convertor combinations would yield the desired 30 kW. The status of this program will be presented. Goals include a specific power in excess of 140 W/kg at the converter level, lifetime in excess of five years and AC output. The initial step is the design and development of a nominal 5 kW per cylinder Stirling converter assembly (SCA) which will serve as a prototype of one or more SCAs that will make up the final 30 kW Stirling Converter Power System. Assumed requirements for this new converter for lunar fission power systems will be presented. The primary objective of this development effort will be to demonstrate a 5 kW SCA that can be tested to validate the viability of Stirling technology for space fission surface power systems.

New 5 Kilowatt Free-Piston Stirling Space Convertor Developments

NASA Technical Reports Server (NTRS)

Brandhorst, Henry W.

2007-01-01

NASA has recently funded development of a 5 kW (or greater) free-piston Stirling conversion system for reactor power systems. A nominal 5 kW convertor allows two of these units to be dynamically balanced. A group of three dual-convertor combinations would yield the desired 30 kW. The status of this program will be presented. Goals include a specific power in excess of 140 W/kg at the convertor level, lifetime in excess of five years and AC output. The initial step is the design and development of a nominal 5 kW per cylinder Stirling convertor assembly (SCA) which will serve as a prototype of one or more SCAs that will make up the final 30 kW Stirling Convertor Power System. Assumed requirements for this new convertor for lunar fission power systems will be presented. The primary objective of this development effort will be to demonstrate a 5 kW SCA that can be tested to validate the viability of Stirling technology for space fission surface power systems.

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.

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...

Power Systems for Future Missions: Appendices A-L

NASA Technical Reports Server (NTRS)

Gill, S. P.; Frye, P. E.; Littman, Franklin D.; Meisl, C. J.

1994-01-01

Selection of power system technology for space applications is typically based on mass, readiness of a particular technology to meet specific mission requirements, and life cycle costs (LCC). The LCC is typically used as a discriminator between competing technologies for a single mission application. All other future applications for a given technology are usually ignored. As a result, development cost of a technology becomes a dominant factor in the LCC comparison. Therefore, it is common for technologies such as DIPS and LMR-CBC to be potentially applicable to a wide range of missions and still lose out in the initial LCC comparison due to high development costs. This collection of appendices (A through L) contains the following power systems technology plans: CBC DIPS Technology Roadmap; PEM PFC Technology Roadmap; NAS Battery Technology Roadmap; PV/RFC Power System Technology Roadmap; PV/NAS Battery Technology Roadmap; Thermionic Reactor Power System Technology Roadmap; SP-100 Power System Technology Roadmap; Dynamic SP-100 Power System Technology Roadmap; Near-Term Solar Dynamic Power System Technology Roadmap; Advanced Solar Dynamic Power System Technology Roadmap; Advanced Stirling Cycle Dynamic Isotope Power System Technology Roadmap; and the ESPPRS (Evolutionary Space Power and Propulsion Requirements System) User's Guide.

Science in Flux: NASA's Nuclear Program at Plum Brook Station 1955-2005

NASA Technical Reports Server (NTRS)

Bowles, Mark D.

2006-01-01

Science in Flux traces the history of one of the most powerful nuclear test reactors in the United States and the only nuclear facility ever built by NASA. In the late 1950's NASA constructed Plum Brook Station on a vast tract of undeveloped land near Sandusky, Ohio. Once fully operational in 1963, it supported basic research for NASA's nuclear rocket program (NERVA). Plum Brook represents a significant, if largely forgotten, story of nuclear research, political change, and the professional culture of the scientists and engineers who devoted their lives to construct and operate the facility. In 1973, after only a decade of research, the government shut Plum Brook down before many of its experiments could be completed. Even the valiant attempt to redefine the reactor as an environmental analysis tool failed, and the facility went silent. The reactors lay in costly, but quiet standby for nearly a quarter-century before the Nuclear Regulatory Commission decided to decommission the reactors and clean up the site. The history of Plum Brook reveals the perils and potentials of that nuclear technology. As NASA, Congress, and space enthusiasts all begin looking once again at the nuclear option for sending humans to Mars, the echoes of Plum Brook's past will resonate with current policy and space initiatives.

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

NASA Astrophysics Data System (ADS)

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

2017-07-01

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

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 50.83 - 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 50.64 - 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...

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...

Economical Production of Pu-238: NIAC Phase I Final Report

NASA Technical Reports Server (NTRS)

Howe, Steven D.; Crawford, Douglas; Navarro, Jorge; O'Brien, Robert C.; Katalenich, Jeff; Ring, Terry

2016-01-01

All space exploration missions traveling beyond Jupiter must use radioisotopic power sources for electrical power. The best isotope to power these sources is plutonium-238 (Pu-238). The US supply of Pu-238 is almost exhausted and will be gone within the next decade. The Department of Energy has initiated a production program with a $10M allocation from NASA but the cost is estimated at over $100M to get to production levels. The Center for Space Nuclear Research (CSNR) has conceived of a potentially better process to produce Pu-238 earlier and for significantly less cost. Potentially, the front end capital costs could be provided by private industry such that the government only had to pay for the product produced. In the Phase I NIAC (NASA Innovative Advanced Concepts) grant, the CSNR has evaluated the feasibility of using a low power, commercially available nuclear reactor to produce 1.5 kg of Pu-238 per year. The impact on the neutronics of the reactor have been assessed, the amount of Neptunium target material estimated, and the production rates calculated. In addition, the size of the post-irradiation processing facility has been established. Finally, as the study progressed, a new method for fabricating the Pu-238 product into the form used for power sources has been identified to reduce the cost of the final product. In short, the concept appears to be viable, can produce the amount of Pu-238 needed to support the NASA missions, can be available within a few years, and will cost significantly less than the current DOE program.

Fusion energy from the Moon for the twenty-first century

NASA Technical Reports Server (NTRS)

Kulcinski, G. L.; Cameron, E. N.; Santarius, J. F.; Sviatoslavsky, I. N.; Wittenberg, L. J.; Schmitt, Harrison H.

1992-01-01

It is shown in this paper that the D-He-3 fusion fuel cycle is not only credible from a physics standpoint, but that its breakeven and ignition characteristics could be developed on roughly the same time schedule as the DT cycle. It was also shown that the extremely low fraction of power in neutrons, the lack of significant radioactivity in the reactants, and the potential for very high conversion efficiencies, can result in definite advantages for the D-He-3 cycle with respect to DT fusion and fission reactors in the twenty-first century. More specifically, the D-He-3 cycle can accomplish the following: (1) eliminate the need for deep geologic waste burial facilities and the wastes can qualify for Class A, near-surface land burial; (2) allow 'inherently safe' reactors to be built that, under the worst conceivable accident, cannot cause a civilian fatality or result in a significant (greater than 100 mrem) exposure to a member of the public; (3) reduce the radiation damage levels to a point where no scheduled replacement of reactor structural components is required, i.e., full reactor lifetimes (approximately 30 FPY) can be credibly claimed; (4) increase the reliability and availability of fusion reactors compared to DT systems because of the greatly reduced radioactivity, the low neutron damage, and the elimination of T breeding; and (5) greatly reduce the capital costs of fusion power plants (compared to DT systems) by as much as 50 percent and present the potential for a significant reduction on the COE. The concepts presented in this paper tie together two of the most ambitious high-technology endeavors of the twentieth century: the development of controlled thermonuclear fusion for civilian power applications and the utilization of outer space for the benefit of mankind on Earth.

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

NASA Technical Reports Server (NTRS)

Mireles, Omar R.; Houts, Michael G.

2011-01-01

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

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

NASA Astrophysics Data System (ADS)

Horak, James A.; Egner, Larry K.

1994-07-01

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

Fusion energy from the Moon for the twenty-first century

NASA Astrophysics Data System (ADS)

Kulcinski, G. L.; Cameron, E. N.; Santarius, J. F.; Sviatoslavsky, I. N.; Wittenberg, L. J.; Schmitt, Harrison H.

1992-09-01

It is shown in this paper that the D-He-3 fusion fuel cycle is not only credible from a physics standpoint, but that its breakeven and ignition characteristics could be developed on roughly the same time schedule as the DT cycle. It was also shown that the extremely low fraction of power in neutrons, the lack of significant radioactivity in the reactants, and the potential for very high conversion efficiencies, can result in definite advantages for the D-He-3 cycle with respect to DT fusion and fission reactors in the twenty-first century. More specifically, the D-He-3 cycle can accomplish the following: (1) eliminate the need for deep geologic waste burial facilities and the wastes can qualify for Class A, near-surface land burial; (2) allow 'inherently safe' reactors to be built that, under the worst conceivable accident, cannot cause a civilian fatality or result in a significant (greater than 100 mrem) exposure to a member of the public; (3) reduce the radiation damage levels to a point where no scheduled replacement of reactor structural components is required, i.e., full reactor lifetimes (approximately 30 FPY) can be credibly claimed; (4) increase the reliability and availability of fusion reactors compared to DT systems because of the greatly reduced radioactivity, the low neutron damage, and the elimination of T breeding; and (5) greatly reduce the capital costs of fusion power plants (compared to DT systems) by as much as 50 percent and present the potential for a significant reduction on the COE. The concepts presented in this paper tie together two of the most ambitious high-technology endeavors of the twentieth century: the development of controlled thermonuclear fusion for civilian power applications and the utilization of outer space for the benefit of mankind on Earth.

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.

10 CFR 2.1115 - Designation of issues for adjudicatory hearing.

Code of Federal Regulations, 2010 CFR

2010-01-01

... at Civilian Nuclear Power Reactors § 2.1115 Designation of issues for adjudicatory hearing. (a) After... reactor already licensed to operate at the site, or any civilian nuclear power reactor for which a... the issuance of a construction permit or operating license for a civilian nuclear power reactor at...

Kilowatt-Class Fission Power Systems for Science and Human Precursor Missions

NASA Technical Reports Server (NTRS)

Mason, Lee S.; Gibson, Marc Andrew; Poston, Dave

2013-01-01

Nuclear power provides an enabling capability for NASA missions that might otherwise be constrained by power availability, mission duration, or operational robustness. NASA and the Department of Energy (DOE) are developing fission power technology to serve a wide range of future space uses. Advantages include lower mass, longer life, and greater mission flexibility than competing power system options. Kilowatt-class fission systems, designated "Kilopower," were conceived to address the need for systems to fill the gap above the current 100-W-class radioisotope power systems being developed for science missions and below the typical 100-k We-class reactor power systems being developed for human exploration missions. This paper reviews the current fission technology project and examines some Kilopower concepts that could be used to support future science missions or human precursors.

Kilowatt-Class Fission Power Systems for Science and Human Precursor Missions

NASA Technical Reports Server (NTRS)

Mason, Lee; Gibson, Marc; Poston, Dave

2013-01-01

Nuclear power provides an enabling capability for NASA missions that might otherwise be constrained by power availability, mission duration, or operational robustness. NASA and the Department of Energy (DOE) are developing fission power technology to serve a wide range of future space uses. Advantages include lower mass, longer life, and greater mission flexibility than competing power system options. Kilowatt-class fission systems, designated "Kilopower," were conceived to address the need for systems to fill the gap above the current 100-Wclass radioisotope power systems being developed for science missions and below the typical 100-kWe-class reactor power systems being developed for human exploration missions. This paper reviews the current fission technology project and examines some Kilopower concepts that could be used to support future science missions or human precursors.

The safety review and approval process for space nuclear power sources

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

Bennett, G.L.

1991-01-01

Over the past 30 yr. the U.S. Government has evolved a process for the safety review and launch approval of nuclear power sources (NPSs) proposed for launch into space. This process, which involves a number of governmental agencies, ensures that the various postulated accident scenarios are considered, that the responses of the NPSs to the accident environments are assessed, and that appropriate elements of the Federal Government are involved in the launch approval. This process has worked very well in the successful launches of 37 radioisotope thermoelectric generators and 1 reactor by the United States since 1961. Particular attention willmore » be focused on the recent launch of the Galileo spacecraft. 19 refs., 12 figs., 4 tabs.« less

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

NASA Technical Reports Server (NTRS)

Anghaie, S.; Chen, G.

1996-01-01

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

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 ENABLER - Based on proven NERVA technology

NASA Astrophysics Data System (ADS)

Livingston, Julie M.; Pierce, Bill L.

The ENABLER reactor for use in a nuclear thermal propulsion engine uses the technology developed in the NERVA/Rover program, updated to incorporate advances in the technology. Using composite fuel, higher power densities per fuel element, improved radiation resistant control components and the advancements in use of carbon-carbon materials; the ENABLER can provide a specific impulse of 925 seconds, an engine thrust to weight (excluding reactor shield) approaching five, an improved initial mass in low Earth orbit and a consequent reduction in launch costs and logistics problems. This paper describes the 75,000 lbs thrust ENABLER design which is a low cost, low risk approach to meeting tommorrow's space propulsion needs.

Reflector and Shield Material Properties for Project Prometheus

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

J. Nash

2005-11-02

This letter provides updated reflector and shield preliminary material property information to support reactor design efforts. The information provided herein supersedes the applicable portions of Revision 1 to the Space Power Program Preliminary Reactor Design Basis (Reference (a)). This letter partially answers the request in Reference (b) to provide unirradiated and irradiated material properties for beryllium, beryllium oxide, isotopically enriched boron carbide ({sup 11}B{sub 4}C) and lithium hydride. With the exception of {sup 11}B{sub 4}C, the information is provided in Attachments 1 and 2. At the time of issuance of this document, {sup 11}B{sub 4}C had not been studied.

Nuclear Theft: Real and Imagined Dangers

DTIC Science & Technology

1976-03-01

are utilized in connection with fossil fuel energy research and development programs and related activities conducted by the Bureau of Mines "energy... development associated with the U.S. nuclear weapons program . Addition- ally, ERDA conducts related programs which include power reactor design... development , nuclear propulsion, and other systems associated with space programs . The military and ERDA enjoy a symbiotic relationship in that nuclear

Shock and vibration tests of a SNAP-8 NaK pump

NASA Technical Reports Server (NTRS)

Stromquist, A. J.; Nelson, R. B.; Hibben, L.

1971-01-01

The pump used for reactor cooling in the SNAP 8 space power system was subjected to the expected vehicle launch vibration, and shock loading in accordance with the SNAP 8 environmental specification. Subsequent disassembly revealed damage to the thrust bearing pins, which should be redesigned and strengthened. The unit was operational, however, when run in a test loop after reassembly.

Overview of space power electronic's technology under the CSTI High Capacity Power Program

NASA Technical Reports Server (NTRS)

Schwarze, Gene E.

1994-01-01

The Civilian Space Technology Initiative (CSTI) is a NASA Program targeted at the development of specific technologies in the areas of transportation, operations and science. Each of these three areas consists of major elements and one of the operation's elements is the High Capacity Power element. The goal of this element is to develop the technology base needed to meet the long duration, high capacity power requirements for future NASA initiatives. The High Capacity Power element is broken down into several subelements that includes energy conversion in the areas of the free piston Stirling power converter and thermoelectrics, thermal management, power management, system diagnostics, and environmental compatibility and system's lifetime. A recent overview of the CSTI High capacity Power element and a description of each of the program's subelements is given by Winter (1989). The goals of the Power Management subelement are twofold. The first is to develop, test, and demonstrate high temperature, radiation-resistant power and control components and circuits that will be needed in the Power Conditioning, Control and Transmission (PCCT) subsystem of a space nuclear power system. The results obtained under this goal will also be applicable to the instrumentation and control subsystem of a space nuclear reactor. These components and circuits must perform reliably for lifetimes of 7-10 years. The second goal is to develop analytical models for use in computer simulations of candidate PCCT subsystems. Circuits which will be required for a specific PCCT subsystem will be designed and built to demonstrate their performance and, also, to validate the analytical models and simulations. The tasks under the Power Management subelement will now be described in terms of objectives, approach and present status of work.

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

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

John Darrell Bess

2008-06-01

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

A new safety channel based on ¹⁷N detection in research reactors.

PubMed

Seyfi, Somayye; Gharib, Morteza

2015-10-01

Tehran research reactor (TRR) is a representative of pool type research reactors using light water, as coolant and moderator. This reactor is chosen as a prototype to demonstrate and prove the feasibility of (17)N detection as a new redundant channel for reactor power measurement. In TRR, similar to other pool type reactors, neutron detectors are immersed in the pool around the core as the main power measuring devices. In the present article, a different approach, using out of water neutron detector, is employed to measure reactor power. This new method is based on (17)O (n,p) (17)N reaction taking place inside the core and subsequent measurement of delayed neutrons emitted due to (17)N disintegration. Count and measurement of neutrons around outlet water pipe provides a reliable redundant safety channel to measure reactor power. Results compared with other established channels indicate a good agreement and shows a linear interdependency with true thermal power. Safety of reactor operation is improved with installation & use of this new power measuring channel. The new approach may equally serve well as a redundant channel in all other types of reactors having coolant comprised of oxygen in its molecular constituents. Contrary to existing channels, this one is totally out of water and thus is an advantage over current instrumentations. It is proposed to employ the same idea on other reactors (nuclear power plants too) to improve safety criteria. Copyright © 2015 Elsevier Ltd. All rights reserved.

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

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

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

2005-02-06

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

Acceptability of reactors in space

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

Buden, D.

1981-04-01

Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it dies not appear that reactors add measurably to the risk associated with the Space Transportation System.

Experimental physics characteristics of a heavy-metal-reflected fast-spectrum critical assembly

NASA Technical Reports Server (NTRS)

Heneveld, W. H.; Paschall, R. K.; Springer, T. H.; Swanson, V. A.; Thiele, A. W.; Tuttle, R. J.

1972-01-01

A zero-power critical assembly was designed, constructed, and operated for the purpose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power electric conversion system. The range of the previous experimental investigations has been expanded to include the reactivity effects of:(1) surrounding the reactor with 15.24 cm (6 in.) of polyethylene, (2) reducing the heights of a portion of the upper and lower axial reflectors by factors of 2 and 4, (3) adding 45 kg of W to the core uniformly in two steps, (4) adding 9.54 kg of Ta to the core uniformly, and (5) inserting 2.3 kg of polyethylene into the core proper and determining the effect of a Ta addition on the polyethylene worth.

Non-Nuclear Testing of Compact Reactor Technologies at NASA MSFC

NASA Technical Reports Server (NTRS)

Houts, Michael G.; Pearson, J. Boise; Godfroy, Thomas J.

2011-01-01

Safe, reliable, compact, autonomous, long-life fission systems have numerous potential applications, both terrestrially and in space. Technologies and facilities developed in support of these systems could be useful to a variety of concepts. At moderate power levels, fission systems can be designed to operate for decades without the need for refueling. In addition, fast neutron damage to cladding and structural materials can be maintained at an acceptable level. Nuclear design codes have advanced to the stage where high confidence in the behavior and performance of a system can be achieved prior to initial testing. To help ensure reactor affordability, an optimal strategy must be devised for development and qualification. That strategy typically involves a combination of non-nuclear and nuclear testing. Non-nuclear testing is particularly useful for concepts in which nuclear operating characteristics are well understood and nuclear effects such as burnup and radiation damage are not likely to be significant. To be mass efficient, a SFPS must operate at higher coolant temperatures and use different types of power conversion than typical terrestrial reactors. The primary reason is the difficulty in rejecting excess heat to space. Although many options exist, NASA s current reference SFPS uses a fast spectrum, pumped-NaK cooled reactor coupled to a Stirling power conversion subsystem. The reference system uses technology with significant terrestrial heritage while still providing excellent performance. In addition, technologies from the SFPS system could be applicable to compact terrestrial systems. Recent non-nuclear testing at NASA s Early Flight Fission Test Facility (EFF-TF) has helped assess the viability of the reference SFPS and evaluate methods for system integration. In July, 2011 an Annular Linear Induction Pump (ALIP) provided by Idaho National Laboratory was tested at the EFF-TF to assess performance and verify suitability for use in a10 kWe technology demonstration unit (TDU). In November, 2011 testing of a 37-pin core simulator (designed in conjunction with Los Alamos National Laboratory) for use with the TDU will occur. Previous testing at the EFFTF has included the thermal and mechanical coupling of a pumped NaK loop to Stirling engines (provided by GRC). Testing related to heat pipe cooled systems, gas cooled systems, heat exchangers, and other technologies has also been performed. Integrated TDU testing will begin at GRC in 2013. Thermal simulators developed at the EFF-TF are capable of operating over the temperature and power range typically of interest to compact reactors. Small and large diameter simulators have been developed, and simulators (coupled with the facility) are able to closely match the axial and radial power profile of all potential systems of interest. A photograph of the TDU core simulator during assembly is provided in Figure 2.

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

NASA Technical Reports Server (NTRS)

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

1973-01-01

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

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...

Heat Pipe Space Nuclear Reactor Design Assessment. Volume 2. Feasibility Study of Upgrading the SP-100 Heat Pipe Space Nuclear Power System.

DTIC Science & Technology

1985-08-01

system thermal -to-electric energy conversion efficiency quite high (-20-25 percent). c. Fuel system--Figure 6 compares the fuel swelling of U02 , UN... high temperature reservoir. Figure 28 shows a schematic of an AMTEC operation. The efficiency of ANTEC is calculated by: IV IV + 1 (L + C AT) + Qloss... thermal energy to electrical energy. The high efficiency and low specific mass (for large systems) are the common advantages of these active systems. In

CHARACTERISTIC QUALITIES OF SOME ATOMIC POWER STATIONS (in Hungarian)

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

Ligeti, G.

1962-04-01

Mostly as the result of economic factors, the current rate of construction of public atomic power stations has slowed down. The use of atomic energy is considered economical only in a few special cases, such as ship propulsion or supplying power to remote regions. For this reason, many reactors were designed especially for the construction of such midget'' power stations, operating at power levels ranging from 10 to 70 Mw. Technical details are given of such already-built or proposed systems, including the following: pressurized- water reactors such as the Babcock and Wilcox 60-Mw reactor, using 2.4% U/sup 235/ fuel; themore » Humphrey-Glasow Company's 20 Mw reactor; the gascooled system of the de Havilland Company; the organicmoderated reactor of the English Electric Company; the organic-moderated system of the Hawker-Siddeley Nuclear Power Company; the boiling-water reactor of the Mitchell Engineering Company and the steam-cooled, heavy-water reactor of the Rolls-Royce & Vickers Company. (TTT)« less

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.

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.

Cryogenic Fluid Management Technology Development for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Taylor, B. D.; Caffrey, J.; Hedayat, A.; Stephens, J.; Polsgrove, R.

2015-01-01

Cryogenic fluid management technology is critical to the success of future nuclear thermal propulsion powered vehicles and long duration missions. This paper discusses current capabilities in key technologies and their development path. The thermal environment, complicated from the radiation escaping a reactor of a nuclear thermal propulsion system, is examined and analysis presented. The technology development path required for maintaining cryogenic propellants in this environment is reviewed. This paper is intended to encourage and bring attention to the cryogenic fluid management technologies needed to enable nuclear thermal propulsion powered deep space missions.

10 CFR 171.3 - Scope.

Code of Federal Regulations, 2010 CFR

2010-01-01

... holding an operating license for a power reactor, test reactor or research reactor issued under part 50 of... authorizes operation of a power reactor. The regulations in this part also apply to any person holding a...

Groundbreaking Ceremony at the NACA's Plum Brook Station

NASA Image and Video Library

1956-09-21

Addison Rothrock, the National Advisory Committee for Aeronautics’s (NACA) Assistant Director of Research, speaks at the groundbreaking ceremony for the Lewis Flight Propulsion Laboratory’s new test reactor at Plum Brook Station. This dedication event was held almost exactly one year after the NACA announced that it would build its $4.5 million nuclear reactor on 500 acres of the army’s 9000-acre Plum Brook Ordnance Works. The site was located in Sandusky, Ohio, approximately 60 miles west of the NACA Lewis laboratory in Cleveland. Lewis Director Raymond Sharp is seated to the left of Rothrock, Congressman Albert Baumhart and NACA Secretary John Victory are to the right. Many government and local officials were on hand for the press conference and ensuing luncheon. In the wake of World War II the military, the Atomic Energy Commission, and the NACA became interested in the use of atomic energy for propulsion and power. A Nuclear Division was established at NACA Lewis in the early 1950s. The division’s request for a 60-megawatt research reactor was approved in 1955. The semi-remote Plum Brook location was selected over 17 other possible sites. Construction of the Plum Brook Reactor Facility lasted five years. By the time of its first trial runs in 1961 the aircraft nuclear propulsion program had been cancelled. The space age had arrived, however, and the reactor would be used to study materials for a nuclear powered rocket.

An Update on the Status of the Supply of Plutonium-238 for Future NASA Missions

NASA Astrophysics Data System (ADS)

Wham, R. M.

2016-12-01

For more than five decades, Radioisotope Power Systems (RPSs) have enabled space missions to operate in locations where the Sun's intensity is too weak, obscured, or otherwise inadequate for solar power or other conventional power‒generation technologies. The natural decay heat (0.57 W/g) from the radioisotope, plutonium-238 (238Pu), provides the thermal energy source used by an RPS to generate electricity for operation of instrumentation, as well as heat to keep key subsystems warm for missions such as Voyagers 1 and 2, the Cassini mission to Saturn, the New Horizons flyby of Pluto, and the Mars Curiosity rover which were sponsored by the National Aeronautics and Space Administration (NASA). Plutonium-238 is produced by irradiation of neptunium-237 in a nuclear reactor a relatively high neutron flux. The United States has not produced new quantities of 238Pu since the early 1990s. RPS‒powered missions have continued since then using existing 238Pu inventory managed by the U.S. Department of Energy (DOE), including material purchased from Russia. A new domestic supply is needed to ensure the continued availability of RPSs for future NASA missions. NASA and DOE are currently executing a project to reestablish a 238Pu supply capability using its existing facilities and reactors, which are much smaller than the large-scale production reactors and processing canyon equipment used previously. The project is led by the Oak Ridge National Laboratory (ORNL). Target rods, containing NpO2, will be fabricated at ORNL and irradiated in the ORNL High Flux Isotope Reactor and the Advanced Test Reactor at Idaho National Laboratory. Irradiated targets will be processed in chemical separations at the ORNL Radiochemical Engineering Center to recover the plutonium product and unconverted neptunium for recycle. The 238PuO2 product will be shipped to Los Alamos National Laboratory for fabrication of heat source pellets. Key activities, such as transport of the neptunium to ORNL, irradiation of neptunium, and chemical processing to recover the newly generated 238Pu, have begun and have been demonstrated with the initial amounts (50-100 g) produced. Product samples have been shipped to LANL for evaluation, including chemical impurity analysis. This paper will provide an overview of the approach to the project and its progress to date.

Supercritical Brayton Cycle Nuclear Power System Concepts

NASA Astrophysics Data System (ADS)

Wright, Steven A.

2007-01-01

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

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

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

Bruce G. Schnitzler

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

Efficient solution of the simplified P N equations

DOE PAGES

Hamilton, Steven P.; Evans, Thomas M.

2014-12-23

We show new solver strategies for the multigroup SPN equations for nuclear reactor analysis. By forming the complete matrix over space, moments, and energy a robust set of solution strategies may be applied. Moreover, power iteration, shifted power iteration, Rayleigh quotient iteration, Arnoldi's method, and a generalized Davidson method, each using algebraic and physics-based multigrid preconditioners, have been compared on C5G7 MOX test problem as well as an operational PWR model. These results show that the most ecient approach is the generalized Davidson method, that is 30-40 times faster than traditional power iteration and 6-10 times faster than Arnoldi's method.

Nuclear electric power for multimegawatt orbit transfer vehicles

NASA Astrophysics Data System (ADS)

Casagrande, R. D.

Multimegawatt nuclear propulsion is an attractive option for orbit transfer vehicles. The masses of these platforms are expected to exceed the capability of a single launch from Earth necessitating assembly in space in a parking orbit. The OTV would transfer the platform from the parking orbit to the operational orbit and then return for the next mission. Electric propulsion is advantageous because of the high specific impulse achieved by the technology, 1000 to 5000 s and beyond, to reduce the propellant required. Nuclear power is attractive as the power system because of the weight savings over solar systems in the multimegawatt regime, and multimegawatts of power are required. A conceptual diagram is shown of an OTV with a command control module using electric thrusters powered from an SP-100 class nuclear reactor power system.

Neutron fluxes in test reactors

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

Youinou, Gilles Jean-Michel

Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

Influence of power supply on the generation of ozone and degradation of phenol in a surface discharge reactor

NASA Astrophysics Data System (ADS)

Zhao, Yan; Shang, Kefeng; Duan, Lijuan; Li, Yue; An, Jiutao; Zhang, Chunyang; Lu, Na; Li, Jie; Wu, Yan

2013-03-01

A surface Dielectric Barrier Discharge (DBD) reactor was utilized to degrade phenol in water. Different power supplies applied to the DBD reactor affect the discharge modes, the formation of chemically active species and thus the removal efficiency of pollutants. It is thus important to select an optimized power supply for the DBD reactor. In this paper, the influence of the types of power supplies including alternate current (AC) and bipolar pulsed power supply on the ozone generation in a surface discharge reactor was measured. It was found that compared with bipolar pulsed power supply, higher energy efficiency of O3 generation was obtained when DBD reactor was supplied with 50Hz AC power supply. The highest O3 generation was approximate 4 mg kJ-1 moreover, COD removal efficiency of phenol wastewater reached 52.3% after 3 h treatment under an AC peak voltage of 2.6 kV.

Phase 1 Space Fission Propulsion Energy Source Design

NASA Technical Reports Server (NTRS)

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

2002-01-01

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

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...

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...

Recent Advances in Power Conversion and Heat Rejection Technology for Fission Surface Power

NASA Technical Reports Server (NTRS)

Mason, Lee

2010-01-01

Under the Exploration Technology Development Program, the National Aeronautics and Space Administration (NASA) and the Department of Energy (DOE) are jointly developing Fission Surface Power (FSP) technology for possible use in human missions to the Moon and Mars. A preliminary reference concept was generated to guide FSP technology development. The concept consists of a liquid-metal-cooled reactor, Stirling power conversion, and water heat rejection, with Brayton power conversion as a backup option. The FSP project has begun risk reduction activities on some key components with the eventual goal of conducting an end-to-end, non-nuclear, integrated system test. Several power conversion and heat rejection hardware prototypes have been built and tested. These include multi-kilowatt Stirling and Brayton power conversion units, titanium-water heat pipes, and composite radiator panels.

Nuclear Electric Propulsion for Deep Space Exploration

NASA Astrophysics Data System (ADS)

Schmidt, G.

Nuclear electric propulsion (NEP) holds considerable promise for deep space exploration in the future. Research and development of this technology is a key element of NASA's Nuclear Systems Initiative (NSI), which is a top priority in the President's FY03 NASA budget. The goal is to develop the subsystem technologies that will enable application of NEP for missions to the outer planets and beyond by the beginning of next decade. The high-performance offered by nuclear-powered electric thrusters will benefit future missions by (1) reducing or eliminating the launch window constraints associated with complex planetary swingbys, (2) providing the capability to perform large spacecraft velocity changes in deep space, (3) increasing the fraction of vehicle mass allocated to payload and other spacecraft systems, and, (3) in some cases, reducing trip times over other propulsion alternatives. Furthermore, the nuclear energy source will provide a power-rich environment that can support more sophisticated science experiments and higher- speed broadband data transmission than current deep space missions. This paper addresses NASA's plans for NEP, and discusses the subsystem technologies (i.e., nuclear reactors, power conversion and electric thrusters) and system concepts being considered for the first generation of NEP vehicles.

Application of a Systems Engineering Approach to Support Space Reactor Development

NASA Astrophysics Data System (ADS)

Wold, Scott

2005-02-01

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

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

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

Afanas'ev, Yurii V; Zavestovskaya, I N; Zvorykin, V D

A review of reports made on the International Forum on Advanced High-Power Lasers and Applications, which was held at the beginning of November 1999 in Osaka (Japan), is presented. Five conferences were held during the forum on High-Power Laser Ablation, High-Power Lasers in Energy Engineering, High-Power Lasers in Civil Engineering and Architecture, High-Power Lasers in Manufacturing, and Advanced High-Power Lasers. The following trends in the field of high-power lasers and their applications were presented: laser fusion, laser applications in space, laser-triggered lightning, laser ablation of materials by short and ultrashort pulses, application of high-power lasers in manufacturing, application of high-powermore » lasers in mining, laser decommissioning and decontamination of nuclear reactors, high-power solid-state and gas lasers, x-ray and free-electron lasers. One can find complete information on the forum in SPIE, vols. 3885-3889. (chronicle)« less

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.

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

NASA Astrophysics Data System (ADS)

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

1991-01-01

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

Manned space flight nuclear system safety. Volume 4: Space shuttle nuclear system transportation. Part 1: Space shuttle nuclear safety

NASA Technical Reports Server (NTRS)

1972-01-01

An analysis of the nuclear safety aspects (design and operational considerations) in the transport of nuclear payloads to and from earth orbit by the space shuttle is presented. Three representative nuclear payloads used in the study were: (1) the zirconium hydride reactor Brayton power module, (2) the large isotope Brayton power system and (3) small isotopic heat sources which can be a part of an upper stage or part of a logistics module. Reference data on the space shuttle and nuclear payloads are presented in an appendix. Safety oriented design and operational requirements were identified to integrate the nuclear payloads in the shuttle mission. Contingency situations were discussed and operations and design features were recommended to minimize the nuclear hazards. The study indicates the safety, design and operational advantages in the use of a nuclear payload transfer module. The transfer module can provide many of the safety related support functions (blast and fragmentation protection, environmental control, payload ejection) minimizing the direct impact on the shuttle.

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

NEUTRONIC REACTOR

DOEpatents

Fraas, A.P.; Mills, C.B.

1961-11-21

A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

Using Additive Manufacturing to Optimize FLiBe Coolant Blanket in Fusion Reactors

NASA Astrophysics Data System (ADS)

Fry, Vincent Michael

Fusion reactors have often been hailed as the holy grail of clean energy generation, though a power-generating reactor has never been built due to a multitude of limiting factors. One such factor is the immense 12-15 MW/m2 heat fluxes experienced by the inner wall of the reactor. Multiple groups have proposed the use of tungsten swirl tubes to withstand the heat generated within the reactor core. The primary focus of this investigation is to parameterize this 'first wall' interior structure to determine the highest achievable heat transfer coefficient given the many tungsten configurations enabled via additive manufacturing. Two general tube structures were considered: an orthogonal three-dimensional mesh of various diameters and spacings, as well as a swirl tube geometry with varying 'tape' thicknesses. The coolant liquid proposed is FLiBe (2LiF-BeF2) due to its high specific heat capacity as well as its ability to breed tritium, the fuel for the reactor. This was accomplished using theoretical calculations; computational fluid dynamics and conjugate heat transfer simulations in ANSYS Workbench; as well as an experimental setup to confirm tube pressure drop along the pipe. It was determined that heat transfer coefficients between upwards of 60,000 W/m 2K were readily achievable, keeping the first wall temperature around 1300 K. A multitude of designs proved to be feasible given the pumping power restrictions, though the suggested design going forward is a swirl tube with 2 mm 'tape' thickness and 3 m/s inlet velocity. Simulated pressure drop with water was accurate to within 30% of experimentally measured values, giving confidence in the credibility of the results.

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.

The ENABLER—based on proven NERVA technology

NASA Astrophysics Data System (ADS)

Livingston, Julie M.; Pierce, Bill L.

1991-01-01

The ENABLER reactor for use in a nuclear thermal propulsion engine uses the technology developed in the NERVA/Rover program, updated to incorporate advances in the technology. Using composite fuel, higher power densities per fuel element, improved radiation resistant control components and the advancements in use of carbon-carbon materials; the ENABLER can provide a specific impulse of 925 seconds, an engine thrust to weight (excluding reactor shield) approaching five, an improved initial Mass In Low Earth Orbit (IMLEO) and a consequent reduction in launch costs and logistics problems. This paper describes the 75,000 lbs thrust ENABLER design which is a low cost, low risk approach to meeting tomorrow's space propulsion needs.

Optical gamma thermometer

DOEpatents

Koster, Glen Peter; Xia, Hua; Lee, Boon Kwee

2013-08-06

An optical gamma thermometer includes a metal mass having a temperature proportional to a gamma flux within a core of a nuclear reactor, and an optical fiber cable for measuring the temperature of the heated metal mass. The temperature of the heated mass may be measured by using one or more fiber grating structures and/or by using scattering techniques, such as Raman, Brillouin, and the like. The optical gamma thermometer may be used in conjunction with a conventional reactor heat balance to calibrate the local power range monitors over their useful in-service life. The optical gamma thermometer occupies much less space within the in-core instrument tube and costs much less than the conventional gamma thermometer.

10 CFR 2.1105 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors § 2.1105 Definitions. As used in this part: (a) Civilian nuclear power reactor means a civilian nuclear power plant required to be licensed... nuclear fuel means fuel that has been withdrawn from a nuclear reactor following irradiation, the...

10 CFR 2.1105 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors § 2.1105 Definitions. As used in this part: (a) Civilian nuclear power reactor means a civilian nuclear power plant required to be licensed... nuclear fuel means fuel that has been withdrawn from a nuclear reactor following irradiation, the...

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.

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

NASA Astrophysics Data System (ADS)

Permana, Sidik

2017-07-01

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

TREAT Neutronics Analysis of Water-Loop Concept Accommodating LWR 9-rod Bundle

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

Hill, Connie M.; Woolstenhulme, Nicolas E.; Parry, James R.

Abstract. Simulation of a variety of transient conditions has been successfully achieved in the Transient Reactor Test (TREAT) facility during operation between 1959 and 1994 to support characterization and safety analysis of nuclear fuels and materials. A majority of previously conducted tests were focused on supporting sodium-cooled fast reactor (SFR) designs. Experiments evolved in complexity. Simulation of thermal-hydraulic conditions expected to be encountered by fuels and materials in a reactor environment was realized in the development of TREAT sodium loop experiment vehicles. These loops accommodated up to 7-pin fuel bundles and served to simulate more closely the reactor environment whilemore » safely delivering large quantities of energy into the test specimen. Some of the immediate TREAT restart operations will be focused on testing light water reactor (LWR) accident tolerant fuels (ATF). Similar to the sodium loop objectives, a water loop concept, developed and analyzed in the 1990’s, aimed at achieving thermal-hydraulic conditions encountered in commercial power reactors. The historic water loop concept has been analyzed in the context of a reactivity insertion accident (RIA) simulation for high burnup LWR 2-pin and 3-pin fuel bundles. Findings showed sufficient energy could be deposited into the specimens for evaluation. Similar results of experimental feasibility for the water loop concept (past and present) have recently been obtained using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. The old water loop concept required only two central TREAT core grid spaces. Preparation for future experiments has resulted in a modified water loop conceptual design designated the TREAT water environment recirculating loop (TWERL). The current TWERL design requires nine TREAT core grid spaces in order to place the water recirculating pump under the TREAT core. Due to the effectiveness of water moderation, neutronics analysis shows that removal of seven additional TREAT fuel elements to facilitate the experiment will not inhibit the ability to successfully simulate a RIA for the 2-pin or 3-pin bundle. This new water loop design leaves room for accommodating a larger fuel pin bundle than previously analyzed. The 7-pin fuel bundle in a hexagonal array with similar spacing of fuel pins in a SFR fuel assembly was considered the minimum needed for one central fuel pin to encounter the most correct thermal conditions. The 9-rod fuel bundle in a square array similar in spacing to pins in a LWR fuel assembly would be considered the LWR equivalent. MCNP analysis conducted on a preliminary LWR 9-rod bundle design shows that sufficient energy deposition into the central pin can be achieved well within range to investigate fuel and cladding performance in a simulated RIA. This is achieved by surrounding the flow channel with an additional annulus of water. Findings also show that a highly significant increase in TREAT to specimen power coupling factor (PCF) within the central pin can be achieved by surrounding the experiment with one to two rings of TREAT upgrade fuel assemblies. The experiment design holds promise for the performance evaluation of PWR fuel at extremely high burnup under similar reactor environment conditions.« less

Promethus Hot Leg Piping Concept

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

AM Girbik; PA Dilorenzo

2006-01-24

The Naval Reactors Prime Contractor Team (NRPCT) recommended the development of a gas cooled reactor directly coupled to a Brayton energy conversion system as the Space Nuclear Power Plant (SNPP) for NASA's Project Prometheus. The section of piping between the reactor outlet and turbine inlet, designated as the hot leg piping, required unique design features to allow the use of a nickel superalloy rather than a refractory metal as the pressure boundary. The NRPCT evaluated a variety of hot leg piping concepts for performance relative to SNPP system parameters, manufacturability, material considerations, and comparison to past high temperature gas reactormore » (HTGR) practice. Manufacturability challenges and the impact of pressure drop and turbine entrance temperature reduction on cycle efficiency were discriminators between the piping concepts. This paper summarizes the NRPCT hot leg piping evaluation, presents the concept recommended, and summarizes developmental issues for the recommended concept.« less

TOPAZ II Anti-Criticality Device Rapid Prototype

NASA Astrophysics Data System (ADS)

Campbell, Donald R.; Otting, William D.

1994-07-01

The Ballistic Missile Defense Organization (BMDO) has been working on a Nuclear Electric Propulsion Space Test Project (NEPSTP) using an existing Russian Topaz II reactor system to power the NEPSTP satellite. Safety investigations have shown that it will be possible to safely launch the Topaz II system in the United States with some modification to preclude water flooded criticality. A ``fuel-out'' water subcriticality concept was selected by the Los Alamos National Laboratory (LANL) as the baseline concept. A fuel-out anti-criticality device (ACD) conceptual design was developed by Rockwell. The concept functions to hold the fuel from the four centermost thermionic fuel elements (TFEs) outside the reactor during launch and reliably inserts the fuel into the reactor once the operational orbit is achieved. A four-tenths scale ACD rapid prototype model, fabricated from the CATIA solids design model, clearly shows in three dimensions the relative size and spatial relationship of the ACD components.

A review of carbide fuel corrosion for nuclear thermal propulsion applications

NASA Astrophysics Data System (ADS)

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

1993-10-01

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

A Review of Carbide Fuel Corrosion for Nuclear Thermal Propulsion Applications

NASA Astrophysics Data System (ADS)

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

1994-07-01

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