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Sample records for 100-kwe-class space reactor

  1. Design and Off-design Performance of 100 Kwe-class Brayton Power Conversion Systems

    NASA Technical Reports Server (NTRS)

    Johnson, Paul K.; Mason, Lee S.

    2005-01-01

    The NASA Glenn Research Center in-house computer model Closed Cycle Engine Program (CCEP) was used to explore the design trade space and off-design performance characteristics of 100 kWe-class recuperated Closed Brayton Cycle (CBC) power conversion systems. Input variables for a potential design point included the number of operating units (1, 2, 4), cycle peak pressure (0.5, 1, 2 MPa), and turbo-alternator shaft speed (30, 45, 60 kRPM). The design point analysis assumed a fixed turbine inlet temperature (1150 K), compressor inlet temperature (400 K), working-fluid molecular weight (40 g/mol), compressor pressure ratio (2.0), recuperator effectiveness (0.95), and a Sodium-Potassium (NaK) pumped-loop radiator. The design point options were compared on the basis of thermal input power, radiator area, and mass. For a nominal design point with defined Brayton components and radiator area, off-design cases were examined by reducing turbine inlet temperature (as low as 900 K), reducing shaft speed (as low as 50% of nominal), and circulating a percentage (up to 20%) of the compressor exit flow back to the gas cooler. The off-design examination sought approaches to reduce thermal input power without freezing the radiator.

  2. Space reactor shielding fabrication

    NASA Technical Reports Server (NTRS)

    Welch, F. H.

    1972-01-01

    The fabrication of space reactor neutron shielding by a melting and casting process utilizing lithium hydride is described. The first neutron shield fabricated is a large pancake shape 86 inches in diameter, containing about 1700 pounds of lithium hydride. This shield, fabricated by the unique melting and casting process, is the largest lithium hydride shield ever built.

  3. Acceptability of reactors in space

    SciTech Connect

    Buden, D.

    1981-01-01

    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 does not appear that reactors add measurably to the risk associated with the Space Transportation System.

  4. Acceptability of reactors in space

    SciTech Connect

    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.

  5. Unique features of space reactors

    SciTech Connect

    Buden, D.

    1990-01-01

    Space reactors are designed to meet a unique set of requirements; they must be sufficiently compact to be launched in a rocket to their operational location, operate for many years without maintenance and servicing, operate in extreme environments, and reject heat by radiation to space. To meet these restrictions, operating temperatures are much greater than in terrestrial power plants, and the reactors tend to have a fast neutron spectrum. Currently, a new generation of space reactor power plants is being developed. The major effort is in the SP-100 program, where the power plant is being designed for seven years of full power, and no maintenance operation at a reactor outlet operating temperature of 1350 K. 8 refs., 3 figs., 1 tab.

  6. SP-100 space reactor safety

    SciTech Connect

    Not Available

    1987-05-01

    The SP-100 space reactor power system is being developed to meet the large electrical power requirements of civilian and military missions planned for the 1990's and beyond. It will remove the restrictions on electrical power generation that have tended to limit missions and will enable the fuller exploration and utilization of space. This booklet describes the SP-100 space reactor power system and its development. Particular emphasis is given to safety. The design aand operational features as well as the design and safety review process that will assure that the SP-100 can be launched nd operated safely are described.

  7. Multimegawatt space power reactors

    SciTech Connect

    Dearien, J.A.; Whitbeck, J.F.

    1989-01-01

    In response to the need of the Strategic Defense Initiative (SDI) and long range space exploration and extra-terrestrial basing by the National Air and Space Administration (NASA), concepts for nuclear power systems in the multi-megawatt levels are being designed and evaluated. The requirements for these power systems are being driven primarily by the need to minimize weight and maximize safety and reliability. This paper will discuss the present requirements for space based advanced power systems, technological issues associated with the development of these advanced nuclear power systems, and some of the concepts proposed for generating large amounts of power in space. 31 figs.

  8. Small Reactor for Deep Space Exploration

    ScienceCinema

    none,

    2016-07-12

    This is the first demonstration of a space nuclear reactor system to produce electricity in the United States since 1965, and an experiment demonstrated the first use of a heat pipe to cool a small nuclear reactor and then harvest the heat to power a Stirling engine at the Nevada National Security Site's Device Assembly Facility confirms basic nuclear reactor physics and heat transfer for a simple, reliable space power system.

  9. Small Reactor for Deep Space Exploration

    SciTech Connect

    none,

    2012-11-29

    This is the first demonstration of a space nuclear reactor system to produce electricity in the United States since 1965, and an experiment demonstrated the first use of a heat pipe to cool a small nuclear reactor and then harvest the heat to power a Stirling engine at the Nevada National Security Site's Device Assembly Facility confirms basic nuclear reactor physics and heat transfer for a simple, reliable space power system.

  10. Zirconium Hydride Space Power Reactor design.

    NASA Technical Reports Server (NTRS)

    Asquith, J. G.; Mason, D. G.; Stamp, S.

    1972-01-01

    The Zirconium Hydride Space Power Reactor being designed and fabricated at Atomics International is intended for a wide range of potential applications. Throughout the program a series of reactor designs have been evaluated to establish the unique requirements imposed by coupling with various power conversion systems and for specific applications. Current design and development emphasis is upon a 100 kilowatt thermal reactor for application in a 5 kwe thermoelectric space power generating system, which is scheduled to be fabricated and ground tested in the mid 70s. The reactor design considerations reviewed in this paper will be discussed in the context of this 100 kwt reactor and a 300 kwt reactor previously designed for larger power demand applications.

  11. Autonomous Control of Space Nuclear Reactors

    NASA Technical Reports Server (NTRS)

    Merk, John

    2013-01-01

    Nuclear reactors to support future robotic and manned missions impose new and innovative technological requirements for their control and protection instrumentation. Long-duration surface missions necessitate reliable autonomous operation, and manned missions impose added requirements for failsafe reactor protection. There is a need for an advanced instrumentation and control system for space-nuclear reactors that addresses both aspects of autonomous operation and safety. The Reactor Instrumentation and Control System (RICS) consists of two functionally independent systems: the Reactor Protection System (RPS) and the Supervision and Control System (SCS). Through these two systems, the RICS both supervises and controls a nuclear reactor during normal operational states, as well as monitors the operation of the reactor and, upon sensing a system anomaly, automatically takes the appropriate actions to prevent an unsafe or potentially unsafe condition from occurring. The RPS encompasses all electrical and mechanical devices and circuitry, from sensors to actuation device output terminals. The SCS contains a comprehensive data acquisition system to measure continuously different groups of variables consisting of primary measurement elements, transmitters, or conditioning modules. These reactor control variables can be categorized into two groups: those directly related to the behavior of the core (known as nuclear variables) and those related to secondary systems (known as process variables). Reliable closed-loop reactor control is achieved by processing the acquired variables and actuating the appropriate device drivers to maintain the reactor in a safe operating state. The SCS must prevent a deviation from the reactor nominal conditions by managing limitation functions in order to avoid RPS actions. The RICS has four identical redundancies that comply with physical separation, electrical isolation, and functional independence. This architecture complies with the

  12. Reactor safety for the Space Exploration Initiative

    NASA Technical Reports Server (NTRS)

    Dix, Terry E.

    1991-01-01

    A task force was created by the National Aeronautics and Space Administration to conduct a 90-day study to support efforts to determine requirements to meet the goals of the Space Exploration Initiative. The task force identified the need for a nuclear reactor to provide the electrical power required as the outpost power demands on the moon and Mars evolve into hundreds of kilowatts. A preliminary hazards analysis has been performed to examine safety aspects of nuclear reactor power systems for representative missions to the moon and Mars. Mission profiles were defined for reference lunar and Martian flights. Potential alternatives to each mission phase were also defined. Accident scenarios were qualitatively defined for the mission phases. The safety issues decay heat removal, reactor control, disposal, criticality, end-of-mission shutdown, radiation exposure, the Martian environment, high speed impact on the surfaces of the moon or Mars, and return flyby trajectories were identified.

  13. Static conversion systems. [for space power reactors

    NASA Technical Reports Server (NTRS)

    Ewell, R.; Mondt, J.

    1985-01-01

    Historically, all space power systems that have actually flown in space have relied on static energy conversion technology. Thus, static conversion is being considered for space nuclear power systems as well. There are four potential static conversion technologies which should be considered. These include: the alkali metal thermoelectric converter (AMTEC), the thermionic converter, the thermoelectric converter, and the thermophotovoltaic converter (TPV). These four conversion technologies will be described in brief detail along with their current status and development needs. In addition, the systems implications of using each of these conversion technologies with a space nuclear reactor power system will be evaluated and some comparisons made.

  14. Autonomous Control of Space Reactor Systems

    SciTech Connect

    Belle R. Upadhyaya; K. Zhao; S.R.P. Perillo; Xiaojia Xu; M.G. Na

    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.

  15. Shielding Analysis of a Small Compact Space Nuclear Reactor

    DTIC Science & Technology

    1987-08-01

    there will be a need for a power source that’s reliable, has a high pow,-- density, and, in some cases, portable. These reasons, and many more, make...Conventional power sources are not able to reasonably obtain these higher power levels. The Space Power 100 kWe (SP-100) space nuclear reactor is the...space nuclear reactors an attractive power source for future space missions. The idea of using space nuclear reactors in space is not new. The United

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

    NASA Technical Reports Server (NTRS)

    Lantz, E.; Mayo, W.

    1972-01-01

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

  17. Space reactor assessment and validation study

    NASA Technical Reports Server (NTRS)

    Gedeon, Stephen; Morey, Dennis

    1987-01-01

    The present difficulties experienced by the United States in launching payloads into space has suggested a number of problems which are associated with the handling of hazardous materials in spacecraft. The question has arisen as to the safety of launching highly radioactive material such as plutonium-238, related to the possibility of its dispersion into the atmosphere during a launch vehicle explosion. An alternative is the use of a small nuclear reactor which is not started until it is in space and contains little or no radioactivity at launch. A first order assessment of six small reactor concepts with power levels up to 100 MWe was performed. Both the nuclear feasibility of these concepts to operate at their rated power levels between 7 and 10 years and the capability of these concepts to remain subcritical both before and during launch and also in the case of water immersion during a potential launch failure or abort were investigated.

  18. Fast critical experiment data for space reactors

    SciTech Connect

    Collins, P.J.; McFarlane, H.F.; Olsen, D.N.; Atkinson, C.A.; Ross, J.R.

    1987-01-01

    Data from a number of previous critical experiments exist that are relevant to the design concepts being considered for SP-100 and MMW space reactors. Although substantial improvements in experiment techniques have since made some of the measured quantities somewhat suspect, the basic criticality data are still useful in most cases. However, the old experiments require recalculation with modern computational methods and nuclear cross section data before they can be applied to today's designs. Recently, we have calculated about 20 fast benchmark critical experiments with the latest ENDF/B data and modern transport codes. These calculations were undertaken as a part of the planning process for a new series of benchmark experiments aimed at supporting preliminary designs of SP-100 and MMW space reactors.

  19. 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.

  20. Space reactor safety, 1985--1995 lessons learned

    SciTech Connect

    Marshall, A.C.

    1995-12-31

    Space reactor safety activities and decisions have evolved over the last decade. Important safety decisions have been made in the SP-100, Space Exploration Initiative, NEPSTP, SNTP, and Bimodal Space Reactor programs. In addition, international guidance on space reactor safety has been instituted. Space reactor safety decisions and practices have developed in the areas of inadvertent criticality, reentry, radiological release, orbital operation, programmatic, and policy. In general, the lessons learned point out the importance of carefully reviewing previous safety practices for appropriateness to space nuclear programs in general and to the specific mission under consideration.

  1. Space reactor safety, 1985-1995 lessons learned

    NASA Astrophysics Data System (ADS)

    Marshall, Albert C.

    1996-03-01

    Space reactor safety activities and decisions have evolved over the last decade. Important safety decisions have been made in the SP-100, Space Exploration Initiative, NEPSTP, SNTP, and Bimodal Space Reactor programs. In addition, international guidance on space reactor safety has been instituted. Space reactor safety decisions and practices have developed in the areas of inadvertent criticality, reentry, radiological release, orbital operation, programmatics, and policy. In general, the lessons learned point out the importance of carefully reviewing previous safety practices for appropriateness to space nuclear programs in general and to the specific mission under consideration.

  2. Nuclear safety as applied to space power reactor systems

    SciTech Connect

    Cummings, G.E.

    1987-01-01

    Current space nuclear power reactor safety issues are discussed with respect to the unique characteristics of these reactors. An approach to achieving adequate safety and a perception of safety is outlined. This approach calls for a carefully conceived safety program which makes uses of lessons learned from previous terrestrial power reactor development programs. This approach includes use of risk analyses, passive safety design features, and analyses/experiments to understand and control off-design conditions. The point is made that some recent accidents concerning terrestrial power reactors do not imply that space power reactors cannot be operated safety.

  3. 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

    SciTech Connect

    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 philosophy 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.

  4. A Methodology for the Neutronics Design of Space Nuclear Reactors

    SciTech Connect

    King, Jeffrey C.; El-Genk, Mohamed S.

    2004-02-04

    A methodology for the neutronics design of space power reactors is presented. This methodology involves balancing the competing requirements of having sufficient excess reactivity for the desired lifetime, keeping the reactor subcritical at launch and during submersion accidents, and providing sufficient control over the lifetime of the reactor. These requirements are addressed by three reactivity values for a given reactor design: the excess reactivity at beginning of mission, the negative reactivity at shutdown, and the negative reactivity margin in submersion accidents. These reactivity values define the control worth and the safety worth in submersion accidents, used for evaluating the merit of a proposed reactor type and design. The Heat Pipe-Segmented Thermoelectric Module Converters space reactor core design is evaluated and modified based on the proposed methodology. The final reactor core design has sufficient excess reactivity for 10 years of nominal operation at 1.82 MW of fission power and is subcritical at launch and in all water submersion accidents.

  5. SP-100 Space Reactor Power System readiness

    NASA Astrophysics Data System (ADS)

    Josloff, A. T.; Matteo, D. N.; Bailey, H. S.

    The SP-100 Space Reactor Power System is being developed by GE, under contract to the U.S. Department of Energy, to provide electrical power in the range of 10's to 100's of kW. The system represents an enabling technology for a wide variety of earth orbital and interplanetary science missions, nuclear electric propulsion (NEP) stages, and lunar/Mars surface power for the Space Exploration Initiative (SEI). An effective infrastructure of Industry, National Laboratories and Government agencies has made substantial progress since the 1988 System Design Review. Hardware development and testing has progressed to the point of resolving all key technical feasibility issues. The technology and design is now at a state of readiness to support the definition of early flight demonstration missions. Of particular importance is that SP-100 meets the demanding U.S. safety, performance, reliability and life requirements. The system is scalable and flexible and can be configured to provide 10's to 100's of kWe without repeating development work and can meet DoD goals for an early, low-power demonstration flight in the 1996 - 1997 time frame.

  6. Assessment of nuclear reactor concepts for low power space applications

    NASA Technical Reports Server (NTRS)

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

    1988-01-01

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

  7. Space reactor ground tests, assessment of facility needs

    SciTech Connect

    Philbin, J.S.; Wemple, R.P.

    1985-01-01

    This paper discusses facility requirements for previously-specified, space-reactor ground tests of operating reactor prototypes. The paper also discusses engineering development tests applicable to fully-integrated space reactors and their subsystems. The development tests were derived by considering the functions and environments that the reactor might encounter in normal or abnormal launch-to-recovery sequences. Surety (safety, safeguards, and reliability) priorities will influence the final character of both the test program and the facilities requirements. Sandia facilities suitable for conducting the above tests are the main focus of this paper.

  8. 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.

  9. A Wide Range Neutron Detector for Space Nuclear Reactor Applications

    SciTech Connect

    Nassif, Eduardo; Sismonda, Miguel; Matatagui, Emilio; Pretorius, Stephan

    2007-01-30

    We propose here a versatile and innovative solution for monitoring and controlling a space-based nuclear reactor that is based on technology already proved in ground based reactors. A Wide Range Neutron Detector (WRND) allows for a reduction in the complexity of space based nuclear instrumentation and control systems. A ground model, predecessor of the proposed system, has been installed and is operating at the OPAL (Open Pool Advanced Light Water Research Reactor) in Australia, providing long term functional data. A space compatible Engineering Qualification Model of the WRND has been developed, manufactured and verified satisfactorily by analysis, and is currently under environmental testing.

  10. State space modeling of reactor core in a pressurized water reactor

    SciTech Connect

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W.; Shamsuddin, Mustaffa; Abdullah, M. Adib

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  11. Fission fragment assisted reactor concept for space propulsion: Foil reactor

    NASA Technical Reports Server (NTRS)

    Wright, Steven A.

    1991-01-01

    The concept is to fabricate a reactor using thin films or foils of uranium, uranium oxide and then to coat them on substrates. These coatings would be made so thin as to allow the escaping fission fragments to directly heat a hydrogen propellant. The idea was studied of direct gas heating and direct gas pumping in a nuclear pumped laser program. Fission fragments were used to pump lasers. In this concept two substrates are placed opposite each other. The internal faces are coated with thin foil of uranium oxide. A few of the advantages of this technology are listed. In general, however, it is felt that if one look at all solid core nuclear thermal rockets or nuclear thermal propulsion methods, one is going to find that they all pretty much look the same. It is felt that this reactor has higher potential reliability. It has low structural operating temperatures, very short burn times, with graceful failure modes, and it has reduced potential for energetic accidents. Going to a design like this would take the NTP community part way to some of the very advanced engine designs, such as the gas core reactor, but with reduced risk because of the much lower temperatures.

  12. Thermophotovoltaic Energy Conversion in Space Nuclear Reactor Power Systems

    DTIC Science & Technology

    2004-12-01

    Franklin Chang-Diaz of NASA Johnson’s Advanced Space Propulsion Laboratory led me pursue this topic when he asked about the best way to get megawatts of...wise to remember the words of ADM Hyman G. Rickover, the first Director of Naval Nuclear Propulsion . An academic reactor or reactor plant almost always

  13. FRINK - A Code to Evaluate Space Reactor Transients

    SciTech Connect

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

    2007-01-30

    One of the biggest needs for space reactor design and development is detailed system modeling. Most proposed space fission systems are very different from previously operated fission power systems, and extensive testing and modeling will be required to demonstrate integrated system performance. There are also some aspects of space reactors that make them unique from most terrestrial application, and require different modeling approaches. The Fission Reactor Integrated Nuclear Kinetics (FRINK) code was developed to evaluate simplified space reactor transients (note: the term ''space reactor'' inherently includes planetary and lunar surface reactors). FRINK is an integrated point kinetic/thermal-hydraulic transient analysis FORTRAN code - ''integrated'' refers to the simultaneous solution of the thermal and neutronic equations. In its current state FRINK is a very simple system model, perhaps better referred to as a reactor model. The ''system'' only extends to the primary loop power removal boundary condition; however this allows the simulation of simplified transients (e.g. loss of primary heat sink, loss of flow, large reactivity insertion, etc.), which are most important in bounding early system conceptual design. FRINK could then be added to a complete system model later in the design and development process as system design matures.

  14. Heat pipe reactors for space power applications

    NASA Technical Reports Server (NTRS)

    Koenig, D. R.; Ranken, W. A.; Salmi, E. W.

    1977-01-01

    A family of heat pipe reactors design concepts has been developed to provide heat to a variety of electrical conversion systems. Three power plants are described that span the power range 1-500 kWe and operate in the temperature range 1200-1700 K. The reactors are fast, compact, heat-pipe cooled, high-temperature nuclear reactors fueled with fully enriched refractory fuels, UC-ZrC or UO2. Each fuel element is cooled by an axially located molybdenum heat pipe containing either sodium or lithium vapor. Virtues of the reactor designs are the avoidance of single-point failure mechanisms, the relatively high operating temperature, and the expected long lifetimes of the fuel element components.

  15. 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.

  16. Space reactors - What is a kilogram

    SciTech Connect

    Buden, D.; Angelo, J.; Ek, D.; Voss, S.

    1984-08-01

    The use of nuclear electric propulsion can triple the payloads to GEO for a single Shuttle launch. Life orbits of 300 years can be used to allow most of the fission and activation products to decay before a reactor reenters the biosphere. Enough radioactive materials remain with very long lifetimes to make it desirable to design the reactor to disperse upon reentry and little additional risk to the biosphere is introduced by initiating NEP operations from 300 km.

  17. Space reactors: What is a kilogram

    SciTech Connect

    Buden, D.; Angelo, J. Jr.; Ek, D.; Voss, S.

    1984-01-01

    The use of nuclear electric propulsion can triple payloads to GEO for a single Shuttle launch. Life orbits of 300 years can be used to allow most of the fission and activation products to decay before a reactor reenters the biosphere. Enough radioactive materials remain with very long lifetimes to make it desirable to design the reactor to disperse upon reentry and little additional risk to the biosphere is introduced by initiating NEP operations from 300 km.

  18. FBR and RBR particle bed space reactors

    SciTech Connect

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

    1983-01-01

    Compact, high-performance nuclear reactor designs based on High-Temperature Gas Reactors (HTGRs) particulate fuel are investigated. The large surface area available with the small-diameter (approx. 500 microns) particulate fuel allows very high power densities (MW's/liter), small temperature differences between fuel and coolant (approx. 10/sup 0/K), high coolant-outlet temperatures (1500 to 3000/sup 0/K, depending on design), and fast reactor startup (approx. 2 to 3 seconds). Two reactor concepts are developed - the Fixed Bed Reactor (FBR), where the fuel particles are packed into a thin annular bed between two porous cylindrical drums, and the Rotating Bed Reactor (RBR), where the fuel particles are held inside a cold rotating (typically approx. 500 rpm) porous cylindrical drum. The FBR can operate steady-state in the closed-cycle He-cooled mode or in the open-cycle H/sub 2/-cooled mode. The RBR will operate only in the open-cycle H/sub 2/-cooled mode.

  19. SP-100 Program: space reactor system and subsystem investigations

    SciTech Connect

    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.

  20. 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.

  1. Sabatier Methanation Reactor for Space Exploration

    NASA Technical Reports Server (NTRS)

    Murdoch, Karen; Goldblatt, Loel; Carrasquillo, Robyn; Harris, Danny

    2005-01-01

    The Sabatier Methanation Reactor technology is of vital importance to the success of the human and robotic exploration program. In order to achieve an affordable program, the logistics supply to support the mission must be minimized to the fullest extent possible. One area of potential reduction with high return on investment is the closure of life support loops, particularly oxygen and water. The Sabatier system accomplishes this by utilizing hydrogen and carbon dioxide, waste products from the life support system, to produce water and methane. The recovered water is then recycled back into the life support system to provide oxygen; while the methane can be used for propulsion, or can be broken down further to recover the hydrogen. This technology is applicable not only to transit phases of exploration, but surface habitats as well as in-situ propellant production. The Sabatier Reactor system has been developed for ground based demonstration experiments extensively over the past 30 years. Over the past three years, NASA has funded development of the Sabatier Carbon Dioxide Reduction Assembly (CRA) for use on the ISS. Currently this system is at TRL 5 and it is expected that the system will be flown on the ISS as a flight experiment, The purpose of the flight experiment is to integrate the Sabatier CRA into a synchronized system with the oxygen generation system and the carbon dioxide concentrator. The flight experiment will verify the integration of the different systems working together plus it will verify the capability of the system to operate, and effectively separate its products in a micro-gravity environment. Subsequent to design validation, the flight experiment can remain onboard the ISS providing valuable water to offset logistics re-supply requirements. Some of the challenges facing the development of the Sabatier system include handling vibration induced particulates, microgravity phase separation and containment of hazardous gases. Plans for

  2. Space reactor shielding: an assessment of the technology

    SciTech Connect

    Bartine, D.E.; Engle, W.W. Jr.

    1982-01-01

    Space power reactor systems require shielding to protect payload and reactor shielding components, and also maintenance and operating personnel. Shield composition, size, and shape are important design considerations, since the shield can dominate the overall weight of the system. Techniques for space reactor shield design analysis and optimization and experimental test facilities are available for design verification. With these tools, a shielding technology in support of current and future space power reactor systems can be developed. Efforts in this direction should begin with a generic shielding program to provide information on materials properties and geometric effects and should be followed by project-specific shielding programs to provide design optimization and prototype shield verification.

  3. Gas-cooled reactor for space power systems

    SciTech Connect

    Walter, C.E.; Pearson, J.S.

    1987-05-01

    Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors.

  4. SOAR: Space Orbiting Advanced Fusion Power Reactor.

    DTIC Science & Technology

    1987-09-01

    this are the possibility of electrostatic direct conversion and the ease of increasing magnetic fields to maintain plasma pressure . The gene- ral...106 0.2 29 290 Known Reserves 3 x 103 0.2 187 1870 MAN-MADE U.S. DOE MRC Sales 1.3/yr MRC Inventory 13.4 CANDU Reactors (by year 2000) Production 2/yr...Research Corp. currently sells up to 1.3 kg/yr and has a 13 kg inventory. The Canadian CANDU reactors produce tritium in their D20 moderators. By 1987, 5

  5. Shielding considerations for advanced space nuclear reactor systems

    NASA Astrophysics Data System (ADS)

    Angelo, J. P., Jr.; Buden, D.

    To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO2) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The status of this advanced heat pipe reactor is reviewed and the radiation environments and shielding requirements for representative manned and unmanned applications are explored.

  6. Shielding considerations for advanced space nuclear reactor systems

    SciTech Connect

    Angelo, J.P. Jr.; Buden, D.

    1982-01-01

    To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO/sub 2/) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications.

  7. Space station prototype Sabatier reactor design verification testing

    NASA Technical Reports Server (NTRS)

    Cusick, R. J.

    1974-01-01

    A six-man, flight prototype carbon dioxide reduction subsystem for the SSP ETC/LSS (Space Station Prototype Environmental/Thermal Control and Life Support System) was developed and fabricated for the NASA-Johnson Space Center between February 1971 and October 1973. Component design verification testing was conducted on the Sabatier reactor covering design and off-design conditions as part of this development program. The reactor was designed to convert a minimum of 98 per cent hydrogen to water and methane for both six-man and two-man reactant flow conditions. Important design features of the reactor and test conditions are described. Reactor test results are presented that show design goals were achieved and off-design performance was stable.

  8. Evaluating Russian space nuclear reactor technology for United States applications

    SciTech Connect

    Polansky, G.F.; Schmidt, G.L.; Voss, S.S.; Reynolds, E.L.

    1994-08-01

    Space nuclear power and nuclear electric propulsion are considered important technologies for planetary exploration, as well as selected earth orbit applications. The Nuclear Electric Propulsion Space Test Program (NEPSTP) was intended to provide an early flight demonstration of these technologies at relatively low cost through extensive use of existing Russian technology. The key element of Russian technology employed in the program was the Topaz II reactor. Refocusing of the activities of the Ballistic Missile Defense Organization (BMDO), combined with budgetary pressures, forced the cancellation of the NEPSTP at the end of the 1993 fiscal year. The NEPSTP was faced with many unique flight qualification issues. In general, the launch of a spacecraft employing a nuclear reactor power system complicates many spacecraft qualification activities. However, the NEPSTP activities were further complicated because the reactor power system was a Russian design. Therefore, this program considered not only the unique flight qualification issues associated with space nuclear power, but also with differences between Russian and United States flight qualification procedures. This paper presents an overview of the NEPSTP. The program goals, the proposed mission, the spacecraft, and the Topaz II space nuclear power system are described. The subject of flight qualification is examined and the inherent difficulties of qualifying a space reactor are described. The differences between United States and Russian flight qualification procedures are explored. A plan is then described that was developed to determine an appropriate flight qualification program for the Topaz II reactor to support a possible NEPSTP launch.

  9. Uranium ARC Fission Reactor for Space Power and Propulsion

    DTIC Science & Technology

    1992-03-01

    thruster or MHD accelerator/generator. Uranium arc technology is being developed for use in space nuclear thermal and electric propulsion reactors. In...specific impulse propulsion or ultrahigh temperature power conversion. Fission events in the nuclear arc plasma provide for additional dissociation and...I Technical Objectives 3 2. URANIUM ARC FISSION REACTOR CONCEPT AND NUCLEAR -AUGMENTED THRUSTER CONCEPT 4 2.1 Physics Basis 4 2.2 Uranium Arc

  10. Shield Design for a Space Based Vapor Core Reactor

    SciTech Connect

    Knight, Travis; Anghaie, Samim

    2002-07-01

    Innovative shielding strategies were sought to reduce the mass of the required shielding for a space based vapor core reactor system with magnetohydrodynamic energy conversion. Gamma-rays directly resultant from fission were found to play no role in the dose rate, while secondary gamma-rays from fission neutron interactions were the dominant contributor to the dose rate. Hydrogen containing materials such as polyethylene were utilized to provide shielding of both radiation from the reactor complex and also solar and galactic cosmic radiation. This shield design was found to contribute 0.125 kg/kWe to the baseline vapor core reactor system specific mass. (authors)

  11. Autonomous Control and Diagnostics of Space Reactor Systems

    SciTech Connect

    Upadhyaya, B.R.; Xu, X.; Perillo, S.R.P.; Na, M.G.

    2006-07-01

    This paper describes three key features of the development of an autonomous control strategy for space reactor systems. These include the development of a reactor simulation model for transient analysis, development of model-predictive control as part of the autonomous control strategy, and a fault detection and isolation module. The latter is interfaced with the control supervisor as part of a hierarchical control system. The approach has been applied to the nodal model of the SP-100 reactor with a thermo-electric generator. The results of application demonstrate the effectiveness of the control approach and its ability to reconfigure the control mode under fault conditions. (authors)

  12. 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.

  13. Space reactors. Progress report, January-March 1981

    SciTech Connect

    Ranken, W.A.

    1982-01-01

    Significant results have been obtained in Space Reactors design and experimental programs. Missions and applications for nuclear reactor space power supplies have been identified through the deliberations of the DOD/NASA/DOE Working Group. Differences in results obtained by two calculational techniques have been resolved for the reactor water immersion criticality studies. The in-pile test capsule is progressing, as is the fabrication of artery heat pipes. Corrosion studies continue on UO/sub 2//Na/Mo, TZM systems; one capsule has failed. Potassium/titanium heat pipe corrosion tests continue successfully. Thermophysical measurements of the GaP-bonded SiGe alloy used in the thermocouple work have shown a higher-than-expected figure of merit value.

  14. Flying Reactors: The Political Feasibility of Nuclear Power in Space

    DTIC Science & Technology

    2004-04-01

    accomplish is summarized well in the following quote from Glenn Seaborg , then the chairman of the Atomic Energy Commission. What we are attempting to...minutes.81 Glenn Seaborg AEC Chairman, 1958 Early Space Propulsion Reactor Programs Initial studies on nuclear rocket propulsion had begun as far back

  15. Gas-cooled reactor power systems for space

    SciTech Connect

    Walter, C.E.

    1987-01-01

    In this paper the characteristics of six designs for power levels of 2, 10, and 20 MWe for operating times of 1 and 7 y are described. The operating conditions for these arbitrary designs were chosen to minimize system specific mass. The designs are based on recent work which benefits from earlier analyses of nuclear space power systems conducted at our Laboratory. Both gas- and liquid-cooled reactors had been considered. Pitts and Walter (1970) reported on the results of a detailed study of a 10-MWe lithium-cooled reactor in a potassium Rankine system. Unpublished results (1966) of a computer analysis provide details of an argon-cooled reactor in an argon Brayton system. The gas-cooled reactor design was based on extensive development work on the 500-MWth reactor for the nuclear ramjet (Pluto) as described by Walter (1964). The designs discussed here draw heavily on the Pluto project experience, which culminated in a successful full-power ground test as reported by Reynolds (1964). At higher power levels gas-cooled reactors coupled with Brayton systems with advanced radiator designs become attractive.

  16. Reactivity Control Schemes for Fast Spectrum Space Nuclear Reactors

    SciTech Connect

    Craft, Aaron E.; King, Jeffrey C.

    2008-01-21

    Several different reactivity control schemes are considered for future space nuclear reactor power systems. Each of these control schemes uses a combination of boron carbide absorbers and/or beryllium oxide reflectors to achieve sufficient reactivity swing to keep the reactor subcritical during launch and to provide sufficient excess reactivity to operate the reactor over its expected 7-15 year lifetime. The size and shape of the control system directly impacts the size and mass of the space reactor's reflector and shadow shield, leading to a tradeoff between reactivity swing and total system mass. This paper presents a trade study of drum, shutter, and petal control schemes based on reactivity swing and mass effects for a representative fast-spectrum, gas-cooled reactor. For each control scheme, the dimensions and composition of the core are constant, and the reflector is sized to provide $5 of cold-clean excess reactivity with each configuration in its most reactive state. The advantages and disadvantages of each configuration are discussed, along with optimization techniques and novel geometric approaches for each scheme.

  17. Methods for Determining Operation Lifetime of Space Reactors

    NASA Astrophysics Data System (ADS)

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

    2007-01-01

    Space fission reactors can provide reliable, high power levels for periods of more than 10 years to support human outposts and for space travel to the farthest planets in the solar system. The operation lifetimes of these reactors depend on many factors chief among which are the hot-clean excess reactivity and the fuel burnup rate (or operation power) and the accumulation and decay of fission products. Other important parameters are the fuel average temperature and fissile inventory and the Doppler reactivity effect. Determining the operation lifetime for space reactors is a critical input to mission planning, requiring the use of sophisticated fuel burnup and criticality computational tools and benchmarking the results against actual data, if readily available. This paper performs parametric and comparative studies using widely used codes and a simplified approach for determining the operation lifetimes of two space reactors: the Sectored, Compact Reactor (SCoRe) that is liquid metal cooled, and the Submersion-Subcritical, Safe Space (S∧4) reactor that is cooled by a He-Xe binary gas mixture. The codes investigated against experimental data from a LWR are: (a) Monteburns 2.0, coupling MCNP5 1.30 to Origen2.2, (b) MCNPX 2.6b's internal burn package incorporating CINDER90, and (c) TRITON a code in the SCALE5 package using NEWT and Origen-S. From the results Monteburns and MCNPX performed the best, and are selected to compare their predictions of the lifetimes of the two space reactors with those of a simplified method. This method couples MCNP5 to a burnup analysis model in Simulink® considering only the 10 most probable low Z and 10 most probable high Z elements of the fission yield peaks plus 149Sm. Results show that the predicted operational lifetimes using the simplified method are within -6.6 to 12.8% of those calculated using the widely used Monteburns 2.0 and MCNPX 2.6bc1 codes.

  18. Methods for Determining Operation Lifetime of Space Reactors

    SciTech Connect

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

    2007-01-30

    Space fission reactors can provide reliable, high power levels for periods of more than 10 years to support human outposts and for space travel to the farthest planets in the solar system. The operation lifetimes of these reactors depend on many factors chief among which are the hot-clean excess reactivity and the fuel burnup rate (or operation power) and the accumulation and decay of fission products. Other important parameters are the fuel average temperature and fissile inventory and the Doppler reactivity effect. Determining the operation lifetime for space reactors is a critical input to mission planning, requiring the use of sophisticated fuel burnup and criticality computational tools and benchmarking the results against actual data, if readily available. This paper performs parametric and comparative studies using widely used codes and a simplified approach for determining the operation lifetimes of two space reactors: the Sectored, Compact Reactor (SCoRe) that is liquid metal cooled, and the Submersion-Subcritical, Safe Space (S and 4) reactor that is cooled by a He-Xe binary gas mixture. The codes investigated against experimental data from a LWR are: (a) Monteburns 2.0, coupling MCNP5 1.30 to Origen2.2, (b) MCNPX 2.6b's internal burn package incorporating CINDER90, and (c) TRITON a code in the SCALE5 package using NEWT and Origen-S. From the results Monteburns and MCNPX performed the best, and are selected to compare their predictions of the lifetimes of the two space reactors with those of a simplified method. This method couples MCNP5 to a burnup analysis model in Simulink registered considering only the 10 most probable low Z and 10 most probable high Z elements of the fission yield peaks plus 149Sm. Results show that the predicted operational lifetimes using the simplified method are within -6.6 to 12.8% of those calculated using the widely used Monteburns 2.0 and MCNPX 2.6bc1 codes.

  19. Space-reactor electric systems: subsystem technology assessment

    SciTech Connect

    Anderson, R.V.; Bost, D.; Determan, W.R.

    1983-03-29

    This report documents the subsystem technology assessment. For the purpose of this report, five subsystems were defined for a space reactor electric system, and the report is organized around these subsystems: reactor; shielding; primary heat transport; power conversion and processing; and heat rejection. The purpose of the assessment was to determine the current technology status and the technology potentials for different types of the five subsystems. The cost and schedule needed to develop these potentials were estimated, and sets of development-compatible subsystems were identified.

  20. Issues in the flight qualification of a space power reactor

    SciTech Connect

    Polansky, G.F.; Schmidt, G.L.; Voss, S.S.; Reynolds, E.L.

    1994-10-01

    This paper presents an overview of the Nuclear Electric Propulsion Space Test Program (NEPSTP). The program goals, the proposed mission, the spacecraft, and the Topaz II space nuclear power system are described. The subject of flight qualification is examined and the inherent difficulties of qualifying a space reactor are described. The differences between US and Russian flight qualification procedures are explored. A plan is then described that was developed to determine an appropriate flight qualification program for the Topaz II reactor to support a possible NEPSTP launch. Refocusing of the activities of the Ballistic Missile Defense Organization (BMDO), combined with budgetary pressures, forced the cancellation of the NEPSTP at the end of the 1993 fiscal year.

  1. Issues in the flight qualification of a space power reactor

    NASA Astrophysics Data System (ADS)

    Polansky, G. F.; Schmidt, G. L.; Voss, S. S.; Reynolds, E. L.

    1994-09-01

    This paper presents an overview of the Nuclear Electric Propulsion Space Test Program (NEPSTP). The program goals, the proposed mission, the spacecraft, and the Topaz II space nuclear power system are described. The subject of flight qualification is examined and the inherent difficulties of qualifying a space reactor are described. The differences between US and Russian flight qualification procedures are explored. A plan is then described that was developed to determine an appropriate flight qualification program for the Topaz II reactor to support a possible NEPSTP launch. Refocusing of the activities of the Ballistic Missile Defense Organization (BMDO), combined with budgetary pressures, forced the cancellation of the NEPSTP at the end of the 1993 fiscal year.

  2. Reactor Lithium Heat Pipes for HP-STMCs Space Reactor Power System

    NASA Astrophysics Data System (ADS)

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

    2004-02-01

    Design and performance analysis of the nuclear reactor's lithium heat pipes for a 110-kWe Heat Pipes-Segmented Thermoelectric Module Converters (HP-STMCs) Space Reactor Power system (SRPS) are presented. The evaporator length of the heat pipes is the same as the active core height (0.45 m) and the C-C finned condenser is of the same length as the STMC panels (1.5 m). The C-C finned condenser section is radiatively coupled to the collector shoes of the STMCs placed on both sides. The lengths of the adiabatic section, the values of the power throughput and the evaporator wall temperature depend on the radial location of the heat pipe in the reactor core and the number and dimensions of the potassium heat pipes in the heat rejection radiator. The reactor heat pipes have a total length that varies from 7.57 to 7.73 m, and a 0.2 mm thick Mo-14%Re wick with an average pore radius of 12 μm. The wick is separated from the Mo-14%Re wall by a 0.5 mm annulus filled with liquid lithium, to raise the prevailing capillary limit. The nominal evaporator (or reactor) temperature varies from 1513 to 1591 K and the thermal power of the reactor is 1.6 MW, which averages 12.7 kW for each of the 126 reactor heat pipes. The power throughput per heat pipe increase to a nominal 15.24 kW at the location of the peak power in the core and to 20.31 kW when an adjacent heat pipe fails. The prevailing capillary limit of the reactor heat pipes is 28.3 kW, providing a design margin >= 28%.

  3. Space reactors. Progress report, July-September 1981

    SciTech Connect

    Ranken, W.A.

    1982-06-01

    Progress in technology development for the Space Power Advanced Reactor (SPAR) project is reported for the period July 1-September 30, 1981. The weights of neutron and gamma shields required to protect the SPAR system payloads for a range of permissible exposures have been determined, and initial results are reported. SPAR reactor safety in the case of water immersion has been modeled. Approval-in-Principle has been received for the SPAR fuel test in the Experimental Breeder Reactor-II (EBR-II); the heat pipe developed for this test is performing well. SPAR system design variations are being examined under the possibility of using long core heat pipes. Testing of the initial molybdenum/sodium artery heat pipe continued, with ambiguous results. Fabrication of the first all-bonded thermoelectric units has been completed and testing has been initiated.

  4. Spectral Shift Absorbers for Fast Spectrum Space Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    King, Jeffrey C.; El-Genk, Mohamed S.

    2005-02-01

    The space nuclear reactors being considered to support the Jupiter Icy Moons Orbiter (JIMO-1) mission-1 sometime in the next decade are compact and fast spectrum with void fractions ranging from 20-40%. In order to secure launch approval, it has to be demonstrated that these reactors will remain sufficiently subcritical when submerged in water or wet sand and subsequently flooded with water, following a launch abort accident. The resulting shift in the neutron spectrum towards thermal increases reactivity, potentially making the reactors supercritical. Incorporating ``Spectral Shift Absorbers'' (or SSAs), elements such as boron, europium, gadolinium or rhenium, which have significantly higher absorption cross-sections for thermal versus fast neutrons, can offset the reactivity increase. It has always been the assumption that the worst-case submersion accident is with a fully flooded reactor; however, this work shows that, depending on the type and amount of SSA in the reactor, a submerged but unflooded reactor core could be more reactive, raising a safety concern. This condition, referred to as ``reactivity inversion'', is investigated for the following SSA elements: boron, boron-10, samarium-149, europium, europium-151, gadolinium, gadolinium-155, gadolinium-157 and rhenium, by varying the ratio of SSA to fissile atoms in the reactor. The effect of placing a coating of SSA material on the outside of the reactor vessel is also investigated. Gadolinium-157 is the most effective SSA material investigated as a core additive, yielding the greatest increase in the safety worth of a submerged space reactor with a comparatively small decrease in excess reactivity. Samarium-149 yields a similar increase in safety worth, but with a much larger decrease in excess reactivity. Natural europium and gadolinium are also promising: however, the decrease in the excess reactivity is greater with europium than with gadolinium. Boron and boron-10 are not particularly attractive as

  5. Space Reactor Launch Safety--An Acceptably Low Risk

    SciTech Connect

    Weitzberg, Abraham; Wright, Steven

    2008-01-21

    Results from previous space reactor and radioisotope power source risk assessments were combined to provide a scoping assessment of the possible risks from the launch of a reactor power system for use on the surface of the moon or Mars. It is assumed that future reactor power system launches would be subject to the same rigorous safety analysis and launch approval process as past nuclear payload launches. Using the same methodology that has gained approval of past launches, it was determined that the mission risk would be 0.029 person-rem worldwide which translates to 1.5*10{sup -5} latent health effects. It is seen that the only significant sources of radiological risks from a non-operating reactor are possible inadvertent criticality accidents and the consequences of such events have been shown to be extremely low. Passive means such as spectral shift poisons or high reactor core length/diameter ratios have been shown to be able to reduce or eliminate the possibility of the more credible criticality accidents, such as flooding or sand burial. This paper advances the premise that, for design purposes, future space reactor surface-power designs should primarily address the credible accidents and not the hypothetical accidents. For launch accidents and other safety assessments, a probabilistic risk assessment approach will have to be used to assess the safety impact of all types of accidents, including the hypothetical accidents. With this approach, the design of the system will not be burdened with design features that are based on hypothetical criticality accidents having negligible risk. Moreover, there is little chance of convincingly demonstrating that these design features can substantially reduce or eliminated the risk associated with hypothetical criticality accidents.

  6. Enabling autonomous control for space reactor power systems

    SciTech Connect

    Wood, R. T.

    2006-07-01

    The application of nuclear reactors for space power and/or propulsion presents some unique challenges regarding the operations and control of the power system. 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 space reactor power system (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. Thus, a SRPS control system must provide for operational autonomy. Oak Ridge National Laboratory (ORNL) has conducted an investigation of the state of the technology for autonomous control to determine the experience base in the nuclear power application domain, both for space and terrestrial use. It was found that control systems with varying levels of autonomy have been employed in robotic, transportation, spacecraft, and manufacturing applications. 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 basic control for a SRPS is clearly feasible under optimum circumstances. However, autonomous control is primarily intended to account for the non optimum circumstances when degradation, failure, and other off-normal events challenge the performance of the reactor and near-term human intervention is not possible. Thus, the development and demonstration of autonomous control capabilities for the specific domain of space nuclear power operations is needed. This paper will discuss the findings of the ORNL study and provide a description of the concept of autonomy, its key characteristics, and a prospective

  7. Space reactors. Progress report, April-June 1981

    SciTech Connect

    Ranken, W.A.

    1982-01-01

    Significant progress has been made in the Space Reactors design and experimental programs. Further analysis of the space reactor reactivity has shown that 2 to 3 kg of B/sub 4/C must be evenly distributed throughout the core to prevent criticality in the case of water immersion. A new core heat pipe processing, filling, and confirmatory testing system has significantly reduced the time required for these operations. The first dual-artery molybdenum/sodium heat pipe has been fabricated successfully. Fabrication of the in-pile test capsules and fuel tiles has begun. Experiments to develop methods of strengthening wicks and reducing wick pore size through chemical vapor deposition have yielded good results using tungsten. Thermoelectric couple redesign for fabrication and assembly methods prove-in has been completed, as has the test facility for these units.

  8. 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.

  9. The design of asymmetric 4 pi shields for space reactors

    NASA Technical Reports Server (NTRS)

    Engle, W. W., Jr.; Childs, R. L.; Mynatt, F. R.

    1972-01-01

    A one dimensional shield optimization program based on the method of discrete ordinates has been developed and is used to determine material thicknesses used in asymmetric 4 pion shields for space power reactors. The two dimensional discrete ordinates program DOT is used to check the design, and the information generated in the DOT calculation is used as a guide in shaping the shield which may be considered a first step in two dimensional shield optimization.

  10. Flying Reactors: The Political Feasibility of Nuclear Power in Space

    DTIC Science & Technology

    2005-04-01

    date , NASA has safely devel- oped, tested, and flown radioisotope power systems on 17 mis- sions and the United States successfully launched a nuclear...subject to a penalty for failing to comply with a collection of information if it does not display a currently valid OMB control number. 1. REPORT DATE ...APR 2005 2. REPORT TYPE 3. DATES COVERED - 4. TITLE AND SUBTITLE Flying Reactors. The Political Feasibility of Nuclear Power in Space 5a

  11. 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.

  12. Autonomous Control Capabilities for Space Reactor Power Systems

    SciTech Connect

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

    2004-02-04

    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.

  13. How Small Can Fast-Spectrum Space Reactors Get?

    NASA Astrophysics Data System (ADS)

    Hatton, Steven A.; El-Genk, Mohamed S.

    2006-01-01

    Fast neutron spectrum space reactors are an appropriate choice for high thermal powers, but for low powers, they may not satisfy the excess reactivity requirement while remaining sub-critical when immersed in wet sand and flooded with seawater following a launch abort accident. This paper identifies the smallest size fast spectrum, Sectored, Compact Reactor loaded with Single UN fuel pins (SCoRe-S7), which satisfy the requirements of cold clean excess reactivity > $4.00 and remains at least $1.00 subcritical at shutdown and in submersion conditions. Results indicate that increasing the diameter of the SCoRe-S core reduces its active height and the UN fuel enrichment, but increases the Spectrum-Shift Absorber (SSA) of 157GdN additive to the fuel. All SCoRe-S cores also have a 0.1 mm thick 157Gd2O3 SSA coating on the outer surface of the reactor vessel to reduce the effect of the wet sand reflector, while the SSA fuel additive reduces the effect on the criticality of the flooded reactor caused by thermal neutron fission. The active core height decreases from 42.4 cm for the smallest SCoRe-S7 to as much as to 37.4 cm for the largest core of SCoRe-S11. For a 1.8 MWth reactor thermal power the UN fuel specific power decreases from 17.0 in the SCoRe-S7 to 11.5 Wth/kg in the -S11. The corresponding reactor total mass, including the BeO reflector, increases from 440 kg to 512 kg.

  14. The MAUS nuclear space reactor with ion propulsion system

    NASA Astrophysics Data System (ADS)

    Mainardi, Enrico

    2006-06-01

    MAUS (Moltiplicatore Avanzato Ultracompatto Spaziale) is a nuclear reactor concept design capable to ensure a reliable, long-lasting, low-mass, compact energy supply needed for advanced, future space missions. The exploration of the solar system and the space beyond requires the development of nuclear energy generators for supplying electricity to space-bases, spacecrafts, probes or satellites, as well as for propelling ships in long space missions. For propulsion, the MAUS nuclear reactor could be used to power electric ion drive engines. An ion engine is able to build up to very high velocities, far greater than chemical propulsion systems, but has high power and long service requirements. The MAUS concept is described, together with the ion propulsion engine and together with the reference thermoionic process used to convert the thermal power into electricity. The design work has been performed at the Nuclear Engineering and Energy Conversion Department of the University of Rome "La Sapienza" starting from 1992 on an issue submitted by the Italian Space Agency (ASI), in cooperation with the research laboratories of ENEA.

  15. Utilizing a Russian space nuclear reactor for a United States space mission: Systems integration issues

    SciTech Connect

    Reynolds, E.; Schaefer, E.; Polansky, G.; Lacy, J.; Bocharov, A.

    1993-09-30

    The Nuclear Electric Propulsion Space Test Program (NEPSTP) has developed a cooperative relationship with several institutes of the former Soviet Union to evaluate Russian space hardware on a US spacecraft One component is the Topaz II Nuclear Power System; a built and flight qualified nuclear reactor that has yet to be tested in space. The access to the Topaz II reactor provides the NEPSTP with a rare opportunity; to conduct an early flight demonstration of nuclear electric propulsion at a relatively low cost. This opportunity, however, is not without challenges. Topaz II was designed to be compatible with Russian spacecraft and launch vehicles. It was manufactured and flight qualified by Russian techniques and standards and conforms to safety requirements of the former Soviet Union, not the United States. As it is desired to make minimal modifications to the Topaz II, integrating the reactor system with a United States spacecraft and launch vehicle presents an engineering challenge. This paper documents the lessons teamed regarding the integration of reactor based spacecraft and also some insight about integrating Russian hardware. It examines the planned integration flow along with specific reactor requirements that affect the spacecraft integration including American-Russian space system compatibility.

  16. Cermet-fueled reactors for advanced space applications

    SciTech Connect

    Cowan, C.L.; Palmer, R.S.; Taylor, I.N.; Vaidyanathan, S.; Bhattacharyya, S.K.; Barner, J.O.

    1987-12-01

    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 were 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.

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

    NASA Technical Reports Server (NTRS)

    Haley, F. A.

    1972-01-01

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

  18. Design and analysis of the SAFE-400 space fission reactor

    NASA Astrophysics Data System (ADS)

    Poston, David I.; Kapernick, Richard J.; Guffee, Ray M.

    2002-01-01

    Ambitious solar system exploration missions in the near future will require robust power sources in the range of 10 to 200 kWe. Fission systems are well suited to provide safe, reliable, and economic power within this range. 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 Safe Affordable Fission Engine (SAFE) is an HPS reactor designed for producing electricity in space. The SAFE-400 is a 400-kWt reactor that has been designed to couple with a 100-kWe Brayton power system. The SAFE-400 contains 127 identical molybdenum (Mo) modules. A Mo/sodium heatpipe is at the center of each module, surrounded by three Mo tubes that each contain a rhenium-clad uranium-nitride fuel pin. Fission energy is conducted from the fuel pins to the heatpipes, which then carry the heat to a heatpipe-to-gas heat exchanger. This paper describes the design and analysis of the current SAFE-400 reactor design. .

  19. Nuclear safety as applied to space power reactor systems

    SciTech Connect

    Cummings, G.E.

    1987-01-01

    To develop a strategy for incorporating and demonstrating safety, it is necessary to enumerate the unique aspects of space power reactor systems from a safety standpoint. These features must be differentiated from terrestrial nuclear power plants so that our experience can be applied properly. Some ideas can then be developed on how safe designs can be achieved so that they are safe and perceived to be safe by the public. These ideas include operating only after achieving a stable orbit, developing an inherently safe design, ''designing'' in safety from the start and managing the system development (design) so that it is perceived safe. These and other ideas are explored further in this paper.

  20. Recommendations for space reactor R&D tasks

    SciTech Connect

    Wiley, R.L.; Marshall, A.C.

    1995-12-01

    A rationale was developed to determine which technologies a space nuclear reactor technology based program pursue based on the fact that budgets would be limited. A preliminary evaluation was conducted to identify key technical issues and to recommend a prioritized set of candidate research projects that could be undertaken as part of the Defense Nuclear Agency (DNA) program in the near term. The recommendations made have not been adopted formally by the DNA`s Topaz International Program process. (TIP), but serve as inputs to the program plannin process.

  1. Space nuclear reactor shields for manned and unmanned applications

    NASA Technical Reports Server (NTRS)

    Mckissock, Barbara I.; Bloomfield, Harvey S.

    1990-01-01

    Missions which use nuclear reactor power systems require radiation shielding of payload and/or crew areas to predetermined dose rates. Since shielding can become a significant fraction of the total mass of the system, it is of interest to show the effect of various parameters on shield thickness and mass for manned and unmanned applications. Algorithms were developed to give the thicknesses needed if reactor thermal power, separation distances, and dose rates are given as input. The thickness algorithms were combined with models for four different shield geometries to allow tradeoff studies of shield volume and mass for a variety of manned and unmanned missions. Shield design tradeoffs presented in this study include the effects of: higher allowable dose rates; radiation hardened electronics; shorter crew exposure times; shield geometry; distance of the payload and/or crew from the reactor; and changes in the size of the shielded area. Specific NASA missions that were considered in this study include unmanned outer planetary exploration, manned advanced/evolutionary Space Station, and advanced manned lunar base.

  2. Space nuclear reactor shields for manned and unmanned applications

    NASA Technical Reports Server (NTRS)

    Mckissock, Barbara I.; Bloomfield, Harvey S.

    1989-01-01

    Missions which use nuclear reactor power systems require radiation shielding of payload and/or crew areas to predetermined dose rates. Since shielding can become a significant fraction of the total mass of the system, it is of interest to show the effect of various parameters on shield thickness and mass for manned and unmanned applications. Algorithms were developed to give the thicknesses needed if reactor thermal power, separation distances, and dose rates are given as input. The thickness algorithms were combined with models for four different shield geometries to allow tradeoff studies of shield volume and mass for a variety of manned and unmanned missions. Shield design tradeoffs presented in this study include the effects of: higher allowable dose rates; radiation hardened electronics; shorter crew exposure times; shield geometry; distance of the payload and/or crew from the reactor; and changes in the size of the shielded area. Specific NASA missions that were considered in this study include unmanned outer planetary exploration, manned advanced/evolutionary space station, and advanced manned lunar base.

  3. Space nuclear reactor shields for manned and unmanned applications

    SciTech Connect

    McKissock, B.I.; Bloomfield, H.S.

    1994-09-01

    Missions which use nuclear reactor power systems require radiation shielding of payload and/or crew areas to predetermined dose rates. Since shielding can become a significant fraction of the total mass of the system, it is of interest to show the effect of various parameters on shield thickness and mass for manned and unmanned applications. Algorithms were developed to give the thicknesses needed if reactor thermal power, separation distances, and dose rates are given as input. The thickness algorithms were combined with models for four different shield geometries to allow tradeoff studies of shield volume and mass for a variety of manned and unmanned missions. Shield design tradeoffs presented in this study include the effects of: Higher allowable dose rates; radiation hardened electronics; shorter crew exposure times; shield geometry; distance of the payload and/or crew from the reactor; and changes in the size of the shielded area. Specific NASA missions that were considered in this study include unmanned outer planetary exploration, manned advanced/evolutionary space station, and advanced manned lunar base.

  4. Space reactors. Progress report, October 1981-March 1982

    SciTech Connect

    Ranken, W.A.

    1983-01-01

    Progress in design studies and technology for the SP-100 Project - successor to the Space Power Advanced Reactor (SPAR) Project - is reported for the period October 1, 1981 to March 31, 1982. The basis for selecting a high-temperature, UO/sub 2/-fueled, heat-pipe-cooled reactor with a thermoelectric conversion system as the 100-kW/sub e/ reference design has been reviewed. Although no change has been made in the general concept, design studies have been done to investigate various reactor/conversion system coupling methods and core design modifications. Thermal and mechanical finite element modeling and three-dimensional Monte Carlo analysis of a core with individual finned fuel elements are reported. Studies of unrestrained fuel irradiation data are discussed that are relevant both to the core modeling work and to the design and fabrication of the first in-pile irradiation test, which is also reported. Work on lithium-filled core heat pipe development is described, including the attainment of 15.6 kW/sub t/ operation at 1525 K for a 2-m-long heat pipe with a 15.7-mm outside diameter. The successful operation of a 5.5-m-long, lightweight potassium/titanium heat pipe at 760 K is described, and test results of a thermoelectric module with GaP-modified SiGe thermoelectric elements are presented.

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

    SciTech Connect

    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 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.

  6. The liquid annular reactor system (LARS) for deep space exploration

    NASA Astrophysics Data System (ADS)

    Maise, George; Paniagua, John; Powell, James R.; Ludewig, Hans; Todosow, Michael

    1999-05-01

    A new propulsion concept for high Δ V space missions, termed LARS (Liquid Annular Reactor System), uses liquid nuclear fuel elements to heat hydrogen propellant to very high temperatures (-6000 K). The molten fuel is contained in a lower-temperature solid container which rotates to stabilize and hold in the liquid layer by centripetal force. Containment of ultra high temperature molten refractories, using this method, has been experimentally demonstrated by A.V. Grosse. The specific impulse of a rocket exhausting hydrogen at 6000 K is 2000 seconds, approximately double that of solid-core nuclear rockets. A LARS-powered space probe could accomplish extra-solar missions to 550 A.U. in approximately 35 years.

  7. The TOPAZ II space reactor response under accident conditions

    SciTech Connect

    Voss, S.S.

    1993-12-31

    The TOPAZ II is a single-cell thermionic space reactor power system developed by the Russians during the period of time from {approximately}1969 to 1989. The TOPAZ II has never been flight demonstrated, but the system was extensively tested on the ground. As part of the development and test program, the response of the TOPAZ II under accident conditions was analyzed and characterized. The US TOPAZ II team has been working closely with the Russian specialists to understand the TOPAZ II system, its operational characteristics, and its response under potential accident conditions. The purpose of the technical exchange is to enable a potential launch of a TOPAZ II by the US. The information is required to integrate the system with a US spacecraft and to support the safety review process. The purpose of this paper is to provide a brief overview of the system and its response under actual and postulated accident conditions.

  8. Heat Pipe Space Nuclear Reactor Design Assessment. Volume 1. Design Status of the SP-100 Heat Pipe Space Nuclear Reactor System

    DTIC Science & Technology

    1985-08-01

    design studies for a higher power , unmanned nuclear reactor space power source with a long design lifetime (7 yr). A 100 kWe , high temperature...reactors are potentially the best source of high power levels (in excess of 100 kWe ) for space. Other possible sources are chemical combustion...configured as a com- pact power source of greater than 10 kWe . Table 1 compares solar to nuclear power at three

  9. Development of a Robust Tri-Carbide Fueled Reactor for Multimegawatt Space Power and Propulsion Applications

    SciTech Connect

    Samim Anghaie; Travis W. Knight; Johann Plancher; Reza Gouw

    2004-08-11

    An innovative reactor core design based on advanced, mixed carbide fuels was analyzed for nuclear space power applications. Solid solution, mixed carbide fuels such as (U,Zr,Nb)c and (U,Zr, Ta)C offer great promise as an advanced high temperature fuel for space power reactors.

  10. A lithium-cooled reactor - Brayton turboelectric power converter design for 100-kWe class space reactor electric systems

    SciTech Connect

    Anderson, R.V.

    1984-08-01

    The conceptual design of a 100-kWe space reactor electric system to satisfy the design goals of the Tri-Agency SP-100 Program has been completed. The system was selected from an initial field of over 500 potential choices covering a wide range of reactor, power converter, shield, heat transport, and radiator subsystems. The selected system -- a lithium-cooled, UN-fueled, refractory-clad reactor coupled to a redundant pair of 110-kWe (gross) Brayton turboelectric power converters -shows strong promise of not only meeting the SP-100 Program design goals but also of providing for substantial growth in power levels for potential future needs.

  11. System simulation of a multicell thermionic space power reactor

    NASA Astrophysics Data System (ADS)

    von Arx, Alan Vincent

    For many years, thermionic power has been considered for space application. The prominent feature of the power conversion system is that there are no moving parts. Although designs have been developed by various organizations, no comprehensive system models are known to exist which can simulate transient behavior of a multicell design nor is there a method to directly couple these models to other codes that can calculate variations in reactivity. Thus, a procedure has been developed to couple the performance calculations of a space nuclear reactor thermal/hydraulics code with a neutron diffusion code to analyze temperature feedback. Thermionic power is based on the thermionic emissions principle where free electrons in a conductor have sufficient energy to escape the surface. Kinetic energy is given to the electrons by heating the conductor. Specifically, a 48 kWe thermionic power converter system model has been developed and used to model startup and other transients. Less than 10% of the fuel heat is converted to electricity, and the rest is rejected to space via a heat pipe radiator. An electromagnetic pump circulates the liquid metal coolant. First, a startup transient model was developed which showed stable operation through ignition of the Thermionic Fuel Elements (TFEs) and thawing of the radiator heat pipes. Also, the model's capability was expanded to include two-phase heat transfer to model boiling using coupled mass and thermal energy conservation equations. The next step incorporated effects of reactivity feedback---showing that various mechanisms will prevent power and temperature run-up for a flow reduction scenario where the reactor control systems fail to respond. In particular, the Doppler effect was shown to counter a positive worth due to partial core voiding although steps must be taken to preclude film boiling in that high superheats will result in TFE failures. Finally, analysis of the core grid spacer location suggests it should be located at

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

    NASA Technical Reports Server (NTRS)

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

    1985-01-01

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

  13. Transient modeling of the thermohydraulic behavior of high temperature heat pipes for space reactor applications

    NASA Technical Reports Server (NTRS)

    Hall, Michael L.; Doster, Joseph M.

    1986-01-01

    Many proposed space reactor designs employ heat pipes as a means of conveying heat. Previous researchers have been concerned with steady state operation, but the transient operation is of interest in space reactor applications due to the necessity of remote startup and shutdown. A model is being developed to study the dynamic behavior of high temperature heat pipes during startup, shutdown and normal operation under space environments. Model development and preliminary results for a hypothetical design of the system are presented.

  14. System aspects of a Space Nuclear Reactor Power System

    SciTech Connect

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

    1988-01-01

    Selected systems aspects of a 300 kW nuclear reactor power system for spacecraft have been studied. The approach included examination of two candidate missions and their associated spacecraft, and a number of special topics dealing with the power system design and operation. The missions considered were a reusable orbital transfer vehicle and a space-based radar. The special topics included: power system configuration and scaling, launch vehicle integration, operating altitude, orbital storage, start-up, thawing, control, load following, procedures in case of malfunction, restart, thermal and nuclear radiation to other portions of the spacecraft, thermal stresses between subsystems, boom and cable designs, vibration modes, altitude control, reliability, and survivability. Among the findings are 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. The power system design evolved during the study and has continued to evolve; the current design differs somewhat from that examined in this paper.

  15. Use of Beryllium and Beryllium Oxide in Space Reactors

    SciTech Connect

    Snead, L. L.; Zinkle, S. J.

    2005-02-06

    Beryllium and beryllium oxide are attractive candidate materials for neutron reflector application in space reactors due to their beneficial combination of low density and high neutron moderation and reflection capabilities. Drawbacks to their use include the expense of working with toxic materials, a limited industrial infrastructure, and material properties that are challenging in the non-irradiated state and seriously degrade under neutron irradiation. As an example of neutron effects, mechanical properties degrade under relevant conditions to the point where encasement in structural alloys is necessary. Such measures are required if neutron fluence exceeds {approx}1x1024 n/m2 (E>0.1 MeV). At high temperatures (>500 deg. C for Be and >600 deg. C for BeO), irradiation-induced swelling may also limit the maximum allowable dose without additional engineering measures. Significant volumetric swelling (>5%) can occur in these materials during neutron irradiation at elevated temperatures for neutron fluences above 1x1025 n/m2. This paper will review Be and BeO fabrication considerations, and summarize the effects of neutron irradiation on material properties.

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

    DTIC Science & Technology

    2007-11-02

    25 Schematic Diagram Of The " Topaz " Moderated TFE Space Reactor (Ref. 6)... 26 Schematic Diagram Of A Moderate TFE Reactor With Driver Fuel...reactor (Ref. 4), (b)a moderated TFE reactor ( TOPAZ , Ref. 5), and (c) a TFE reactor with SNAP driver fuel (Ref. 6). All of these concepts...or driver fuel permits coolant temperatures up to 1000 K or more. The TOPAZ reactor (Figure 2.5) incorporates a Zirconium- Hydride moderator

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

    NASA Technical Reports Server (NTRS)

    Davis, H. S.

    1984-01-01

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

  18. A long term radiological risk model for plutonium-fueled and fission reactor space nuclear system

    SciTech Connect

    Bartram, B.W.; Dougherty, D.K.

    1987-01-01

    This report describes the optimization of the RISK III mathematical model, which provides risk assessment for the use of a plutonium-fueled, fission reactor in space systems. The report discusses possible scenarios leading to radiation releases on the ground; distinctions are made for an intact reactor and a dispersed reactor. Also included are projected dose equivalents for various accident situations. 54 refs., 31 figs., 11 tabs. (TEM)

  19. Space reactor fuel element testing in upgraded TREAT

    SciTech Connect

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

    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. 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.

  20. Space reactor fuel element testing in upgraded TREAT

    SciTech Connect

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

    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. 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.

  1. Sufficient condition for velocity-space stability of the alpha particle distribution in a tokamak reactor

    SciTech Connect

    Cordey, J.G.; Goldston, R.J.; Mikkelsen, D.R.

    1980-08-01

    A condition is derived for the velocity-space distribution of suprathermal alpha particles to be monotonically decreasing with energy - and hence stable to homogeneous plasma instabilities - during the heating phase of a tokamak reactor. This stability condition is easily satisfied for presently envisaged neutral injection heating of reactors, but may be violated in strong heating of smaller plasmas or during fast compressional heating.

  2. A Comparison of Fast-Spectrum and Moderated Space Fission Reactors

    SciTech Connect

    Poston, David I.

    2005-02-06

    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.

  3. An assessment of space reactor technology needs and recommendations for development

    SciTech Connect

    Marshall, A.C.; Wiley, R.L.

    1995-11-01

    In order to provide a strategy for space reactor technology development, the Defense Nuclear Agency (DNA) has authorized a brief review of potential national needs that may be addressed by space reactor systems. a systematic approach was used to explore needs at several levels that are increasingly specific. Level 0 -- general trends and issues; Level 1 -- generic space capabilities to address trends; Level 2 -- requirements to support capabilities; Level 3 -- system types capable of meeting requirements; Level 4 --generic reactor system types; and Level 5 -- specific baseline systems. Using these findings, a strategy was developed to support important space reactor technologies within a limited budget. A preliminary evaluation identified key technical issues and provide a prioritized set of candidate research projects. The evaluation of issues and the recommended research projects are presented in a companion paper.

  4. Reactor Start-up and Control Methodologies: Consideration of the Space Radiation Environment

    NASA Astrophysics Data System (ADS)

    Bragg-Sitton, Shannon M.; Holloway, James Paul

    2004-02-01

    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 the accomplishment of ambitious space exploration missions. The natural radiation environment in space 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. Initial investigation using MCNPX 2.5.b for proton transport through the SAFE-400 reactor indicates a secondary neutron net current of 1.4×107 n/s at the core-reflector interface, with an incoming current of 3.4×106 n/s due to neutrons produced in the Be reflector alone. This neutron population could provide a reliable startup source for a space reactor. Additionally, this source must be considered in developing a reliable control strategy during reactor startup, steady-state operation, and power transients. An autonomous control system is developed and analyzed for application during reactor startup, accounting for fluctuations in the radiation environment that result from changes in vehicle location (altitude, latitude, position in solar system) or due to temporal variations in the radiation field, as may occur in the case of solar flares. One proposed application of a nuclear electric propulsion vehicle is in a tour of the Jovian system, where the time required for communication to Earth is significant. Hence, it is important that a reactor control system be designed with feedback mechanisms to automatically adjust to changes in reactor temperatures, power levels, etc., maintaining nominal operation without user intervention. This paper will evaluate the potential use of secondary neutrons produced by proton interactions in the reactor vessel as a startup source for a space reactor and will present

  5. Power monitoring in space nuclear reactors using silicon carbide radiation detectors

    NASA Technical Reports Server (NTRS)

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

    2005-01-01

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

  6. Small space reactor power systems for unmanned solar system exploration missions

    SciTech Connect

    Bloomfield, H.S.

    1987-12-01

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

  7. Small space reactor power systems for unmanned solar system exploration missions

    NASA Technical Reports Server (NTRS)

    Bloomfield, Harvey S.

    1987-01-01

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

  8. Effects of space-dependent cross sections on core physics parameters for compact fast spectrum space power reactors

    NASA Astrophysics Data System (ADS)

    Lell, R. M.; Hanan, N. A.

    Effects of multigroup neutron cross section generation procedures on core physics parameters for compact fast spectrum reactors were examined. Homogeneous and space dependent multigroup cross section set were generated in 11 and 27 groups for a representative fast reactor core. These cross sections were used to compute various reactor physics parameters for the reference core. Coarse group structure and neglect of space dependence in the generation procedure resulted in inaccurate computations of reactor flux and power distributions and in significant errors regarding estimates of core reactivity and control system worth. Delayed neutron fraction was insensitive to cross section treatment, and computed reactivity coefficients were only slightly sensitive. However, neutron lifetime was found to be very sensitive to cross section treatment. Deficiencies in multigroup cross sections are reflected in core nuclear design and, consequently, in system mechanical design.

  9. 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.

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

    SciTech Connect

    Carlson, W F; Bitten, E J

    1992-06-01

    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 feature 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.

  11. 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.

  12. The role of integral experiments and nuclear cross section evaluations in space nuclear reactor design

    NASA Astrophysics Data System (ADS)

    Moses, David L.; McKnight, Richard D.

    The importance of the nuclear and neutronic properties of candidate space reactor materials to the design process has been acknowledged as has been the use of benchmark reactor physics experiments to verify and qualify analytical tools used in design, safety, and performance evaluation. Since June 1966, the Cross Section Evaluation Working Group (CSEWG) has acted as an interagency forum for the assessment and evaluation of nuclear reaction data used in the nuclear design process. CSEWG data testing has involved the specification and calculation of benchmark experiments which are used widely for commercial reactor design and safety analysis. These benchmark experiments preceded the issuance of the industry standards for acceptance, but the benchmarks exceed the minimum acceptance criteria for such data. Thus, a starting place has been provided in assuring the accuracy and uncertainty of nuclear data important to space reactor applications.

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

    SciTech Connect

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

  14. 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.

  15. 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).

  16. 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.

  17. A computer program for engineering simulations of space reactor system performance

    SciTech Connect

    Dobranich, D.

    1992-07-01

    Nuclear thermal propulsion systems are envisioned as a fast and efficient form of transportation for the exploration of space. Several nuclear reactor concepts have been proposed. This document discusses SAFSIM (System Analysis Flow SIMulator) which is an engineering computer program that allows the fluid mechanic, heat transfer, and reactor dynamic simulation of the entire propulsion system. SAFSIM currently contains three basic physics modules: (1) fluid mechanics, (2) heat transfer, and (3) reactor dynamics. All three modules are coupled to allow the prediction of system performance. The analyst can employ any or all of the physics modules as the problem dictates.

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

    SciTech Connect

    Berg, Thomas A.; Disney, Richard K.

    2004-02-04

    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.

  19. 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.

  20. Analysis and evaluation of ZPPR (Zero Power Physics Reactor) critical experiments for a 100 kilowatt-electric space reactor

    SciTech Connect

    McFarlane, H.F.; Collins, P.J.; Carpenter, S.G.; Olsen, D.N.; Smith, D.M.; Schaefer, R.W. ); Doncals, R.A.; Andre, S.V.; Porter, C.A. ); Cowan, C.L; Stewart, S.L.; Protsik, R. . Astro Space Div.)

    1990-01-01

    ZPPR critical experiments were used for physics testing the reactor design of the SP-100, a 100-kW thermoelectric LMR that is being developed to provide electrical power for space applications. These tests validated all key physics characteristics of the design, including the ultimate safety in the event of a launch or re-entry accident. Both the experiments and the analysis required the use of techniques not previously applied to fast reactor designs. A few significant discrepancies between the experimental and calculated results leave opportunities for further optimization. An initial investigation has been made into application of the ZPPR-20 results, along with those of other relevant integral data, to the SP-100 design. 13 refs., 5 figs., 7 tabs.

  1. 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.

  2. 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.

  3. 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…

  4. Heat pipe cooled reactors for multi-kilowatt space power supplies

    NASA Astrophysics Data System (ADS)

    Ranken, W. A.; Houts, M. G.

    Three nuclear reactor space power system designs are described that demonstrate how the use of high temperature heat pipes for reactor heat transport, combined with direct conversion of heat to electricity, can result in eliminating pumped heat transport loops for both primary reactor cooling and heat rejection. The result is a significant reduction in system complexity that leads to very low mass systems with high reliability, especially in the power range of 1 to 20 kWe. In addition to removing heat exchangers, electromagnetic pumps, and coolant expansion chambers, the heat pipe/direct conversion combination provides such capabilities as startup from the frozen state, automatic rejection of reactor decay heat in the event of emergency or accidental reactor shutdown, and the elimination of single point failures in the reactor cooling system. The power system designs described include a thermoelectric system that can produce 1 to 2 kWe, a bimodal modification of this system to increase its power level to 5 kWe and incorporate high temperature hydrogen propulsion capability, and a moderated thermionic reactor concept with 5 to 20 kWe power output that is based on beryllium modules that thermally couple cylindrical thermionic fuel elements (TFE's) to radiator heat pipes.

  5. A spherical torus nuclear fusion reactor space propulsion vehicle concept for fast interplanetary travel

    NASA Astrophysics Data System (ADS)

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

    1999-01-01

    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>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 major systems including payload, central truss, nuclear reactor (including diverter 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, and component design.

  6. 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.

  7. Sliding mode control of the space nuclear reactor system TOPAZ II

    SciTech Connect

    Shtessel, Y.B.; Wyant, F.J.

    1996-03-01

    The Automatic Control System (ACS) of the space nuclear reactor power system TOPAZ II that generates electricity from nuclear heat using in-core thermionic converters is considered. Sliding Mode Control Technique was applied to the reactor system controller design in order to provide the robust high accuracy following of a neutron (thermal) power reference profile in a start up regime and a payload electric power (current) reference profile following in an operation regime. Extensive simulations of the TOPAZ II reactor system with the designed sliding mode controllers showed improved accuracy and robustness of the reactor system performances in a start up regime and in an electric power supply regime as well. {copyright} {ital 1996 American Institute of Physics.}

  8. 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.

  9. Thermionic reactor power system: Effects of radiation on integration with Manned Space Station

    NASA Technical Reports Server (NTRS)

    Gietzen, A. J.; Heath, C. A.; Perry, L. W.

    1972-01-01

    The application of a thermionic reactor power system to the modular space station is described. The nominal net power is 40 kWe, with the power system designed to be applicable over the power range from 25 to 60 kWe. The power system is designed to be launched by the space shuttle. Radiation protection is provided by LiH neutron shielding and W gamma shielding in a shaped 4 pion configuration, i.e., the reactor is shielded on all sides but not to equal extent. Isodose contours are presented for the region around the modular space station. Levels and spectral distribution of radiation are given for later evaluation of effects on space station experiments. Parametric data on the effects of separation distance on power system mass are presented.

  10. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  11. PBF Reactor Building (PER620). Piping in basement fills space. Secondary ...

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

    PBF Reactor Building (PER-620). Piping in basement fills space. Secondary coolant flowed through carbon steel pipe; primary coolant, through stainless steel. Photographer: Larry Page. Date: April 30, 1970. INEEL negative no. 70-2080 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  12. Space radiation studies at the White Sands Missile Range Fast Burst Reactor

    NASA Technical Reports Server (NTRS)

    Delapaz, A.

    1972-01-01

    The operation of the White Sands Missile Range Fast Burst Reactor is discussed. Space radiation studies in radiobiology, dosimetry, and transient radiation effects on electronic systems and components are described. Proposed modifications to increase the capability of the facility are discussed.

  13. Applying design principles to fusion reactor configurations for propulsion in space

    NASA Technical Reports Server (NTRS)

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

    1993-01-01

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

  14. SUSEE: A Compact, Lightweight Space Nuclear Power System Using Present Water Reactor Technology

    SciTech Connect

    Maise, George; Powell, James; Paniagua, John

    2006-01-20

    The SUSEE space reactor system uses existing nuclear fuels and the standard steam cycle to generate electrical and thermal power for a wide range of in-space and surface applications, including manned bases, sub-surface mobile probes to explore thick ice deposits on Mars and the Jovian moons, and mobile rovers. SUSEE cycle efficiency, thermal to electric, ranges from {approx}20 to 24%, depending on operating parameters. Rejection of waste heat is by a lightweight condensing radiator that can be launched as a compact rolled-up package and deployed into flat panels when appropriate. The 50 centimeter diameter SUSEE reactor can provide power over the range of 10 kW(e) to 1 MW(e) for a period of 10 years. Higher power outputs are possible using slightly larger reactors. System specific weight (reactor, turbine, generator, piping, and radiator) is {approx}3 kg/kW(e). Two SUSEE reactor options are described, based on the existing Zr/O2 cermet and the UH3/ZrH2 TRIGA nuclear fuels.

  15. Lightweight power bus for a baseload nuclear reactor in space

    NASA Astrophysics Data System (ADS)

    Massie, Lowell D.; Hoffman, Dennis J.; Oberly, Charles E.

    Metallic superconductors requiring refrigeration in the range of 4 to 10 K are of no benefit to the baseload space power system bus because the refrigeration and insulation constraints are too severe. The ceramic superconductors that can operate in the range of 20 to 100 K alleviate a great deal of the refrigeration problem and can compete with conventional hot bus distribution systems on the basis of mass for a bus exceeding a few meters in length. The ultimate benefit of the superconducting bus to the space power system will not be the mass savings. The great benefit of the superconducting bus will be the enormous reduction in bus voltage requirements due to the zero voltage drop along the bus. Low bus voltage (less than 100 Vdc) will permit a conventional dielectric insulation technology to be utilized as baseload powers are forced above 10 kW on spacecraft.

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

    DTIC Science & Technology

    2007-11-02

    which of their combinations would lead to the optimal compact reactor configuration. These parameters and materials are shown in Table 2.1. Three...TRIGA, EBR-2 and FFTF reactors. The life of these TFE elements has already been proven to be in excess of three years. The irradiation of the...the Lawrence Livermore Laboratory. The spectrum used in generating these cross-sections was that of the Godiva critical assembly (a 93 percent

  17. The effects of spectral shift absorbers on the design and safety of fast spectrum space reactors

    NASA Astrophysics Data System (ADS)

    King, Jeffrey Charles

    Spectral Shift Absorbers (SSAs) are incorporated into space reactors to maintain them sufficiently subcritical when submerged in seawater or wet sand and subsequently flooded, following a launch abort accident. The effect of four SSAs (samarium-149, europium-151, gadolinium-155, and gadolinium-157) on the submersion criticality, operation, and temperature reactivity feedback of the thermal spectrum reactors developed in the Systems for Nuclear Auxilary Power (SNAP) program is extensively documented. Recent work on SSAs in fast spectrum space reactors, preferred for compactness and higher powers, has focused on rhenium as the primary SSA. In addition to identifying additional SSAs, the present work investigates the effects of SSAs on the overall size and mass, temperature reactivity feedback, and operational lifetime of fast spectrum space reactors. The fast spectrum S4 reactor has a sectored Mo-14%Re solid-core, loadedwith UN fuel, cooled by He-30%Xe, and designed to avoid single point failures at a steady thermal power of 550 kWth. The addition of SSAs to the reactor core increases the fuel enrichment and decreases the size and mass of the reactor and the radiation shadow shield. SSA additions of boron-10, europium-151, gadolinium-155 and iridium result in the smallest and lightest S4 reactors. The effects of SSA additions on the operational lifetime and the temperature and burnup reactivity coefficients of the S^4 reactor are studied. An increasein fuel enrichment with SSAs markedly increases the operational lifetime by decreasing the burnup reactivity coefficient with only a slight decrease in the temperature reactivity feedback coefficient. With no SSAs, the UN fuel enrichment is lowest (58.5 wt%), the temperature and burnup reactivity coefficients are the highest (-0.2709 ¢/K and -1.3470 /atom%), and the estimated operating lifetime is the shortest (7.6 years). The temperature and burnup reactivity coefficients decrease to -0.2649 ¢/K and -1.0230 /atom%, and

  18. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  19. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  20. Advanced Fusion Reactors for Space Propulsion and Power Systems

    NASA Technical Reports Server (NTRS)

    Chapman, John J.

    2011-01-01

    In recent years the methodology proposed for conversion of light elements into energy via fusion has made steady progress. Scientific studies and engineering efforts in advanced fusion systems designs have introduced some new concepts with unique aspects including consideration of Aneutronic fuels. The plant parameters for harnessing aneutronic fusion appear more exigent than those required for the conventional fusion fuel cycle. However aneutronic fusion propulsion plants for Space deployment will ultimately offer the possibility of enhanced performance from nuclear gain as compared to existing ionic engines as well as providing a clean solution to Planetary Protection considerations and requirements. Proton triggered 11Boron fuel (p- 11B) will produce abundant ion kinetic energy for In-Space vectored thrust. Thus energetic alpha particles "exhaust" momentum can be used directly to produce high ISP thrust and also offer possibility of power conversion into electricity. p- 11B is an advanced fusion plant fuel with well understood reaction kinematics but will require some new conceptual thinking as to the most effective implementation.

  1. Advanced Fusion Reactors for Space Propulsion and Power Systems

    SciTech Connect

    Chapman, John J.

    2011-06-15

    In recent years the methodology proposed for conversion of light elements into energy via fusion has made steady progress. Scientific studies and engineering efforts in advanced fusion systems designs have introduced some new concepts with unique aspects including consideration of Aneutronic fuels. The plant parameters for harnessing aneutronic fusion appear more exigent than those required for the conventional fusion fuel cycle. However aneutronic fusion propulsion plants for Space deployment will ultimately offer the possibility of enhanced performance from nuclear gain as compared to existing ionic engines as well as providing a clean solution to Planetary Protection considerations and requirements. Proton triggered 11Boron fuel (p- 11B) will produce abundant ion kinetic energy for In-Space vectored thrust. Thus energetic alpha particles' exhaust momentum can be used directly to produce high Isp thrust and also offer possibility of power conversion into electricity. p-11B is an advanced fusion plant fuel with well understood reaction kinematics but will require some new conceptual thinking as to the most effective implementation.

  2. Fabricating niobium test loops for the SP-100 space reactor

    NASA Technical Reports Server (NTRS)

    Bryhan, Anthony J.; Chan, Ricky C.

    1993-01-01

    This article describes the successful fabrication, operation, and evaluation of a series of niobium-alloy (Nb-1 Zr and PWC-11) thermal convection loops designed to contain and circulate molten lithium at 1,350 K. These loops were used to establish the fabrication variables of significance for a nuclear power supply for space. Approximately 200 weldments were evaluated for their tendency to be attacked by lithium as a function of varying atmospheric contamination. No attack occurred for any weldment free of contamination, with or without heat treatment, and no welds accidentally deviated from purity. The threshold oxygen content for weldment attack was determined to be 170-200 ppm. Attack varied directly with weldment oxygen and nitrogen contents.

  3. Nuclear reactors using fine-particulate fuel for primary power in space

    SciTech Connect

    Botts, T.E.; Powell, J.R.; Usher, J.L.; Horn, F.L.

    1982-01-01

    Large future power requirements in space, include power beaming to earth, airplanes, and solar-powered satellites in eclipse; industrial processing; and space colonies. The Rotating Bed Nuclear Reactor (RBR) and Fixed Bed Reactor (FBR) are multi-megawatt power systems which are light, compact and suited to operation in space. Both are cavity reactors, with an annular fuel region (e.g., a bed of 500 ..mu.. HTGR fuel particulates made of UC with ceramic coating) surrounded by a reflector that moderates fast neutrons from the /sup 235/U fuel. A porous metal drum holds the fuel. In the RBR, rotation of the drum allows the particulate fuel bed to fluidize as cooling gas passes through. In the FBR, an inner porous carbon drum holds the packed fuel bed, which is not fluidized. The RBR and FBR have many important features for space nuclear power: very high power density (up to thousands of MW(th)/m/sup 3/ of fuel); very small size and weight, excellent thermal shock and fatigue resistance; short start/stop times (sec); high gas outlet temperatures (to 3000/sup 0/K), good neutron economy, low critical mass; and simple/reliable construction.

  4. Thermal-Hydraulic Analyses of the Submersion-Subcritical Safe Space (S and 4) Reactor

    SciTech Connect

    King, Jeffrey C.; El-Genk, Mohamed S.

    2007-01-30

    Detailed thermal-hydraulic analyses of the S and 4 reactor are performed to reduce the maximum fuel temperature of the Submersion-Subcritical Safe Space (S and 4) reactor to below 1300 K. The fuel pellet diameter is reduced from 1.315 cm to 1.25 cm, decreasing the thermal resistance of the pellets and each of the 1.54 cm diameter coolant channels in the reactor core are replaced with several 0.3 cm ID channels to increase the effective heat transfer area and to encourage mixing of the flowing helium-28% xenon coolant. The calculated maximum fuel temperature decreased from more than 1900 K to 1302 K and the relative pressure drop across the reactor core increased from 1.98% to 2.57% of the inlet pressure. Moving the concentric inlet and outlet pipes 1 cm towards the center of the reactor core encouraged more flow through the center region, further reducing the maximum fuel temperature by 14 degrees to 1288 K, with a negligible effect on the core pressure losses.

  5. 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.

  6. Technological implications of SNAP reactor power system development on future space nuclear power systems

    SciTech Connect

    Anderson, R.V.

    1982-11-16

    Nuclear reactor systems are one method of satisfying space mission power needs. The development of such systems must proceed on a path consistent with mission needs and schedules. This path, or technology roadmap, starts from the power system technology data base available today. Much of this data base was established during the 1960s and early 1970s, when government and industry developed space nuclear reactor systems for steady-state power and propulsion. One of the largest development programs was the Systems for Nuclear Auxiliary Power (SNAP) Program. By the early 1970s, a technology base had evolved from this program at the system, subsystem, and component levels. There are many implications of this technology base on future reactor power systems. A review of this base highlights the need for performing a power system technology and mission overview study. Such a study is currently being performed by Rockwell's Energy Systems Group for the Department of Energy and will assess power system capabilities versus mission needs, considering development, schedule, and cost implications. The end product of the study will be a technology roadmap to guide reactor power system development.

  7. Critical evaluation of molybdenum and its alloys for use in space reactor core heat pipes

    SciTech Connect

    Lundberg, L.B.

    1981-01-01

    The choice of pure molybdenum as the prime candidate material for space reactor core heat pipes is examined, and the advantages and disadvantages of this material are brought into focus. Even though pure molybdenum heat pipes have been built and tested, this metal's high ductile-brittle transition temperature and modest creep strength place significant design restrictions on a core heat pipe made from it. Molybdenum alloys are examined with regard to their promise as potential replacements for pure molybdenum. The properties of TZM and molybdenum-rhenium alloys are examined, and it appears that Mo-Re alloys with 10 to 15 wt % rhenium offer the most advantage as an alternative to pure molybdenum in space reactor core heat pipes.

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

    SciTech Connect

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

    2004-02-04

    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.

  9. Development of Liquid-Vapor Core Reactors with MHD Generator for Space Power and Propulsion Applications

    SciTech Connect

    Samim Anghaie

    2002-08-13

    Any reactor that utilizes fuel consisting of a fissile material in a gaseous state may be referred to as a gaseous core reactor (GCR). Studies on GCRs have primarily been limited to the conceptual phase, mostly due to budget cuts and program cancellations in the early 1970's. A few scientific experiments have been conducted on candidate concepts, primarily of static pressure fissile gas filling a cylindrical or spherical cavity surrounded by a moderating shell, such as beryllium, heavy water, or graphite. The main interest in this area of nuclear power generation is for space applications. The interest in space applications has developed due to the promise of significant enhancement in fuel utilization, safety, plant efficiency, special high-performance features, load-following capabilities, power conversion optimization, and other key aspects of nuclear power generation. The design of a successful GCR adapted for use in space is complicated. The fissile material studied in the pa st has been in a fluorine compound, either a tetrafluoride or a hexafluoride. Both of these molecules have an impact on the structural material used in the making of a GCR. Uranium hexafluoride as a fuel allows for a lower operating temperature, but at temperatures greater than 900K becomes essentially impossible to contain. This difficulty with the use of UF6 has caused engineers and scientists to use uranium tetrafluoride, which is a more stable molecule but has the disadvantage of requiring significantly higher operating temperatures. Gas core reactors have traditionally been studied in a steady state configuration. In this manner a fissile gas and working fluid are introduced into the core, called a cavity, that is surrounded by a reflector constructed of materials such as Be or BeO. These reactors have often been described as cavity reactors because the density of the fissile gas is low and criticality is achieved only by means of the reflector to reduce neutron leakage from the core

  10. 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.

  11. Power beaming to space using a nuclear reactor-pumped laser

    SciTech Connect

    Lipinski, R.J.; Monroe, D.K.; Pickard, P.S.

    1993-10-01

    The present political and environmental climate may slow the inevitable direct utilization of nuclear power in space. In the meantime, there is another approach for using nuclear energy for space power. That approach is to let nuclear energy generate a laser beam in a ground-based nuclear reactor-pumped laser (RPL), and then beam the optical energy into space. Potential space applications for a ground-based RPL include (1) illuminating geosynchronous communication satellites in the earth`s shadow to extend their lives, (2) beaming power to orbital transfer vehicles, (3) providing power (from earth) to a lunar base during the long lunar night, and (4) removing space debris. FALCON is a high-power, steady-state, nuclear reactor-pumped laser (RPL) concept that is being developed by the Department of Energy with Sandia National Laboratories as the lead laboratory. The FALCON program has experimentally demonstrated reactor-pumped lasing in various mixtures of xenon, argon, neon, and helium at wavelengths of 0.585, 0.703, 0.725, 1.271, 1.733, 1.792, 2.032, 2.63, 2.65, and 3.37 {mu}m with intrinsic efficiency as high as 2.5%. Frequency-doubling the 1.733{minus}{mu}m line would yield a good match for photovoltaic arrays at 0.867 {mu}m. Preliminary designs of an RPL suitable for power beaming have been completed. The MWclass laser is fairly simple in construction, self-powered, closed-cycle (no exhaust gases), and modular. This paper describes the FALCON program accomplishments and power-beaming applications.

  12. RSMASS-D models: An improved method for estimating reactor and shield mass for space reactor applications

    SciTech Connect

    Marshall, A.C.

    1997-10-01

    Three relatively simple mathematical models have been developed to estimate minimum reactor and radiation shield masses for liquid-metal-cooled reactors (LMRs), in-core thermionic fuel element (TFE) reactors, and out-of-core thermionic reactors (OTRs). The approach was based on much of the methodology developed for the Reactor/Shield Mass (RSMASS) model. Like the original RSMASS models, the new RSMASS-derivative (RSMASS-D) models use a combination of simple equations derived from reactor physics and other fundamental considerations, along with tabulations of data from more detailed neutron and gamma transport theory computations. All three models vary basic design parameters within a range specified by the user to achieve a parameter choice that yields a minimum mass for the power level and operational time of interest. The impact of critical mass, fuel damage, and thermal limitations are accounted for to determine the required fuel mass. The effect of thermionic limitations are also taken into account for the thermionic reactor models. All major reactor component masses are estimated, as well as instrumentation and control (I&C), boom, and safety system masses. A new shield model was developed and incorporated into all three reactor concept models. The new shield model is more accurate and simpler to use than the approach used in the original RSMASS model. The estimated reactor and shield masses agree with the mass predictions from separate detailed calculations within 15 percent for all three models.

  13. 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

  14. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  15. Fabrication and testing of uranium nitride fuel for space power reactors

    NASA Astrophysics Data System (ADS)

    Matthews, R. B.; Chidester, K. M.; Hoth, C. W.; Mason, R. E.; Petty, R. L.

    1988-02-01

    Uranium nitride fuel was selected for previous space power reactors because of its attractive thermal and physical properties; however, all UN fabrication and testing activities were terminated over ten years ago. An accelerated irradiation test, SP-1, was designed to demonstrate the irradiation performance of Nb-1 Zr clad UN fuel pins for the SP-100 program. A carbothermic-reduction/nitriding process was developed to synthesize UN powders. These powders were fabricated into fuel pellets by conventional cold-pressing and sintering. The pellets were loaded into Nb-1 Zr cladding tubes, irradiated in a fast-test reactor, and destructively examined after 0.8 at% burnup. Preliminary postirradiation examination (PIE) results show that the fuel pins behaved as designed. Fuel swelling, fission-gas release, and microstructural data are presented, and suggestions to enhance the reliability of UN fuel pins are discussed.

  16. The Application of U-Np Fuel and {sup 6}Li Burnable Poison for Space Reactors

    SciTech Connect

    Nikitin, Konstantin L.; Saito, Masaki; Artisyuk, Vladimir V.

    2003-11-15

    The possible application of {sup 6}Li as a burnable poison and U-Np nitride as a fuel for space nuclear reactors has been studied. The analysis was performed for an infinite lattice with a leakage in the form of buckling and (U-Np)N fuel with 20% uranium enrichment. The combination of {sup 7}Li as a coolant and {sup 6}Li as a burnable poison results in a favorable criticality behavior during burnup. The parameters taken into consideration include the different fuel and coolant compositions, the form of absorber material, and the various absorber mass and concentrations. It was found that absorption properties of {sup 6}Li allow reaching the burnup value up to 67 GWd/tHM while reactivity swing is comparable with {beta}{sub eff}. The corresponding reactor lifetime is {approx}10 to 30 yr.

  17. Start-up simulation of a thermionic space nuclear reactor system

    SciTech Connect

    El-Genk, M.S.; Xue, H.; Paramonov, D. )

    1993-01-15

    The Thermionic Transient Analysis Model (TITAM) is used in this paper to simulate the start-up of the TOPAZ-II space nuclear power system in orbit. The start-up procedures simulated herein are assumed for the purpose of demonstrating the capabilities of the model and may not represent an accurate account of the actual start-up procedures of the TOPAZ-II system. The temperature reactivity feedback effects of the moderator, UO[sub 2] fuel, electrodes, coolant, and other components in the core are calculated and their effects on the thermal and criticality conditions of the reactor are investigated. Also, estimates of the time constants of the temperature reactivity feedback for the UO[sub 2] fuel and the ZrH moderator during start-up, as well as of the total temperature reactivity feedback as a function of the reactor steady-state thermal power, are obtained.

  18. Transient analysis and startup simulation of a thermionic space nuclear reactor system

    SciTech Connect

    El-Genk, M.S.; Xue, Huimin; Paramonov, D. . Dept. of Chemical and Nuclear Engineering)

    1994-01-01

    The thermionic transient analysis model is used to simulate the startup of the TOPAZ-2 space nuclear power system in orbit. The simulated startup procedures are assumed for the purpose of demonstrating the capabilities of the model and may not represent an accurate account of the actual startup procedures of the TOPAZ-2 system. The temperature reactivity feedback effects of the moderator, UO[sub 2] fuel, electrodes, coolant, and other components in the core are calculated, and their effects on the thermal and criticality conditions of the reactor are investigated. Also, estimates of the time constants of the temperature reactivity feedback for the UO[sub 2] fuel and the ZrH moderator during startup, as well as of the total temperature reactivity feedback as a function of the reactor steady-state thermal power, are obtained.

  19. Space reactor/Stirling cycle systems for high power Lunar applications

    SciTech Connect

    Schmitz, P.D.; Mason, L.S.

    1994-09-01

    NASA`s Space Exploration Initiative (SEI) has proposed the use of high power nuclear power systems on the lunar surface as a necessary alternative to solar power. Because of the long lunar night ({approximately} 14 earth days) solar powered systems with the requisite energy storage in the form of regenerative fuel cells or batteries becomes prohibitively heavy at high power levels ({approximately} 100 kWe). At these high power levels nuclear power systems become an enabling technology for variety of missions. One way of producing power on the lunar surface is with an SP-100 class reactor coupled with Stirling power converters. In this study, analysis and characterization of the SP-100 class reactor coupled with Free Piston Stirling Power Conversion (FPSPC) system will be performed. Comparison of results with previous studies of other systems, particularly Brayton and Thermionic, are made.

  20. SP-100, the US Space Nuclear Reactor Power Program. Technical information report

    SciTech Connect

    Truscello, V.C.

    1983-11-01

    DARPA, in conjunction with DOE`s Office of Nuclear Energy, and NASA`s Office of Aeronautics and Space Technology are jointly sponsoring a space nuclear reactor power system program known as the Space Power-100 (SP-100) Development Project. The program is presently in the critical technology phase. This phase, better known as technology assessment and advancement, includes mission requirements definition, system conceptual designs, and critical technology development. A ground test phase decision is scheduled for July 1985. If the decision is positive, the next phase would begin in fiscal year 1986. An overriding concern in conducting this program is to ensure that nuclear safety is being properly addressed even in these early stages.

  1. Monodisperse Latex Reactor (MLR): A materials processing space shuttle mid-deck payload

    NASA Technical Reports Server (NTRS)

    Kornfeld, D. M.

    1985-01-01

    The monodisperse latex reactor experiment has flown five times on the space shuttle, with three more flights currently planned. The objectives of this project is to manufacture, in the microgravity environment of space, large particle-size monodisperse polystyrene latexes in particle sizes larger and more uniform than can be manufactured on Earth. Historically it has been extremely difficult, if not impossible to manufacture in quantity very high quality monodisperse latexes on Earth in particle sizes much above several micrometers in diameter due to buoyancy and sedimentation problems during the polymerization reaction. However the MLR project has succeeded in manufacturing in microgravity monodisperse latex particles as large as 30 micrometers in diameter with a standard deviation of 1.4 percent. It is expected that 100 micrometer particles will have been produced by the completion of the the three remaining flights. These tiny, highly uniform latex microspheres have become the first material to be commercially marketed that was manufactured in space.

  2. Development and characterization of lubricants for use near nuclear reactors in space vehicles

    NASA Technical Reports Server (NTRS)

    Robinson, G. L.; Akawie, R. I.; Gardos, M. N.; Krening, K. C.

    1972-01-01

    The synthesis and evaluation program was conducted to develop wide-temperature range lubricants suitable for use in space vehicles particularly in the vicinity of nuclear reactors. Synthetic approaches resulted in nonpolymeric, large molecular weight materials, all based on some combination of siloxane and aromatic groups. Evaluation of these materials indicated that certain tetramethyl and hexamethyl disiloxanes containing phenyl thiophenyl substituents are extremely promising with respect to radiation stability, wide temperature range, good lubricity, oxidation resistance and additive acceptance. The synthesis of fluids is discussed, and the equipment and methods used in evaluation are described, some of which were designed to evaluate micro-quantities of the synthesized lubricants.

  3. 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.

  4. 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.

  5. 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.

  6. Startup control of the TOPAZ-II space nuclear reactor. Master`s thesis

    SciTech Connect

    Astrin, C.D.

    1996-09-01

    The Russian designed and manufactured TOPAZ-II Thermionic Nuclear Space Reactor has been supplied to the Ballistic Missile Defense Organization for study as part of the TOPAZ International Program. A Preliminary Nuclear Safety Assessment investigated the readiness to use the TOPAZ-II in support of a Nuclear Electric Propulsion Space Test Mission (NEPSTP). Among the anticipated system modifications required for launching the TOPAZ-II system within safety goals is for a U.S. designed Automatic Control System. The requirements and desired features of such a control system are developed based upon U.S. safety standards. System theory and design are presented in order to establish the basis for development of a hybrid control model from available simulations. The model is verified and then used in exploration of various control schemes and casualty analysis providing groundwork for future Automatic Control System design.

  7. 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.

  8. 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.

  9. 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)

  10. Development of a steady state creep behavior model of polycrystalline tungsten for bimodal space reactor application

    SciTech Connect

    Purohit, A.; Hanan, N.A.; Bhattacharyya, S.K.; Gruber, E.E.

    1995-02-01

    The fuel element for one of the many reactor concepts being currently evaluated for bimodal applications in space consists of spherical fuel particles clad with tungsten or alloys of tungsten. The fuel itself consists of stabilized UO{sub 2}. One of the life limiting phenomena for the fuel element is failure of the cladding because of creep deformation. This report summarizes the information available in literature regarding the creep deformation of tungsten and its alloys and proposes a relation to be used for calculating the creep strains for elevated temperatures in the low stress region ({sigma} {le} 20 MPa). Also, results of the application of this creep relation to one of the reactor design concepts (NEBA-3) are discussed. Based on the traditional definition of creep deformation, the temperatures of 1500 K to 2900 K for tungsten and its alloys are considered to be in the {open_quotes}high{close_quotes} temperature range. In this temperature range, the rate controlling mechanisms for creep deformation are believed to be non-conservative motion of screw dislocations and short circuit diffusional paths. Extensive theoretical work on creep and in particular for creep of tungsten and its alloys have been reported in the literature. These theoretical efforts have produced complex mathematical models that require detailed materials properties. These relations, however, are not presently suitable for the creep analysis because of lack of consistent material properties required for their use. Variations in material chemistry and thermomechanical pre-treatment of tungsten have significant effects on creep and the mechanical properties. Analysis of the theoretical models and limited data indicates that the following empirical relation originally proposed by M. Jacox of INEL and the Air Force Phillips Laboratory, for calculating creep deformation of tungsten cladding, can be used for the downselection of preliminary bimodal reactor design concepts.

  11. A bifurcation analysis of boiling water reactor on large domain of parametric spaces

    NASA Astrophysics Data System (ADS)

    Pandey, Vikas; Singh, Suneet

    2016-09-01

    The boiling water reactors (BWRs) are inherently nonlinear physical system, as any other physical system. The reactivity feedback, which is caused by both moderator density and temperature, allows several effects reflecting the nonlinear behavior of the system. Stability analyses of BWR is done with a simplified, reduced order model, which couples point reactor kinetics with thermal hydraulics of the reactor core. The linear stability analysis of the BWR for steady states shows that at a critical value of bifurcation parameter (i.e. feedback gain), Hopf bifurcation occurs. These stable and unstable domains of parametric spaces cannot be predicted by linear stability analysis because the stability of system does not include only stability of the steady states. The stability of other dynamics of the system such as limit cycles must be included in study of stability. The nonlinear stability analysis (i.e. bifurcation analysis) becomes an indispensable component of stability analysis in this scenario. Hopf bifurcation, which occur with one free parameter, is studied here and it formulates birth of limit cycles. The excitation of these limit cycles makes the system bistable in the case of subcritical bifurcation whereas stable limit cycles continues in an unstable region for supercritical bifurcation. The distinction between subcritical and supercritical Hopf is done by two parameter analysis (i.e. codimension-2 bifurcation). In this scenario, Generalized Hopf bifurcation (GH) takes place, which separates sub and supercritical Hopf bifurcation. The various types of bifurcation such as limit point bifurcation of limit cycle (LPC), period doubling bifurcation of limit cycles (PD) and Neimark-Sacker bifurcation of limit cycles (NS) have been identified with the Floquet multipliers. The LPC manifests itself as the region of bistability whereas chaotic region exist because of cascading of PD. This region of bistability and chaotic solutions are drawn on the various

  12. 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.

  13. Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

    SciTech Connect

    M. Chen; CM Regan; D. Noe

    2006-01-09

    Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

  14. Design of a Simplified Closed Brayton Cycle for a Space Reactor Application

    SciTech Connect

    Guimaraes, Lamartine N. F.; Camillo, Giannino Ponchio; Placco, Guilherme Moreira

    2009-03-16

    The Nuclear Energy Division (ENU) of the Institute for Advanced Studies (IEAv) has started a preliminary design study for a Closed Brayton Cycle Loop (CBCL) aimed at a space reactor application. The main objectives of the study are: 1) to establish a starting concept for the CBCL components specifications, and 2) to build a demonstrative simulator of CBCL. This preliminary design study is been developed around the NOELLE 60290 turbo machine. The actual nuclear reactor study is being conducted independently. Because of that, a conventional heat source is being used for the CBCL, in this preliminary design phase. This paper describes details of the CBCL mechanical design and the steady state simulator of the CBCL operating with NOELLE 60290 turbo machine. In principle, several gases are being considered as working fluid, as for instance: air, helium, nitrogen, CO2 and gas mixtures such as helium and xenon. However, for this first application pure helium will be used as working fluid. Simplified models of heat and mass transfer were developed to simulate thermal components. A new graphical interface was developed for the simulator to display the thermal process variables in steady state and to keep track of the modifications being implemented at the NOELLE 60290 turbo machine in order to build the CBCL. A set of new results are being produced. These new results help to establish the hot and cold source geometry allowing for price estimating costs for building the actual device. These fresh new results will be presented and discussed.

  15. 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.

  16. 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.

  17. Physics effects of accidental submersion of space power reactors in water

    NASA Astrophysics Data System (ADS)

    Bhattacharyya, S. K.; Lell, R. M.

    A study has been performed of the generic aspects of the accidental submersion phenomenon in fast spectrum space power reactors. Since the net effect of submersion is a balance of competing effects of neutron spectrum softening and neutron reflection, a systematic study using rigorous analytical methods is necessary to extract the trends. Monte Carlo calculations were used to determine essential flooding characteristics in realistic infinite lattice cells. For practical clad/structural materials, computed flooding reactivity worths in infinite lattices were found to be negative because parasitic absorption rises faster than neutron production over the neutron energy range of importance. The flooding worth depends strongly on cell design, becoming increasingly negative as the coolant volume fraction increases. Negative flooding worth was calculated for all fissile enrichments considered (24 to 93 percent). Transport theory analyses of idealized cores of realistic sizes showed that computed flooding worths are positive or slightly negative for small cores because reflection of thermalized neutrons by the water outside dominates other flooding effects. As core size increases, infinite lattice behavior in the interior dominates, and flooding worth becomes increasingly negative. The loss of control drums was found to increase the core radius for crossover to negative flooding worths. Monte Carlo and Sn calculations for realistic core designs confirmed the dependence of sign and magnitude of the flooding worth on core size and on the presence or absence of control drums. Experimental evidence from terrestrial reactor programs have demonstrated the need for critical experiments to benchmark and calibrate analytical procedures for flooding calculations.

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

    SciTech Connect

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

    2008-01-21

    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 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.

  19. A Preliminary and Simplified Closed Brayton Cycle Modeling for a Space Reactor Application

    SciTech Connect

    Guimaraes, Lamartine Nogueira Frutuoso; Camillo, Giannino Ponchio

    2008-01-21

    The Nuclear Energy Division (ENU) of the Institute for Advanced Studies (IEAv) has started a preliminary design study for a Closed Brayton Cycle Loop (CBCL) aimed at a space reactor application. The main objectives of the study are: 1) to establish a starting concept for the CBCL components specifications, and 2) to build a demonstrative simulator of CBCL. This preliminary design study is developing the CBCL around the NOELLE 60290 turbo machine. The actual nuclear reactor study is being conducted independently. Because of that, a conventional heat source is being used for the CBCL, in this preliminary design phase. This paper describes the steady state simulator of the CBCL operating with NOELLE 60290 turbo machine. In principle, several gases are being considered as working fluid, as for instance: air, helium, nitrogen, CO{sub 2} and gas mixtures such as helium and xenon. However, for this first application pure helium will be used as working fluid. Simplified models of heat and mass transfer were developed to simulate thermal components. Future efforts will focus on implementing a graphical interface to display the thermal process variables in steady state and to keep track of the modifications being implemented at the NOELLE 60290 turbo machine in order to build the CBCL.

  20. Application of a Systems Engineering Approach to Support Space Reactor Development

    SciTech Connect

    Wold, Scott

    2005-02-06

    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.

  1. A Review of Tribological Coatings for Control Drive Mechanisms in Space Reactors

    SciTech Connect

    CJ Larkin; JD Edington; BJ Close

    2006-02-21

    Tribological coatings must provide lubrication for moving components of the control drive mechanism for a space reactor and prevent seizing due to friction or diffusion welding to provide highly reliable and precise control of reflector position over the mission lifetime. Several coatings were evaluated based on tribological performance at elevated temperatures and in ultrahigh vacuum environments. Candidates with proven performance in the anticipated environment are limited primarily to disulfide materials. Irradiation data for these coatings is nonexistent. Compatibility issues between coating materials and structural components may require the use of barrier layers between the solid lubricant and structural components to prevent deleterious interactions. It would be advisable to consider possible lubricant interactions prior to down-selection of structural materials. A battery of tests was proposed to provide the necessary data for eventual solid lubricant/coating selection.

  2. Survey of electronics capability for SP-100 space reactor power system applications

    NASA Technical Reports Server (NTRS)

    Manvi, Ram; Fujita, Tosh

    1991-01-01

    Because of reports indicating improvements in the radiation tolerance of some electronic parts, a survey was recently performed by SP-100 project personnel to determine the advisability of revising SP-100 SRPS (space reactor power systems) allowable neutron and gamma dose rates in order to reduce the size and mass of the radiation shield and thereby achieve system mass reductions. The survey results indicate that recent developments to increase the radiation tolerance of a limited set of electronics parts do not justify increasing the allowable SP-100 dose rates for electronic components. Specifically, the recent improvements on a limited set of parts do not justify increasing the current SP-100 allowable specifications of 5 x 10 exp 5 rads gamma dose and 1 x 10 exp 13 neutrons/sq cm fluence. However, if the improvements of 108 rads for gammas and 10 exp 15 neutrons/sq cm can be extended to a wide range of parts, significant mass savings would result.

  3. Determination and Assessment of Graded Safeguards Categories for Potential Space Reactor Fuels

    SciTech Connect

    Edelmann, Paul G.

    2005-02-06

    Research was conducted in the classification of lower-enrichment space nuclear reactor fuel in terms of its graded safeguards category and attractiveness level. Specifically, fuel with a 235U enrichment of less than 50% and greater than 20% (i.e., 49.9%) meets the criteria for a lower category of special-nuclear-material control and accountability. This fuel is considered to have an attractiveness level of 'D' and therefore is classified as Category II material rather than the highest and most rigorous Category I. The net result is a more cost-effective and facilitated approach to the protection, control, and accountability of this material. The views expressed in this document are those of the author and do not necessarily reflect agreement by the Government.

  4. Reentry safety for the Topaz II Space Reactor: Issues and analyses

    SciTech Connect

    Connell, L.W.; Trost, L.C.

    1994-03-01

    This report documents the reentry safety analyses conducted for the TOPAZ II Nuclear Electric Propulsion Space Test Program (NEPSTP). Scoping calculations were performed on the reentry aerothermal breakup and ground footprint of reactor core debris. The calculations were used to assess the risks associated with radiologically cold reentry accidents and to determine if constraints should be placed on the core configuration for such accidents. Three risk factors were considered: inadvertent criticality upon reentry impact, atmospheric dispersal of U-235 fuel, and the Special Nuclear Material Safeguards risks. Results indicate that the risks associated with cold reentry are very low regardless of the core configuration. Core configuration constraints were therefore not established for radiologically cold reentry accidents.

  5. Oxidation/corrosion of Nb-1 Zr and PWC-11 in lithium for space reactor applications

    SciTech Connect

    DiStefano, J.R.; Hendricks, J.W.

    1995-04-01

    The materials Nb-1 Zr and Nb-1 Zr-0.1 C (PWC-11) are candidate structural materials for compact lithium-cooled reactor systems for space applications. Because the mechanical and corrosion properties of niobium alloys can be significantly altered by oxygen, components must be protected from reactions with oxygen during ground testing. The effect of simultaneous oxidation/corrosion behavior in lithium was determined under several conditions, as well as the conditions under which a thin molybdenum coating was effective in preventing oxidation/corrosion. The effects of oxygen can be mitigated by its reaction with zirconium, and studies were also conducted to evaluate time-temperature relationships for this reaction.

  6. 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

  7. Key Factors Influencing the Decision on the Number of Brayton Units for the Prometheus Space Reactor

    NASA Astrophysics Data System (ADS)

    Ashcroft, John; Belanger, Sean; Burdge, Wayne; Clementoni, Eric; Jensen, Krista; Proctor, N. Beth; Zemo-Fulkerson, Annie

    2007-01-01

    The Naval Reactors (NR) Program and its DOE Laboratories, KAPL and Bettis, were assigned responsibility to develop space reactor systems for the Prometheus Program. After investigating all of the potential reactor and energy conversion options, KAPL and Bettis selected a direct gas Brayton system as the reference approach for the nuclear electric propulsion missions, including the Jupiter Icy Moons Orbiter (JIMO). In order to determine the optimal plant architecture for the direct gas system, KAPL and Bettis investigated systems with one or two active Brayton units and up to two spare units. No final decision was made on the optimal system configuration for the NEP gas-Brayton system prior to closeout of the project. The two most promising options appear to be a single system without spares and a three Brayton system with two operating units, each producing half of the required load, with a single spare unit. The studies show that a single Brayton system, without spares, offers the lowest mass system, with potential for lower operating temperature, and a minimum of system and operational complexity. The lower required mass and increased system efficiency inherent in the single Brayton system may be exploited to satisfy other design objectives such as reduced reactor and radiator operating temperatures. While Brayton system lifetimes applicable to a JIMO or other nuclear electric propulsion (NEP) mission have not been demonstrated, there is no fundamental limit on the lifetime of the Brayton hardware. Use of additional Brayton units with installed spares will allow for continued operation in the event of a failure of an individual Brayton unit. However, preliminary system reliability evaluations do not point to any substantial reliability benefit provided by carrying spare Brayton units. If a spare unit is used, operating two of the units at full power with an unpowered spare proved more efficient than operating all three units at a reduced power and temperature

  8. 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

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

    NASA Astrophysics Data System (ADS)

    Wetch, J. R.

    1988-09-01

    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.

  10. 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.

  11. A quantitative assessment of reliability of the TOPAZ-2 space NPS reactor unit based on ground development results

    SciTech Connect

    Ponomarev-Stepnoi, N.N.; Nechaev, Y.A.; Khazanovich, I.M.; Samodelov, V.N.; Zakharov, S.M.

    1997-01-01

    The paper discusses life-limiting factors (parameters) and statistics of random sudden failures, revealed in the course of ground development, for 4 given subsystems of the TOPAZ-2 space NPS reactor unit. Results are presented of a quantitative assessment of the lower confidence limits of the probability of failure-free operation. {copyright} {ital 1997 American Institute of Physics.}

  12. A quantitative assessment of reliability of the TOPAZ-2 space NPS reactor unit based on ground development results

    SciTech Connect

    Ponomarev-Stepnoi, Nikolai N.; Nechaev, Yuri A.; Khazanovich, Igor M.; Samodelov, Victor N.; Zakharov, Sergei M.

    1997-01-10

    The paper discusses life-limiting factors (parameters) and statistics of random sudden failures, revealed in the course of ground development, for 4 given subsystems of the TOPAZ-2 space NPS reactor unit. Results are presented of a quantitative assessment of the lower confidence limits of the probability of failure-free operation.

  13. 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.

  14. Testimony of Fred R. Mynatt before the Energy Research and Development Subcommittee of the Committee on Science, Space, and Technology, US House of Representatives. [Advanced fuel technology, gas-cooled reactor technology, and liquid metal-cooled reactor technology programs

    SciTech Connect

    Mynatt, F.R.

    1987-03-18

    This report provides a description of the statements submitted for the record to the committee on Science, Space, and Technology of the United States House of Representatives. These statements describe three principal areas of activity of the Advanced Reactor Technology Program of the Department of Energy (DOE). These areas are advanced fuel cycle technology, modular high-temperature gas-cooled reactor technology, and liquid metal-cooled reactor. The areas of automated reactor control systems, robotics, materials and structural design shielding and international cooperation were included in these statements describing the Oak Ridge National Laboratory's efforts in these areas. (FI)

  15. Modeling and Analysis of a Lunar Space Reactor with the Computer Code RELAP5-3D/ATHENA

    SciTech Connect

    Carbajo, Juan J; Qualls, A L

    2008-01-01

    The transient analysis 3-dimensional (3-D) computer code RELAP5-3D/ATHENA has been employed to model and analyze a space reactor of 180 kW(thermal), 40 kW (net, electrical) with eight Stirling engines (SEs). Each SE will generate over 6 kWe; the excess power will be needed for the pumps and other power management devices. The reactor will be cooled by NaK (a eutectic mixture of sodium and potassium which is liquid at ambient temperature). This space reactor is intended to be deployed over the surface of the Moon or Mars. The reactor operating life will be 8 to 10 years. The RELAP5-3D/ATHENA code is being developed and maintained by Idaho National Laboratory. The code can employ a variety of coolants in addition to water, the original coolant employed with early versions of the code. The code can also use 3-D volumes and 3-D junctions, thus allowing for more realistic representation of complex geometries. A combination of 3-D and 1-D volumes is employed in this study. The space reactor model consists of a primary loop and two secondary loops connected by two heat exchangers (HXs). Each secondary loop provides heat to four SEs. The primary loop includes the nuclear reactor with the lower and upper plena, the core with 85 fuel pins, and two vertical heat exchangers (HX). The maximum coolant temperature of the primary loop is 900 K. The secondary loops also employ NaK as a coolant at a maximum temperature of 877 K. The SEs heads are at a temperature of 800 K and the cold sinks are at a temperature of ~400 K. Two radiators will be employed to remove heat from the SEs. The SE HXs surrounding the SE heads are of annular design and have been modeled using 3-D volumes. These 3-D models have been used to improve the HX design by optimizing the flows of coolant and maximizing the heat transferred to the SE heads. The transients analyzed include failure of one or more Stirling engines, trip of the reactor pump, and trips of the secondary loop pumps feeding the HXs of the

  16. Swelling and tensile properties of EBR-II-irradiated tantalum alloys for space reactor applications

    SciTech Connect

    Grossbeck, M.L.; Wiffen, F.W.

    1985-01-01

    The tantalum alloys T-111, ASTAR-811C, Ta-10 W, and unalloyed tantalum were examined following EBR-II irradiation to a fluence of 1.7 x 10/sup 26/ neutrons/m/sup 2/ (E > 0.1 MeV) at temperatures from 650 to 950 K. Swelling was found to be negligible for all alloys; only tantalum was found to exhibit swelling, 0.36%. Tensile testing revealed that irradiated T-111 and Ta-10 W are susceptible to plastic instability, but ASTAR-811C and tantalum were not. The tensile properties of ASTAR-811C appeared adequate for current SP-100 space nuclear reactor designs. Irradiated, oxygen-doped T-111 exhibited no plastic deformation, and the abrupt failure was intergranular in nature. The absence of plastic instability in ASTAR-811C is encouraging for alloys containing carbide precipitates. These fine precipitates might prevent dislocation channeling, which leads to plastic instability in many bcc metals after irradiation. 10 refs., 13 figs., 8 tabs.

  17. Evaluation of HEU-Beryllium Benchmark Experiments to Improve Computational Analysis of Space Reactors

    SciTech Connect

    Bess, John; Bledsoe, Keith C; Rearden, Bradley T

    2011-01-01

    An assessment was previously performed to evaluate modeling capabilities and quantify preliminary biases and uncertainties associated with the modeling methods and data utilized in designing a nuclear reactor such as a beryllium-reflected, highly-enriched-uranium (HEU)-O2 fission surface power (FSP) system for space nuclear power. The conclusion of the previous study was that current capabilities could preclude the necessity of a cold critical test of the FSP; however, additional testing would reduce uncertainties in the beryllium and uranium cross-section data and the overall uncertainty in the computational models. A series of critical experiments using HEU metal were performed in the 1960s and 1970s in support of criticality safety operations at the Y-12 Plant. Of the hundreds of experiments, three were identified as fast-fission configurations reflected by beryllium metal. These experiments have been evaluated as benchmarks for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE). Further evaluation of the benchmark experiments was performed using the sensitivity and uncertainty analysis capabilities of SCALE 6. The data adjustment methods of SCALE 6 have been employed in the validation of an example FSP design model to reduce the uncertainty due to the beryllium cross section data.

  18. Evaluation of HEU-Beryllium Benchmark Experiments to Improve Computational Analysis of Space Reactors

    SciTech Connect

    John D. Bess; Keith C. Bledsoe; Bradley T. Rearden

    2011-02-01

    An assessment was previously performed to evaluate modeling capabilities and quantify preliminary biases and uncertainties associated with the modeling methods and data utilized in designing a nuclear reactor such as a beryllium-reflected, highly-enriched-uranium (HEU)-O2 fission surface power (FSP) system for space nuclear power. The conclusion of the previous study was that current capabilities could preclude the necessity of a cold critical test of the FSP; however, additional testing would reduce uncertainties in the beryllium and uranium cross-section data and the overall uncertainty in the computational models. A series of critical experiments using HEU metal were performed in the 1960s and 1970s in support of criticality safety operations at the Y-12 Plant. Of the hundreds of experiments, three were identified as fast-fission configurations reflected by beryllium metal. These experiments have been evaluated as benchmarks for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE). Further evaluation of the benchmark experiments was performed using the sensitivity and uncertainty analysis capabilities of SCALE 6. The data adjustment methods of SCALE 6 have been employed in the validation of an example FSP design model to reduce the uncertainty due to the beryllium cross section data.

  19. Independent Safety Assessment of the TOPAZ-II space nuclear reactor power system (Revised)

    SciTech Connect

    1993-09-01

    The Independent Safety Assessment described in this study report was performed to assess the safety of the design and launch plans anticipated by the U.S. Department of Defense (DOD) in 1993 for a Russian-built, U.S.-modified, TOPAZ-II space nuclear reactor power system. Its conclusions, and the bases for them, were intended to provide guidance for the U.S. Department of Energy (DOE) management in the event that the DOD requested authorization under section 91b. of the Atomic Energy Act of 1954, as amended, for possession and use (including ground testing and launch) of a nuclear-fueled, modified TOPAZ-II. The scientists and engineers who were engaged to perform this assessment are nationally-known nuclear safety experts in various disciplines. They met with participants in the TOPAZ-II program during the spring and summer of 1993 and produced a report based on their analysis of the proposed TOPAZ-II mission. Their conclusions were confined to the potential impact on public safety and did not include budgetary, reliability, or risk-benefit analyses.

  20. Numerical analysis of the effect of electrode spacing on deposition rate profiles in a capacitively coupled plasma reactor

    NASA Astrophysics Data System (ADS)

    Kim, Ho Jun; Lee, Hae June

    2016-12-01

    The effect of reactor dimension on deposition rate profiles is analyzed with a two-dimensional (2D) fluid simulation of a capacitively coupled plasma (CCP) reactor to deposit a hydrogenated silicon nitride (SiN x H y ) film with a SiH4/NH3/N2/He gas mixture. We focus on the complex function of electrode spacing to reveal the physical relation between reactor geometry and deposition rate profiles. The simulation demonstrates that the localization of electron density is concentrated close to the powered electrode periphery for electrode spacing of 9 mm. However, the plasma distribution becomes bulk dominated with electrode spacing of 15 mm by relaxing the localization. As a result, the increase in the electrode spacing creates a more uniform electron power density profile, and the deposition rate profile of SiN x H y film changes from convex to concave in a radial direction. The change in the deposition rate profile is validated through comparison with the experimental observation, which agrees well with the simulation results with errors of less than 5%. The deposition rate profile with electrode spacing of 9 mm is very sensitive to the non-uniform gas density condition applied to the showerhead inlet. However, the deposition rate profile with electrode spacing of 15 mm is not sensitive to the inlet gas profile because of the increasing residence time. The increase of the electrode spacing promotes molecule-molecule gas phase reactions and consequently weakens the effect of the inlet boundary condition.

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

    SciTech Connect

    Patton, Bruce; Sorensen, Kirk

    2002-07-01

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

  2. 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.

  3. Problems of a thermionic space NPS reactor unit quantitative reliability assessment on the basis of ground development results

    SciTech Connect

    Ponomarev-Stepnoi, Nikolai N.; Nechaev, Yuri A.; Khazanovich, Igor M.; Samodelov, Victor N.; Pavlov, Konstantin A.

    1997-01-10

    The paper sets forth major problems that arose in the course of a quantitative assessment of reliability of a TOPAZ-2 space NPS reactor unit performed on the basis of ground development results. Proposals are made on the possible ways to solve those problems through development and introduction of individual standards especially for the ground development stage, which would specify the assessment algorithm and censoring rules, and exclude a number of existing uncertainties when making a decision on going to flight testing.

  4. Radiator Heat Pipes with Carbon-Carbon Fins and Armor for Space Nuclear Reactor Power Systems

    NASA Astrophysics Data System (ADS)

    Tournier, Jean-Michel; El-Genk, Mohamed

    2005-02-01

    Technologies for Space Reactor Power Systems are being developed to enable future NASA's missions early next decade to explore the farthest planets in the solar system. The choices of the energy conversion technology for these power systems require radiator temperatures that span a wide range, from 350 K to 800 K. Heat pipes with carbon-carbon fins and armor are the preferred choice for these radiators because of inherent redundancy and efficient spreading and rejection of waste heat into space at a relatively small mass penalty. The performance results and specific masses of radiator heat pipes with cesium, rubidium, and potassium working fluids are presented and compared in this paper. The heat pipes operate at 40% of the prevailing operation limit (a design margin of 60%), typically the sonic and/or capillary limit. The thickness of the carbon-carbon fins is 0.5 mm but the width is varied, and the evaporator and condenser sections are 0.15 and 1.35 m long, respectively. The 400-mesh wick and the heat pipe thin metal wall are titanium, and the carbon-carbon armor (~ 2 mm-thick) provides both structural strength and protection against meteoroids impacts. The cross-section area of the D-shaped radiator heat pipes is optimized for minimum mass. Because of the low vapor pressure of potassium and its very high Figure-Of-Merit (FOM), radiator potassium heat pipes are the best performers at temperatures above 800 K, where the sonic limit is no longer an issue. On the other hand, rubidium heat pipes are limited by the sonic limit below 762 K and by the capillary limit at higher temperature. The transition temperature between these two limits for the cesium heat pipes occurs at a lower temperature of 724 K, since cesium has lower FOM than rubidium. The present results show that with a design margin of 60%, the cesium heat pipes radiator is best at 680-720 K, the rubidium heat pipes radiator is best at 720-800 K, while the potassium heat pipes radiator is the best

  5. Performance Analysis of Potassium Heat Pipes Radiator for HP-STMCs Space Reactor Power System

    SciTech Connect

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

    2004-02-04

    A detailed design and performance results of C-C finned, and armored potassium heat pipes radiator for a 110 kWe Heat Pipes-Segmented Thermoelectric Module Converters (HP-STMCs) Space Reactor Power system (SRPS) are presented. The radiator consists of two sections; each serves an equal number of STMCs and has 162 longitudinal potassium heat pipes with 0.508 mm thick C-C fins. The width of the C-C fins at the minor diameter of the radiator is almost zero, but increases with distance along the radiator to reach 3.7 cm at the radiator's major diameter. The radiator's heat pipes (OD = 2.42 cm in front and 3.03 cm in rear) have thin titanium (0.0762 mm thick) liners and wicks (0.20 mm thick with an effective pore radius of 12-16 {mu}m) and a 1.016 mm thick C-C wall. The wick is separated from the titanium liner by a 0.4 mm annulus filled with liquid potassium to increase the capillary limit. The outer surfaces of the heat pipes in the front and rear sections of the radiator are protected with a C-C armor that is 2.17 mm and 1.70 mm thick, respectively. The inside surface of the heat pipes in the front radiator is thermally insulated while the C-C finned condensers of the rear heat pipes are exposed, radiating into space through the rear opening of the radiator cavity. The heat pipes in both the front and the rear radiators have a 1.5 m long evaporator section and each dissipates 4.47 kW while operating at 43.6% of the prevailing sonic limit. The front and rear radiator sections are 5.29 m and 2.61 m long with outer surface area and mass of 47.1 m2 and 314.3 kg, and 39.9 m2 and 243.2 kg, respectively. The total radiator is 7.63 m long and has minor and major diameters of 1.48 m and 5.57 m, respectively, and a total surface area of 87 m2; however, the effective radiator area, after accounting for heat rejection through the rear of the radiator cavity, is 98.8 m2. The radiator's total mass including the C-C armor is 557.5 kg and the specific area and specific mass are 6

  6. Radiator Heat Pipes with Carbon-Carbon Fins and Armor for Space Nuclear Reactor Power Systems

    SciTech Connect

    Tournier, Jean-Michel; El-Genk, Mohamed

    2005-02-06

    Technologies for Space Reactor Power Systems are being developed to enable future NASA's missions early next decade to explore the farthest planets in the solar system. The choices of the energy conversion technology for these power systems require radiator temperatures that span a wide range, from 350 K to 800 K. Heat pipes with carbon-carbon fins and armor are the preferred choice for these radiators because of inherent redundancy and efficient spreading and rejection of waste heat into space at a relatively small mass penalty. The performance results and specific masses of radiator heat pipes with cesium, rubidium, and potassium working fluids are presented and compared in this paper. The heat pipes operate at 40% of the prevailing operation limit (a design margin of 60%), typically the sonic and/or capillary limit. The thickness of the carbon-carbon fins is 0.5 mm but the width is varied, and the evaporator and condenser sections are 0.15 and 1.35 m long, respectively. The 400-mesh wick and the heat pipe thin metal wall are titanium, and the carbon-carbon armor ({approx} 2 mm-thick) provides both structural strength and protection against meteoroids impacts. The cross-section area of the D-shaped radiator heat pipes is optimized for minimum mass. Because of the low vapor pressure of potassium and its very high Figure-Of-Merit (FOM), radiator potassium heat pipes are the best performers at temperatures above 800 K, where the sonic limit is no longer an issue. On the other hand, rubidium heat pipes are limited by the sonic limit below 762 K and by the capillary limit at higher temperature. The transition temperature between these two limits for the cesium heat pipes occurs at a lower temperature of 724 K, since cesium has lower FOM than rubidium. The present results show that with a design margin of 60%, the cesium heat pipes radiator is best at 680-720 K, the rubidium heat pipes radiator is best at 720-800 K, while the potassium heat pipes radiator is the best

  7. 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. .

  8. 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

  9. Comparison of Reactivity Control Systems for the Submersion Subcritical Safe Space (S∧4) Reactor

    NASA Astrophysics Data System (ADS)

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

    2008-01-01

    This paper compares the effectiveness of two control mechanisms for the S∧4 reactor, namely: (a) rotating BeO drums with 120° thin segments of enriched B4C in the radial reflector; and (b) sliding windows in the radial reflector. Investigated are the effects of using these control mechanisms on the differential reactor control worth, power generation profiles, and spatial neutrons flux distributions is the S∧4 reactor. For both control mechanism, the radial reflector has the same dimensions and volume. Results show that the difference in reactor performance with the two control mechanisms is small. The sliding reflector configuration features slightly lower mass and power peaking, and relatively more even fission power profiles in the core. The differential control worth for the sliding reflector segments is almost constant compared to that using rotating control drums, potentially simplifying the reactor control operation. The presence of a strong neutron absorber in the rotating drums slightly decreases the amount of excess reactivity at BOL compared with the sliding reflector configuration. However, the higher rate of reactivity depletion in the S∧4 reactor with the latter may negate this advantage.

  10. Comparison of Reactivity Control Systems for the Submersion Subcritical Safe Space (S and 4) Reactor

    SciTech Connect

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

    2008-01-21

    This paper compares the effectiveness of two control mechanisms for the S and 4 reactor, namely: (a) rotating BeO drums with 120 deg. thin segments of enriched B{sub 4}C in the radial reflector; and (b) sliding windows in the radial reflector. Investigated are the effects of using these control mechanisms on the differential reactor control worth, power generation profiles, and spatial neutrons flux distributions is the S and 4 reactor. For both control mechanism, the radial reflector has the same dimensions and volume. Results show that the difference in reactor performance with the two control mechanisms is small. The sliding reflector configuration features slightly lower mass and power peaking, and relatively more even fission power profiles in the core. The differential control worth for the sliding reflector segments is almost constant compared to that using rotating control drums, potentially simplifying the reactor control operation. The presence of a strong neutron absorber in the rotating drums slightly decreases the amount of excess reactivity at BOL compared with the sliding reflector configuration. However, the higher rate of reactivity depletion in the S and 4 reactor with the latter may negate this advantage.

  11. Partitionable-space enhanced coagulation (PEC) reactor and its working mechanism: a new prospective chemical technology for phosphorus pollution control.

    PubMed

    Zhang, Meng; Zheng, Ping; Abbas, Ghulam; Chen, Xiaoguang

    2014-02-01

    Phosphorus pollution control and phosphorus recycling, simultaneously, are focus of attention in the wastewater treatment. In this work, a novel reactor named partitionable-space enhanced coagulation (PEC) was invented for phosphorus control. The working performance and process mechanism of PEC reactor were investigated. The results showed that the PEC technology was highly efficient and cost-effective. The volumetric removal rate (VRR) reached up to 2.86 ± 0.04 kg P/(m(3) d) with a phosphorus removal rate of over 97%. The precipitant consumption was reduced to 2.60-2.76 kg Fe(II)/kg P with low operational cost of $ 0.632-0.673/kg P. The peak phosphorus content in precipitate was up to 30.44% by P2O5, which reveal the benefit of the recycling phosphorus resource. The excellent performance of PEC technology was mainly attributed to the partitionable-space and 'flocculation filter'. The partition limited the trans-regional back-mixing of reagents along the reactor, which promoted the precipitation reaction. The 'flocculation filter' retained the microflocs, enhancing the flocculation process.

  12. NEUTRONIC REACTOR

    DOEpatents

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  13. NEUTRONIC REACTORS

    DOEpatents

    Wigner, E.P.

    1960-11-22

    A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

  14. NEUTRONIC REACTORS

    DOEpatents

    Vernon, H.C.

    1959-01-13

    A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

  15. SCoRe - Concepts of Liquid Metal Cooled Space Reactors for Avoidance of Single-Point Failure

    SciTech Connect

    El-Genk, Mohamed; Hatton, Steven; Fox, Charles; Tournier, Jean-Michel

    2005-02-06

    Space nuclear Reactor Power Systems (SRPSs) are being developed to meet electrical power requirements for NASA's planetary exploration missions early next decade. In addition to enjoying some degree of autonomy, these systems need to operate reliably through the end of the mission, which could not be realized solely through a redundancy in the reactor's coolant loop. Besides increasing the total system mass, such hardware redundancy does not eliminate a single-point failure in the reactor and subsequent loss of coolant. This paper presents three concepts of the liquid metal cooled. Sectored, Compact Reactor (SCoRe) for the avoidance of single-point failure. The SCoRe-S, ScoRe-M, and SCoRe-L concepts are for small, medium, and large reactor cores, covering a wide range of electrical power requirements, from 10's of kWe to a few MWe. As a common feature in all SCoRe concepts, the reactor core is divided into six sectors that are neutronically coupled but thermal-hydraulically decoupled. The dividers of the sectors are liquid metal heat pipes, which facilitate cooling a sector experiencing a Loss of Coolant (LOC) by passively transporting the fission power generated in it to the two adjacent sectors without losing the mission. At the same time, the fission power of the reactor is reduced to avoid overheating the fuel in the sector experiencing a LOC. The SCoRe concepts have compact, hexagonal cores surrounded by a relatively thick (10 cm minimum) BeO reflector and axial BeO reflector that is 4 cm thick. The SCoRe is placed directly in front of the radiation shield, thus reducing the shield mass and that of the power system. In SCoRe-S cores, the UN fuel pins are arranged in a triangular lattice while in the SCoRe-M and SCoRe-L cores, the UN fuel pins arranged in a triangular lattice are assembled in 19-pin and 37-pin shrouded bundles, respectively.

  16. A molten Salt Am242M Production Reactor for Space Applications

    NASA Technical Reports Server (NTRS)

    Emrich, William

    2005-01-01

    The use of Am242m holds great promise for increasing the efficiency nuclear thermal rocket engines. Because Am242m has the highest fission cross section of any known isotope (1000's of barns), its extremely high reactivity may be used to directly heat a propellant gas with fission fragments. Since this isotope does not occur naturally, it must be bred in special production reactors designed for that purpose. The primary advantage to using molten salt reactors for breeding Am242m is that the reactors can be reprocessed continually yielding a constant rate of production of the isotope. Once built and initially fueled, the reactor will continually breed the additional fuel it needs to remain critical. The only feedstock required is a salt of U238. No enriched fuel is required during normal operation and all fissile material, except the Am242m, is maintained in a closed loop. For a reactor operating at 200 MW several kilograms of Am242m may be bred each year.

  17. 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

    design of the core and the fuel system of space nuclear reactors. Such a design must permit operation at both low (-100 kWe ) and high electrical power ...line design of the HPSNR system (Refs. 1,6-8) was developel to generate 100 kWe of continuous power for -3-7 yr lifetime while keeping the system...determine the physical size of the HPSNR power system for an electrical power output higher than the [00 kWe established for the oas - line design , the

  18. 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.

  19. Problems of a thermionic space NPS reactor unit quantitative reliability assessment on the basis of ground development results

    SciTech Connect

    Ponomarev-Stepnoi, N.N.; Nechaev, Y.A.; Khazanovich, I.M.; Samodelov, V.N.; Pavlov, K.A.

    1997-01-01

    The paper sets forth major problems that arose in the course of a quantitative assessment of reliability of a TOPAZ-2 space NPS reactor unit performed on the basis of ground development results. Proposals are made on the possible ways to solve those problems through development and introduction of individual standards especially for the ground development stage, which would specify the assessment algorithm and censoring rules, and exclude a number of existing uncertainties when making a decision on going to flight testing. {copyright} {ital 1997 American Institute of Physics.}

  20. Space nuclear power system based on thermionic reactor with single-cell TFEs and zirconium hydride moderator

    SciTech Connect

    Ponomarev-Stepnoi, N.N.; Usov, V.A. ); Nickitin, V.P.; Ogloblin, B.G.; Lutov, J.I.; Luppov, A.N.; Gabrusev, V.N.; Klimov, A.V. ); Nicolaev, J.V.; Kucherov, R.J.; Eremin, S.A. )

    1993-01-15

    The results of research and development work on the space nuclear power system with net electric power of 40 kW performed with cooperation of Russian research and engineering organizations in 1991--92 are presented in this paper. Development work was carried out on the basis of experience gained in the course of Topaz-2 SNPS creation. As the basis for SNPS the 127-channel intermediate reactor is used with in-core single-cell TFEs having improved parameters. The SNPS parameters, distinctive features of decision on design and layout, safety ensuring concepts, proposals on the future stages of SNPS development are also considered in the paper.

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

    SciTech Connect

    Zubrin, R.M.; Sulmeisters, T.K. )

    1993-01-20

    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.

  2. Fast Neutron Spectrum Potassium Worth for Space Power Reactor Design Validation

    SciTech Connect

    Bess, John D.; Marshall, Margaret A.; Briggs, J. Blair; Tsiboulia, Anatoli; Rozhikhin, Yevgeniy; Mihalczo, John T.

    2015-03-01

    A variety of critical experiments were constructed of enriched uranium metal (oralloy ) during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Of the hundreds of delayed critical experiments, one was performed that consisted of uranium metal annuli surrounding a potassium-filled, stainless steel can. The outer diameter of the annuli was approximately 13 inches (33.02 cm) with an inner diameter of 7 inches (17.78 cm). The diameter of the stainless steel can was 7 inches (17.78 cm). The critical height of the configurations was approximately 5.6 inches (14.224 cm). The uranium annulus consisted of multiple stacked rings, each with radial thicknesses of 1 inch (2.54 cm) and varying heights. A companion measurement was performed using empty stainless steel cans; the primary purpose of these experiments was to test the fast neutron cross sections of potassium as it was a candidate for coolant in some early space power reactor designs.The experimental measurements were performed on July 11, 1963, by J. T. Mihalczo and M. S. Wyatt (Ref. 1) with additional information in its corresponding logbook. Unreflected and unmoderated experiments with the same set of highly enriched uranium metal parts were performed at the Oak Ridge Critical Experiments Facility in the 1960s and are evaluated in the International Handbook for Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) with the identifier HEU MET FAST 051. Thin

  3. Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs

    SciTech Connect

    Yoder, G.L.

    2005-10-03

    This report documents the work performed during the first phase of the National Aeronautics and Space Administration (NASA), National Research Announcement (NRA) Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs. The document includes an optimization of both 100-kW{sub e} and 250-kW{sub e} (at the propulsion unit) Rankine cycle power conversion systems. In order to perform the mass optimization of these systems, several parametric evaluations of different design options were investigated. These options included feed and reheat, vapor superheat levels entering the turbine, three different material types, and multiple heat rejection system designs. The overall masses of these Nb-1%Zr systems are approximately 3100 kg and 6300 kg for the 100- kW{sub e} and 250-kW{sub e} systems, respectively, each with two totally redundant power conversion units, including the mass of the single reactor and shield. Initial conceptual designs for each of the components were developed in order to estimate component masses. In addition, an overall system concept was presented that was designed to fit within the launch envelope of a heavy lift vehicle. A technology development plan is presented in the report that describes the major efforts that are required to reach a technology readiness level of 6. A 10-year development plan was proposed.

  4. 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.

  5. Space reactor/Stirling cycle systems for high power lunar applications

    NASA Technical Reports Server (NTRS)

    Schmitz, Paul C.; Mason, Lee S.

    1991-01-01

    An analysis is performed to mathematically model a 550 kWe lunar base power supply which uses a SP-100 reactor coupled with Stirling converters. The reactor is placed in an excavation to keep activated coolant in the hole and to allow maintance of the components outside the hole. Two technology levels are considered. They are 1050 and 1300 K heater head Stirling converts. It is found that for a 1050 K converter the total mass which provided 1000 volts dc at 250 m is 14,366 kg while the 1300 K system mass is 12,104 kg. The radiation area of the 1050 and 1300 K systems are 641 and 356 sq m respectively. Comparisons are made with Brayton and thermionic systems with both near term and advanced technology considered.

  6. Space reactor/Stirling cycle systems for high power lunar application

    NASA Technical Reports Server (NTRS)

    Schmitz, Paul C.; Mason, Lee S.

    1991-01-01

    An analysis is performed to mathematically model a 550 kWe lunar base power supply which uses a SP-100 reactor coupled with Stirling converters. The reactor is placed in an excavation to keep activated coolant in the hole and to allow maintenance of the components outside the hole. Two technology levels are considered. They are 1050 and 1300 K heater head Stirling converts. It is found that for a 1050 K converter the total mass which provided 1000 volts DC at 250 m is 14,366 kg while the 1300 K system mass is 12,104 kg. The radiation area of the 1050 and 1300 K systems are 641 and 356 sq m respectively. Comparisons are made with Brayton and thermionic systems with both near term and advanced technology considered.

  7. REACTOR SHIELD

    DOEpatents

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  8. Thermal-Mechanical Studies for Gas-Cooled Space Reactor Designs

    SciTech Connect

    Kapernick, Richard J.; Creamer, William C.

    2006-01-20

    Los Alamos National Laboratory has been involved in the development of reactor concepts to be used as a power source for nuclear electric propulsion and/or for surface power sources. As part of this effort, a high fidelity thermal-mechanical analysis method has been developed for rapid performance assessments of these designs. This method has been used to study several concept alternatives, including both annular and multi-hole monolithic block designs. This paper presents the analysis method developed and results of analyses performed for a gas-cooled reactor. Key results are 1) the annular block design is lower mass than the multi-hole block design, 2) fuel temperatures are effectively controlled by adjusting the number of fuel pins in the core, 3) large thermal-hydraulic performance enhancements are produced by increasing coolant pressure and/or helium mole fraction, and 4) manufacturing and assembly parameters have relatively small effects on thermal-hydraulic performance and care should be taken to balance mechanical design complexity and reliability issues with thermal-hydraulic performance.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. Compact Reactor

    SciTech Connect

    Williams, Pharis E.

    2007-01-30

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date.

  14. NEUTRONIC REACTOR

    DOEpatents

    Metcalf, H.E.

    1957-10-01

    A reactor of the type which preferably uses plutonium as the fuel and a liquid moderator, preferably ordinary water, and which produces steam within the reactor core due to the heat of the chain reaction is described. In the reactor shown the fuel elements are essentially in the form of trays and are ventically stacked in spaced relationship. The water moderator is continuously supplied to the trays to maintain a constant level on the upper surfaces of the fuel element as it is continually evaporated by the heat. The steam passes out through the spaces between the fuel elements and is drawn off at the top of the core. The fuel elements are clad in aluminum to prevent deterioration thereof with consequent contamimation of the water.

  15. A Common Methodology for Safety and Reliability Analysis for Space Reactor Missions

    NASA Astrophysics Data System (ADS)

    Frank, Michael V.

    2006-01-01

    The thesis of this paper is that the methodology of probabilistic risk management (PRM) has the capability to integrate both safety and reliability analyses for space nuclear missions. Practiced within a decision analysis framework, the concept of risk and the overall methodology of PRM are not dependent on whether the outcome affects mission success or mission safety. This paper presents the methodology by means of simplified examples.

  16. A Common Methodology for Safety and Reliability Analysis for Space Reactor Missions

    SciTech Connect

    Frank, Michael V.

    2006-01-20

    The thesis of this paper is that the methodology of probabilistic risk management (PRM) has the capability to integrate both safety and reliability analyses for space nuclear missions. Practiced within a decision analysis framework, the concept of risk and the overall methodology of PRM are not dependent on whether the outcome affects mission success or mission safety. This paper presents the methodology by means of simplified exampl0008.

  17. Load-carrying capabilities of refractory alloys for space reactor power applications

    SciTech Connect

    Horak, J.A.

    1985-01-01

    To achieve sufficient thermodynamic efficiency, space nuclear power systems must operate above 1000/sup 0/C. A quantitative evaluation of the existing mechanical properties data for the refractory alloys relevant to space nuclear power systems design lifetimes up to seven years at temperatures up to 1400/sup 0/C is being conducted. The most important properties for space nuclear power systems are long-term high-temperature (>1000/sup 0/C) creep strength and ductility, low-temperature (<400/sup 0/C) fracture toughness (including ductile-to-brittle transition temperature, (DBTT)), and ductility at high strain rates; of special concern are the above properties for weldments of refractory alloys, composition, applied stress, test temperature, test environment (e.g., vacuum, lithium), and thermomechanical treatment (TMT) history. Currently being evaluated are, in order of ascending mp, selected alloys of niobium (e.g., Nb-1% Zr, Nb-1% Zr-0.1% C), molybdenum (e.g., Mo-13% Re), tantalum (e.g., ASTAR-811C), and tungsten (e.g., CVD W and W-25% Re). Creep properties of these alloys have been correlated via Larson-Miller, Manson-Hafered, and other empirical parameters; creep equations have been developed from these correlations. 12 figs., 8 tabs.

  18. NUCLEAR REACTORS

    DOEpatents

    Long, E.; Ashby, J.W.

    1958-09-16

    ABS>A graphite moderator structure is presented for a nuclear reactor compriscd of an assembly of similarly orientated prismatic graphite blocks arranged on spaced longitudinal axes lying in common planes wherein the planes of the walls of the blocks are positioned so as to be twisted reintive to the planes of said axes so thatthe unlmpeded dtrect paths in direction wholly across the walls of the blocks are limited to the width of the blocks plus spacing between the blocks.

  19. Reactor vessel support system

    DOEpatents

    Golden, Martin P.; Holley, John C.

    1982-01-01

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  20. Benchmark Evaluation of Fuel Effect and Material Worth Measurements for a Beryllium-Reflected Space Reactor Mockup

    SciTech Connect

    Marshall, Margaret A.; Bess, John D.

    2015-02-01

    The critical configuration of the small, compact critical assembly (SCCA) experiments performed at the Oak Ridge Critical Experiments Facility (ORCEF) in 1962-1965 have been evaluated as acceptable benchmark experiments for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. The initial intent of these experiments was to support the design of the Medium Power Reactor Experiment (MPRE) program, whose purpose was to study “power plants for the production of electrical power in space vehicles.” The third configuration in this series of experiments was a beryllium-reflected assembly of stainless-steel-clad, highly enriched uranium (HEU)-O2 fuel mockup of a potassium-cooled space power reactor. Reactivity measurements cadmium ratio spectral measurements and fission rate measurements were measured through the core and top reflector. Fuel effect worth measurements and neutron moderating and absorbing material worths were also measured in the assembly fuel region. The cadmium ratios, fission rate, and worth measurements were evaluated for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. The fuel tube effect and neutron moderating and absorbing material worth measurements are the focus of this paper. Additionally, a measurement of the worth of potassium filling the core region was performed but has not yet been evaluated Pellets of 93.15 wt.% enriched uranium dioxide (UO2) were stacked in 30.48 cm tall stainless steel fuel tubes (0.3 cm tall end caps). Each fuel tube had 26 pellets with a total mass of 295.8 g UO2 per tube. 253 tubes were arranged in 1.506-cm triangular lattice. An additional 7-tube cluster critical configuration was also measured but not used for any physics measurements. The core was surrounded on all side by a beryllium reflector. The fuel effect worths were measured by removing fuel tubes at various radius. An accident scenario

  1. Preliminary nuclear safety assessment of the NEPST (Topaz 2) space reactor program

    NASA Astrophysics Data System (ADS)

    Marshall, A. C.

    The United States (US) Strategic Defense Initiative Organization (SDIO) decided to investigate the possibility of launching a Russian Topaz 2 space nuclear power system. A preliminary nuclear safety assessment was conducted to determine whether or not a space mission could be conducted safely and within budget constraints. As part of this assessment, a safety policy and safety functional requirements were developed to guide both the safety assessment and future Topaz 2 activities. A review of the Russian flight safety program was conducted and documented. Our preliminary nuclear safety assessment included a number of deterministic analyses, such as the following: neutronic analysis of normal and accident configurations, an evaluation of temperature coefficients of reactivity, a reentry and disposal analysis, an analysis of postulated launch abort impact accidents, and an analysis of postulated propellant fire and explosion accidents. Based on the assessment to date, it appears that it will be possible to safely launch the Topaz 2 system in the US with a modification to preclude water flooded criticality. A full scale safety program is now underway.

  2. Phase space optimisation of the USANS instrument Kookaburra at the ANSTO OPAL reactor

    NASA Astrophysics Data System (ADS)

    Freund, A. K.; Rehm, C.

    2011-04-01

    An optimisation of the USANS instrument Kookaburra has been conducted determining the most efficient type of premonochromator crystal and scattering geometry. The approach has been based on representations in phase space combining direct and reciprocal space diagrams. Results are presented for several scenarios involving flat and curved, perfect and mosaic crystals, horizontal and vertical scattering planes. The highest peak current density of 1.7×105 cm-2 s-1 is provided for a wavelength of 4.43 Å by a doubly curved pyrolytic graphite premonochromator diffracting in a horizontal plane and producing a beam size of about 50 cm2 at the sample position. The smallest background is achieved for a wavelength of 2.56 Å and a vertical scattering plane using a bent perfect premonochromator crystal and optionally a collimator to decrease the divergence of the beam exiting the cold neutron guide. For this wavelength a peak current density of 6.6×104 cm-2 s-1 is collected into a sample area of 10 cm2. The bent perfect crystal design is of particular advantage for small samples.

  3. Preliminary nuclear safety assessment of the NEPST (Topaz II) space reactor program

    SciTech Connect

    Marshall, A.C.

    1993-01-01

    The United States (US) Strategic Defense Initiative Organization (SDIO) decided to investigate the possibility of launching a Russian Topaz II space nuclear power system. A preliminary nuclear safety assessment was conducted to determine whether or not a space mission could be conducted safely and within budget constraints. As part of this assessment, a safety policy and safety functional requirements were developed to guide both the safety assessment and future Topaz II activities. A review of the Russian flight safety program was conducted and documented. Our preliminary nuclear safety assessment included a number of deterministic analyses, such as; neutronic analysis of normal and accident configurations, an evaluation of temperature coefficients of reactivity, a reentry and disposal analysis, an analysis of postulated launch abort impact accidents, and an analysis of postulated propellant fire and explosion accidents. Based on the assessment to date, it appears that it will be possible to safely launch the Topaz II system in the US with a modification to preclude water flooded criticality. A full scale safety program is now underway.

  4. Assessing the Effects of Radiation Damage on Ni-base Alloys for the Prometheus Space Reactor System

    SciTech Connect

    T. Angeliu

    2006-01-19

    Ni-base alloys were considered for the Prometheus space reactor pressure vessel with operational parameters of {approx}900 K for 15 years and fluences up to 160 x 10{sup 20} n/cm{sup 2} (E > 0.1 MeV). This paper reviews the effects of irradiation on the behavior of Ni-base alloys and shows that radiation-induced swelling and creep are minor considerations compared to significant embrittlement with neutron ,exposure. While the mechanism responsible for radiation-induced embrittlement is not fully understood, it is likely a combination of helium embrittlement and solute segregation that can be highly dependent on the alloy composition and exposure conditions. Transmutation calculations show that detrimental helium levels would be expected at the end of life for the inner safety rod vessel (thimble) and possibly the outer pressure vessel, primarily from high energy (E > 1 MeV) n,{alpha} reactions with {sup 58}Ni. Helium from {sup 10}B is significant only for the outer vessel due to the proximity of the outer vessel to the Be0 control elements. Recommendations for further assessments of the material behavior and methods to minimize the effects of radiation damage through alloy design are provided.

  5. Assessing the Effects of Radiation Damage on Ni-base Alloys for the Prometheus Space Reactor System

    SciTech Connect

    T Angeliu; J Ward; J Witter

    2006-04-04

    Ni-base alloys were considered for the Prometheus space reactor pressure vessel with operational parameters of {approx}900 K for 15 years and fluences up to 160 x 10{sup 20} n/cm{sup 2} (E > 0.1 MeV). This paper reviews the effects of irradiation on the behavior of Ni-base alloys and shows that radiation-induced swelling and creep are minor considerations compared to significant embrittlement with neutron exposure. While the mechanism responsible for radiation-induced embrittlement is not fully understood, it is likely a combination of helium embrittlement and solute segregation that can be highly dependent on the alloy composition and exposure conditions. Transmutation calculations show that detrimental helium levels would be expected at the end of life for the inner safety rod vessel (thimble) and possibly the outer pressure vessel, primarily from high energy (E > 1 MeV) n,{alpha} reactions with {sup 58}Ni. Helium from {sup 10}B is significant only for the outer vessel due to the proximity of the outer vessel to the BeO control elements. Recommendations for further assessments of the material behavior and methods to minimize the effects of radiation damage through alloy design are provided.

  6. Bellows-Type Accumulators for Liquid Metal Loops of Space Reactor Power Systems

    NASA Astrophysics Data System (ADS)

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

    2006-01-01

    In many space nuclear power systems, the primary and/or secondary loops use liquid metal working fluids, and require accumulators to accommodate the change in the liquid metal volume and maintain sufficient subcooling to avoid boiling. This paper developed redundant and light-weight bellows-type accumulators with and without a mechanical spring, and compared the operating condition and mass of the accumulators for different types of liquid metal working fluids and operating temperatures: potassium, NaK-78, sodium and lithium loops of a total capacity of 50 liters and nominal operating temperatures of 840 K, 860 K, 950 K and 1340 K, respectively. The effects of using a mechanical spring and different structural materials on the design, operation and mass of the accumulators are also investigated. The structure materials considered include SS-316, Hastelloy-X, C-103 and Mo-14Re. The accumulator without a mechanical spring weighs 23 kg and 40 kg for a coolant subcooling of 50 K and 100 K, respectively, following a loss of the fill gas. The addition of a mechanical spring comes with a mass penalty, in favor of higher redundancy and maintaining a higher liquid metal subcooling.

  7. Development and test of a space-reactor-core heat pipe

    SciTech Connect

    Merrigan, M.A.; Runyan, J.E.; Martinez, H.E.; Keddy, E.S.

    1983-01-01

    A heat pipe designed to meet the heat transfer requirements of a 100-kW/sub e/ space nuclear power system has been developed and tested. General design requirements for the device included an operating temperature of 1500/sup 0/K with an evaporator radial flux density of 100 w/cm/sup 2/. The total heat-pipe length of 2 m comprised an evaporator length of 0.3 m, a 1.2-m adiabatic section, and a condenser length of 0.5 m. A four-artery design employing screen arteries and distribution wicks was used with lithium serving as the working fluid. Molybdenum alloys were used for the screen materials and tube shell. Hafnium and zirconium gettering materials were used in connection with a pre-purified distilled lithium charge to ensure internal chemical compatibility. After initial performance verification, the 14.1-mm i.d. heat pipe was operated at 15 kW throughput at 1500/sup 0/K for 100 hours. No performance degradation was observed during the test.

  8. Bellows-Type Accumulators for Liquid Metal Loops of Space Reactor Power Systems

    SciTech Connect

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

    2006-01-20

    In many space nuclear power systems, the primary and/or secondary loops use liquid metal working fluids, and require accumulators to accommodate the change in the liquid metal volume and maintain sufficient subcooling to avoid boiling. This paper developed redundant and light-weight bellows-type accumulators with and without a mechanical spring, and compared the operating condition and mass of the accumulators for different types of liquid metal working fluids and operating temperatures: potassium, NaK-78, sodium and lithium loops of a total capacity of 50 liters and nominal operating temperatures of 840 K, 860 K, 950 K and 1340 K, respectively. The effects of using a mechanical spring and different structural materials on the design, operation and mass of the accumulators are also investigated. The structure materials considered include SS-316, Hastelloy-X, C-103 and Mo-14Re. The accumulator without a mechanical spring weighs 23 kg and 40 kg for a coolant subcooling of 50 K and 100 K, respectively, following a loss of the fill gas. The addition of a mechanical spring comes with a mass penalty, in favor of higher redundancy and maintaining a higher liquid metal subcooling.

  9. Bioconversion reactor

    DOEpatents

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  10. Thermal and neutron-physical features of the nuclear reactor for a power pulsation plant for space applications

    NASA Astrophysics Data System (ADS)

    Gordeev, É. G.; Kaminskii, A. S.; Konyukhov, G. V.; Pavshuk, V. A.; Turbina, T. A.

    2012-05-01

    We have explored the possibility of creating small-size reactors with a high power output with the provision of thermal stability and nuclear safety under standard operating conditions and in emergency situations. The neutron-physical features of such a reactor have been considered and variants of its designs preserving the main principles and approaches of nuclear rocket engine technology are presented.

  11. 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.

  12. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1958-08-19

    A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

  13. 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.

  14. REACTOR CONTROL

    DOEpatents

    Ruano, W.J.

    1957-12-10

    This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.

  15. An analysis of thermionic space nuclear reactor power system: I. Effect of disassembling radial reflector, following a reactivity initiated accident

    SciTech Connect

    El-Genk, M.S.; Paramonov, D. )

    1993-01-10

    An analysis is performed to determine the effect of disassembling the radial reflector of the TOPAZ-II reactor, following a hypothetical severe Reactivity Initiated Accident (RIA). Such an RIA is assumed to occur during the system start-up in orbit due to a malfunction of the drive mechanism of the control drums, causing the drums to rotate the full 180[degree] outward at their maximum speed of 1.4[degree]/s. Results indicate that disassembling only three of twelve radial reflector panels would successfully shutdown the reactor, with little overheating of the fuel and the moderator.

  16. Nuclear reactor

    DOEpatents

    Wade, Elman E.

    1979-01-01

    A nuclear reactor including two rotatable plugs and a positive top core holddown structure. The top core holddown structure is divided into two parts: a small core cover, and a large core cover. The small core cover, and the upper internals associated therewith, are attached to the small rotating plug, and the large core cover, with its associated upper internals, is attached to the large rotating plug. By so splitting the core holddown structures, under-the-plug refueling is accomplished without the necessity of enlarging the reactor pressure vessel to provide a storage space for the core holddown structure during refueling. Additionally, the small and large rotating plugs, and their associated core covers, are arranged such that the separation of the two core covers to permit rotation is accomplished without the installation of complex lifting mechanisms.

  17. 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.

  18. NUCLEAR REACTORS

    DOEpatents

    Long, E.; Ashley, J.W.

    1958-12-16

    A graphite moderator structure is described for a gas-cooled nuclear reactor having a vertical orlentation wherein the structure is physically stable with regard to dlmensional changes due to Wigner growth properties of the graphite, and leakage of coolant gas along spaces in the structure is reduced. The structure is comprised of stacks of unlform right prismatic graphite blocks positioned in layers extending in the direction of the lengths of the blocks, the adjacent end faces of the blocks being separated by pairs of tiles. The blocks and tiles have central bores which are in alignment when assembled and are provided with cooperatlng keys and keyways for physical stability.

  19. 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.

  20. Reactor vessel support system. [LMFBR

    DOEpatents

    Golden, M.P.; Holley, J.C.

    1980-05-09

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  1. NEUTRONIC REACTOR

    DOEpatents

    McGarry, R.J.

    1958-04-22

    Fluid-cooled nuclear reactors of the type that utilize finned uranium fuel elements disposed in coolant channels in a moderater are described. The coolant channels are provided with removable bushings composed of a non- fissionable material. The interior walls of the bushings have a plurality of spaced, longtudinal ribs separated by grooves which receive the fins on the fuel elements. The lands between the grooves are spaced from the fuel elements to form flow passages, and the size of the now passages progressively decreases as the dlstance from the center of the core increases for the purpose of producing a greater cooling effect at the center to maintain a uniform temperature throughout the core.

  2. 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.

  3. Fast Reactors

    NASA Astrophysics Data System (ADS)

    Esposito, S.; Pisanti, O.

    The following sections are included: * Elementary Considerations * The Integral Equation to the Neutron Distribution * The Critical Size for a Fast Reactor * Supercritical Reactors * Problems and Exercises

  4. An analysis of thermionic space nuclear reactor power system: II. Merits of using safety drums for backup control

    SciTech Connect

    El-Genk, M.S.; Huimin Xue )

    1993-01-10

    An analysis is performed to investigate the merits of using the TOPAZ-II safety drums for a backup control to prevent a reactivity excursion, stabilize the reactor, and achieve steady-state power operation, following a severe hypothetical reactivity initiated accident (RIA). Such an RIA is assumed to occur during the system start-up in orbit due to a malfunction of the drive mechanism of the control drums, causing the nine drums to accidentally rotate the full 180[degree] outward. Results show that an immediate, inward rotation of the three safety drums to an angle of 80[degree] will shutdown the reactor, however, a delay time of 10 s will not only prevents a reactivity excursion, but also enables operating the reactor at a steady-state thermal power of about 33.3 kW (0.9 kW per TFE). Conversely, when the immediate rotation of the safety drums is to a larger angle of 100[degree], a steady-state operation at about 37 kW can be achieved, but a delay of 10 s causes a reactivity excursion and overheating of the TFEs. It is therefor concluded that, should the drive mechanism be modified to enable rotating the safety drums for TOPAZ-II reactor at variable speeds of and below 22.5[degree]/s, the three safety drums could be used successfully for a backup control, following an RIA. However, since the reactivity worth of the three safety drums is only $2.0, the maximum steady-state electric power achievable for the system is limited to approximately 0.25 kW, at which the fission power is about 37 kW and the emitter temperature is approximaely 1500 K. To alleviate such a limitation and enable operation at nominal design conditions (fission power of about 107 kW or a system's total electric power of 5.6 kW), the reactivity worth of the safety drums would have to be increased by at least $0.24.

  5. Effect of Neutron Absorbers Mixed in or Coating the Fuel of a 1-MWt Lithium-Cooled Space Reactor

    SciTech Connect

    Amiri, Benjamin W.; Poston, David I.

    2005-02-06

    The goal of this study was to determine the effect of various neutron poisons (boron, dysprosium, erbium, and gadolinium) on a 1-MWt, lithium-cooled liquid-metal reactor. The isotopes were considered to be in-fuel poisons, as well as poisons coating the fuel. One way to quantify the effectiveness of a poison in meeting accident-condition requirements is by defining the safety margin as the difference between keff at the beginning of life and keff during the accident scenarios. The isotope that showed the most potential in increasing the safety margin for the wet-sand/water case was 157Gd. The safety margin was 10%-20% greater using 157Gd as an in-fuel poison as opposed to a coating, depending on the poison quantity. However, the most limiting condition (i.e., the accident scenario with the highest keff, thus the lowest safety margin) is when the reactor is submerged in wet sand. None of the isotopes considered significantly affected the safety margin for the dry-sand case. However, the poison isotopes considered may have applicability for meeting the wet-sand/water keff requirements or as burnable poisons in a moderated system. The views expressed in this document are those of the author and do not necessarily reflect agreement by the government.

  6. U.S.-Russian Cooperation in Science and Technology: A Case Study of the TOPAZ Space-Based Nuclear Reactor International Program

    NASA Astrophysics Data System (ADS)

    Dabrowski, Richard S.

    2014-08-01

    The TOPAZ International Program (TIP) was the final name given to a series of projects to purchase and test the TOPAZ-II, a space-based nuclear reactor of a type that had been further developed in the Soviet Union than in the United States. In the changing political situation associated with the break-up of the Soviet Union it became possible for the United States to not just purchase the system, but also to employ Russian scientists, engineers and testing facilities to verify its reliability. The lessons learned from the TIP illuminate some of the institutional and cultural challenges to U.S. - Russian cooperation in technology research which remain true today.

  7. NEUTRONIC REACTOR

    DOEpatents

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

    1958-09-16

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

  8. Nuclear reactor

    DOEpatents

    Pennell, William E.; Rowan, William J.

    1977-01-01

    A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

  9. Nuclear reactor

    DOEpatents

    Thomson, Wallace B.

    2004-03-16

    A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

  10. 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.

  11. BOILING REACTORS

    DOEpatents

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  12. NEUTRONIC REACTOR

    DOEpatents

    Daniels, F.

    1959-10-27

    A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

  13. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1961-09-01

    A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

  14. Study of Compatibility of Stainless Steel Weld Joints with Liquid Sodium-Potassium Coolants for Fission Surface Power Reactors for Lunar and Space Applications

    SciTech Connect

    Grossbeck, Martin; Qualls, Louis

    2015-07-31

    To make a manned mission to the surface of the moon or to Mars with any significant residence time, the power requirements will make a nuclear reactor the most feasible source of energy. To prepare for such a mission, NASA has teamed with the DOE to develop Fission Surface Power technology with the goal of developing viable options. The Fission Surface Power System (FSPS) recommended as the initial baseline design includes a liquid metal reactor and primary coolant system that transfers heat to two intermediate liquid metal heat transfer loops. Each intermediate loop transfers heat to two Stirling heat exchangers that each power two Stirling converters. Both the primary and the intermediate loops will use sodium-potassium (NaK) as the liquid metal coolant, and the primary loop will operate at temperatures exceeding 600°C. The alloy selected for the heat exchangers and piping is AISI Type 316L stainless steel. The extensive experience with NaK in breeder reactor programs and with earlier space reactors for unmanned missions lends considerable confidence in using NaK as a coolant in contact with stainless steel alloys. However, the microstructure, chemical segregation, and stress state of a weld leads to the potential for corrosion and cracking. Such failures have been experienced in NaK systems that have operated for times less than the eight year goal for the FSPS. For this reason, it was necessary to evaluate candidate weld techniques and expose welds to high-temperature, flowing NaK in a closed, closely controlled system. The goal of this project was to determine the optimum weld configuration for a NaK system that will withstand service for eight years under FSPS conditions. Since the most difficult weld to make and to evaluate is the tube to tube sheet weld in the intermediate heat exchangers, it was the focus of this research. A pumped loop of flowing NaK was fabricated for exposure of candidate weld specimens at temperatures of 600°C, the expected

  15. CONVECTION REACTOR

    DOEpatents

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

    1960-03-22

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

  16. Research reactors

    SciTech Connect

    Tonneson, L.C.; Fox, G.J.

    1996-04-01

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world`s research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted.

  17. Recovery of resources for advanced life support space applications: effect of retention time on biodegradation of two crop residues in a fed-batch, continuous stirred tank reactor

    NASA Technical Reports Server (NTRS)

    Strayer, R. F.; Finger, B. W.; Alazraki, M. P.; Cook, K.; Garland, J. L.

    2002-01-01

    Bioreactor retention time is a key process variable that will influence costs that are relevant to long distance space travel or long duration space habitation. However. little is known about the effects of this parameter on the microbiological treatment options that are being proposed for Advanced Life Support (ALS) systems. Two bioreactor studies were designed to examine this variable. In the first one, six retention times ranging from 1.3 to 21.3 days--were run in duplicate, 81 working-volume continuous stirred tank reactors (CSTR) that were fed ALS wheat residues. Ash-free dry weight loss, carbon mineralization, soluble TOC reduction, changes in fiber content (cellulose, hemicellulose, and lignin), bacterial numbers, and mineral recoveries were monitored. At short retention times--1.33 days--biodegradation was poor (total: 16-20%, cellulose - 12%, hemicellulose - 28%) but soluble TOC was decreased by 75-80% and recovery of major crop inorganic nutrients was adequate, except for phosphorus. A high proportion of the total bacteria (ca. 83%) was actively respiring. At the longest retention time tested, 21.3 days, biodegradation was good (total: 55-60%, cellulose ca. 70%, hemicellulose - ca. 55%) and soluble TOC was decreased by 80%. Recovery of major nutrients, except phosphorus, remained adequate. A very low proportion of total bacteria was actively respiring (ca. 16%). The second bioreactor study used potato residue to determine if even shorter retention times could be used (range 0.25-2.0 days). Although overall biodegradation deteriorated, the degradation of soluble TOC continued to be ca. 75%. We conclude that if the goal of ALS bioprocessing is maximal degradation of crop residues, including cellulose, then retention times of 10 days or longer will be needed. If the goal is to provide inorganic nutrients with the smallest volume/weight bioreactor possible, then a retention time of 1 day (or less) is sufficient.

  18. The Evaluation of Lithium Hydride for Use in a Space Nuclear Reactor Shield, Including a Historical Perspective

    SciTech Connect

    D. Poeth

    2005-12-09

    LiH was one of the five primary shield materials the NRPCT intended to develop (along with beryllium, boron carbide, tungsten, and water) for potential Prometheus application. It was also anticipated that {sup 10}B metal would be investigated for feasibility at a low level of effort. LiH historically has been selected as a low mass, neutron absorption material for space shields (Systems for Nuclear Auxiliary Power (SNAP), Topaz, SP-100). Initial NRPCT investigations did not produce convincing evidence that LiH was desirable or feasible for a Prometheus mission due to material property issues (primarily swelling and hydrogen cover gas containment), and related thermal design complexity. Furthermore, if mass limits allowed, an option to avoid use of LiH was being contemplated to lower development costs and associated risks. However, LiH remains theoretically the most efficient neutron shield material per unit mass, and, with sufficient testing and development, could be an optimal material choice for future flights.

  19. Nuclear reactor construction with bottom supported reactor vessel

    DOEpatents

    Sharbaugh, John E.

    1987-01-01

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

  20. 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)

  1. NERVA Reactor Based on NRX A1

    NASA Technical Reports Server (NTRS)

    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.

  2. COMPARTMENTED REACTOR FUEL ELEMENT

    DOEpatents

    Cain, F.M. Jr.

    1962-09-11

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

  3. REACTOR COOLING

    DOEpatents

    Quackenbush, C.F.

    1959-09-29

    A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

  4. NEUTRONIC REACTOR

    DOEpatents

    Metcalf, H.E.; Johnson, H.W.

    1961-04-01

    BS>A nuclear reactor incorporating fuel rods passing through a moderator and including tubes of a material of higher Thermal conductivity than the fuel in contact with the fuel is described. The tubes extend beyond the active portion of the reactor into contant with a fiuld coolant.

  5. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1958-04-22

    A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

  6. 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.

  7. NUCLEAR REACTOR

    DOEpatents

    Moore, R.V.; Bowen, J.H.; Dent, K.H.

    1958-12-01

    A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

  8. Nuclear reactor vessel fuel thermal insulating barrier

    DOEpatents

    Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

    2013-03-19

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

  9. Plasma core reactor applications

    NASA Technical Reports Server (NTRS)

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

    1976-01-01

    Analytical and experimental investigations were conducted to demonstrate the feasibility of fissioning uranium plasma core reactors and to characterize space and terrestrial applications for such reactors. Uranium hexafluoride fuel is injected into core cavities and confined away from the surface by argon buffer gas injected tangentially from the peripheral walls. Radiant heat transfer calculations were performed for a six-cavity reactor configuration. Axial working fluid channels are located along a fraction of each cavity peripheral wall. Results of calculations for outward-directed radiant energy fluxes corresponding to radiating temperatures of 2000 to 5000 K indicate total operating pressures from 80 to 650 atm, centerline temperatures from 6900 to 30,000 K, and total radiated powers from 25 to 2500 MW, respectively. Applications are described for this type of reactor such as (1) high-thrust, high specific impulse space propulsion, (2) highly efficient systems for generation of electricity, and (3) hydrogen or synthetic fuel production systems using the intense radiant energy fluxes.

  10. Cermet fuel reactors

    SciTech Connect

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

    1987-09-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.

  11. NUCLEAR REACTOR

    DOEpatents

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  12. NEUTRONIC REACTOR

    DOEpatents

    Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

    1959-03-24

    A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

  13. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.

    1960-04-01

    A nuclear reactor is described consisting of blocks of graphite arranged in layers, natural uranium bodies disposed in holes in alternate layers of graphite blocks, and coolant tubes disposed in the layers of graphite blocks which do not contain uranium.

  14. NUCLEAR REACTOR

    DOEpatents

    Sherman, J.; Sharbaugh, J.E.; Fauth, W.L. Jr.; Palladino, N.J.; DeHuff, P.G.

    1962-10-23

    A nuclear reactor incorporating seed and blanket assemblies is designed. Means are provided for obtaining samples of the coolant from the blanket assemblies and for varying the flow of coolant through the blanket assemblies. (AEC)

  15. Broad-Application Test Reactor

    SciTech Connect

    Motloch, C.G.

    1992-05-01

    This report is about a new, safe, and operationally efficient DOE reactor of nuclear research and testing proposed for the early to mid- 21st Century. Dubbed the Broad-Application Test Reactor (BATR), the proposed facility incorporates a multiple-application, multiple-mission design to support DOE programs such as naval reactors and space power and propulsion, as well as research in medical, science, isotope, and electronics arenas. DOE research reactors are aging, and implementing major replacement projects requires long lead times. Primary design drivers include safety, low risk, minimum operation cost, mission flexibility, waste minimization, and long life. Scientists and engineers at the Idaho National Engineering Laboratory are evaluating possible fuel forms, structural materials, reactor geometries, coolants, and moderators.

  16. NUCLEAR REACTOR

    DOEpatents

    Breden, C.R.; Dietrich, J.R.

    1961-06-20

    A water-soluble non-volatile poison may be introduced into a reactor to nullify excess reactivity. The poison is removed by passing a side stream of the water containing the soluble poison to an evaporation chamber. The vapor phase is returned to the reactor to decrease the concentration of soluble poison and the liquid phase is returned to increase the concentration of soluble poison.

  17. NUCLEAR REACTOR

    DOEpatents

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  18. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1959-07-14

    High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

  19. MEANS FOR COOLING REACTORS

    DOEpatents

    Wheeler, J.A.

    1957-11-01

    A design of a reactor is presented in which the fuel elements may be immersed in a liquid coolant when desired without the necessity of removing them from the reactor structure. The fuel elements, containing the fissionable material are in plate form and are disposed within spaced slots in a moderator material, such as graphite to form the core. Adjacent the core is a tank containing the liquid coolant. The fuel elements are mounted in spaced relationship on a rotatable shaft which is located between the core and the tank so that by rotation of the shaft the fuel elements may be either inserted in the slots in the core to sustain a chain reaction or immersed in the coolant.

  20. Neutronic reactor construction

    DOEpatents

    Huston, Norman E.

    1976-07-06

    1. A neutronic reactor comprising a moderator including horizontal layers formed of horizontal rows of graphite blocks, alternate layers of blocks having the rows extending in one direction, the remaining alternate layers having the rows extending transversely to the said one direction, alternate rows of blocks in one set of alternate layers having longitudinal ducts, the moderator further including slotted graphite tubes positioned in the ducts, the reactor further comprising an aluminum coolant tube positioned within the slotted tube in spaced relation thereto, bodies of thermal-neutron-fissionable material, and jackets enclosing the bodies and being formed of a corrosion-resistant material having a low neutron-capture cross section, the bodies and jackets being positioned within the coolant tube so that the jackets are spaced from the coolant tube.

  1. Reactor power system/spacecraft integration

    NASA Technical Reports Server (NTRS)

    Elms, R. V.

    1985-01-01

    The new national initiative in space reactor technology evaluation and development is strongly tied to mission applications and to spacecraft and space transportation system (STS) compatibility. This paper discusses the power system integration interfaces with potential using spacecraft and the STS, and the impact of these requirements on the design. The integration areas of interest are mechanical, thermal, electrical, attitude control, and mission environments. The mission environments include space vacuum, solar input, heat sink, space radiation, weapons effects, and reactor power system radiation environments. The natural, reactor, and weapons effects radiation must be evaluated and combined to define the design requirements for spacecraft electronic equipment.

  2. FOOD IRRADIATION REACTOR

    DOEpatents

    Leyse, C.F.; Putnam, G.E.

    1961-05-01

    An irradiation apparatus is described. It comprises a pressure vessel, a neutronic reactor active portion having a substantially greater height than diameter in the pressure vessel, an annular tank surrounding and spaced from the pressure vessel containing an aqueous indium/sup 1//sup 1//sup 5/ sulfate solution of approximately 600 grams per liter concentration, means for circulating separate coolants through the active portion and the space between the annular tank and the pressure vessel, radiator means adapted to receive the materials to be irradiated, and means for flowing the indium/sup 1//sup 1//sup 5/ sulfate solution through the radiator means.

  3. High flux reactor

    DOEpatents

    Lake, James A.; Heath, Russell L.; Liebenthal, John L.; DeBoisblanc, Deslonde R.; Leyse, Carl F.; Parsons, Kent; Ryskamp, John M.; Wadkins, Robert P.; Harker, Yale D.; Fillmore, Gary N.; Oh, Chang H.

    1988-01-01

    A high flux reactor is comprised of a core which is divided into two symetric segments housed in a pressure vessel. The core segments include at least one radial fuel plate. The spacing between the plates functions as a coolant flow channel. The core segments are spaced axially apart such that a coolant mixing plenum is formed between them. A channel is provided such that a portion of the coolant bypasses the first core section and goes directly into the mixing plenum. The outlet coolant from the first core segment is mixed with the bypass coolant resulting in a lower inlet temperature to the lower core segment.

  4. Comparison of 20-kWe space nuclear power systems: Impact of reactor temperature and technology upon system design and performances

    NASA Astrophysics Data System (ADS)

    Carre, F.; Proust, E.; Chaudourne, S.; Keirle, P.; Tilliette, Z.; Trouve, J.; Vrillon, B.

    The design studies dedicated to three concepts of 20 kWe turboelectric power systems, purposely selected to cover a wide range of technologies applicable to low power systems are presented. The system performances will be analyzed as a function of reactor temperature and of the relevant technology.

  5. NEUTRONIC REACTOR

    DOEpatents

    Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

    1958-09-01

    This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

  6. POWER REACTOR

    DOEpatents

    Zinn, W.H.

    1958-07-01

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

  7. REACTOR CONTROL

    DOEpatents

    Fortescue, P.; Nicoll, D.

    1962-04-24

    A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

  8. NUCLEAR REACTOR

    DOEpatents

    Christy, R.F.

    1958-07-15

    A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

  9. NEUTRONIC REACTORS

    DOEpatents

    Wigner, E.P.; Young, G.J.

    1958-10-14

    A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

  10. Catalytic reactor

    SciTech Connect

    Aaron, Timothy Mark; Shah, Minish Mahendra; Jibb, Richard John

    2009-03-10

    A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

  11. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Szilard, L.

    1957-09-24

    Reactors of the type employing plates of natural uranium in a moderator are discussed wherein the plates are um-formly disposed in parallel relationship to each other thereby separating the moderator material into distinct and individual layers. Each plate has an uninterrupted sunface area substantially equal to the cross-sectional area of the active portion of the reactor, the particular size of the plates and the volume ratio of moderator to uranium required to sustain a chain reaction being determinable from the known purity of these materials and other characteristics such as the predictable neutron losses due to the formation of radioactive elements of extremely high neutron capture cross section.

  12. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  13. Nuclear reactor building

    DOEpatents

    Gou, Perng-Fei; Townsend, Harold E.; Barbanti, Giancarlo

    1994-01-01

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

  14. Fossil fuel furnace reactor

    DOEpatents

    Parkinson, William J.

    1987-01-01

    A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

  15. Nuclear reactor building

    DOEpatents

    Gou, P.F.; Townsend, H.E.; Barbanti, G.

    1994-04-05

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed there above. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define there between an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin. 4 figures.

  16. REACTOR UNLOADING MEANS

    DOEpatents

    Cooper, C.M.

    1957-08-20

    A means for remotely unloading irradiated fuel slugs from a neutronic reactor core and conveying them to a remote storage tank is reported. The means shown is specifically adapted for use with a reactor core wherein the fuel slugs are slidably held in end to end abutting relationship in the horizontal coolant flow tubes, the slugs being spaced from tae internal walls of the tubes to permit continuous circulation of coolant water therethrough. A remotely operated plunger at the charging ends of the tubes is used to push the slugs through the tubes and out the discharge ends into a special slug valve which transfers the slug to a conveying tube leading into a storage tank. Water under pressure is forced through the conveying tube to circulate around the slug to cool it and also to force the slug through the conveving tube into the storage tank. The slug valve and conveying tube are shielded to prevent amy harmful effects caused by the radioactive slug in its travel from the reactor to the storage tank. With the disclosed apparatus, all the slugs in the reactor core can be conveyed to the storage tank shortly after shutdown by remotely located operating personnel.

  17. Neutron shielding panels for reactor pressure vessels

    SciTech Connect

    Singleton, Norman R

    2011-11-22

    In a nuclear reactor neutron panels varying in thickness in the circumferential direction are disposed at spaced circumferential locations around the reactor core so that the greatest radial thickness is at the point of highest fluence with lesser thicknesses at adjacent locations where the fluence level is lower. The neutron panels are disposed between the core barrel and the interior of the reactor vessel to maintain radiation exposure to the vessel within acceptable limits.

  18. Neutronic reactor

    DOEpatents

    Wende, Charles W. J.

    1976-08-17

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.

  19. Neutronic reactor

    DOEpatents

    Carleton, John T.

    1977-01-25

    A graphite-moderated nuclear reactor includes channels between blocks of graphite and also includes spacer blocks between adjacent channeled blocks with an axis of extension normal to that of the axis of elongation of the channeled blocks to minimize changes in the physical properties of the graphite as a result of prolonged neutron bombardment.

  20. Sonochemical Reactors.

    PubMed

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation.

  1. NEUTRONIC REACTORS

    DOEpatents

    Anderson, H.L.

    1958-10-01

    The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

  2. NEUTRONIC REACTOR

    DOEpatents

    Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.

    1959-10-27

    BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

  3. Role of nuclear reactors in future military satellites

    SciTech Connect

    Buden, D.; Angelo, J.A. Jr.

    1982-01-01

    Future military capabilities will be profoundly influenced by emerging Shuttle Era space technology. Regardless of the specific direction or content of tomorrow's military space program, it is clear that advanced space transportation systems, orbital support facilities, and large-capacity power subsystems will be needed to create the generally larger, more sophisticated military space systems of the future. This paper explores the critical role that space nuclear reactors should play in America's future space program and reviews the current state of nuclear reactor power plant technology. Space nuclear reactor technologies have the potential of satisfying power requirements ranging from 10 kW/sub (e)/ to 100 MW/sub (e)/.

  4. Computational Modeling of Multiphase Reactors.

    PubMed

    Joshi, J B; Nandakumar, K

    2015-01-01

    Multiphase reactors are very common in chemical industry, and numerous review articles exist that are focused on types of reactors, such as bubble columns, trickle beds, fluid catalytic beds, etc. Currently, there is a high degree of empiricism in the design process of such reactors owing to the complexity of coupled flow and reaction mechanisms. Hence, we focus on synthesizing recent advances in computational and experimental techniques that will enable future designs of such reactors in a more rational manner by exploring a large design space with high-fidelity models (computational fluid dynamics and computational chemistry models) that are validated with high-fidelity measurements (tomography and other detailed spatial measurements) to provide a high degree of rigor. Understanding the spatial distributions of dispersed phases and their interaction during scale up are key challenges that were traditionally addressed through pilot scale experiments, but now can be addressed through advanced modeling.

  5. NEUTRONIC REACTOR CORE INSTRUMENT

    DOEpatents

    Mims, L.S.

    1961-08-22

    A multi-purpose instrument for measuring neutron flux, coolant flow rate, and coolant temperature in a nuclear reactor is described. The device consists essentially of a hollow thimble containing a heat conducting element protruding from the inner wall, the element containing on its innermost end an amount of fissionsble materinl to function as a heat source when subjected to neutron flux irradiation. Thermocouple type temperature sensing means are placed on the heat conducting element adjacent the fissionable material and at a point spaced therefrom, and at a point on the thimble which is in contact with the coolant fluid. The temperature differentials measured between the thermocouples are determinative of the neutron flux, coolant flow, and temperature being measured. The device may be utilized as a probe or may be incorporated in a reactor core. (AE C)

  6. AIR COOLED NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Szilard, L.

    1958-05-27

    A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

  7. NEUTRONIC REACTOR

    DOEpatents

    Wade, E.J.

    1958-09-16

    This patent relates to a reflector means for a neutronic reactor. A reflector comprised of a plurality of vertically movable beryllium control members is provided surrounding the sides of the reactor core. An absorber of fast neutrons comprised of natural uramum surrounds the reflector. An absorber of slow neutrons surrounds the absorber of fast neutrons and is formed of a plurality of beryllium blocks having natural uranium members distributcd therethrough. in addition, a movable body is positioned directly below the core and is comprised of a beryllium reflector and an absorbing member attached to the botiom thereof, the absorbing member containing a substance selected from the goup consisting of natural urantum and Th/sup 232/.

  8. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1959-12-15

    A reactor which is particularly adapted tu serve as a heat source for a nuclear powered alrcraft or rocket is described. The core of this reactor consists of a porous refractory modera;or body which is impregnated with fissionable nuclei. The core is designed so that its surface forms tapered inlet and outlet ducts which are separated by the porous moderator body. In operation a gaseous working fluid is circulated through the inlet ducts to the surface of the moderator, enters and passes through the porous body, and is heated therein. The hot gas emerges into the outlet ducts and is available to provide thrust. The principle advantage is that tremendous quantities of gas can be quickly heated without suffering an excessive pressure drop.

  9. REACTOR UNLOADING

    DOEpatents

    Leverett, M.C.

    1958-02-18

    This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

  10. NUCLEAR REACTORS

    DOEpatents

    Koch, L.J.; Rice, R.E. Jr.; Denst, A.A.; Rogers, A.J.; Novick, M.

    1961-12-01

    An active portion assembly for a fast neutron reactor is described wherein physical distortions resulting in adverse changes in the volume-to-mass ratio are minimized. A radially expandable locking device is disposed within a cylindrical tube within each fuel subassembly within the active portion assembly, and clamping devices expandable toward the center of the active portion assembly are disposed around the periphery thereof. (AEC)

  11. Neutronic reactor

    DOEpatents

    Lewis, Warren R.

    1978-05-30

    A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.

  12. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1961-01-24

    A core structure for neutronic reactors adapted for the propulsion of aircraft and rockets is offered. The core is designed for cooling by gaseous media, and comprises a plurality of hollow tapered tubular segments of a porous moderating material impregniated with fissionable fuel nested about a common axis. Alternate ends of the segments are joined. In operation a coolant gas passes through the porous structure and is heated.

  13. 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.

  14. Reactors for nuclear electric propulsion

    SciTech Connect

    Buden, D.; Angelo, J.A. Jr.

    1981-01-01

    Propulsion is the key to space exploitation and power is the key to propulsion. This paper examines the role of nuclear fission reactors as the primary power source for high specific impulse electric propulsion systems for space missions of the 1980s and 1990s. Particular mission applications include transfer to and a reusable orbital transfer vehicle from low-Earth orbit to geosynchronous orbit, outer planet exploration and reconnaissance missions, and as a versatile space tug supporting lunar resource development. Nuclear electric propulsion is examined as an indispensable component in space activities of the next two decades.

  15. NUCLEAR REACTOR FUEL ELEMENT

    DOEpatents

    Anderson, W.F.; Tellefson, D.R.; Shimazaki, T.T.

    1962-04-10

    A plate type fuel element which is particularly useful for organic cooled reactors is described. Generally, the fuel element comprises a plurality of fissionable fuel bearing plates held in spaced relationship by a frame in which the plates are slidably mounted in grooves. Clearance is provided in the grooves to allow the plates to expand laterally. The plates may be rigidly interconnected but are floatingly supported at their ends within the frame to allow for longi-tudinal expansion. Thus, this fuel element is able to withstand large temperature differentials without great structural stresses. (AEC)

  16. SODIUM DEUTERIUM REACTOR

    DOEpatents

    Oppenheimer, E.D.; Weisberg, R.A.

    1963-02-26

    This patent relates to a barrier system for a sodium heavy water reactor capable of insuring absolute separation of the metal and water. Relatively cold D/sub 2/O moderator and reflector is contained in a calandria into which is immersed the fuel containing tubes. The fuel elements are cooled by the sodium which flows within the tubes and surrounds the fuel elements. The fuel containing tubes are surrounded by concentric barrier tubes forming annular spaces through which pass inert gases at substantially atmospheric pressure. Header rooms above and below the calandria are provided for supplying and withdrawing the sodium and inert gases in the calandria region. (AEC)

  17. REACTOR COOLANT TUBE SEAL

    DOEpatents

    Morris, W.J.

    1958-12-01

    A plle-flattenlng control element and a fluid seal therefore to permit movement of the element into a liquld contnining region of a neutronlc reactor are described. The device consists of flattened, thin-walled aluminum tubing contalnlng a uniform mixture of thermal neutron absorbing material, and a number of soft rubber closures for the process tubes, having silts capable of passing the flattened elements therethrough, but effectively sealing the process tubes against fluld leaknge by compression of the rubber. The flattened tubing is sufficiently flexible to enable it to conform to the configuratlon of the annular spacing surrounding the fuel elements ln the process tubes.

  18. Fuel elements of research reactor CM

    SciTech Connect

    Kozlov, A.V.; Morozov, A.V.; Vatulin, A.V.; Ershov, S.A.

    2013-07-01

    In 1961 the CM research reactor was commissioned at the Research Institute of Atomic Reactors (Dimitrovgrad, Russia), it was intended to carry on investigations and the production of transuranium nuclides. The reactor is of a tank type. Original fuel assembly contained plate fuels that were spaced with vanes and corrugated bands. Nickel was used as a cladding material, fuel meat was produced from UO{sub 2} + electrolytic nickel composition. Fuel plates have been replaced by self-spacing cross-shaped dispersion fuels clad in stainless steel. In 2005 the reactor was updated. The purpose of this updating was to increase the quantity of irradiation channels in the reactor core and to improve the neutron balance. The updating was implemented at the expense of 20 % reduction in the quantity of fuel elements in the core which released a space for extra channels and decreased the mass of structural materials in the core. The updated reactor is loaded with modified standard fuel elements with 20 % higher uranium masses. At the same time stainless steel in fuel assembly shrouds was substituted by zirconium alloy. Today in progress are investigations and work to promote the second stage of reactor updating that involve developments of cross-shaped fuel elements having low neutron absorption matrix materials. This article gives an historical account of the design and main technical changes that occurred for the CM reactor since its commissioning.

  19. PID Control Effectiveness for Surface Reactor Concepts

    SciTech Connect

    Dixon, David D.; Marsh, Christopher L.; Poston, David I.

    2007-01-30

    Control of space and surface fission reactors should be kept as simple as possible, because of the need for high reliability and the difficulty to diagnose and adapt to control system failures. Fortunately, compact, fast-spectrum, externally controlled reactors are very simple in operation. In fact, for some applications it may be possible to design low-power surface reactors without the need for any reactor control after startup; however, a simple proportional, integral, derivative (PID) controller can allow a higher performance concept and add more flexibility to system operation. This paper investigates the effectiveness of a PID control scheme for several anticipated transients that a surface reactor might experience. To perform these analyses, the surface reactor transient code FRINK was modified to simulate control drum movements based on bulk coolant temperature.

  20. Nuclear Reactors. Revised.

    ERIC Educational Resources Information Center

    Hogerton, John F.

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

  1. Pressurized water reactor flow skirt apparatus

    SciTech Connect

    Kielb, John F.; Schwirian, Richard E.; Lee, Naugab E.; Forsyth, David R.

    2016-04-05

    A pressurized water reactor vessel having a flow skirt formed from a perforated cylinder structure supported in the lower reactor vessel head at the outlet of the downcomer annulus, that channels the coolant flow through flow holes in the wall of the cylinder structure. The flow skirt is supported at a plurality of circumferentially spaced locations on the lower reactor vessel head that are not equally spaced or vertically aligned with the core barrel attachment points, and the flow skirt employs a unique arrangement of hole patterns that assure a substantially balanced pressure and flow of the coolant over the entire underside of the lower core support plate.

  2. REACTOR COMPONETN

    DOEpatents

    Creutz, E.C.

    1959-10-27

    A reactor fuel element comprised of a slug of fissionable material disposed in a sheath of corrosion resistantmaterial is described. The sheath is in the form of a tubular container closed at one end and is in tight-fitting engagement with the peripheral sunface of the slug. An inner cap is insented into the open end of the sheath against the slug, which end is then bent around the inner cap and welded thereto. An outer cap is then welded around its peripheny to the bent portion of the container.

  3. Photocatalytic reactor

    DOEpatents

    Bischoff, B.L.; Fain, D.E.; Stockdale, J.A.D.

    1999-01-19

    A photocatalytic reactor is described for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane. 4 figs.

  4. Application of k- and q-space encoding NMR techniques on granular media in a 3D model fluidized bed reactor

    NASA Astrophysics Data System (ADS)

    Harms, Silke; Stapf, Siegfried; Blümich, Bernhard

    2006-02-01

    A combination of PFG-NMR imaging and velocity encoding methods was applied to investigate the dynamic behavior of a bed of poppy seeds subjected to air flow, representing a model setup for fluidized bed reactors. The particle motion is described both from a statistical point of view, by determining propagators and dispersion coefficients representing an average over the whole bed volume, as well as combined with spatial resolution by generating velocity maps. Velocity images of different horizontal slices in the bed confirm the notion of a toroidal particle flow pattern inside the shallow granular bed. Despite the need of considerable averaging due to the random motion of the relatively few particles in the bed, quantitative velocity images and statistical information about the random particle motion can be obtained from monitoring the fluid component in the seeds by conventional spin-echo techniques.

  5. A fast spectrum dual path flow cermet reactor

    SciTech Connect

    Anghaie, S.; Feller, G.J. ); Peery, S.D.; Parsley, R.C. )

    1993-01-15

    A cermet fueled, dual path fast reactor for space nuclear propulsion applications is conceptually designed. The reactor utilizes an outer annulus core and an inner cylindrical core with radial and axial reflector. The dual path flow minimizes the impact of power peaking near the radial reflector. Basic neutronics and core design aspects of the reactor are discussed. The dual path reactor is integrated into a 25000 lbf thrust nuclear rocket.

  6. Hybrid adsorptive membrane reactor

    NASA Technical Reports Server (NTRS)

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

    2011-01-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  7. Hybrid adsorptive membrane reactor

    SciTech Connect

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  8. ATFSR: a small torsatron reactor

    SciTech Connect

    Houlberg, W.A.; Lacatski, J.T.; Uckan, N.A.

    1985-01-01

    A small (average minor radius anti a approx. = 1 m), moderate-aspect-ratio torsatron reactor based on the Advanced Toroidal Facility (ATF) is proposed as a starting point for improved stellarator reactor designs. The major limitation of the compact size is the lack of space under the helical coils for the blanket and shield. Neoclassical confinement models for helically trapped particles show that a large electric potential (radial electric field) is necessary to achieve ignition in a device of this size, although high-Q operation is still attainable with more modest potentials.

  9. Control Means for Reactor

    DOEpatents

    Manley, J. H.

    1961-06-27

    An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

  10. Thermionic Reactor Design Studies

    SciTech Connect

    Schock, Alfred

    1994-06-01

    During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic losses in the electrodes of full

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

    DOEpatents

    Hunsbedt, Anstein; Busboom, Herbert J.

    1991-01-01

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

  12. NEUTRONIC REACTOR

    DOEpatents

    Daniels, F.

    1962-12-18

    A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)

  13. NEUTRONIC REACTOR

    DOEpatents

    Stewart, H.B.

    1958-12-23

    A nuclear reactor of the type speclfically designed for the irradiation of materials is discussed. In this design a central cyllndrical core of moderating material ls surrounded by an active portlon comprlsed of an annular tank contalning fissionable material immersed ln a liquid moderator. The active portion ls ln turn surrounded by a reflector, and a well ls provided in the center of the core to accommodate the materlals to be irradiated. The over-all dimensions of the core ln at least one plane are equal to or greater than twice the effective slowing down length and equal to or less than twlce the effective diffuslon length for neutrons in the core materials.

  14. Nuclear reactor

    DOEpatents

    Yant, Howard W.; Stinebiser, Karl W.; Anzur, Gregory C.

    1977-01-01

    A nuclear reactor, particularly a liquid-metal breeder reactor, whose upper internals include outlet modules for channeling the liquid-metal coolant from selected areas of the outlet of the core vertically to the outlet plenum. The modules are composed of a highly-refractory, high corrosion-resistant alloy, for example, INCONEL-718. Each module is disposed to confine and channel generally vertically the coolant emitted from a subplurality of core-component assemblies. Each module has a grid with openings, each opening disposed to receive the coolant from an assembly of the subplurality. The grid in addition serves as a holdown for the assemblies of the corresponding subplurality preventing their excessive ejection upwardly from the core. In the region directly over the core the outlet modules are of such peripheral form that they nest forming a continuum over the core-component assemblies whose outlet coolant they confine. Each subassembly includes a chimney which confines the coolant emitted by its corresponding subassemblies to generally vertical flow between the outlet of the core and the outlet plenum. Each subplurality of assemblies whose emitted coolant is confined by an outlet module includes assemblies which emit lower-temperature coolant, for example, a control-rod assembly, or fertile assemblies, and assemblies which emit coolant of substantially higher temperature, for example, fuel-rod assemblies. The coolants of different temperatures are mixed in the chimneys reducing the effect of stripping (hot-cold temperature fluctuations) on the remainder of the upper internals which are composed typically of AISI-304 or AISI-316 stainless steel.

  15. Control system for a small fission reactor

    DOEpatents

    Burelbach, James P.; Kann, William J.; Saiveau, James G.

    1986-01-01

    A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired. In another embodiment, a plurality of flexible hollow tubes each containing a neutron absorber are positioned adjacent to one another in spaced relation around the periphery of the reactor vessel and inside the outer neutron reflector with reactivity controlled by the extension and compression of all or some of the coiled hollow tubes. Yet another embodiment of the invention envisions the neutron reflector in the form of an expandable coil spring positioned in an annular space between the reactor vessel and an outer neutron absorbing structure for controlling the neutron flux reflected back into the reactor vessel.

  16. Reactor and method of operation

    DOEpatents

    Wheeler, John A.

    1976-08-10

    A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor.

  17. Neutronic reactor thermal shield

    DOEpatents

    Lowe, Paul E.

    1976-06-15

    1. The combination with a plurality of parallel horizontal members arranged in horizontal and vertical rows, the spacing of the members in all horizontal rows being equal throughout, the spacing of the members in all vertical rows being equal throughout; of a shield for a nuclear reactor comprising two layers of rectangular blocks through which the members pass generally perpendicularly to the layers, each block in each layer having for one of the members an opening equally spaced from vertical sides of the block and located closer to the top of the block than the bottom thereof, whereby gravity tends to make each block rotate about the associated member to a position in which the vertical sides of the block are truly vertical, the openings in all the blocks of one layer having one equal spacing from the tops of the blocks, the openings in all the blocks of the other layer having one equal spacing from the tops of the blocks, which spacing is different from the corresponding spacing in the said one layer, all the blocks of both layers having the same vertical dimension or length, the blocks of both layers consisting of relatively wide blocks and relatively narrow blocks, all the narrow blocks having the same horizontal dimension or width which is less than the horizontal dimension or width of the wide blocks, which is the same throughout, each layer consisting of vertical rows of narrow blocks and wide blocks alternating with one another, each vertical row of narrow blocks of each layer being covered by a vertical row of wide blocks of the other layer which wide blocks receive the same vertical row of members as the said each vertical row of narrow blocks, whereby the rectangular perimeters of each block of each layer is completely out of register with that of each block in the other layer.

  18. Launch of Russian reactor postponed

    SciTech Connect

    Not Available

    1993-02-05

    Astronomers and weapons scientists seemed heated on a collision course a few months ago over the military's plans to send a Russian nuclear reactor into space. But an agreement reached in late January has prevented a pile-up, at least for 6 months. The astronomers, led by Donald Lamb of the University of Chicago, were objecting to plans by the Strategic Defense Initiative Office (SDIO) to launch Topaz 2, an experimental Russian nuclear reactor, arguing that rogue particles from it might ruin sensitive gamma ray experiments. The reactor is designed to propel itself in space with a jet of xenon ions. One worry was that leaking gamma rays and positrons, which can travel in the earth's magnetic field and pop up in the darndest places, might cause false signals in gamma ray monitors (Science, 18 December 1992, p. 1878). The worry has abated now that SDI officials will postpone choosing a rocket and mission altitutde for Topaz 2 for 6 months, while experts study how its emissions at various altitudes might affect instruments aboard the Gamma Ray Observatory and other satellites. In effect, the SDIO has agreed to an environmental impact study for space, following an unusual meeting organized by former Russian space official Roald Sagdeev at the University of Maryland on 19 January. There the Russian designers of Topaz 2, its new owners at the SDIO, and critics in the astronomy community achieved common ground: that more study was needed.

  19. NASA Reactor Facility Hazards Summary. Volume 1

    NASA Technical Reports Server (NTRS)

    1959-01-01

    The Lewis Research Center of the National Aeronautics and Space Administration proposes to build a nuclear research reactor which will be located in the Plum Brook Ordnance Works near Sandusky, Ohio. The purpose of this report is to inform the Advisory Committee on Reactor Safeguards of the U. S. Atomic Energy Commission in regard to the design Lq of the reactor facility, the characteristics of the site, and the hazards of operation at this location. The purpose of this research reactor is to make pumped loop studies of aircraft reactor fuel elements and other reactor components, radiation effects studies on aircraft reactor materials and equipment, shielding studies, and nuclear and solid state physics experiments. The reactor is light water cooled and moderated of the MTR-type with a primary beryllium reflector and a secondary water reflector. The core initially will be a 3 by 9 array of MTR-type fuel elements and is designed for operation up to a power of 60 megawatts. The reactor facility is described in general terms. This is followed by a discussion of the nuclear characteristics and performance of the reactor. Then details of the reactor control system are discussed. A summary of the site characteristics is then presented followed by a discussion of the larger type of experiments which may eventually be operated in this facility. The considerations for normal operation are concluded with a proposed method of handling fuel elements and radioactive wastes. The potential hazards involved with failures or malfunctions of this facility are considered in some detail. These are examined first from the standpoint of preventing them or minimizing their effects and second from the standpoint of what effect they might have on the reactor facility staff and the surrounding population. The most essential feature of the design for location at the proposed site is containment of the maximum credible accident.

  20. SIMPLIFIED SODIUM GRAPHITE REACTOR SYSTEM

    DOEpatents

    Dickinson, R.W.

    1963-03-01

    This patent relates to a nuclear power reactor comprising a reactor vessel, shielding means positioned at the top of said vessel, means sealing said reactor vessel to said shielding means, said vessel containing a quantity of sodium, a core tank, unclad graphite moderator disposed in said tank, means including a plurality of process tubes traversing said tank for isolating said graphite from said sodium, fuel elements positioned in said process tubes, said core tank being supported in spaced relation to the walls and bottom of said reactor vessel and below the level of said sodium, neutron shielding means positioned adjacent said core tank between said core tank and the walls of said vessel, said neutron shielding means defining an annuiar volume adjacent the inside wall of said reactor vessel, inlet plenum means below said core tank for providing a passage between said annular volume and said process tubes, heat exchanger means removably supported from the first-named shielding means and positioned in said annular volume, and means for circulating said sodium over said neutron shielding means down through said heat exchanger, across said inlet plenum and upward through said process tubes, said last-named means including electromagnetic pumps located outside said vessel and supported on said vessel wall between said heat exchanger means and said inlet plenum means. (AEC)

  1. Reactor safety method

    DOEpatents

    Vachon, Lawrence J.

    1980-03-11

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

  2. Reactor Physics Methods and Analysis Capabilities in SCALE

    SciTech Connect

    DeHart, Mark D; Bowman, Stephen M

    2011-01-01

    The TRITON sequence of the SCALE code system provides a powerful, robust, and rigorous approach for performing reactor physics analysis. This paper presents a detailed description of TRITON in terms of its key components used in reactor calculations. The ability to accurately predict the nuclide composition of depleted reactor fuel is important in a wide variety of applications. These applications include, but are not limited to, the design, licensing, and operation of commercial/research reactors and spent-fuel transport/storage systems. New complex design projects such as next-generation power reactors and space reactors require new high-fidelity physics methods, such as those available in SCALE/TRITON, that accurately represent the physics associated with both evolutionary and revolutionary reactor concepts as they depart from traditional and well-understood light water reactor designs.

  3. Reactor Physics Methods and Analysis Capabilities in SCALE

    SciTech Connect

    Mark D. DeHart; Stephen M. Bowman

    2011-05-01

    The TRITON sequence of the SCALE code system provides a powerful, robust, and rigorous approach for performing reactor physics analysis. This paper presents a detailed description of TRITON in terms of its key components used in reactor calculations. The ability to accurately predict the nuclide composition of depleted reactor fuel is important in a wide variety of applications. These applications include, but are not limited to, the design, licensing, and operation of commercial/research reactors and spent-fuel transport/storage systems. New complex design projects such as next-generation power reactors and space reactors require new high-fidelity physics methods, such as those available in SCALE/TRITON, that accurately represent the physics associated with both evolutionary and revolutionary reactor concepts as they depart from traditional and well-understood light water reactor designs.

  4. Fluidized bed coal combustion reactor

    NASA Technical Reports Server (NTRS)

    Moynihan, P. I.; Young, D. L. (Inventor)

    1981-01-01

    A fluidized bed coal reactor includes a combination nozzle-injector ash-removal unit formed by a grid of closely spaced open channels, each containing a worm screw conveyor, which function as continuous ash removal troughs. A pressurized air-coal mixture is introduced below the unit and is injected through the elongated nozzles formed by the spaces between the channels. The ash build-up in the troughs protects the worm screw conveyors as does the cooling action of the injected mixture. The ash layer and the pressure from the injectors support a fluidized flame combustion zone above the grid which heats water in boiler tubes disposed within and/or above the combustion zone and/or within the walls of the reactor.

  5. Passive cooling safety system for liquid metal cooled nuclear reactors

    DOEpatents

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

    1991-01-01

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

  6. Indirect passive cooling system for liquid metal cooled nuclear reactors

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1990-01-01

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

  7. Zero Power Reactor Simulation

    SciTech Connect

    2010-01-01

    Ever wanted to see a nuclear reactor core in action? Here's a detailed simulation of the Zero Power Reactor experiment, run by Argonne's unique "UNIC" code. Here, we use VisIt to visualize a numerical model of the ZPR-6 Assembly 6a experiment simulated using the Argonne UNIC code. 0:00-0:06: The fuel and other plates are dropped into an example drawer, and the drawer is inserted into the matrix tube. 0:06-0:21: The two matrix halves are brought together to make a critical (self-sustaining) assembly. 0:21-0:35: The fission power is revealed to be centralized in the thin, enriched Uranium plates. 0:35-0:53: Returning to our example drawer, we show the detailed local plate powers and their relation to the drawer composition. 0:53-1:11: Our model shows that each plate can have widely-varying local changes in the space and energy neutron densities. Read more at http://www.anl.gov/Media_Center/News/2010/news100121.html

  8. REACTOR MODERATOR STRUCTURE

    DOEpatents

    Fraas, A.P.; Tudor, J.J.

    1963-08-01

    An improved moderator structure for nuclear reactors consists of moderator blocks arranged in horizontal layers to form a multiplicity of vertically stacked columns of blocks. The blocks in each vertical column are keyed together, and a ceramic grid is disposed between each horizontal layer of blocks. Pressure plates cover- the lateral surface of the moderator structure in abutting relationship with the peripheral terminal lengths of the ceramic grids. Tubular springs are disposed between the pressure plates and a rigid external support. The tubular springs have their axes vertically disposed to facilitate passage of coolant gas through the springs and are spaced apart a selected distance such that at sonae preselected point of spring deflection, the sides of the springs will contact adjacent springs thereby causing a large increase in resistance to further spring deflection. (AEC)

  9. Mechanical cutting of irradiated reactor internal components

    SciTech Connect

    Anderson, Michael G.

    2008-01-15

    Mechanical cutting methods to volume reduce and package reactor internal components are now a viable solution for stakeholders challenged with the retirement of first generation nuclear facilities. The recent completion of the removal of the Reactor Vessel Internals (RVI) from within the Sacramento Municipal Utility District's (SMUD) Rancho Seco Nuclear Power Plant demonstrates that unlike previous methods, inclusive of plasma arc and abrasive water-jet cutting, mechanical cutting minimizes exposure to workers, costly water cleanup, and excessive secondary waste generation. Reactor internal components were segmented, packaged, and removed from the reactor building for shipment or storage, allowing the reactor cavity to be drained and follow-on reactor segmentation activities to proceed in the dry state. Area exposure rates at the work positions during the segmentation process were generally 1 mR per hr. Radiological exposure documented for the underwater segmentation processes totaled 13 person rem. The reactor internals weighing 343,000 pounds were segmented into over 200 pieces for maximum shipping package efficiency and produced 5,600 lb of stainless steel chips and shavings which were packaged in void spaces of existing disposal containers, therefore creating no additional disposal volume. Because no secondary waste was driven into suspension in the reactor cavity water, the water was free released after one pass through a charcoal bed and ion exchange filter system. Mechanical cutting techniques are capable of underwater segmentation of highly radioactive components on a large scale. This method minimized radiological exposure and costly water cleanup while creating no secondary waste.

  10. STEAM STIRRED HOMOGENEOUS NUCLEAR REACTOR

    DOEpatents

    Busey, H.M.

    1958-06-01

    A homogeneous nuclear reactor utilizing a selfcirculating liquid fuel is described. The reactor vessel is in the form of a vertically disposed tubular member having the lower end closed by the tube walls and the upper end closed by a removal fianged assembly. A spherical reaction shell is located in the lower end of the vessel and spaced from the inside walls. The reaction shell is perforated on its lower surface and is provided with a bundle of small-diameter tubes extending vertically upward from its top central portion. The reactor vessel is surrounded in the region of the reaction shell by a neutron reflector. The liquid fuel, which may be a solution of enriched uranyl sulfate in ordinary or heavy water, is mainiained at a level within the reactor vessel of approximately the top of the tubes. The heat of the reaction which is created in the critical region within the spherical reaction shell forms steam bubbles which more upwardly through the tubes. The upward movement of these bubbles results in the forcing of the liquid fuel out of the top of these tubes, from where the fuel passes downwardly in the space between the tubes and the vessel wall where it is cooled by heat exchangers. The fuel then re-enters the critical region in the reaction shell through the perforations in the bottom. The upper portion of the reactor vessel is provided with baffles to prevent the liquid fuel from splashing into this region which is also provided with a recombiner apparatus for recombining the radiolytically dissociated moderator vapor and a control means.

  11. Heterogeneous Recycling in Fast Reactors

    SciTech Connect

    Forget, Benoit; Pope, Michael; Piet, Steven J.; Driscoll, Michael

    2012-07-30

    Current sodium fast reactor (SFR) designs have avoided the use of depleted uranium blankets over concerns of creating weapons grade plutonium. While reducing proliferation risks, this restrains the reactor design space considerably. This project will analyze various blanket and transmutation target configurations that could broaden the design space while still addressing the non-proliferation issues. The blanket designs will be assessed based on the transmutation efficiency of key minor actinide (MA) isotopes and also on mitigation of associated proliferation risks. This study will also evaluate SFR core performance under different scenarios in which depleted uranium blankets are modified to include minor actinides with or without moderators (e.g. BeO, MgO, B4C, and hydrides). This will be done in an effort to increase the sustainability of the reactor and increase its power density while still offering a proliferation resistant design with the capability of burning MA waste produced from light water reactors (LWRs). Researchers will also analyze the use of recycled (as opposed to depleted) uranium in the blankets. The various designs will compare MA transmutation efficiency, plutonium breeding characteristics, proliferation risk, shutdown margins and reactivity coefficients with a current reference sodium fast reactor design employing homogeneous recycling. The team will also evaluate the out-of-core accumulation and/or burn-down rates of MAs and plutonium isotopes on a cycle-by-cycle basis. This cycle-by-cycle information will be produced in a format readily usable by the fuel cycle systems analysis code, VISION, for assessment of the sustainability of the deployment scenarios.

  12. CHIMNEY FOR BOILING WATER REACTOR

    DOEpatents

    Petrick, M.

    1961-08-01

    A boiling-water reactor is described which has vertical fuel-containing channels for forming steam from water. Risers above the channels increase the head of water radially outward, whereby water is moved upward through the channels with greater force. The risers are concentric and the radial width of the space between them is somewhat small. There is a relatively low rate of flow of water up through the radially outer fuel-containing channels, with which the space between the risers is in communication. (AE C)

  13. Tokamak reactor studies

    SciTech Connect

    Baker, C.C.

    1981-01-01

    This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features.

  14. Hybrid plasmachemical reactor

    SciTech Connect

    Lelevkin, V. M. Smirnova, Yu. G.; Tokarev, A. V.

    2015-04-15

    A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

  15. Attrition reactor system

    SciTech Connect

    Scott, Charles D.; Davison, Brian H.

    1993-01-01

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  16. Attrition reactor system

    SciTech Connect

    Scott, C.D.; Davison, B.H.

    1993-09-28

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

  17. NEUTRONIC REACTOR POWER PLANT

    DOEpatents

    Metcalf, H.E.

    1962-12-25

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

  18. Period meter for reactors

    DOEpatents

    Rusch, Gordon K.

    1976-01-06

    An improved log N amplifier type nuclear reactor period meter with reduced probability for noise-induced scrams is provided. With the reactor at low power levels a sampling circuit is provided to determine the reactor period by measuring the finite change in the amplitude of the log N amplifier output signal for a predetermined time period, while at high power levels, differentiation of the log N amplifier output signal provides an additional measure of the reactor period.

  19. NUCLEAR REACTOR CONTROL SYSTEM

    DOEpatents

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

    1959-11-01

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

  20. Efficient Silicon Reactor

    NASA Technical Reports Server (NTRS)

    Bates, H. E.; Hill, D. M.; Jewett, D. N.

    1983-01-01

    High-purity silicon efficiently produced and transferred by continuous two-cycle reactor. New reactor operates in relatively-narrow temperature rate and uses large surfaces area to minimize heat expenditure and processing time in producing silicon by hydrogen reduction of trichlorosilane. Two cycles of reactor consists of silicon production and removal.

  1. NEUTRONIC REACTOR SHIELDING

    DOEpatents

    Borst, L.B.

    1961-07-11

    A special hydrogenous concrete shielding for reactors is described. In addition to Portland cement and water, the concrete essentially comprises 30 to 60% by weight barytes aggregate for enhanced attenuation of fast neutrons. The biological shields of AEC's Oak Ridge Graphite Reactor and Materials Testing Reactor are particular embodiments.

  2. High solids fermentation reactor

    DOEpatents

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-01-01

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  3. High solids fermentation reactor

    DOEpatents

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-03-02

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  4. Improved vortex reactor system

    DOEpatents

    Diebold, James P.; Scahill, John W.

    1995-01-01

    An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

  5. Advanced Test Reactor Tour

    SciTech Connect

    Miley, Don

    2011-01-01

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  6. Advanced Test Reactor Tour

    ScienceCinema

    Miley, Don

    2016-07-12

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  7. 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.

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

    SciTech Connect

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

    2006-01-20

    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.

  9. Assessing pretreatment reactor scaling through empirical analysis

    DOE PAGES

    Lischeske, James J.; Crawford, Nathan C.; Kuhn, Erik; ...

    2016-10-10

    Pretreatment is a critical step in the biochemical conversion of lignocellulosic biomass to fuels and chemicals. Due to the complexity of the physicochemical transformations involved, predictively scaling up technology from bench- to pilot-scale is difficult. This study examines how pretreatment effectiveness under nominally similar reaction conditions is influenced by pretreatment reactor design and scale using four different pretreatment reaction systems ranging from a 3 g batch reactor to a 10 dry-ton/d continuous reactor. The reactor systems examined were an Automated Solvent Extractor (ASE), Steam Explosion Reactor (SER), ZipperClave(R) reactor (ZCR), and Large Continuous Horizontal-Screw Reactor (LHR). To our knowledge, thismore » is the first such study performed on pretreatment reactors across a range of reaction conditions (time and temperature) and at different reactor scales. The comparative pretreatment performance results obtained for each reactor system were used to develop response surface models for total xylose yield after pretreatment and total sugar yield after pretreatment followed by enzymatic hydrolysis. Near- and very-near-optimal regions were defined as the set of conditions that the model identified as producing yields within one and two standard deviations of the optimum yield. Optimal conditions identified in the smallest-scale system (the ASE) were within the near-optimal region of the largest scale reactor system evaluated. A reaction severity factor modeling approach was shown to inadequately describe the optimal conditions in the ASE, incorrectly identifying a large set of sub-optimal conditions (as defined by the RSM) as optimal. The maximum total sugar yields for the ASE and LHR were 95%, while 89% was the optimum observed in the ZipperClave. The optimum condition identified using the automated and less costly to operate ASE system was within the very-near-optimal space for the total xylose yield of both the ZCR and the LHR, and

  10. Assessing pretreatment reactor scaling through empirical analysis

    SciTech Connect

    Lischeske, James J.; Crawford, Nathan C.; Kuhn, Erik; Nagle, Nicholas J.; Schell, Daniel J.; Tucker, Melvin P.; McMillan, James D.; Wolfrum, Edward J.

    2016-10-10

    Pretreatment is a critical step in the biochemical conversion of lignocellulosic biomass to fuels and chemicals. Due to the complexity of the physicochemical transformations involved, predictively scaling up technology from bench- to pilot-scale is difficult. This study examines how pretreatment effectiveness under nominally similar reaction conditions is influenced by pretreatment reactor design and scale using four different pretreatment reaction systems ranging from a 3 g batch reactor to a 10 dry-ton/d continuous reactor. The reactor systems examined were an Automated Solvent Extractor (ASE), Steam Explosion Reactor (SER), ZipperClave(R) reactor (ZCR), and Large Continuous Horizontal-Screw Reactor (LHR). To our knowledge, this is the first such study performed on pretreatment reactors across a range of reaction conditions (time and temperature) and at different reactor scales. The comparative pretreatment performance results obtained for each reactor system were used to develop response surface models for total xylose yield after pretreatment and total sugar yield after pretreatment followed by enzymatic hydrolysis. Near- and very-near-optimal regions were defined as the set of conditions that the model identified as producing yields within one and two standard deviations of the optimum yield. Optimal conditions identified in the smallest-scale system (the ASE) were within the near-optimal region of the largest scale reactor system evaluated. A reaction severity factor modeling approach was shown to inadequately describe the optimal conditions in the ASE, incorrectly identifying a large set of sub-optimal conditions (as defined by the RSM) as optimal. The maximum total sugar yields for the ASE and LHR were 95%, while 89% was the optimum observed in the ZipperClave. The optimum condition identified using the automated and less costly to operate ASE system was within the very-near-optimal space for the total xylose yield of both the ZCR and the LHR, and was

  11. Control system for a small fission reactor

    DOEpatents

    Burelbach, J.P.; Kann, W.J.; Saiveau, J.G.

    1985-02-08

    A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired.

  12. Nuclear reactor overflow line

    DOEpatents

    Severson, Wayne J.

    1976-01-01

    The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.

  13. HEAVY WATER MODERATED NEUTRONIC REACTOR

    DOEpatents

    Szilard, L.

    1958-04-29

    A nuclear reactor of the type which utilizes uranium fuel elements and a liquid coolant is described. The fuel elements are in the form of elongated tubes and are disposed within outer tubes extending through a tank containing heavy water, which acts as a moderator. The ends of the fuel tubes are connected by inlet and discharge headers, and liquid bismuth is circulated between the headers and through the fuel tubes for cooling. Helium is circulated through the annular space between the outer tubes in the tank and the fuel tubes to cool the water moderator to prevent boiling. The fuel tubes are covered with a steel lining, and suitable control means, heat exchange means, and pumping means for the coolants are provided to complete the reactor assembly.

  14. Nuclear reactor alignment plate configuration

    SciTech Connect

    Altman, David A; Forsyth, David R; Smith, Richard E; Singleton, Norman R

    2014-01-28

    An alignment plate that is attached to a core barrel of a pressurized water reactor and fits within slots within a top plate of a lower core shroud and upper core plate to maintain lateral alignment of the reactor internals. The alignment plate is connected to the core barrel through two vertically-spaced dowel pins that extend from the outside surface of the core barrel through a reinforcement pad and into corresponding holes in the alignment plate. Additionally, threaded fasteners are inserted around the perimeter of the reinforcement pad and into the alignment plate to further secure the alignment plate to the core barrel. A fillet weld also is deposited around the perimeter of the reinforcement pad. To accomodate thermal growth between the alignment plate and the core barrel, a gap is left above, below and at both sides of one of the dowel pins in the alignment plate holes through with the dowel pins pass.

  15. Spinning fluids reactor

    DOEpatents

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  16. High temperature reactors

    NASA Astrophysics Data System (ADS)

    Dulera, I. V.; Sinha, R. K.

    2008-12-01

    With the advent of high temperature reactors, nuclear energy, in addition to producing electricity, has shown enormous potential for the production of alternate transport energy carrier such as hydrogen. High efficiency hydrogen production processes need process heat at temperatures around 1173-1223 K. Bhabha Atomic Research Centre (BARC), is currently developing concepts of high temperature reactors capable of supplying process heat around 1273 K. These reactors would provide energy to facilitate combined production of hydrogen, electricity, and drinking water. Compact high temperature reactor is being developed as a technology demonstrator for associated technologies. Design has been also initiated for a 600 MWth innovative high temperature reactor. High temperature reactor development programme has opened new avenues for research in areas like advanced nuclear fuels, high temperature and corrosion resistant materials and protective coatings, heavy liquid metal coolant technologies, etc. The paper highlights design of these reactors and their material related requirements.

  17. Reactor water cleanup system

    DOEpatents

    Gluntz, Douglas M.; Taft, William E.

    1994-01-01

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

  18. Reactor water cleanup system

    DOEpatents

    Gluntz, D.M.; Taft, W.E.

    1994-12-20

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

  19. SNAP (Space Nuclear Auxiliary Power) Reactor Overview

    DTIC Science & Technology

    1984-08-01

    Although the program did not proceed past the developmental stage, Pratt and Whitney were commended by the Chairman of the AEC, Dr. Glenn Seaborg , for a...34Shift is Urged from CANEL on Nuclear Project," HTFG Courant, May 1965. 17. Seaborg Says CANEL Halt No Reflection on Its Work, Middletown Press, 7-6-65

  20. Safety analysis of the advanced thermionic initiative reactor

    NASA Astrophysics Data System (ADS)

    Lee, Hsing H.; Klein, Andrew C.

    1995-01-01

    Previously, detailed analysis was conducted to assess the technology developed for the Advanced Thermionic Initiative reactor. This analysis included the development of an overall system design code capability and the improvement of analytical models necessary for the assessment of the use of single cell thermionic fuel elements in a low power space nuclear reactor. The present analysis extends this effort to assess the nuclear criticality safety of the ATI reactor for various different scenarios. The analysis discusses the efficacy of different methods of reactor control such as control rods, and control drums.

  1. 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.

  2. ETR, TRA642. ON BASEMENT FLOOR. REACTOR VESSEL WILL BE PLACED ...

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

    ETR, TRA-642. ON BASEMENT FLOOR. REACTOR VESSEL WILL BE PLACED WITHIN THE INNER METAL FORM. WHEN CONCRETE IS POURED OUTSIDE THIS FORM, CONDUIT HOLES WILL BE PRESERVE SPACE THROUGH HOLES. INL NEGATIVE NO. 56-1507. Jack L. Anderson, Photographer, 5/8/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  3. METHOD OF FABRICATING A GRAPHITE MODERATED REACTOR

    DOEpatents

    Kratz, H.R.

    1963-05-01

    S>A nuclear reactor formed of spaced bodies of uranium and graphite blocks is improved by diffusing helium through the graphite blocks in order to replace the air in the pores of the graphite with helium. The helium-impregnated graphite conducts heat better, and absorbs neutrons less, than the original air- impregnated graphite. (AEC)

  4. Reactors Save Energy, Costs for Hydrogen Production

    NASA Technical Reports Server (NTRS)

    2014-01-01

    While examining fuel-reforming technology for fuel cells onboard aircraft, Glenn Research Center partnered with Garrettsville, Ohio-based Catacel Corporation through the Glenn Alliance Technology Exchange program and a Space Act Agreement. Catacel developed a stackable structural reactor that is now employed for commercial hydrogen production and results in energy savings of about 20 percent.

  5. HORIZONTAL BOILING REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1958-11-18

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

  6. Hybrid reactors. [Fuel cycle

    SciTech Connect

    Moir, R.W.

    1980-09-09

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of /sup 233/U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m/sup -2/, and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid.

  7. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Goett, J.J.

    1961-01-24

    A system is described which includes a neutronic reactor containing a dispersion of fissionable material in a liquid moderator as fuel and a conveyor to which a portion of the dispersion may be passed and wherein the self heat of the slurry evaporates the moderator. Means are provided for condensing the liquid moderator and returning it to the reactor and for conveying the dried fissionable material away from the reactor.

  8. NEUTRONIC REACTOR CONTROL

    DOEpatents

    Dreffin, R.S.

    1959-12-15

    A control means for a nuclear reactor is described. Particularly a device extending into the active portion of the reactor consisting of two hollow elements coaxially disposed and forming a channel therebetween, the cross sectional area of the channel increasing from each extremity of the device towards the center thereof. An element of neutron absorbing material is slidably positionable within the inner hollow element and a fluid reactor poison is introduced into the channel defined by the two hollow elements.

  9. The Integral Fast Reactor

    SciTech Connect

    Chang, Y.I.

    1988-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab.

  10. Improved vortex reactor system

    DOEpatents

    Diebold, J.P.; Scahill, J.W.

    1995-05-09

    An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

  11. FLOW SYSTEM FOR REACTOR

    DOEpatents

    Zinn, W.H.

    1963-06-11

    A reactor is designed with means for terminating the reaction when returning coolant is below a predetermined temperature. Coolant flowing from the reactor passes through a heat exchanger to a lower reservoir, and then circulates between the lower reservoir and an upper reservoir before being returned to the reactor. Means responsive to the temperature of the coolant in the return conduit terminate the chain reaction when the temperature reaches a predetermined minimum value. (AEC)

  12. University Reactor Sharing Program

    SciTech Connect

    Dr. W.D. Reece

    1999-09-01

    The University Reactor Sharing Program provides funding for reactor experimentation to institutions that do not normally have access to a research reactor. Research projects supported by the program include items such as dating geological material to producing high current super conducting magnets. The funding also gives small colleges and universities the opportunity to use the facility for teaching courses in nuclear processes; specifically neutron activation analysis and gamma spectroscopy.

  13. 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.-

  14. Reactor pressure vessel nozzle

    DOEpatents

    Challberg, Roy C.; Upton, Hubert A.

    1994-01-01

    A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough.

  15. Reactor pressure vessel nozzle

    DOEpatents

    Challberg, R.C.; Upton, H.A.

    1994-10-04

    A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough. 2 figs.

  16. HOMOGENEOUS NUCLEAR POWER REACTOR

    DOEpatents

    King, L.D.P.

    1959-09-01

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

  17. Membrane reactors at Degussa.

    PubMed

    Wöltinger, Jens; Karau, Andreas; Leuchtenberger, Wolfgang; Drauz, Karlheinz

    2005-01-01

    The review covers the development of membrane reactor technologies at Degussa for the synthesis of fine chemicals. The operation of fed-batch or continuous biocatalytic processes in the enzyme membrane reactor (EMR) is well established at Degussa. Degussa has experience of running EMRs from laboratory gram scale up to a production scale of several hundreds of tons per year. The transfer of the enzyme membrane reactor from biocatalysis to chemical catalysis in the chemzyme membrane reactor (CMR) is discussed. Various homogeneous catalysts have been investigated in the CMR, and the scope and limitation of this new technique is discussed.

  18. Pressurized fluidized bed reactor

    DOEpatents

    Isaksson, J.

    1996-03-19

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

  19. Pressurized fluidized bed reactor

    DOEpatents

    Isaksson, Juhani

    1996-01-01

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

  20. Remote Reactor Monitoring

    SciTech Connect

    Bernstein, Adam; Dazeley, Steve; Dobie, Doug; Marleau, Peter; Brennan, Jim; Gerling, Mark; Sumner, Matthew; Sweany, Melinda

    2014-10-21

    The overall goal of the WATCHMAN project is to experimentally demonstrate the potential of water Cerenkov antineutrino detectors as a tool for remote monitoring of nuclear reactors. In particular, the project seeks to field a large prototype gadolinium-doped, water-based antineutrino detector to demonstrate sensitivity to a power reactor at ~10 kilometer standoff using a kiloton scale detector. The technology under development, when fully realized at large scale, could provide remote near-real-time information about reactor existence and operational status for small operating nuclear reactors out to distances of many hundreds of kilometers.

  1. A Potential NASA Research Reactor to Support NTR Development

    NASA Technical Reports Server (NTRS)

    Eades, Michael; Gerrish, Harold; Hardin, Leroy

    2013-01-01

    In support of efforts for research into the design and development of a man rated Nuclear Thermal Rocket (NTR) engine, the National Aeronautics and Space Administration (NASA), Marshall Space Flight Center (MSFC), is evaluating the potential for building a Nuclear Regulatory Commission (NRC) licensed research reactor. The proposed reactor would be licensed by NASA and operated jointly by NASA and university partners. The purpose of this reactor would be to perform further research into the technologies and systems needed for a successful NTR project and promote nuclear training and education.

  2. Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

    SciTech Connect

    Not Available

    1992-04-01

    This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

  3. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOEpatents

    Schreiber, Roger B.; Fero, Arnold H.; Sejvar, James

    1997-01-01

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor.

  4. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOEpatents

    Schreiber, R.B.; Fero, A.H.; Sejvar, J.

    1997-12-16

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor. 8 figs.

  5. Making Latex Microspheres in Space

    NASA Technical Reports Server (NTRS)

    Kornfeld, D. M.; Vanderhoff, J. W.; El-Aasser, M. S.; Micale, F. J.; Sudol, E. D.; Tseng, C. M.; Silwanowicz, A.

    1986-01-01

    Equipment yields larger, more uniform particles. Two NASA reports describe first commercial product to be manufactured in space. Product monodisperse latex, suspension of spherical particles of essentially same diameter. Carried aboard Space Shuttle on its orbital missions, monodisperse latex reactor (MLR) produces spheres of much larger size than possible on Earth. Mircospheres 30 micrometers in diameter produced, whereas 5 micrometers is limit for Earthbound reactors. Microspheres as large as 100 micrometers scheduled for production in MLR.

  6. 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.

  7. NEUTRONIC REACTOR SHIELD AND SPACER CONSTRUCTION

    DOEpatents

    Wigner, E.P.; Ohlinger, L.A.

    1958-11-18

    Reactors of the heterogeneous, graphite moderated, fluid cooled type and shielding and spacing plugs for the coolant channels thereof are reported. In this design, the coolant passages extend horizontally through the moderator structure, accommodating the fuel elements in abutting end-to-end relationship, and have access openings through the outer shield at one face of the reactor to facilitate loading of the fuel elements. In the outer ends of the channels which extend through the shields are provided spacers and shielding plugs designed to offer minimal reslstance to coolant fluid flow while preventing emanation of harmful radiation through the access openings when closed between loadings.

  8. Polymerization Reactor Engineering.

    ERIC Educational Resources Information Center

    Skaates, J. Michael

    1987-01-01

    Describes a polymerization reactor engineering course offered at Michigan Technological University which focuses on the design and operation of industrial polymerization reactors to achieve a desired degree of polymerization and molecular weight distribution. Provides a list of the course topics and assigned readings. (TW)

  9. N Reactor hydrogen control

    SciTech Connect

    Shuford, D.H.; Kripps, L.J.

    1988-08-01

    Following the accident at the Chernobyl nuclear power reactor in the Soviet Union, a number of reviews were conducted of the N Reactor. Hydrogen generation during postulates severe accidents and the possibility of resulting hydrogen deflagrations/detonations that could affect confinement integrity were issues raised in several reviews, along with recommendations for adding hydrogen mitigation features. To respond to these reviews, an N Reactor Safety Enhancement Program and a subsequent Accelerated Safety Enhancement Program were initiated to address all post-Chernobyl N Reactor review findings. The Safety Enhancement Program and Accelerated Safety Enhancement Program efforts involving hydrogen control included the following: Calculate the potential hydrogen source for a range of severe accidents at the N Reactor to establish an acceptable design basis for the hydrogen mitigation system; Analyze the N Reactor confinement hydrogen mixing capability to identify areas of concern and to the verify effectiveness of the hydrogen mitigation system; Select, design, and construct a hydrogen mitigation system to enhance the N Reactor capability to accommodate possible hydrogen generation from postulated severe accidents; Provide post-accident hydrogen monitoring as an operator aid in assessing confinement conditions. In additions, it was necessary to verify that incorporation of the hydrogen mitigation system had no adverse impact N Reactor safety (e.g., radiological consequence analyses). 77 refs., 25 figs., 10 tabs.

  10. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1959-02-10

    A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

  11. NEUTRONIC REACTOR BURIAL ASSEMBLY

    DOEpatents

    Treshow, M.

    1961-05-01

    A burial assembly is shown whereby an entire reactor core may be encased with lead shielding, withdrawn from the reactor site and buried. This is made possible by a five-piece interlocking arrangement that may be easily put together by remote control with no aligning of bolt holes or other such close adjustments being necessary.

  12. REFLECTOR FOR NEUTRONIC REACTORS

    DOEpatents

    Fraas, A.P.

    1963-08-01

    A reflector for nuclear reactors that comprises an assembly of closely packed graphite rods disposed with their major axes substantially perpendicular to the interface between the reactor core and the reflector is described. Each graphite rod is round in transverse cross section at (at least) its interface end and is provided, at that end, with a coaxial, inwardly tapering hole. (AEC)

  13. The Integral Fast Reactor

    SciTech Connect

    Till, C.E.; Chang, Y.I. ); Lineberry, M.J. )

    1990-01-01

    Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 5 refs., 3 figs.

  14. Light water reactor program

    SciTech Connect

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  15. REACTOR FUEL SCAVENGING MEANS

    DOEpatents

    Coffinberry, A.S.

    1962-04-10

    A process for removing fission products from reactor liquid fuel without interfering with the reactor's normal operation or causing a significant change in its fuel composition is described. The process consists of mixing a liquid scavenger alloy composed of about 44 at.% plutoniunm, 33 at.% lanthanum, and 23 at.% nickel or cobalt with a plutonium alloy reactor fuel containing about 3 at.% lanthanum; removing a portion of the fuel and scavenger alloy from the reactor core and replacing it with an equal amount of the fresh scavenger alloy; transferring the portion to a quiescent zone where the scavenger and the plutonium fuel form two distinct liquid layers with the fission products being dissolved in the lanthanum-rich scavenger layer; and the clean plutonium-rich fuel layer being returned to the reactor core. (AEC)

  16. Status of French reactors

    SciTech Connect

    Ballagny, A.

    1997-08-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

  17. Models for multimegawatt space power systems

    SciTech Connect

    Edenburn, M.W.

    1990-06-01

    This report describes models for multimegawatt, space power systems which Sandia's Advanced Power Systems Division has constructed to help evaluate space power systems for SDI's Space Power Office. Five system models and models for associated components are presented for both open (power system waste products are exhausted into space) and closed (no waste products) systems: open, burst mode, hydrogen cooled nuclear reactor -- turboalternator system; open, hydrogen-oxygen combustion turboalternator system; closed, nuclear reactor powered Brayton cycle system; closed, liquid metal Rankine cycle system; and closed, in-core, reactor therminonic system. The models estimate performance and mass for the components in each of these systems. 17 refs., 8 figs., 15 tabs.

  18. Slurry reactor design studies

    SciTech Connect

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. ); Akgerman, A. ); Smith, J.M. )

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  19. Nuclear reactor control column

    DOEpatents

    Bachovchin, Dennis M.

    1982-01-01

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

  20. Reactor Safety Research Programs

    SciTech Connect

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  1. REACTOR BASE, SOUTHEAST CORNER. INTERIOR WILL CONTAIN REACTOR TANK, COOLING ...

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

    REACTOR BASE, SOUTHEAST CORNER. INTERIOR WILL CONTAIN REACTOR TANK, COOLING WATER PIPES, COOLING AIR DUCTS, AND SHIELDING. INL NEGATIVE NO. 776. Unknown Photographer, 10/1950 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  2. Reactor Simulator Testing

    NASA Technical Reports Server (NTRS)

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

    2013-01-01

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

  3. Fast Breeder Reactor studies

    SciTech Connect

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  4. REACTOR CONTROL SYSTEM

    DOEpatents

    MacNeill, J.H.; Estabrook, J.Y.

    1960-05-10

    A reactor control system including a continuous tape passing through a first coolant passageway, over idler rollers, back through another parallel passageway, and over motor-driven rollers is described. Discrete portions of fuel or poison are carried on two opposed active sections of the tape. Driving the tape in forward or reverse directions causes both active sections to be simultaneously inserted or withdrawn uniformly, tending to maintain a more uniform flux within the reactor. The system is particularly useful in mobile reactors, where reduced inertial resistance to control rod movement is important.

  5. NUCLEAR REACTOR FUEL SYSTEMS

    DOEpatents

    Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

    1959-09-15

    Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

  6. COOLED NEUTRONIC REACTOR

    DOEpatents

    Binner, C.R.; Wilkie, C.B.

    1958-03-18

    This patent relates to a design for a reactor of the type in which a fluid coolant is flowed through the active portion of the reactor. This design provides for the cooling of the shielding material as well as the reactor core by the same fluid coolant. The core structure is a solid moderator having coolant channels in which are disposed the fuel elements in rod or slug form. The coolant fluid enters the chamber in the shield, in which the core is located, passes over the inner surface of said chamber, enters the core structure at the center, passes through the coolant channels over the fuel elements and out through exhaust ducts.

  7. Microfluidic electrochemical reactors

    DOEpatents

    Nuzzo, Ralph G [Champaign, IL; Mitrovski, Svetlana M [Urbana, IL

    2011-03-22

    A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

  8. Nuclear reactor reflector

    DOEpatents

    Hopkins, Ronald J.; Land, John T.; Misvel, Michael C.

    1994-01-01

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

  9. Nuclear reactor reflector

    DOEpatents

    Hopkins, R.J.; Land, J.T.; Misvel, M.C.

    1994-06-07

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.

  10. NEUTRONIC REACTOR CONTROL

    DOEpatents

    Metcalf, H.E.

    1958-10-14

    Methods of controlling reactors are presented. Specifically, a plurality of neutron absorber members are adjustably disposed in the reactor core at different distances from the center thereof. The absorber members extend into the core from opposite faces thereof and are operated by motive means coupled in a manner to simultaneously withdraw at least one of the absorber members while inserting one of the other absorber members. This feature effects fine control of the neutron reproduction ratio by varying the total volume of the reactor effective in developing the neutronic reaction.

  11. CONTROL FOR NEUTRONIC REACTOR

    DOEpatents

    Lichtenberger, H.V.; Cameron, R.A.

    1959-03-31

    S>A control rod operating device in a nuclear reactor of the type in which the control rod is gradually withdrawn from the reactor to a position desired during stable operation is described. The apparatus is comprised essentially of a stop member movable in the direction of withdrawal of the control rod, a follower on the control rod engageable with the stop and means urging the follower against the stop in the direction of withdrawal. A means responsive to disengagement of the follower from the stop is provided for actuating the control rod to return to the reactor shut-down position.

  12. Spherical torus fusion reactor

    DOEpatents

    Martin Peng, Y.K.M.

    1985-10-03

    The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

  13. Lateral restraint assembly for reactor core

    DOEpatents

    Gorholt, Wilhelm; Luci, Raymond K.

    1986-01-01

    A restraint assembly for use in restraining lateral movement of a reactor core relative to a reactor vessel wherein a plurality of restraint assemblies are interposed between the reactor core and the reactor vessel in circumferentially spaced relation about the core. Each lateral restraint assembly includes a face plate urged against the outer periphery of the core by a plurality of compression springs which enable radial preloading of outer reflector blocks about the core and resist low-level lateral motion of the core. A fixed radial key member cooperates with each face plate in a manner enabling vertical movement of the face plate relative to the key member but restraining movement of the face plate transverse to the key member in a plane transverse to the center axis of the core. In this manner, the key members which have their axes transverse to or subtending acute angles with the direction of a high energy force tending to move the core laterally relative to the reactor vessel restrain such lateral movement.

  14. RSMASS: A preliminary reactor/shield mass model for SDI applications

    SciTech Connect

    Marshall, A.C.

    1986-08-01

    A simple mathematical model (RSMASS) has been developed to provide rapid estimates of reactor and shield masses for space-based reactor power systems. Approximations are used rather than correlations or detailed calculations to estimate the reactor fuel mass and the masses of the moderator, structure, reflector, pressure vessel, miscellaneous components, and the reactor shield. The fuel mass is determined either by neutronics limits, specific power limits, or fuel burnup limits - whichever yields the largest mass. RSMASS requires the reactor power and energy, 24 reactor parameters, and 20 shield parameters to be specified. This parametric approach should provide good mass estimates for a very broad range of reactor types. Reactor and shield masses calculated by RSMASS were found to be in good agreement with the masses obtained from detailed calculations.

  15. NEUTRONIC REACTOR FUEL COMPOSITION

    DOEpatents

    Thurber, W.C.

    1961-01-10

    Uranium-aluminum alloys in which boron is homogeneously dispersed by adding it as a nickel boride are described. These compositions have particular utility as fuels for neutronic reactors, boron being present as a burnable poison.

  16. NEUTRONIC REACTOR STRUCTURE

    DOEpatents

    Daniels, F.

    1961-10-24

    A reactor core, comprised of vertical stacks of hexagonal blocks of beryllium oxide having axial cylindrical apertures extending therethrough and cylindrical rods of a sintered mixture of uranium dioxide and beryllium oxide, is described. (AEC)

  17. Reactor Neutrino Spectra

    NASA Astrophysics Data System (ADS)

    Hayes, Anna C.; Vogel, Petr

    2016-10-01

    We present a review of the antineutrino spectra emitted from reactors. Knowledge of these spectra and their associated uncertainties is crucial for neutrino oscillation studies. The spectra used to date have been determined either by converting measured electron spectra to antineutrino spectra or by summing over all of the thousands of transitions that make up the spectra, using modern databases as input. The uncertainties in the subdominant corrections to β-decay plague both methods, and we provide estimates of these uncertainties. Improving on current knowledge of the antineutrino spectra from reactors will require new experiments. Such experiments would also address the so-called reactor neutrino anomaly and the possible origin of the shoulder observed in the antineutrino spectra measured in recent high-statistics reactor neutrino experiments.

  18. Packed Bed Reactor Experiment

    NASA Video Gallery

    The purpose of the Packed Bed Reactor Experiment in low gravity is to determine how a mixture of gas and liquid flows through a packed bed in reduced gravity. A packed bed consists of a metal pipe ...

  19. 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.

  20. Technology and applications of space nuclear power

    NASA Technical Reports Server (NTRS)

    Reck, Gregory M.; Rosen, Robert; Bennett, Gary L.; Schnyer, A. D.

    1991-01-01

    Requirements for a number of potential NASA civil space missions are addressed, and the nuclear power technology base to meet these requirements is described. Particular attention is given to applications of space nuclear power to lunar, Mars, and science missions and the technology status of space nuclear power with emphasis on dynamic isotope and space nuclear reactor power systems.

  1. Molten metal reactors

    DOEpatents

    Bingham, Dennis N; Klingler, Kerry M; Turner, Terry D; Wilding, Bruce M

    2013-11-05

    A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.

  2. Compact power reactor

    DOEpatents

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

    1978-01-01

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

  3. Future reactor experiments

    SciTech Connect

    Wen, Liangjian

    2015-07-15

    The non-zero neutrino mixing angle θ{sub 13} has been discovered and precisely measured by the current generation short-baseline reactor neutrino experiments. It opens the gate of measuring the leptonic CP-violating phase and enables the neutrino mass ordering. The JUNO and RENO-50 proposals aim at resolving the neutrino mass ordering using reactors. The experiment design, physics sensitivity, technical challenges as well as the progresses of those two proposed experiments are reviewed in this paper.

  4. K-Reactor readiness

    SciTech Connect

    Rice, P.D.

    1991-12-04

    This document describes some of the more significant accomplishments in the reactor restart program and details the magnitude and extent of the work completed to bring K-Reactor to a state of restart readiness. The discussion of restart achievements is organized into the three major categories of personnel, programs, and plant. Also presented is information on the scope and extent of internal and external oversight of the efforts, as well as some details on the startup plan.

  5. NEUTRONIC REACTOR CONSTRUCTION AND OPERATION

    DOEpatents

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

    1960-03-15

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

  6. F Reactor Inspection

    SciTech Connect

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-10-29

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

  7. F Reactor Inspection

    ScienceCinema

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2016-07-12

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

  8. Moon base reactor system

    NASA Technical Reports Server (NTRS)

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

    1989-01-01

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

  9. Reactor Safety Research Programs

    SciTech Connect

    Dotson, CW

    1980-08-01

    This document summarizes the work performed by Pacific Northwest laboratory from October 1 through December 31, 1979, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, lspra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  10. Space nuclear power and man's extraterrestrial civilization

    SciTech Connect

    Angelo, J.J.; Buden, D.

    1983-01-01

    This paper examines leading space nuclear power technology candidates. Particular emphasis is given the heat-pipe reactor technology currently under development at the Los Alamos National Laboratory. This program is aimed at developing a 10-100 kWe, 7-year lifetime space nuclear power plant. As the demand for space-based power reaches megawatt levels, other nuclear reactor designs including: solid core, fluidized bed, and gaseous core, are considered.

  11. 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.

  12. Pin-Type Gas Cooled Reactor for Nuclear Electric Propulsion

    NASA Astrophysics Data System (ADS)

    Wright, Steven A.; Lipinski, Ronald J.

    2003-01-01

    This paper describes a point design for a pin-type Gas-Cooled Reactor concept that uses a fuel pin design similar to the SP100 fuel pin. The Gas-Cooled Reactor is designed to operate at 100 kWe for 7 years plus have a reduced power mode of 20% power for a duration of 5 years. The power system uses a gas-cooled, UN-fueled, pin-type reactor to heat He/Xe gas that flows directly into a recuperated Brayton system to produce electricity. Heat is rejected to space via a thermal radiator that unfolds in space. The reactor contains approximately 154 kg of 93.15 % enriched UN in 313 fuel pins. The fuel is clad with rhenium-lined Nb-1Zr. The pressures vessel and ducting are cooled by the 900 K He/Xe gas inlet flow or by thermal radiation. This permits all pressure boundaries to be made of superalloy metals rather than refractory metals, which greatly reduces the cost and development schedule required by the project. The reactor contains sufficient rhenium (a neutron poison) to make the reactor subcritical under water immersion accidents without the use of internal shutdown rods. The mass of the reactor and reflectors is about 750 kg.

  13. Recent BeO-reflector-controlled reactor experiments in ZPPR

    SciTech Connect

    McFarlane, H.F.; Brumbach, S.B.; Carpenter, S.G.; Collins, P.J.; McKnight, R.D.

    1986-01-01

    Integral reactor physics measurements were performed on a BeO-reflected fast reactor assembly in the ZPPR facility during January and February of 1985. The measurements emphasized power distributions and reflector control worths in two different critical states. The measurements have been analyzed using three-dimensional deterministic and Monte Carlo methods and the ENDF/B-V.2 nuclear data library. Together the measurements and analyses form a modern, reliable, benchmark data set for testing calculational methods that will be used in predicting some of the design parameters for future space reactors.

  14. The temperature effect and safety of the TOPAZ reactor

    NASA Astrophysics Data System (ADS)

    Artiukhov, G. Ia.; Zelentsov, S. N.; Ionkin, V. I.; Kudriavtsev, V. P.; Makarenkov, Iu. D.; Marin, S. N.; Pupko, V. Ia.; Raskach, F. P.

    The TOPAZ reactor-converter is a thermionic space nuclear-power system of less than 50 kW capacity; its reactor employs a high-temperature moderator containing hydrogen and highly enriched fuel. The designers of TOPAZ gave special attention to the determination of temperature effects (TEs) and temperature reactivity coefficient (TRC). A critical assembly that was heated isothermally by an electric furnace was used to conduct measurements of TEs, TRC, energy-release fields, control and safety system performance, and reactor spectrum and dynamic characteristics.

  15. Dismantling the nuclear research reactor Thetis

    SciTech Connect

    Michiels, P.

    2013-07-01

    The research reactor Thetis, in service since 1967 and stopped in 2003, is part of the laboratories of the institution of nuclear science of the University of Ghent. The reactor, of the pool-type, was used as a neutron-source for the production of radio-isotopes and for activation analyses. The reactor is situated in a water pool with inner diameter of 3 m. and a depth of 7.5 m. The reactor core is situated 5.3 m under water level. Besides the reactor, the pool contains pneumatic loops, handling tools, graphite blocks for neutron moderation and other experimental equipment. The building houses storage rooms for fissile material and sources, a pneumatic circuit for transportation of samples, primary and secondary cooling circuits, water cleaning resin circuits, a ventilation system and other necessary devices. Because of the experimental character of the reactor, laboratories with glove boxes and other tools were needed and are included in the dismantling program. The building is in 3 levels with a crawl-space. The ground-floor contains the ventilation installation, the purification circuits with tanks, cooling circuits and pneumatic transport system. On the first floor, around the reactor hall, the control-room, visiting area, end-station for pneumatic transport, waste-storage room, fuel storage room and the labs are located. The second floor contains a few laboratories and end stations of the two high speed transfer tubes. The lowest level of the pool is situated under ground level. The reactor has been operated at a power of 150 kW and had a max operating power of 250 kW. Belgoprocess has been selected to decommission the reactor, the labs, storage halls and associated circuits to free release the building for conventional reuse and for the removal of all its internals as legal defined. Besides the dose-rate risk and contamination risk, there is also an asbestos risk of contamination. During construction of the installation, asbestos-containing materials were

  16. HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING

    SciTech Connect

    Austin, W.; Brinkley, D.

    2011-10-13

    The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D&D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment

  17. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    SciTech Connect

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  18. REACTOR GROUT THERMAL PROPERTIES

    SciTech Connect

    Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

    2011-01-28

    Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

  19. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Dickson, J.J.

    1963-09-24

    A method is described whereby fuel tubes or pins are cut, loaded with fuel pellets and a heat transfer medium, sealed at each end with slotted fittings, and assembled into a rectangular tube bundle to form a fuel element. The tubes comprising the fuel element are laterally connected between their ends by clips and tabs to form a linear group of spaced parallel tubes, which receive their vertical support by resting on a grid. The advantages of this method are that it permits elimination of structural material (e.g., fuel-element cans) within the reactor core, and removal of at least one fuel pin from an element and replacement thereof so that a burnable poison may be utilized during the core lifetime. (AEC)

  20. NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY

    DOEpatents

    Stengel, F.G.

    1963-12-24

    A method of fabricating nuclear reactor fuel element assemblies having a plurality of longitudinally extending flat fuel elements in spaced parallel relation to each other to form channels is presented. One side of a flat side plate is held contiguous to the ends of the elements and a welding means is passed along the other side of the platertransverse to the direction of the longitudinal extension of the elements. The setting and speed of travel of the welding means is set to cause penetration of the side plate with welds at bridge the gap in each channel between adjacent fuel elements with a weld-through bubble of predetermined size. The fabrication of a high strength, dependable fuel element is provided, and the reduction of distortion and high production costs are facilitated by this method. (AEC)

  1. Nuclear reactor internals alignment configuration

    DOEpatents

    Gilmore, Charles B.; Singleton, Norman R.

    2009-11-10

    An alignment system that employs jacking block assemblies and alignment posts around the periphery of the top plate of a nuclear reactor lower internals core shroud to align an upper core plate with the lower internals and the core shroud with the core barrel. The distal ends of the alignment posts are chamfered and are closely received within notches machined in the upper core plate at spaced locations around the outer circumference of the upper core plate. The jacking block assemblies are used to center the core shroud in the core barrel and the alignment posts assure the proper orientation of the upper core plate. The alignment posts may alternately be formed in the upper core plate and the notches may be formed in top plate.

  2. Nuclear reactor composite fuel assembly

    DOEpatents

    Burgess, Donn M.; Marr, Duane R.; Cappiello, Michael W.; Omberg, Ronald P.

    1980-01-01

    A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  3. NEP Space Test Program Objective

    NASA Technical Reports Server (NTRS)

    1992-01-01

    The objective of the Nuclear Electric Propulsion (NEP) space test program is to launch an NEP satellite powered by the Russion Topaz 2 reactor by Dec. 1995. The primary goals of the NEP space test program are as follows: (1) demonstrate the feasibility of launching a space nuclear power system; (2) demonstrate and orbit adjust capability using NEP; (3) evaluate the in-orbit performance of the Topaz 2 reactor and selected electric thrusters; and (4) measure, analyze, and model the NEP self-induced environment. The discussion is presented in vugraph form.

  4. A look at the Soviet space nuclear power program

    NASA Technical Reports Server (NTRS)

    Bennett, Gary L.

    1989-01-01

    For the most part Soviet nuclear power sources have been low-power nuclear reactors using a thermoelectric conversion principle. Recently the Soviet Union has flown two satellites using a higher power reactor that employs a thermionic conversion system. Despite reentry of two of the earlier reactors on board Cosmos 954 and Cosmos 1402 and the recent potential accident involving Cosmos 1900, the evidence points toward a continued Soviet use of nuclear power sources in space. Information in the open literature on the Soviet space nuclear power program, including the Romashka Topaz, the new reactor based on the Topaz program, and the RORSAT reactor experience, is summarized.

  5. REACTOR AND NOVEL METHOD

    DOEpatents

    Young, G.J.; Ohlinger, L.A.

    1958-06-24

    A nuclear reactor of the type which uses a liquid fuel and a method of controlling such a reactor are described. The reactor is comprised essentially of a tank for containing the liquid fuel such as a slurry of discrete particles of fissionnble material suspended in a heavy water moderator, and a control means in the form of a disc of neutron absorbirg material disposed below the top surface of the slurry and parallel thereto. The diameter of the disc is slightly smaller than the diameter of the tank and the disc is perforated to permit a flow of the slurry therethrough. The function of the disc is to divide the body of slurry into two separate portions, the lower portion being of a critical size to sustain a nuclear chain reaction and the upper portion between the top surface of the slurry and the top surface of the disc being of a non-critical size. The method of operation is to raise the disc in the reactor until the lower portion of the slurry has reached a critical size when it is desired to initiate the reaction, and to lower the disc in the reactor to reduce the size of the lower active portion the slurry to below criticality when it is desired to stop the reaction.

  6. EBT reactor analysis

    SciTech Connect

    Uckan, N. A.; Jaeger, E. F.; Santoro, R. T.; Spong, D. A.; Uckan, T.; Owen, L. W.; Barnes, J. M.; McBride, J. B.

    1983-08-01

    This report summarizes the results of a recent ELMO Bumpy Torus (EBT) reactor study that includes ring and core plasma properties with consistent treatment of coupled ring-core stability criteria and power balance requirements. The principal finding is that constraints imposed by these coupling and other physics and technology considerations permit a broad operating window for reactor design optimization. Within this operating window, physics and engineering systems analysis and cost sensitivity studies indicate that reactors with <..beta../sub core/> approx. 6 to 10%, P approx. 1200 to 1700 MW(e), wall loading approx. 1.0 to 2.5 MW/m/sup 2/, and recirculating power fraction (including ring-sustaining power and all other reactors auxiliaries) approx. 10 to 15% are possible. A number of concept improvements are also proposed that are found to offer the potential for further improvement of the reactor size and parameters. These include, but are not limited to, the use of: (1) supplementary coils or noncircular mirror coils to improve magnetic geometry and reduce size, (2) energetic ion rings to improve ring power requirements, (3) positive potential to enhance confinement and reduce size, and (4) profile control to improve stability and overall fusion power density.

  7. Methanation assembly using multiple reactors

    DOEpatents

    Jahnke, Fred C.; Parab, Sanjay C.

    2007-07-24

    A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

  8. Looking Southwest at Reactor Box Furnaces With Reactor Boxes and ...

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

    Looking Southwest at Reactor Box Furnaces With Reactor Boxes and Repossessed Uranium in Recycle Recovery Building - Hematite Fuel Fabrication Facility, Recycle Recovery Building, 3300 State Road P, Festus, Jefferson County, MO

  9. Reactor for exothermic reactions

    DOEpatents

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

    1993-01-01

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

  10. Reactor for exothermic reactions

    DOEpatents

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

    1993-03-02

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

  11. MERCHANT MARINE SHIP REACTOR

    DOEpatents

    Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

    1961-05-01

    A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

  12. Merchant Marine Ship Reactor

    DOEpatents

    Sankovich, M. F.; Mumm, J. F.; North, Jr, D. C.; Rock, H. R.; Gestson, D. K.

    1961-05-01

    A nuclear reactor for use in a merchant marine ship is described. The reactor is of pressurized, light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements that are confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass. (AEC)

  13. Heat dissipating nuclear reactor

    DOEpatents

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

  14. Nuclear reactor safety device

    DOEpatents

    Hutter, Ernest

    1986-01-01

    A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

  15. Thermionic Reactor Design Studies

    SciTech Connect

    Schock, Alfred

    1994-08-01

    Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

  16. Dynamic bed reactor

    SciTech Connect

    Stormo, K.E.

    1996-07-02

    A dynamic bed reactor is disclosed in which a compressible open cell foam matrix is periodically compressed and expanded to move a liquid or fluid through the matrix. In preferred embodiments, the matrix contains an active material such as an enzyme, biological cell, chelating agent, oligonucleotide, adsorbent or other material that acts upon the liquid or fluid passing through the matrix. The active material may be physically immobilized in the matrix, or attached by covalent or ionic bonds. Microbeads, substantially all of which have diameters less than 50 microns, can be used to immobilize the active material in the matrix and further improve reactor efficiency. A particularly preferred matrix is made of open cell polyurethane foam, which adsorbs pollutants such as polychlorophenol or o-nitrophenol. The reactors of the present invention allow unidirectional non-laminar flow through the matrix, and promote intimate exposure of liquid reactants to active agents such as microorganisms immobilized in the matrix. 27 figs.

  17. Dynamic bed reactor

    SciTech Connect

    Stormo, Keith E.

    1996-07-02

    A dynamic bed reactor is disclosed in which a compressible open cell foam matrix is periodically compressed and expanded to move a liquid or fluid through the matrix. In preferred embodiments, the matrix contains an active material such as an enzyme, biological cell, chelating agent, oligonucleotide, adsorbent or other material that acts upon the liquid or fluid passing through the matrix. The active material may be physically immobilized in the matrix, or attached by covalent or ionic bonds. Microbeads, substantially all of which have diameters less than 50 microns, can be used to immobilize the active material in the matrix and further improve reactor efficiency. A particularly preferred matrix is made of open cell polyurethane foam, which adsorbs pollutants such as polychlorophenol or o-nitrophenol. The reactors of the present invention allow unidirectional non-laminar flow through the matrix, and promote intimate exposure of liquid reactants to active agents such as microorganisms immobilized in the matrix.

  18. Heat dissipating nuclear reactor

    DOEpatents

    Hunsbedt, Anstein; Lazarus, Jonathan D.

    1987-01-01

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extends from the metal base plate downwardly and outwardly into the earth.

  19. REACTOR CONTROL DEVICE

    DOEpatents

    Graham, R.H.

    1962-09-01

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

  20. COMSOL-based Nuclear Reactor Kinetics Studies at the HFIR

    SciTech Connect

    Chandler, David; Freels, James D; Maldonado, G Ivan; Primm, Trent

    2011-01-01

    The computational ability to accurately predict the dynamic behavior of a nuclear reactor core in response to reactivity-induced perturbations is an important subject in reactor physics. Space-time and point kinetics methodologies were developed for the purpose of studying the transient-induced behavior of the High Flux Isotope Reactor s (HFIR) compact core. The space-time simulations employed the three-energy-group neutron diffusion equations, and transients initiated by control cylinder and hydraulic tube rabbit ejections were studied. The work presented here is the first step towards creating a comprehensive multiphysics methodology for studying the dynamic behavior of the HFIR core during reactivity perturbations. The results of these studies show that point kinetics is adequate for small perturbations in which the power distribution is assumed to be time-independent, but space-time methods must be utilized to determine localized effects.